BackgroundThe floating nuclear power plant (FNPP) is a vital energy supply method for future ocean exploitation and island construction. The typical fuel type of FNPP is similar to the onshore nuclear power plant, i.e., the rod bundle fuel assembly. Due to the effect of ocean waves and wind, the FNNP would be in continuous motion. Rolling is one of the most common types of motion. It can induce the periodical change of the inertial force field of the coolant and the change of flow and heat transfer characteristics of the rod bundle channel. Coupling with neutronic and thermohydraulic, as a result, the operation characteristics, safety, and economics of the FNPP can be affected.PurposeThis study aims to investigate the impact of neutronic-thermo-hydraulic coupling on natural circulation characteristics with the rod bundle channel under rolling motion condition.MethodsA natural circulation system with a 5×5 square array basic rod bundles channel was taken as research object, and it was designed and built on a mechanical rolling platform. Then, based on the point reactor kinetic model, the coupling of neutronic-thermohydraulic-motion was achieved by real-time data acquisition of thermal parameters and calculation of real-time nuclear power, and the effects of fuel temperature feedback and coolant temperature feedback on single-phase natural circulation were considered. Finally, an experimental study on the low-pressure single-phase natural circulation under rolling motion condition was carried out.ResultsUnder static conditions, neutronic-thermo-hydraulic coupling makes the power fluctuate slightly, reactivity and power fluctuation amplitude increase with the increase of temperature feedback coefficient. When the feedback coefficient is lower than -5×10-4 ℃-1, fuel temperature feedback has a greater impact on power than coolant temperature feedback. Increasing the fuel temperature feedback coefficient reduces system stability. Under rolling motion conditions, the smaller the rolling amplitude or the shorter the rolling period, the smaller the amplitude of the introduced power fluctuations. During the initiation of the rolling motion, neutronic-thermo-hydraulic coupling significantly increases 50% of the time for the system to establish a new stable circulation.ConclusionsThese results provide a valuable application for further investigations of the FNNP's design and validation.

BackgroundThe primary heat exchanger (PHX) used in the 10 MWt Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL), is a U-tube heat exchanger, where the shell side (hot side) contains the fuel salt from the primary loop and the tube side (cold side) carries the coolant salt from the secondary loop.PurposeThis study aims to deepen the understanding and mastery of the operational characteristics of molten salt heat exchangers, and to accumulate experience in their design and operation within molten salt reactors.MethodsFirstly, based on the design parameters, the MSRE-PHX was modeled, and theoretical calculations for shell and tube hear exchanger were conducted using the Kern method and the Bell-Delaware method. Then, software simulations were performed using HTRI Xchanger Suite, and computational fluid dynamics (CFD) simulations were also carried out with Ansys Fluent. Finally, critical performance metrics, such as the heat transfer coefficient, the pressure drop, and the heat transfer power, were obtained and compared to the MSRE operation data.ResultsThe comparison results indicate that the discrepancies from theoretical calculations, HTRI software, and CFD simulations, are all within acceptable margins to the experimental data. Notably, the greatest variance is found with the Kern method, which showed a deviation in heat transfer quantity of about 15%, while the smallest discrepancy is observed in the overall heat transfer coefficient calculated using HTRI software, differing by merely 0.16% from the experimental data.ConclusionsAll of the methods are suitable and applicable for designing and studying a molten salt shell and tube heat exchanger. Moreover, the CFD simulation can provide fine localized details of the heat transfer and flow of the molten salt fluid. This offers substantial theoretical support and practical guidance for the future design and improvement of molten salt heat exchangers.

BackgroundReactor core computational fluid dynamics (CFD) plays a crucial role in identifying core vulnerabilities, optimizing feature structures, and improving safety and economic in nuclear reactors. However, conventional pressurized water reactor fuel assemblies often feature a multitude of spacer grids with mixing vanes, leading to challenges in mesh generation and numerical solution instability, excessive computational resource requirements. The current momentum source model established on the basis of the mechanism of fluid-structure interaction has not considered the effect of the low-pressure region on the fluid downstream of the mixing vanes, leading to significant errors in predicting the axial flow distribution downstream of the mixing vanes. Furthermore, it is challenging to identify the solid domain of the mixing vanes and to add momentum source terms.PurposeThis study aims to present a joint simulation scheme based on detailed porous media and momentum source modeling to simulate coolant flow in 5×5 rod bundle channels with mixing vanes, hence to reduce cells, lower mesh generation difficulty, and enhance numerical stability during the CFD solving process.MethodsThis scheme employed a detailed porous media approach in the spacer zone, while adopting Global Momentum Source Model in the vane zone. Simultaneously, a domain identification scheme was developed to determine the placement of momentum sources and detailed porous media models. The position of mixing vanes within the fluid domain was accurately located by this approach and established detailed porous media and momentum source models based on the fluid-structure interactions in the grid spacer zone and leeward side and windward side of mixing vanes. To simulate the flow field distribution within the spacer zone, a detailed porous media model was employed to enhance local flow resistance, thereby achieving an accurate simulation of the flow field distribution in the spacer zone. Finally, validation against experimental and body-fitted mesh simulations was performed to examine the effectiveness of this scheme in simulating flow blockage, fluid flow, mixing, and vortex shedding.ResultsThis scheme, compared to the momentum source scheme, exhibits stronger numerical stability. In the vane zone, the established momentum source model simultaneously considers the effects of the leeward side and the windward side of mixing vanes, leading to a more accurate prediction of axial flow velocity and heat transfer downstream of the mixing vanes. This approach allows for modeling without needing to consider the structure of the spacer grids with mixing vanes, thus greatly simplifying mesh generation. It achieves complete structured mesh modeling, significantly reducing the number of cells, and enhancing computational efficiency. Validation confirm the effectiveness of this scheme and results in a 90% reduction in cells and a 60% decrease in computational time for modeling and simulation of a 5×5 rod bundle channel with mixing vanes.ConclusionsThis scheme offers simplicity in modeling, reduces CFD computation time, insensitivity to mesh, and superior robustness. Furthermore, when identifying larger-scale components, the approach involves identifying the multi-span fuel components, since the mixing vanes form a regular array in both axial and transverse directions. Therefore, domain identification at a larger scale can be achieved by modifying coordinates, applying momentum source model developed in this paper.

BackgroundThe molten salt reactor (MSR) is one of the six advanced reactors identified by the Generation IV International Forum. The MSR exhibits unique characteristics, such as intrinsic safety, sustainable development, nuclear nonproliferation, natural resource protection, and economic efficiency. After a liquid-fuel molten salt reactor shuts down, residual decay heat in the reactor core is passively dissipated to the environment through a natural circulation loop of the molten salt. However, the decay heat from the molten salt in the main circuit can impact the thermal capacity of the overall system.PurposeThis study aims to determine the thermal characteristics of the passive system by establishing an analysis basis for a natural cycle model to examine the effects of natural circulation loop on the physical properties of the molten salt, the loop structure, and equipment resistance coefficient K.MethodsBased on the direct reactor auxiliary cooling system (DRACS) for MSR, the natural circulation model of the passive residual heat-removal loop of a liquid-fuel molten salt reactor was established using self-developed Python analysis program, and the temperature distribution of the molten salt in the loop was explored using the numerical model. Then, the effects of different physical properties, loop structures, and core and heat exchanger resistance coefficients K on the heat transfer and flow characteristics of the residual heat-removal system were analyzed. Finally, the effects of critical factors on the residual heat-removal capacity of the reactor core were analyzed using self-developed Python program for natural circulation calculation equations code (NCCC) and validated by using the natural cycle experimental results of the DRACS circuit in CIET1.0.ResultsThe findings indicate that the presence of decay heat from molten salt in a system loop decreases the natural circulation-driven heat transfer by the molten salt within the core. Based on the significance analysis results show that fuel salt density, specific heat capacity, and height difference between the hot and cold cores are parameters that significantly influence the natural circulation capacity. When these three parameter values are increased by 15% separately, the residual heat removal capacities increase by 26.02%, 15.00%, and 18.59%, respectively.ConclusionsResults of this study demonstrate that the molten salt properties, circuit structure, and equipment resistance coefficient all affect the natural circulation heat-removal capacity.

BackgroundSubchannel analysis of fuel assemblies is critical for the development of lead-bismuth reactors.PurposeThis study aims to modify and optimize the COBRA subchannel program to make it suitable for lead-bismuth reactors and validate its performance.MethodsModifications were made to the COBRA subchannel program, involving adjusting physical properties, convective heat transfer models, friction models, and turbulence mixing models. The performance of the modified program was evaluated by comparing its numerical calculation results to experimental data. To optimize results over a wide range of mass flow rate conditions, an optimization method based on a subchannel model and coupled with a neural network was proposed, and the influence of mass flow rate on calculation accuracy was analyzed.ResultsThe comparison results demonstrate that the modified subchannel program performs well under experimental conditions, with an error of no more than 5% compared with experimental results and no more than 3% compared with FLUENT results. The application of neural networks is found to improve accuracy and reduce errors by an order of magnitude.ConclusionsThe optimized subchannel analysis method, derived from the modifications and neural network coupling, can accurately predict outlet temperatures for lead-bismuth reactors under a wide range of mass flow rate conditions. This method provides valuable guidance for the design of such reactors.

Background252Cf is a high-intensity isotope neutron source in great demand for scientific research and device development. Currently, it is produced only in high-flux reactors in the United States and Russia, which rely on imports in China for a long time.PurposeThis study aims to analyze key factors of 252Cf production by irradiation based on the preliminary design scheme of a high-flux fast reactor.MethodsFirstly, an irradiation target design was implemented, and the fission deposition energies and energy spectra were calculated for three irradiated target designs using different zirconium hydride and Eu2O3 absorbers. Then, burnup calculations for heavy and light curium targets were performed using a burnup calculation program STEP, which was developed by China Institute of Atomic Energy. The experimental values for 252Cf produced by irradiation in the United States were then compared. Finally, the calculation results were analyzed using the energy spectrum and cross-section.ResultsComparison results between simulation and experiment indicate that 245Cm is the key nuclide affecting the production of 252Cf. Utilizing the hard-energy spectral characteristics of high-flux fast reactors can effectively reduce the fission loss of the target and increase the production of 252Cf.ConclusionsThe calculations and analysis in this study can provide theoretical and technical support for the high-flux fast neutron research reactor irradiation production of 252Cf.

BackgroundHeat pipe cooled reactor (HPR) has many characteristics, such as reliability, inherent safety, small volume, modularity, and solid core. The nuclear fuel of solid core is seriously affected by high temperature, strong irradiation, and solid constraint when operating, which affect the heat transfer performance and mechanical properties of the core seriously. The stress and gap heat transfer caused by the contact between monolith and other components change nonlinearly with the increase of burnup, and they influence each other. Therefore, the coupled irradiation-thermal-mechanical behavior of the monolith is a complex multi-physics phenomena.PurposeThis study aims to develop a coupled irradiation-thermal-mechanical model to explore the characteristics of gap variation, heat transfer and mechanics during the lifetime of solid core.MethodsFirst of all, based on the geometric parameter and material of a typical solid core of HPR with fuel rod composed of UO2 pellets and 316 stainless steel cladding, a coupled irradiation-thermal-mechanical model was developed and applied to the finite element multi-physics field analysis software COMSOL. The calculation parameter settings mainly referred to the design parameters of the MegaPower reactor. Then, a thermal conductivity model changing with the increase of burnup for UO2, the gap heat transfer model and mechanical contact were introduced in the gaps in the solid core, and both irradiation-induced deformation effect including densification and fission product swelling, and creep effect of UO2 pellets and 316 stainless steel monolith were taken into account. Finally, the model was applied to calculating the typical HPR and the characteristics of gap variation, heat transfer and mechanics were analyzed.ResultsAnalysis results show that pellet temperature and creep of monolith and cladding increase after complete contact between monolith and cladding. A smaller average number of heat pipes around the fuel rod result in higher temperature and stress distribution in the nearby area, and the cladding in this area has a risk of creep failure during its lifetime caused by internal pressure of the fuel rod and contact pressure between the monolith and cladding.ConclusionsThe gap contact can affect the heat transfer and mechanical properties of the solid core of HPR, and even result in an increase in the risk of cladding failure.

BackgroundThe quantification of uncertainty has become a common requirement in reactor physics analysis whilst the covariance data of nuclear data serves as the foundational data for conducting uncertainty quantification.PurposeThis study aims to develop a covariance data generation module, named covar_calc, embedded in the nuclear data processing software NECP-Atlas, a nuclear data processing program independently developed by the laboratory of nuclear engineering computational physics (NECP) of Xi'an Jiaotong University, to produce continuous energy covariance data for Monte Carlo programs and multi-group covariance data for deterministic programs.MethodsWithin the framework of NECP-Atlas, covar_calc module was developed to process all covariance data provided in evaluated nuclear data, according to the different storage formats of nuclear data and different computational methods. Covariance data of various parameters such as average fission neutron multiplicities, cross sections, angular distributions of secondary particles, fission spectrum, resonance parameters, and neutron activation cross-sections, were all be processed by covar_calc. Comparative verification was carried out with the covariance data production module in the nuclear data processing software NJOY21. Finally, sensitivity coefficients for different benchmarks were calculated using both the Monte Carlo calculation code NECP-MCX and uncertainty analysis code NECP-UNICORN, and the final uncertainties were computed by incorporating both continuous energy covariance data and multi-group covariance data, and utilizing the "Sandwich formula".ResultsComparison results demonstrate that the accuracy of the multi-group covariance data produced by NECP-Atlas is equivalent to that of NJOY21 and the maximum bias is less than 0.1%. The uncertainties calculated using the multi-group covariance data generated by NECP-Atlas exhibit comparable accuracy to those obtained with NJOY21.ConclusionsThe precision in creating both continuous energy covariance and multi-group covariance presented in this study meets the requirements for usage in Monte Carlo programs and deterministic programs, validating the efficacy of covar_calc module within NECP-Atlas for uncertainty quantification in reactor physics analysis.

Background3C-SiC (β-SiC) exhibits outstanding electrochemical properties and radiation resistance, surpassing hexagonal-phase silicon carbide in irradiation resistance. As a promising candidate for the next generation of structural materials in nuclear applications and high-performance precision electronic devices for challenging reactor environments, the material has been garnering significant attention in recent decades. Within this realm, the exploration of one-dimensional silicon carbide nanomaterials has become a focal point in silicon carbide materials research. However, their practical applications have been hindered by challenges such as the absence of effective nanomaterial processing methods and processing complexities. Notably, ultrasonic processing technology has demonstrated effectiveness in addressing these challenges.PurposeThis study aims to synthesize and study 3C-SiC nanowires (NWs), investigating their ultrasonic fracture behavior for comprehensive understanding of the ultrasonic fracture characteristics of 3C-SiC NWs, laying the groundwork for basic research in the processing of one-dimensional SiC nanomaterials.MethodsFirstly, silicon carbide nanowires were prepared by chemical vapor deposition. Then the silicon carbide nanowires were characterized by microstructure observed by scanning electron microscope (SEM), transmission electron microscope (TEM), X-Ray diffraction (XRD) and Raman spectrum. Subsequently, the 3C-SiC nanowires were subjected to ultrasonic treatment, and the average length-to-diameter ratios of the ultrasonically treated nanowires were statistically analyzed to elucidate the effect of ultrasonic treatment on the nanowires. Finally, the strength of the silicon carbide nanowires was estimated by combining the bubble jet model and statistical data.ResultsThe findings reveal that the synthesized 3C-SiC NWs are predominantly of the 3C-SiC phase, exhibiting a notable presence of stacking faults. Ultrasonic treatment significantly influences the SiC NWs, leading to a noticeable reduction in the average Length-Diameter ratio, stabilizing at 18 post-treatment.ConclusionsThe observed results align with the effects of bubble jetting and are corroborated by the ultrasonic fragmentation behavior of 3C-SiC NWs. These findings offer valuable insights for the manipulation of nanomaterial size and morphology. This study provides a new perspective for the ultrasonic cutting of silicon carbide nanowires and the strength research of nanowires, and is of great significance for the future application of silicon carbide nanowires in the field of nuclear energy.

BackgroundElement capture logging can be used to determine the elemental contents of rocks in formations.PurposeThis study aims to obtain accurate elemental composition, the content of shale reservoirs and the inaugural parameter well for shale gas in the Cambrian Niutitang Formation of the Baojing Block, located in the Middle Yangtze region of China, with emphasis on the developmental and distributional characteristics of shale gas reservoirs in this formation.MethodsThe shale gas parameter well BY2 was taken as the research object, by interpreting and processing elemental capture logging data, precise elemental compositions of the shale reservoirs were determined. This analysis led to the creation of a comprehensive geochemical index profile for the Niutitang Formation. Additionally, elemental geochemical indicators were used to identify and reconstruct the paleosedimentary environments.ResultsThe analysis results reveal that the predominant elements in the Niutitang Formation's shale are Si, Al, and Fe, accompanied by lower amounts of K, Ca, Mg, and S. The shale features relatively high concentrations of Si, Fe, and S, which contributed to its enhanced fracturing ability. The sedimentation process of this shale can be categorized as active continental margin sedimentation. The source material for the sedimentary rocks is originated from the Kangdian ancient land, located in the northwest. The sedimentation is primarily normal but was influenced by the presence of hydrothermal fluids in the region's active tectonic zone.ConclusionsThe upper section of the Niutitang formation is subject to a dry climate during its depositional period, featuring gentle slope sedimentation at the periphery of a stagnant basin and a lack of oxygen, characterizing with high water salinity, ample land supply, and low water body paleoproductivity this region. Conversely, the lower section is experienced a humid climate and served as a deep-water retention basin, with limited the land supply, but high water salinity and paleoproductivity, leading to the accumulation of organic matter. The aquatic setting is primarily anaerobic, conditions that are conducive to the preservation of organic matter, and provides an optimal sedimentary environment for the generation and concentration of shale gas.

BackgroundThe neutron scintillator detector is one of the main detectors for China Spallation Neutron Source (CSNS). High detection efficiency is the design goal of the second-generation neutron scintillator detector, which is in process of development, composed of neutron scintillation screen, wavelength transfer optical fiber and photoelectric converter. CSNS_VASD chip is an application-specific integrated circuit (ASIC) developed for China Spallation Neutron Source as the second-generation neutron scintillator detector.PurposeThis study aims to evaluate the performance of the front-end ASIC chip CSNS_VASD by experimental test.MethodsThe combined architecture of "ASIC test board plus digital readout board" was adopted for overall test system design. The ASIC test board was used to amplify, shape and distinguish signals whilst the digital readout board took the roles of configuring the ASIC and its auxiliary circuits, packing and caching the test data, and sending it to the back end for processing and analysis through optical fiber Ethernet. The relevant performance parameters of the chip were tested in the laboratory using exponential like wave test signal with amplitude of 10~120 MV, pulse width of 1 μs and frequency of 100 K as the input signal of ASIC chip, and the neutron beamline of CSNS.ResultsThe results show that the nonlinear error of CSNS_VASD is better than 1%, the equivalent voltage noise is better than 0.63 mV, the crosstalk value is better than 0.89%, and the detection efficiency of the detector is 40.7%@0.1 nm.ConclusionsAll of the test indexes meet the requirement of the design target. The successful development of the CSNS_VASD chip provides a reliable technical guarantee for the smooth construction of China Spallation Neutron Source.

BackgroundWith the development of nuclear medicine, the amount of medical radioactive waste increase rapidly. The radioactivity of nuclides in medical radioactive liquid waste must be monitored to meet the relevant standards before discharge of the radioactive liquid waste. The volume of the waste sample and its distribution around the detector have a direct impact on the detection efficiency.PurposeThis study aims to explore the variation law of the size parameters of the optimal Marlin cup sample box and provide a basis for subsequent monitoring methods.MethodsThe LaBr3(Ce) crystal was applied to the detection of nuclide activity in medical radioactive waste liquid. Geant4 tool was employed to establish a LaBr3(Ce) crystal detection model. The changing rules of the optimal Marin cup sample box size parameters were explored using a Ø25.4 mm×25.4 mm LaBr3(Ce) detector, and 3D printed photosensitive resin samples were used in the laboratory box for verification experiments.ResultsExperimental results show that simply increasing the sample volume cannot improve the detection efficiency, and the change trend of the detection efficiency in the depth direction of the annular part of the sample container tends to be flat with increase of the sample volume. The optimal size ratio of the Marinelli beaker is that the depth of the annular portion (h2) and the radius (r) are approximately two times the length of the detector crystal and the diameter of the hollow cavity, respectively, and the ratio of the radius (r) to the height of the sample container (H) is approximately 0.5. The experimental results of the full energy peak detection efficiency with optimized sample container size are consistent with the simulation results, and the relative deviation is better than 2.5%.ConclusionsThe results of the study provide an important technical reference for detector selection, sampling container design, and processing and traceability methods of medical radioactive liquid waste monitoring devices.

BackgroundIn cave disposal environment, surrounding rock is the last barrier to prevent radionuclides from entering the environment, and the colloid of surrounding rock produced in the long-term disposal process increases the risk of radionuclides migration.PurposeThis study aims to explore the adsorption performance of surrounding rock colloid for Sr2+ disposed the repository, and the stability of surrounding rock colloid.MethodsFirst of all, the surrounding rock colloid sample was prepared with approximate mass concentration of 0.03 g·L-1, zeta potential of 21.53 mV and particle size of 205.7 nm. Then, the effects of time, pH value, various ions under different concentrations, and other factors on the adsorption properties, such as the zeta potential and average particle size of the surrounding rock colloids were investigated by experimental measurements. Finally, the adsorption kinetics and adsorption isothermal model were analyzed.ResultsExperimental results show that the colloid adsorption Sr2+ reaches equilibrium at 12 h, and the equilibrium adsorption capacity is 41.79 mg·g-1. In alkaline environment, the adsorption capacity increases with the increase of pH. The adsorption of Sr2+ by colloid of surrounding rock is inhibited by different ions, and the inhibitory effect of cations is greater than that of anions.ConclusionsResults of this study indicate that pH value, temperature and partial ions in the groundwater of the cave disposal facility all have influence on the stability of surrounding rock, and the surrounding rock colloid has good stability in the groundwater environment of the repository. The adsorption of Sr2+ by surrounding rock colloid conforms to the quasi-second-order kinetic model and the Freundlich adsorption isothermal model.

BackgroundMost of domestic radiation therapy doses can only be traced back to air kerma of 60Co γ-ray at present in China. The uncertainty of directly tracing radiation therapy dose to absorbed dose to water is much smaller compared to tracing radiation therapy dose to air kerma, which must be converted to absorbed dose to water.PurposeThis study aims to solve the problem of traceability and transfer of 60Co γ-ray absorbed dose to water, as well as the problem that the dose measurement of radiotherapy cannot be directly traced through the absorbed dose to water, NIMTT (National Institute of Measurement and Testing Technology) has established an absolute measurement device of 60Co γ-ray absorbed dose to water.MethodsAn absolute measurement device of 60Co γ-ray absorbed dose to water was established by National Institute of Measurement and Testing Technology (NIMTT) using water calorimeterfor direct measurement of absorbed dose to water. The correction factors for water absorbed dose measurement were obtained by experimental and simulation methods. A comparison of laboratory reproducing results was made between NIMTT and NRC (National Research Council, Canada) using the method of transmitting standard to further verify the accuracy and consistency of the measurement of 60Co γ-ray absorbed dose to water. Two PTW ionization chambers were used as the transfer standard.ResultsThe reproduced results of two laboratories for absolute measurement of 60Co γ-ray absorbed dose to water are consistent within the relative standard uncertainty of 0.71%, and the normalized error value of En is -0.45.ConclusionsThe comparison results have verified that NIMTT laboratory has the ability of traceability and transmission of absorbed dose to water. The results of this study provide a reference for the absolute measurement of absorbed dose to water for 60Co γ-ray.

BackgroundRuthenium (Ru) exhibits outstanding mechanical properties and chemical stability, along with excellent optical properties such as high reflectivity and a larger critical angle. These attributes make it a good candidate coating material for optical components in X-ray free electron laser facilities.PurposeThis study aims to analyze the fundamental thermal evolutions of the Ru electron and lattice subsystems under irradiation of X-ray free electron laser pulses for better understanding of the laser-optics interaction mechanism under operating conditions.MethodsThe numerical method was adopted to solve the two-temperature model (TTM) in terms of simulating the interaction process between X-ray free electron laser pulses and thin ruthenium films. The laser energy incident on the Ru target material was the actual energy absorbed by the material and the pulse laser incidence was in the x-direction with a film thickness of 50 nm, spatial discretization step size of 0.2 nm and the time step size of 0.02 fs. Simutanously, the influence of different laser source parameters, such as pulse width, penetration depth, energy density, etc., on the thermal effects during the interaction between laser and metal Ru was explored, and the temperature evolutions of the Ru electron and lattice subsystems were obtained, as well as the corresponding dependency relationships of the heat transfer processes on source parameters.ResultsNumerical solution of the TTM indicate that the system equilibrium temperature of Ru is identified to be positively correlated with energy density. The peak electron temperature of Ru decreases with the increasing of pulse width. Besides, as penetration depth increases, the equilibrium temperature of Ru will decrease until a stable level. As for the front surface, the time taken to achieve electron-lattice equilibrium decrease with the increasing of electron-phonon coupling coefficient.ConclusionsSimulation data and theoretical analyses presented in this study throw an insight into the thermal response of the optical material (Ru) to laser irradiation.

The nuclear heating rate is a critical parameter for reactor core design and irradiation tests, it is typically determined via experimental measurements in a reactor experimental channel. In-pile calibration of calorimeter is an important method for measuring the nuclear heating rate in a fission reactor. This paper summarizes the working principle of the common in-pile calorimeter and reviews the current situation and research progress with regard to in-pile calorimetry employed worldwide. The structural design and performance characteristics of single-cell and multi-cell (differential) calorimeters are compared, and the design method of multimodal integrated measurement device, which represents one of development directions of calorimeters, is introduced. Moreover, the in- and out-pile calibration principles and application methods are described. The advantages and disadvantages of the calibration methods are analyzed, and the developing trend and direction for next-generation in-pile calorimeters are prospected.

BackgroundUranium dioxide (UO2) has been broadly employed as nuclear fuel in nuclear reactors. However, the poor thermal conductivity of UO2 compromises the safety of the reactor owing to possible sharp temperature gradients. Graphene oxide (GO) is a promising additive to improve the thermal conductivity of UO2 owing to its excellent thermal performance.PurposeThis study aims to achieve uniform distribution of GO in UO2 pellets, effectively controlling the doping amounts to enhance the thermal conductivity of UO2 pellets.MethodsGO-doped UO2 powders with different doping amounts were prepared using solid-liquid mixing and ammonium diuranate (ADU) co-precipitation methods. After establishing the optimized powdering process, UO2-GO composite fuel pellets were prepared by spark plasma sintering (SPS). Properties of the UO2-GO composite fuel pellets, such as density, grain size, physical phase, and thermal conductivity, were examined using scanning electron microscope (SEM), energy dispersion spectrometer (EDS), metallographic microscope, laser pulse thermal conductivity meter, etc., and compared with those of conventional pure UO2 pellets.ResultsThe results showed that the density of UO2-GO pellets can reach up to 97.6% T.D.. The thermal conductivity of UO2-GO pellets with 1.5 wt.% doped GO is 85.9% higher than that of conventional UO2 pellets at 1 000 ℃. The grain size of the UO2-GO pellets is uniform, and the GO is homogeneously distributed at the grain boundary to form a bridging thermal conduction network.ConclusionsThe thermal conductivity of UO2 pellets is successfully improved through GO doping.

BackgroundThe enrichment of nuclear fuel may significantly influence the fuel performance of a reactor.PurposeThis study aims to explore the effect of U3Si2-Al plate-type fuel enrichment on its performance,MethodsThe fuel performance analysis code BEEs-Plates, neutronics Monte Carlo code OpenMC, and the one-dimensional system analysis code ZEBRA were coupled together within the MOOSE (Multiphysics Object-Oriented Simulation Environment) framework. Then the coupling code was employed to conduct the multiphysics coupling calculation for JRR-3 (Japanese Research Reactor No.3) fuel assembly enriched at 15%, 20%, and 25%. The data exchange among the three codes were realized with the help of interpolation transfer method implemented in MOOSE. Additionally, neutron physical parameters and fuel performance parameters after 231 d of operation were analyzed when the average fuel consumption of the module was 125.71 GWd?tU-1.ResultsThe calculation results indicate that the max power density of fuel assembly enriched at 25% is 18% higher than the assembly enriched at 15%. Due to the high thermal conductivity of aluminum, the temperature difference in the fuel assembly is almost negligible whilst there is a significant difference in the fast neutron fluence. The results of fuel temperature and fast neutron fluence show that the volumetric strain is more obviously affected by the fuel temperature. Specifically, the plastic strain of the assembly with 25% enrichment is approximately 40% higher than that of the assembly with 15% enrichment.ConclusionThe analysis results of this study suggest that the assembly with a higher enrichment is more prone to failure.

BackgroundPassive safety system reliability is generally evaluated through best estimation plus uncertainty (BEPU) analysis. An important step in the evaluation process is sensitivity analysis of parameters, which is used to identify key system parameters to reduce the complexity of the model. However, local sensitivity analysis methods based on linear or monotonic assumptions may yield incorrect sensitivity results for complex nuclear power systems. Meanwhile, applying the global sensitivity method is difficult in practical engineering because of its high calculation cost.PurposeThis study aims to develop an efficient and low-cost global sensitivity analysis method for passive systems under ocean conditions.MethodsFirstly, the low-rank approximation (LRA) method was employed to improve the Sobol method based on Monte Carlo simulation. The number of unknown coefficients was significantly reduced by using the multivariate-based tensor product. Then, the LRA coefficients were used to calculate the sensitivity index, and the validity of the proposed method was verified by addressing several sensitivity analysis benchmark questions. Finally, taking an integrated pressurized water reactor including the passive residual heat removal system as the object, a simulation program for thermal-hydraulics analysis under ocean conditions was developed. And its sensitivity analysis was conducted using the proposed method.ResultsThe results show that the proposed method can accurately identify system key parameters after only 200 simulation calculations taking about 55 min, and the sensitivity ranking results are consistent with those obtained by Sobol method after 1.0×105 simulation calculations taking about 19 d.ConclusionsThe efficient global sensitivity analysis method established in this study can provide effective guidance for reliability analysis and design optimization of passive systems.

BackgroundCAT-1 (China Astro-Torus 1) is a levitated dipole field magnetic confinement device, which mainly used for dipole plasma physics experiments, requiring a central floating superconducting coil to be stably levitated for at least 5 h without cooling or power supply.PurposeThis study aims to design a levitation control system of coupling superconducting levitation coil and floating coil for CAT-1 dipole device.MethodsAccording to the design parameters of the suspension magnet system of the CAT-1 device, Simulink model of the control system was established and applied to the simulation. Based on Routh-Hurwitz stability criterion, the influence of PID (Proportion-Integral-Derivative) control strategy on stability control was studied. The selection range of stability control parameters was determined to ensure the stable levitation of 1 200 kg, 5 MA floating magnet.ResultsSimulation results show that under ideal conditions, delay time of the PD (Proportion-Derivative) control system is 0.046 3 s, rise time is 0.154 5 s, peak time is 0.628 3 s, adjustment time is 0.084 8 s, and overshot δ=1.6. It means that PD can restore the levitated superconducting ring to the preset balance position in a short time, and the load of the circuit can be greatly reduced by adopting the appropriate starting mode.ConclusionThe results provide key technical support for the design and development of levitated superconducting dipole field devices.

BackgroundSteam generator tube rupture (SGTR) accidents in lead-bismuth cooled fast reactors result in the generation of numerous steam bubbles owing to the interaction of the high-temperature liquid lead-bismuth eutectic (LBE) from the primary circuit and high-pressure subcooled water of the secondary circuit. These steam bubbles carried by the LBE may enter the core, causing local heat transfer deterioration and a power transient, seriously affecting reactor safety.PurposeThis study aims to elucidate the movement and dynamic behaviors of steam bubbles in liquid LBE and develop a drag coefficient model applicable to bubble migration in LBE for assessing the safety of the core during SGTR accidents.MethodsBased on the coupled level-set and volume-of-fluid (CLSVOF) method, a three-dimensional numerical model for calculating the drag coefficient of steam bubbles in LBE was established to study their movement and dynamic behaviors. Firstly, the deformation, velocity, and trajectory characteristics of bubbles in LBE were analyzed, and their drag coefficients were estimated. Then, the simulated drag coefficient values were compared with those from existing models whilst the prediction performance of the model proposed by Tomiyama was superior to those of the others. Finally, Tomiyama's drag model was optimized by introducing the We number, and its applicability was analyzed.ResultsAnalysis results show that the errors of the existing drag coefficient models are large under the condition of bubble–LBE two-phase flow, and the calculation error of the optimized model for the bubble drag coefficient in LBE is within 15%.ConclusionsThe feasibility of drag calculation model optimized in this study is verified, demonstrating its suitability for calculating the drag coefficient of steam bubbles in LBE.

BackgroundFast neutron pulsed reactors are sensitive to wall scattering neutrons, and their waveforms are changed by reflected neutrons. In addition, their operation may be adversely affected when there are too many reflected neutrons.PurposeThis study aims to solve the problem of fast neutron pulse reactor wall reflection neutron.MethodThe point reactor kinetic method containing the reflection effect, Monte Carlo method, and ANSYS were combined to analyze the Godiva-I transient process in fast neutron pulsed reactor with wall-reflected neutron effect. Firstly, the quench coefficient of the fast neutron pulsed reactor was calculated. Then, the point reactor kinetic method containing the reflection effect was established. Finally, the thermal power obtained from the neutronics calculation was combined with the ANSYS thermal-mechanical module to establish the thermal-mechanical calculation method for fast neutron pulsed reactor, and the effect of wall-reflected neutrons was analyzed and calculated.ResultsThe results show that the reflected neutrons increased the rear edge of the pulse. The reactivity decreases when the flat is washed whilst the core displacement and stress are improved.ConclusionsThe method established in this study can reasonably explain the phenomenon of reduction in attenuation and the increase in power after pulse.

BackgroundMolten salt reactors (MSRs) feature high temperature, low pressure, high chemical stability, and nuclear non-proliferation, which make them promising for a wide range of applications. However, owing to the fluid nature of the fuel, changes in the core structure of MSRs are bound to affect the fuel distribution and loading, consequently affecting the physical parameters of the core. Currently, the research on the impact of structural changes of the MSR core, composed of dispersed graphite components, on reactivity is insufficient.PurposeThis study aims to explores the influence of core structure changes on reactivity.MethodsThe core design model of new solid hexagonal graphite module MSR was taken as a reference, the small geometric changes, such as deformation and displacement of the core components, caused by factors including thermal expansion, fluid erosion, core vibration, and graphite irradiation, were analyzed using MCNP code. Additionally, the impact of these structural changes in the core on reactivity was investigated in details.ResultsThe results indicate that the structural changes in the core caused by thermal expansion introduce negative reactivity. The reactivity introduced by fluid erosion and core vibration, causing graphite component displacement, fluctuates. However, the overall trend shows that a shift of the graphite components towards the center of the core introduces positive reactivity, whereas shift towards the outer periphery of the core introduces negative reactivity. Graphite irradiation-induced deformation initially decreases reactivity and then increases it. At the end of the core's lifespan, the reactivity is still lower than at the beginning of the lifespan. This reactivity change can be compensated for by online feed or control rod movement, which has a limited impact on the operation of the MSR. However, the critical issue of control should be considered when this batch of fuel is reinserted into a new core.ConclusionsResults of this study suggest the necessity of constraining the graphite components within a certain range to ensure the safe operation of the reactor, providing an important reference for the design, operation, and maintenance of MSRs.

BackgroundLaser cladding, recognized for its cost-effectiveness and high efficiency, has become a focal point in the field of laser remanufacturing. GX4CrNi13-4 martensitic stainless steel produced by laser cladding is a widely used structural material in nuclear power plants.PurposeThis study aims to enhance the mechanical properties of GX4CrNi13-4 martensitic stainless steel, fabricated using laser cladding technology, through different heat treatments that cause microstructure modification.MethodsThe GX4CrNi13-4 stainless steel sample was prepared using laser cladding technology, and its heat treatment microstructure was studied in details. Firstly, thermal expansion experiments identified the onset temperature of austenitic phase transformation of sample at 620 °C, serving as a pivotal reference for developing heat treatment schemes. Two distinct heat treatment processes, i.e., solution treatment plus aging (STPA) at 1 050 °C for 1 h followed by a similar treatment at 550 °C for 4 h and single aging (SA) at 620 °C for 2 h, were applied to experiments. The effects of these treatments on the microstructure and mechanical performance of the cladding were comparatively analyzed by using X-ray diffraction (XRD), optical microscopy, scanning electron microscopy (SEM), and transmission electron microscopy (TEM) were employed to characterize the post-treatment microstructure and phase distribution. Tensile tests at room temperature were performed on samples before and after heat treatment.ResultsExperimental results indicate that the as-cladded GX4CrNi13-4 stainless steel exhibits a dual-phase microstructure primarily comprising martensite and ferrite, with continuous network-like ferrite precipitated along martensitic boundaries, accompanied by a minor presence of residual austenite. Post STPA, the matrix still predominantly comprises martensite and ferrite, but the continuous network-like ferrite decomposes, and numerous micrometer-scale transgranular precipitates within the martensite are observed. This led to a slight improvement in plasticity but a significant decrease in strength. The SA treatment of the cladded samples, performed at the critical temperature for austenitic phase transformation, induces the formation of the reversed austenitic phase. This phase, during tensile deformation, triggers the transformation induced plasticity (TRIP) effect. Furthermore, the network-like ferrite precipitated along the martensite decomposes into a dispersed distribution post-SA. The combined effect of TRIP and ferrite decomposition notably enhances the plasticity of the laser-cladded GX4CrNi13-4 stainless steel while effectively maintaining its strength.ConclusionsThe use of austenitic phase transition temperature for aging in this study, coupled with the synergistic effect of reversed austenite TRIP and ferrite decomposition, successfully achieves a balanced strength-plasticity performance in laser-cladded GX4CrNi13-4 stainless steel. Appropriate heat treatment and microstructural control emerge as effective strategies to improve the comprehensive mechanical properties of materials.

BackgroundMo leaching rate is one of the key product properties for high molybdenum radioactive waste vitrification. Furthermore, soluble molybdenum yellow phase is easy to precipitate, resulting in the enhancement of Mo leaching. It was reported that ZnO can promote the network stability and chemical durability in borosilicate glass to some extent. Whereas, if the composition of solidification glass is complex, then whether ZnO can still improve the chemical stability is worth studying. On the other hand, due to the long period of chemical stability test, it is hoped that the Mo leaching rate can be predicted by mathematical simulation to accelerate the research process.PurposeThis study aims to investigate the effect of ZnO in borosilicate glass on Mo leaching rate in simulated high-level radioactive waste (HLW), and establish the prediction model of Mo leaching rate.MethodsFirstly, a series multi-component borosilicate solidification glass samples with simulated high MoO3 (mass ~3%) HLW and designed mass fractions of ZnO content of 1%, 1.5%, 3.5%, 4%, 4.5%, and 5%, respectively, were prepared for experimental test of Mo leaching rate. Then the glass structure gene modeling (GSgM) were employed to accurately simulate the glass properties using a small amount of data to establish the corresponding structural prediction model. Finally, Infrared spectrometer (IR), X-ray Diffraction (XRD), Differential Scanning Calorimetry (DSC) and thermal dilatometer (DIL) were used to test the transformation temperature, heating rate, coefficient of thermal expansion, etc. of the samples, and the model validation was also performed.ResultsExperimental results show that Mo leaching is mainly affected by Si-O-Si rocking vibration around ~530 cm-1 and νSi-O-Si vibration in Q4 at ~1 180 cm-1, furthermore, B2O3, ZrO2 and alkali variation influence the relative concentration of these two structural units, and then affect the Mo leaching. Mo leaching rate is enhanced with mass of ZnO over 3.5% in this complex glass system.ConclusionsModel validation in this study proves that the structure model of Mo leaching rate has good simulation accuracy and statistical reliability, which are further improved after model iteration, hence can be used for the Mo-leaching prediction in the quick-screening of glass composition development.

BackgroundIn the nuclear industry, zirconium alloys, employed as structural materials, undergo irradiation-induced deformation owing to the combined effects of thermo, mechanical stress, and neutron irradiation during the irradiation process, thereby affecting their reliability in use.PurposeThis study aims to explore and predict the irradiation deformation of zirconium alloys to ensure the safety and economical operation of reactors.MethodsFirst of all, a database of experimental data on irradiation deformation of zirconium alloys from publicly available literature was developed by collection, cleaning, re-organization and placement of these data into a well-designed database, accompanied by detailed descriptions of these data. Then, data patterns, model validation and data mining were conducted by combining the database with our own developed mesoscopic model. Specifically, over fifty sets of preliminary data from the data mining library were subjected to data mining to analyze the correlation between model parameters and control variables.ResultsData mining results indicate that the critical resolved shear stress (CRSS) of the slip system increases with Nb content whilst the irradiation creep compliance of zirconium alloys increases with temperature.ConclusionsCompared with conventional data induction methods, this study introduces a novel approach for fitting and data mining using a physics-based mesoscopic model, offering fresh perspectives and methodologies for studying irradiation-induced deformation in zirconium alloys. The impact of this approach will be boosted as the collected dataset expands.

BackgroundNeutron localization technology can be used for monitoring spent fuel reprocessing to prevent nuclear criticality accidents.PurposeThis study aims to investigate the effects of different neutron emission modes, 3He pressure, tube diameter, and moderators on the position resolution and detection efficiency of a position-sensitive 3He proportional counter.MethodsFirstly, a 3He proportional counter with a tube diameter, tube length, and 3He pressure of 2.54 cm, 80 cm, and 405 300 Pa, respectively, was taken as object of study, and the Monte Carlo (MC) algorithm was employed to simulate the effects of neutrons with different energies, neutron emission modes, 3He pressures, tube diameters, and moderating bodies on the position resolution and detection efficiency of a position-sensitive 3He proportional counter. Then, the effects of the time difference of arrival (TDOA) on the location of a source in an active zone, source to tube distances, thickness of moderating body, and position resolution of the 3He proportional counter were investigated. Finally, a linear fitting analysis of the active region was performed, and the locations of inactive regions were estimated to determine the positioning ability of the 3He tube.ResultsSimulation results show that the position resolution of this 3He proportional counter is equal in all axial sections, and the position-resolution limit of thermal neutron source is 0.17 cm. When the thickness of the moderating body is 5 cm, and the 244Cm source is close to the external wall of the moderating body, the neutron detection efficiency is 3.0%, and the theoretical position resolution is 9.25 cm whilst experimental result of position resolution is 18.50 cm, half of the theoretical value, and the location resolution limit is 2.64 cm after processed by the linear fitting curve of the active location and a time-difference crest.ConclusionsMC simulation and experimental tests show that the position-sensitive 3He proportional counter exhibits better position resolution and detection efficiency for the 244Cm source when the thickness of the moderating body is 2~3 cm, hence suitable for practical applications.

BackgroundGaussian shaping is commonly used to filter and extract the amplitude of nuclear pulse signals owing to its high signal-to-noise ratio, resistance to ballistic loss, and ease of amplitude extraction. However, the problems of asymmetry and downwash exist in Gaussian-like signals generated by Sallen-Key and CR-(RC)mfilters.PurposeThis study aims to address the problems of asymmetry and downwash in Gaussian-like signals generated by Sallen-Key and CR-(RC)m filters, this paper presents a trigonometric function-based Gaussian-like pulse shaping algorithm for dual-exponential signals.MethodsFirst, the impact of shaping parameters on the shaping pulse and filtering performance of the algorithm was examined by applying Gaussian-like pulse shaping to simulated nuclear pulse signals. Then, the filtering performances of Gaussian-like and sin pulse shaping algorithms were compared and analyzed using the same shaping width. Finally, the characteristic X-ray signals of manganese (FAST-SDD detector) were collected using a self-made digital nuclear signal acquisition board under various tube flows. The energy resolution and count rate characteristics of the energy spectra obtained from Gaussian-like, sin, and trapezoidal shaping algorithms were comparatively analyzed for different tube flows and shaping times.ResultsComparison results show that Gaussian pulse shaping exhibits superior denoising performance to sin pulse shaping, with an 8.95% improvement in the SNR for the same peak arrival time. Additionally, the energy resolution of the spectrum obtained using Gaussian pulse shaping exceeds that of sin pulse shaping, and its stacking pulse separation ability outperforms that of trapezoidal pulse shaping, making it highly applicable.ConclusionsGaussian-like pulse shaping presented in this study demonstrats better stacking pulse separation capability than trapezoidal pulse shaping, indicating promising application prospects.

BackgroundWith the development of nuclear energy, spent nuclear fuel reprocessing has generated high-level liquid waste (HLLW) containing large quantities of minor actinides (e.g., Np, Am, and Cm). N,N,N',N'-tetraoctyl-diglycolamide (TODGA), an amide ether extraction agent, has immense potential for the extraction and separation of actinides from HLLW. However, HLLW is characterized by high radioactivity, which can destroy the molecular structure of TODGA and diminish its extraction capacity. In addition, the type of organic medium selected to dilute TODGA has a significant impact on its extraction ability and radiolysis.PurposeThis study aims to evaluate the gamma radiolysis and extraction performance of TODGA in a kerosene medium.MethodsFirstly, several experimental reagent systems were prepared. Organic phase: 50 mmol?L-1 TODGA/kerosene; aqueous phase: 0~4 mol?L-1 HNO3; absorbed doses: 5~1 000 kGy by a 60Co gamma-irradiator (3.7×1015 Bq) at room temperature; absorbed dose rates: 0.6 kGy?h-1 (5~15 kGy) and 5.8 kGy?h-1 (50~1 000 kGy). Then, ultra-performance liquid chromatography (UPLC) and mass spectrometer were used to identify the components in the organic phase and assess the relative concentrations of TODGA and radiolysis products before and after irradiation. Lanthanides (substitutes for actinides) and alkaline earth metals were extracted using TODGA at different absorbed doses (5~100 kGy) and the concentration of metal ions for extraction was 100 μg?mL-1 in 3 mol?L-1 HNO3. Finally, after extraction and dilution, the aqueous phase was detected using an inductively coupled plasma-optical emission spectrometer to estimate the lanthanide and alkaline earth metal concentrations.ResultsThe experimental results indicate that the concentration of TODGA steadily increases with increasing concentrations of HNO3 at a radiation dose of 1 000 kGy. The radiolysis rate of TODGA in the acid-free solution is >80%. Compared to the acid-free solution, the degree of radiolysis can be inhibited at approximately 8% (adding 0.5~3 mol?L-1 HNO3) and 18% (adding 4 mol?L-1 HNO3). Moreover, with an increase in the absorbed doses, the concentration of TODGA decreases in the studied systems. The radiolysis rate of TODGA is estimated to be 10%~40% when the absorbed dose was less than 100 kGy. However, it increases dramatically and reaches over 70% at 1 000 kGy. Breaking the ether bond results in the radiolysis of TODGA, generating two types of radiolysis products (C18H37NO2 and C18H37NO). HNO3 alters the radiolysis path, and breaking of the octyl side chain occurs with the appearance of another radiolysis product (C28H56N2O3). TODGA exhibits extremely poor extraction affinity for Sr, and the extraction rate decreases from 25% to non-extraction at an absorbed dose of 100 kGy. However, TODGA achieves approximately 100% lanthanide extraction (Ce, Eu, and Dy).ConclusionsTODGA has low radiation resistance in a kerosene medium at an absorbed dose of 1 000 kGy. Its radiolysis can be inhibited by the addition of HNO3; a higher concentration of HNO3 leads to a stronger inhibitory effect. The radiolysis of TODGA is insignificant within a 100 kGy absorbed dose, particularly in the range of 50~100 kGy. The molecular structure of TODGA is susceptible to breaking of its ether bond owing to gamma rays or radicals produced by kerosene. These findings demonstrate that TODGA can maintain the ability to extract lanthanides after irradiation at an absorbed dose of up to 100 kGy. Thus, it can be inferred that TODGA has a good extraction capacity for actinides in kerosene under the studied conditions.

BackgroundThe remediation technology for U(VI), Ca2+ and other pollutants in the groundwater of the mining area of in-situ leaching of uranium has become a key technical bottleneck restricting the decommission of uranium mine, and the microbial-induced calcium carbonate precipitation (MICP) technology can well mineralize and remove heavy metals in the groundwater.PurposeThis study aims to remove the pollutants in groundwater of uranium mining areas with MICP.MethodsFirst of all, bacillus pasteurianus was selected, and its acid resistance and tolerance to uranium were analyzed. Then, the effects of initial U(VI) concentration, initial Ca2+ concentration, initial pH value, concentration of bacterial solution, and the time of mineralization on the removal of pollutants by MICP were explored. In addition, scanning electron microscopy (SEM) and X-ray diffraction spectroscopy (XRD) were employed to characterize the composition and microstructure of the mineralization products of MICP, in order to reveal the mechanism of mineralization and decontamination. Finally, water with low uranium concentration was treated by bacillus pasteurianus to verify the application of MICP.ResultsThe results show that Bartonella pasteurii has good urease activity at pH value of 4 and can adapt to the groundwater containing U(VI) concentration of 100 mg?L-1. When the initial pH value is 4, the initial uranium concentration is 50 mg?L-1, the Ca2+ concentration is 8 000 mg?L-1, and the time of mineralization by MICP is 48 h, the removal rates of U(VI) and Ca2+ in the groundwater are 61% and 54%, respectively. The removal rate of U(VI) further increases to 91% when the initial pH value is increased to 7. Meanwhile, the increase of Ca2+ concentration promotes the removal of U(VI) in groundwater by MICP, however, the removal rate of Ca2+ is relatively low if the concentration of Ca2+ is too high.ConclusionsThe composition of MICP mineralization products is mainly uranium-containing calcium carbonate, in which the uranium is immobilized mainly by co-precipitation. Therefore, Bacillus pasteurianus can well remove U(VI) and Ca2+ in the groundwater of the mining area of in-situ leaching of uranium, and has a good prospect of application.

BackgroundMolten salt electrochemical separation of spent fuel in molten salt medium is one of the most widely studied technologies in pyroprocessing. It is based on the use of electrolysis to achieve the separation.PurposeThis study aims to investigate the separation of actinides and lanthanides in situ using high-temperature molten salt media.MethodsCyclic voltammetry (CV), square wave voltammetry (SWV), and constant potential electrolysis were employed to study the electrochemical behavior and separation feasibility of UF4 and LnF3 (Ln=Eu, Sm, Yb, La, Ce, and Ho) in FLiBe molten salt. Firstly, the eutectic salt was prepared, followed by the precise measurement and mixing of FLiBe molten salt and UF4 or LnF3 powder in accordance with the desired concentration ratio. Then, the electrochemical behavior of uranium and lanthanide elements in FLiBe molten salt was experimentally studied using electrochemical CV and SWV. Subsequently, the constant potential electrolysis method was employed to achieve the electrochemical deposition of uranium or lanthanide elements. Finally, the feasibility of separating uranium and lanthanide elements in FLiBe molten salt was evaluated through X-ray diffraction (XRD) and energy-dispersive X-ray spectroscopy (EDS) analysis of the electrolysis products.ResultsExperimental results show that U4+ undergoes two electron transfer reductions on an inert tungsten electrode, namely U4++e-→U3+ and U3++3e-→U0, while Yb3+, Sm3+, and Eu3+ have only one electron transfer reduction, leading to the +2 valence state of Yb2+, Sm2+, and Eu2+ ions, respectively. In addition, La3+, Ce3+, and Ho3+ ions do not exhibit significant electrochemical signals within the electrochemical window of FLiBe molten salt. The findings from XRD and EDS analyses verify that the electrolysis products presented on the W electrode consist of metal U, UF3, and entrain molten salt, with no lanthanide (Ln=Eu, Sm, Yb, La, Ce, and Ho) electrolytic deposition product detected.ConclusionsThe suitable electrolytic separation method developed in this study provides basic data to separate actinides and lanthanides from molten salts.

BackgroundThe study of irradiated samples is of considerable importance. Owing to these samples being radioactive, the applicability of conventional characterization methods is limited. Because of the high sensitivity of 3He detectors to neutrons, the small-angle neutron scattering (SANS) technique is nearly unaffected by radiation such as gamma, beta rays, and sample preparation is simple.PurposeThis study aims to develop a device for SANS measurement of nanostructures in radioactive samples.MethodsThe shielding thickness of the device was optimized through Monte Carlo simulations. Experimental efficiency and safety of the device were improved by combining remote-control functionality and automated sample switching among up to 12 samples loaded simultaneously. Finally, the device was used to carry out a SANS experiment for characterizing the radioactive A508-III steel sample.ResultsThe optimized thickness of the lead shielding layer of the device is 7.5 cm, the corresponding maximum dose rate of the measurable radioactive sample is 1.4 mSv·h-1. The results of successful SANS experiment on a A508-III steel irradiation surveillance specimen indicate that low-dose, long-term irradiation has a minimal impact on the nanostructure of pressure vessel steel.ConclusionsThe device and associated technical methods can be applied for nanostructure characterization of radioactive samples.

BackgroundDuring the reconstruction of radionuclide activity of waste by conventional tomography gamma scanning (TGS), the measurement time is long due to the division of a large number of invalid voxels, and the reconstruction accuracy is low due to the iterative solution of the pathology equation. The traditional division method of encrypting only circumferential voxels does not significantly improve the reconstruction accuracy.PurposeThis study aims to solve the problem of long measurement time and low accuracy of TGS by elaborating the influence law of voxel division method on reconstruction accuracy.MethodsFor a 400 L cement waste drum with a radius of 35 cm, a coaxial HPGe detector with a detection efficiency of 40% (crystal diameter 6.09 cm, length 5.18 cm) was used at a distance of 74 cm from the center of the drum to measure two types of nuclides, 60Co and 137Cs. The error situation of the optimized method of voxel partitioning and the traditional voxel partitioning method were compared by random point source activity reconstruction experiments. By calculating the condition number and count rate deviation of different voxel partitioning methods, the influence law of voxel partitioning methods on the reconstruction accuracy was investigated.ResultsComparison results show that the radial encryption method leads to an increase in the number of very small value points and a decrease in the value of very large value points in the count rate deviation curve, which improves the reconstruction accuracy. Compared with the conventional division method, the voxel division optimization method reduces the number of voxels and thus the measurement time by about 9/10; at the same time, it reduces the number of conditions and the count rate deviation, which reduces the maximum reconstruction error by about 2/3.ConclusionAn optimized voxel division method proposed in this study makes use of the influence law of different voxel divisions on the reconstruction error through theory and experiment, hence improves the measurement accuracy and reduce the measurement time.

BackgroundThere is increased demand for high spatial resolution in domestic and international nano-imaging experimental stations utilizing synchrotron radiation. Nano or nanoradian level positioning accuracy and stability are necessary for focusing mirrors or monochromators on the beamlines.PurposeThis study aims to design and optimize a redundant parallel flexible hinge rotating device to meet this demand.MethodsFirst of all, the kinematics of the flexible mechanism were initially analyzed, and the virtual displacement principle was employed to derive the overall static rotational stiffness of the mechanism. Then, the characteristics of the mechanism, as well as the impact of hinge parameters on stiffness, were investigated. Subsequently, the dynamic model of the mechanism was built by employing the Lagrange equation, and the natural frequency in the motion direction was deduced. Finally, an optimization model was developed for the static and dynamic dual purpose mechanism design, solved through a genetic algorithm accounting for nonlinear constraints. In addition, the first four resonance frequencies and vibration modes of the flexible mechanism were examined using a finite element method for modal analysis followed by the creation and assembly of a high-precision flexible hinge mechanism along with a rotary adjustment device for experimental testing.ResultsTest results indicate that the flexible angular displacement adjustment mechanism achieves a rotation angle of 0.668° and bidirectional repeatability of ±8.91 nrad during fine-tuning. In addition, the angle resolution of 15 nrad and stability of 2.72 nrad (root mean square value) are achieved over a 30 min test in the frequency range of 1~500 Hz. The first natural frequency of the mechanism is approximately 295 Hz, which aligns with the theoretical calculations and finite element analysis results.ConclusionsThe effectiveness and reliability of the optimized flexible mechanism in achieving nanoradian level high-precision angular displacement adjustment is confirmed by this study.

BackgroundElectron cyclotron resonance heating (ECRH) is an important heating and plasma current control method for the HL-3 tokamak. Microwave inject into plasma through the launcher, which is an important part of the ECRH system. The ECRH launcher system of HL-3 tokamak consists of three launchers, including a mid equatorial launcher and two upper launchers. All launchers are located in the same sector, and can work together to complete functions such as heating and Neoclassical Tearing Mode (NTM) control.PurposeThis study aims to design and test the ECRH launcher system of the HL-3 tokamak.MethodsFirst of all, the overall planning was carried out for the launcher system. Subsequently, the transmission path and structure of the launcher were designed, and the effect of microwave injection were calculated by simulation. Then, detail design and implementation for the mid-equatorial launcher and No.1 upper launcher were completed, so did the optical path design for No.2 upper launcher. Finally, the transmission angle and rotation speed of the launcher were tested, and the rotation angle of the launcher was calibrated.ResultsThe full range response time of the mid-equatorial launcher is less than 90 ms. The full range response time of the No.1 upper launcher is less than 190 ms.ConclusionsThe control of the equatorial launcher and the No.1 upper launcher is precise and fast, meeting the requirements for experimental use of the tokamak.

BackgroundThe high voltage power supply is an important part of the neutral beam injection heating system, which determines the beam energy and the quality of the extraction beam current. With the gradual increase of voltage level, the pulse step modulation (PSM) high voltage power supply cannot meet the experimental requirements.PurposeThis study aims to design an inverter high voltage power supply based on super capacitor energy storage to achieve fast switching of injected power for neutral beam modulation.MethodsSuper capacitor energy storage was adopted to reduce the required grid capacity and minimize the impact on the grid. The DC-DC resonant converter structure with soft-switching technology was used to improve the response speed of the power supply and reduce the switching loss of the switching devices. After the design of power module circuit topology,system modeling and calculation based on power supply performance specifications, the charging circuit and main loop PSIM simulation model were established, and the power supply performance indexes were simulated and verified. Finally, a test prototype of inverter power supply module was built to conduct the test of relevant performance indexes. [Results and Conclusions] After simulation and experimental verification, the power module is able to achieve a stable output of 1 600 V/50 A, which meets the design requirements of 6 MW/120 kV.

BackgroundLower hybrid current drive (LHCD) is one of the main auxiliary heating and current driving methods for Tokamak, and studies have shown that the boundary parasitics effect of lower hybrid (LH) wave in scrap-off layer (SOL) significantly decreases current drive efficiency of LH wave. Among these, wave scattering caused by density fluctuation at the boundary results in spectral changes within the SOL, which changes the power deposition location and current drive efficiency.PurposeThis study aims to explore the lower hybrid scattering caused by density fluctuation in the scrap-off layer on the Experimental Advanced Superconducting Tokamak (EAST) device.MethodsBased on the cold plasma dispersion theory and the blob density fluctuation theory, COMSOL Multiphysics was employed to establish a full-wave solution model of lower hybrid scattering in the SOL. Based on the parameters of the EAST device, the influence of low-frequency electron density fluctuation with different characteristics on wave scattering was analyzed.ResultsThe simulation results show that the backscattering direction caused by the density wave packet "blob" is more obvious than the forward scattering, and scattering induced by blob leads to the change of spatial structure of low hybrid power flow. The density fluctuation in the blob mainly affects the amplitude of the wave field disturbance, the radius of the blob mainly affects the spatial range of the wave scattering, the full-field disturbance caused by multiple blobs increases significantly.ConclusionsThe full-wave solution model of the SOL established in this study is reliable. The simulation results using this model show that the density fluctuation will cause lower hybrid scattering, which will change the spatial structure of the power flow.

BackgroundData-driven methods for power fault diagnosis heavily rely on the signal data quality of power sensors. The power systems in Tokamak fusion devices often operate in environments with complex electromagnetic field coupling, leading to the mixing of physical characteristic signals with a significant amount of inseparable noise in the collected data.PurposeThis study aims to mitigate the impact of noise on the final diagnostic results by proposing a multi-branch denoising network, termed Hierarchy Branch Denoising Convolutional Neural Network (HBD-CNN) that utilizes noise-resistant wavelet enhancement in conjunction with one-dimensional convolutional neural networks to accomplish power system fault diagnosis tasks under the influence of noise interference.MethodsFirstly, the signal decomposition function of discrete wavelet transform (DWT) was incorporated into the network layer of the convolutional neural network (CNN), and the optimization of the traditional 1D-CNN network structure was deepened alongside the more robust exponentially linear unit (ELU) for noise. Then, a data multi-level structure was constructed based on prior knowledge to leverage and couple it with hierarchical classification modules within the network, hence the generalization capability of HBD-CNN was enhanced. Finally, preliminary validation of the architecture of this model was conducted based on the simulated power supply dataset.ResultsValidation results show that the fault diagnosis accuracy for the power converter reaches 98.31% when the signal-to-noise ratio (SNR) is 10 dB. Even at an SNR of 2 dB, the accuracy remains above 92%.ConclusionsThe results of this study indicate that HBD-CNN demonstrates excellent fault diagnosis performance and potential under noisy conditions.

BackgroundDue to its pollution-free nature and non consumption of fossil fuels, nuclear fusion is the most ideal future energy source. China is preparing to build a China Fusion Engineering Test Reactor (CFETR) with independent intellectual property rights, and plans to build a commercial thermonuclear fusion power plant that can generate electricity externally by the mid-20th century. However, there is contradiction between the instability of nuclear fusion heating power output and the smooth operation of steam turbine generators, hence thermal storage technology is used for peak shaving and valley smoothing of power output in nuclear fusion reactors.PurposeThis study aims to compare heat storage technologies applied to CFETR nuclear fusion power plants to reduce its the peak and valley power output.MethodsThe parameters of helium cooled ceramic breeder cladding in nuclear fusion reactors was selected as the boundary conditions for thermal storage technology. By evaluated the applicable temperature range of thermal storage technology, three potential thermal storage technologies, i.e., chemical heat storage, sensible heat storage technology and phase change heat storage, for CFETR nuclear fusion power plants were preliminarily analyzed, and their costs were preliminarily predicted.ResultsThe three major types of heat storage technologies can all select heat storage media suitable for the temperature parameters of the helium cooled breeder blanket in CFETR nuclear fusion power plants. However, chemical heat storage has the potential to be applied in CFETR nuclear fusion power plants due to the temperature difference between its heat absorption and release, which is not conducive to the stability of the system and causes energy loss. Sensible heat storage technology and phase change heat storage technology have smaller temperature differences between their heat absorption and release. The preliminary economic analysis results show that the cost of phase change heat storage is the lowest, followed by molten salt heat storage, and the use of silicon bricks as the heat storage medium in solid-phase sensible heat storage technology. The use of cast steel as the heat storage medium in solid-phase sensible heat storage technology has the highest cost.ConclusionIn thermal storage technology, molten salt thermal storage technology has a high degree of maturity and has a large number of engineering applications, with a cost between phase change thermal storage and solid-phase sensible thermal storage, and has great potential for application. The cost of phase change heat storage is the lowest, and the parameters are suitable for nuclear fusion power generation. However, its technological maturity is relatively low, and it is expected to become a focus of future research.

BackgroundENN Science and Technology Development Co., Ltd. (ENN Fusion Technology R&D Center) is upgrading its compact fusion research facility EXL-50 to EXL-50U. Both devices are the conventional conductor tokamak, on which the magnet power supply system is composed by 1 TF (Toroidal Field) power supply, 1 CS (Center Solenoid) power supply and 10 PF (Poloidal Field) power supplies PF1-10. All 12 sets of power supply system are powered by 2 AC pulse generators and output DC current through thyristor-based converters.PurposeThis study aims to design EXL-50U magnet power supply for satisfying high parameter requirements of EXL-50U.MethodsPower supply capacity was the first concern for upgrading and the corresponding protection strategies under high parameter conditions was taken into account as well. The configuration of AC pulse generator was introduced at the beginning. Then transformers and converters were listed and designed in scheme. Control system and protection process were implemented respectively, followed by detail power supply system illustration and commissioning waveforms display for each power supply.ResultsThe reliability and controllability of developed power supply system are verified by the waveforms that forms plasma current under the condition of CS breakdown.ConclusionIt is proved that this power supply system can work stably, and output waveforms can be repeated no matter it works alone or under complex condition of joint debugging.

BackgroundIn neutral beam injectors (NBIs) of Tokamak device, the calorimetric target is one of the most important water-cooled components, responsible for receiving and measuring beam power. In addition, by using a built-in thermocouple array, the temperature rise at different positions of the target plate can be monitored in real-time, thereby obtaining the power density distribution of the extracted ion beam or neutral beam.PurposeThis study aims to develop a calorimetric target for the neutral beam injector in the HL-3 device, which can meet the requirements of target plate lifting and thermal load absorption.MethodsA linear push rod mechanism was applied to the design of the calorimetry target to achieve lifting and lowering, and a "W" - shaped target plate structure was adopted to achieve absorption of neutral beam energy. According to the design parameters, the fluid calculation module of Ansys Workbench was employed to simulate the temperature distribution of the calorimetry target under full power operation.ResultsCalculation results show that the maximum temperature rise of the calorimetry target is 526.4 oC when the deflecting magnet is opened under full power operating conditions, and the temperature can be lowered to room temperature within 30 s, satisfying the requirements for the use of the beam line.ConclusionsThe successfully developed calorimetric target of this study meets application requirements on second neutral beam injection beamline of HL-3 device, providing references for further engineering design of calorimetric targets in other NBIs.

BackgroundNeutral beam injection heating is a very effective auxiliary heating method in magnetic confinement fusion experiment. The built 5 MW neutral beam injection plays an important role in the H-mode operation with 1 MA plasma current of HL-3 tokamak. In order to achieve the H-mode operation with 2.5 MA, three neutral beam injection heating systems will be constructed with a total heating power of 20 MW.PurposeThis study aims to investigate the transmission efficiency of 7 MW neutral beam injector designed for HL-3 tokamak.MethodsThe beam transmission efficiency was studied from two aspects of neutralization efficiency and power deposition. The initial target thickness of the neutralizer under two beam parameters was analyzed based on Monte Carlo method. The optimal neutralization efficiency was achieved by adding neutralizer gas supply, and the target thickness required for the optimal neutralization efficiency was given. Finally, power deposition method was employed to analyze the transmission performance of the two neutral beams with different parameters at different beam divergence angles.ResultsThe results show that the optimal neutralizer efficiency is achieved when the neutralizer gas supply is 0.6 Pa?m3?s-1 and 1.1 Pa?m3?s-1, respectively. The neutral beam power obtained under two beam parameters with a divergence angle of 1.2° is 7.2 MW and 6.8 MW, respectively, and the corresponding transmission efficiency is 0.4 and 0.35, respectively.ConclusionsThe new neutral beam injector of HL-3 tokamak will be built according to the design of this study, with optimized parameter of 100 kV/45 A, while still retaining the ability to deliver a 120 keV deuterium beam.

BackgroundElectric propulsion systems, compared to traditional chemical propulsion, offer longer operational lifespans and lower fuel consumption in space missions, garnering significant attention in recent years. However, for high-power Hall effect electric thrusters developed based on controlled fusion concepts, measuring thrust proves challenging due to the high-temperature environments required for ionizing propellants in electric thruster, resulting in the generation of hot plasma plumes during operation. As a result, traditional thrust measurement methods are unable to accurately measure the thrust.PurposeThis study aims to accurately measure thrust during ground testing of electric thrusters for precise control of the spacecraft's attitude and orbit maintenance.MethodsFirstly, a thrust measurement platform based on flexible beam structure was designed and implemented with capability of measuring the thrust generated by electric thrusters in high-temperature. Simultaneously, a calibration system was built to verify the stability and repeatability of the thrust measurement results. Then, simulation analysis was conducted on the structural mechanics and thermal coupling field of the thrust measurement system under different operating conditions. Finally, the variable specific impulse magnetoplasma rocket (VASIMR) was taken for experimental thrust measurement of its electric propulsion system.ResultsThe experimental results of VASIMR indicate the thrust is 266.5 mN measured in real time at the central magnetic field intensity of 0.2 T with a mass flow rate of 20 mg?s-1.ConclusionsThe thrust measurement platform based on the bending beam structure can meet the measurement requirements of VASIMR, providing valuable references for subsequent experiments.

BackgroundWith the development of human aerospace industry, it is necessary to develop propulsion systems suitable for different space mission scenarios. MegnetoPlasmaDynamic thruster (MPDT), which is similar to the principle of magnetic confinement fusion, is a typical representative of electromagnetic thruster, which stands out among many electric thrusters because of its superior performance in thrust power ratio and specific impulse. Anode power deposition is the result of the interaction between plasma and wall during MPDT operation. It is one of the main mechanisms of power loss of this type of thruster, accounting for 40%~90% of the total power, which seriously reduces the efficiency of thruster.PurposeThis study aims to solve the problem of low efficiency of thruster by investigating the influence of anode radius on the efficiency of thruster from the perspective of anode power deposition.MethodsFirst of all, based on MagnetoHydroDynamic (MHD) equations, numerical models for radial discharge parameters and physical model for anode power deposition were established. Then, the influences of anode radius on discharge parameters, anode power deposition and anode power deposition fraction were studied on the basis of these models by numerical calculation method. Finally, a water-cooled structure anode was designed, and the effectiveness of its heat dissipation structure was verified by thermal simulation.ResultsThe results show that with the increase of anode radius, the electron density and ion velocity are increased, the anode power deposition fraction is decreased whilst the anode power deposition is increased. The thruster efficiency is improved by increasing the anode radius. The thermal simulation results show that when the input power deposition of the water-cooled structure anode is about 3 kW, the corresponding temperature difference of the anode cooling water is 5 K.ConclusionsThis study verifies the reliability of the physical model of anode power deposition, indicating that increasing the anode radius is an effective means to improve the efficiency of the thruster.

BackgroundWhen fusion nuclear reactor is in operation, neutron activation causes a large number of radioisotopes. Activation calculation is a very important step in both of reactor shielding calculation and radiation safety analysis.PurposeThis study aims to develop the capability of transport-activation coupling calculation for fusion reactor based on the Monte Carlo code cosRMC.MethodsFirstly, the built-in burnup solver "Depth" of cosRMC was employed to develop transport-activation internal coupling calculation function under fixed source mode with embedded calculation of activation related nuclide single group reaction cross-sections in neutron transport process without the transmission of neutron spectra to external activation programs. Then, the developed code was applied to the activation calculations of the first wall (FW) material steel and plasma facing component (PFC) material tungsten of the Chinese Fusion Engineering Testing Reactor (CFETR) using continuous energy cross-sections and multi group cross-sections, respectively, and calculated results were compared and verified with that of the activation program ALARA.ResultsThe comparison results of activation calculation of FW steel and PFC tungsten in CFETR show the consistency between cosRMC-based internal coupling method and ALARA program, which preliminarily verifies the correctness of the transport activation internal coupling calculation function of the developed cosRMC program.ConclusionThe developed cosRMC-based internal coupling method can dynamically update neutron spectra and material information, and use continuous energy cross-sections for reaction rate calculation, obtaining reaction cross-sections related to the geometry and energy spectra of actual problems, thus accurately considering the influence of resonance zone nuclear cross-sections.

BackgroundThe control of edge localized modes (ELMs) is a key issue for the safety operation of ITER and future magnetic confinement fusion reactors. In recent years, extensive theoretical simulations and experimental studies have demonstrated that resonant magnetic perturbation (RMP) is a promising method for controlling edge localized modes (ELMs) in H-mode plasmas.PurposeThis study aims to investigate drift kinetic resonant effects of thermal particles on the plasma response to the applied RMP and compare with that of the fluid model for better understanding of ELMs control in HL-2A.MethodsBased on experimental plasma and RMP coil configurations in HL-2A, the MARS-F/K codes were employed to compute the drift dynamic response of plasma to RMP under high constraint mode and compared it with the results of fluid model. Further sensitivity studies were conducted on key parameters including the plasma equilibrium pressure, toroidal flow as well as the thermal particle collision effects.ResultsThe fluid response model predicts an initially relatively weak enhancement of the plasma response amplitude with the equilibrium pressure parameter βN, followed by a strong variation as βN approaches the Troyon no-wall beta limit. The latter is unphysical and is eliminated by the kinetic response model. Including kinetic response and considering particle collision, the non-adiabatic resonant contribution of trapped thermal ions is found to be significant. Neglecting the particle collision, however, the trapped thermal electron contribution becomes more pronounced.ConclusionsThe importance of considering kinetic effects for high-beta plasma response is emphasized by this study. It is important to include the non-adiabatic resonant contribution of trapped particles to the kinetic response while taking into account the particle collision effect.

BackgroundNegative ion sources driven by radio frequency (RF) waves have become the preferred solution for future neutral beam injection systems.PurposeThis study aims to monitor the plasma discharge state of each exciter of a high power RF negative ion source by developing a plasma luminescence monitoring system based on a photodiode is designed and constructed.MethodsThe intensity of plasma emission was closely related to the number of specific collisions, collision particle density, and collision particle energy. Therefore, the intensity of plasma emission was applied to monitoring plasma parameters qualitatively, and a photodiode-based multichannel plasma luminescence monitoring system was designed and implemented to monitor the plasma discharge state of each exciter of a high power RF negative ion source. Based on a reasonable collision radiation model, plasma parameters were quantitatively obtained by analyzing the intensity of plasma characteristic spectral lines, and the influence of the filtering magnetic field generated by plasma current on plasma emission signals as experimental tested.ResultsExperimental results this monitoring system demonstrate that the real-time intensity information of plasma emission from different positions is successfully collected, and subsequently presented and saved in the form of voltage signals for real-time monitoring and post-data processing by the host computer. The intensity of plasma emission has a good linearity with RF discharge power.ConclusionsThe plasma light monitoring system of this study can accurately and real-time measure the excitation, maintenance, and extinction processes of plasma in the RF exciter.

BackgroundIn a tokamak, when fast electrons are deconfined by the tokamak magnetic field constraint and lost to the vacuum wall or limiter, the device may become damaged and the discharge may be affected.PurposeThis study aims to explore the loss behavior of fast electrons during discharge using a diagnostic system based on a ZnS(Ag) scintillator probe for detecting the loss of fast electrons on the Experimental Advanced Superconducting Tokamak (EAST).MethodsThe Geant4 simulation program was employed to simulate the interaction between electrons in different initial states and the scintillator probe of the diagnostic system. Firstly, the probe model and the filling material model of stainless steel and ZnS(Ag) coating were established in Geant4. Then, the interaction between electron beam and scintillator probe under different incident conditions (incident energy, angle, scintillator thickness, magnetic field size, etc.) were simulated, and the physical processes were recorded. Finally, the recorded data were accessed by MATLAB programming for analysis.ResultsThe results show that the contribution of secondary electrons and initial electrons to the luminescence intensity of scintillators occupies different dominant energy ranges. The luminescence intensity first increases and then decreases with the increase of incident electron energy, with a peak value around 12 MeV, and the number of emitted photons at oblique incidence is greater than that at vertical incidence. When the electron energy is lower than 4.3 MeV, secondary particles dominate the scintillation, and when the electron energy is higher than 4.3 MeV, primary particles dominate. The thickness of the scintillator has no significant effect on the peak position. After, the luminous intensity is considerably affected by the magnetic field angle and electron pitch angle after adding a magnetic field.ConclusionsThe results of this study contribute to the understanding of the fast electron loss signal detected by the scintillator probe in the EAST experiments, providing a basis for further study of fast electron loss.

BackgroundHL-2M tokamak is a new-generation magnetic confinement fusion plasma device in China, which has realized the high parameter operation mode with 1 MA plasma current.PurposeThis study aims to investigate the turbulent transport associated with the internal transport barrier (ITB) using gyrokinetic calculations.MethodsNumerical simulation was performed based on the gyrokinetic theory. The turbulent transport relevant to ITB was studied using the gyrokinetic toroidal code (GTC) combined with the equilibrium of HL-2M tokamak. A filtering zonal flow was considered in analyzing the influence of zonal flow on turbulent saturation level. A time evolution analysis of turbulent poloidal spectrum was conducted to investigate the effect of different wavelength modes on turbulent transport.ResultsThe results show that the turbulent transport at ITB saturates twice in succession, and the calculated average ion heat transport diffusivity is approximately twice that of the first saturation level. Moreover, the short-wave mode kθρi~2.15 dominates the first turbulent transport saturation, whereas the long-wave mode kθρi~0.49 dominates the second turbulent transport saturation. Specifically, an "M" shape distribution of the radial heat transport diffusivities is obtained at the transport barrier position during the turbulent saturation period. Finally, the minimum radial heat transport diffusivity during the turbulent saturation period occurs near the ITB where a maximum plasma temperature and density gradient occurs.ConclusionsThe turbulent transport at ITB may be dominated by two types of microinstabilities at different stages of turbulence development. Turbulent energy of the system is inversely cascaded from the modes with short wavelengths to those with long-wavelengths. The results are in good agreement with the theoretical prediction of the ITB.

BackgroundDuring severe reactor accidents, melten core causes the radioactive source term material to be no longer retained in the fuel but is released into the environment, causing serious radioactive contamination in the surrounding areas.PurposeThis study aims to investigate the release of fission products, both inside and outside the pressure vessel, using different models in order to analyze the effectiveness of the spray system in controlling the source term release and the decay heat generated.MethodsBased on a typical mega kilowatt pressurized water reactor (PWR) nuclear power plant model, the integrated security analysis program MAAP was applied to modeling reactor rupture accident. Then, the accident sequence and consequences of the reactor primary circuit heat pipe breakage superimposed high and low voltage safety injection failure were calculated and analyzed under CORSOR-M, CORSOR-O and ORNL-BOOTH source term release models.ResultsThe findings indicate that the fission product source term is mainly released in the pressure vessel, and the release amount is significantly higher than that released outside the pressure vessel. Under the CORSOR-O model, the pressure vessel fails the last whereas the containment fails the first; although the pressure vessel fails first in the ORNL-BOOTH model, the containment vessel fails much later than that in the other two models. The difference in source term release leads to different decay heat phenomena in different models, and the main heat source is the volatile fission product. Turning on the spray can not only keeps the suspended iodide in the containment vessel but also effectively removes the decay heat generated by the source items and reduces the pressure of containment.ConclusionsThe ORNL-BOOTH model results in lesser release of source term in the pressure vessel, the greatest variety of source term released, and the maximum time to containment failure. In addition, the opening of the spraying system effectively ensures the integrity of the containment.

BackgroundTraditional text-based system engineering in the design and application of passive residual heat removal systems (PRHRS) for lead-cooled fast reactors has several requirement problems, such as low development efficiency, a long iteration cycle, and model ambiguity.PurposeThis study aims to effectively address the aforementioned problems encountered in the PRHRS of lead-cooled fast reactors by adopting a model-based system engineering method.MethodsThe model-based system engineering (MBSE) method was preliminarily applied to the design requirement analysis of a PRHRS for lead-cooled fast reactors. During the design requirements study, the design process was combined with the preliminary design of the system architecture, comprising three parts: requirements analysis, functional analysis, and design synthesis.ResultsThe generated requirement and use case diagrams describe the system requirements and determine the top-level use cases of the system in the requirement analysis stage. The time sequence, activity, and state machine diagrams form the system function model and provide early confirmation and verification in the functional analysis stage. Finally, the white box model realizes the analysis and design of the system architecture in the design synthesis stage.ConclusionsThe system architecture designed by this method ensures the consistency of the design requirements. In the future, it will further reduce the design risk, improve the design efficiency, and provide an application reference for the design and optimization of digital lead-cooled fast reactors' passive residual heat removal systems.

BackgroundThe printed circuit heat exchanger (PCHE) has high heat transfer efficiency and compact structure, which can be used as the key equipment for heat transfer of small modular molten salt reactor, It is of great significance to the study of its flow heat transfer characteristics.PurposeThis study aims to explore the heat transfer characteristics of printed circuit heat exchanger and compare numerical simulation result with experimental data.MethodsThe flow heat transfer characteristics of molten salt (FNaBe) -helium heat exchangers with different flow channel structures, fin types and pitch were obtained by CFD numerical simulation, and the results were compared with those of traditional straight channel structures for comprehensive evaluation. The main thermal performance parameters of the heat exchanger with various airfoil fin structures were verified by experiment to find the best structure.ResultsThe results show that the numerical simulation results are in good agreement with the experimental results. Compared with the traditional straight channel structure, the fins can strengthen the heat transfer characteristics of the heat exchanger and reduce the flow resistance. NACA0025-8 mm airfoil fin structure has the best flow heat transfer characteristics.ConclusionsThe numerical simulation method established in this study can be used to predict the flow heat transfer characteristics of printed circuit heat exchanger. The empirical correlation formula of NACA0025-8 mm airfoil fin structure is fitted, which provides a theoretical basis for the design of subsequent heat exchangers.

BackgroundWhen the suspended dipole field coils experience significant disturbances during the operation of the suspended dipole field device, the Tilt-Slide-Rotate (TSR) coils are used to control the attitude of the dipole field coils, thereby preventing them from losing control.PurposeThis study aims at attitude control of dipole fields by coil forces of the TSR coils with relatively complex geometric structures.MethodsBased on the line segment approximation method, the modeling and analysis were conducted for scenarios where the dipole field coils return to their balanced position under conditions of dipole field tilt and offset by controlling the magnetic forces generated by the TSR coils on both opposite and same sides. The coupling magnetic field between the TSR coils and the dipole field coils was calculated, and the magnetic field line structure was analyzed. The excitation magnetic field of the TSR coils, in the form of sinusoidal currents, was simulated to induce azimuthal magnetic disturbances. Analysis was performed on the magnetic field line structure and phase of the perturbation magnetic field on the Poincaré section at the equatorial plane. [Results and Conclusions] If only one group of TSR coils on the same or opposite sides is used to control the dipole field coils to restore equilibrium, exhibited mechanical instability needs necessitating coordinated control by multiple sets of TSR coils. Applying pseudo-sinusoidal currents to all TSR coils induces compression and stretching effects on the background dipole magnetic field. By controlling the current of the TSR coils to control the phase of the compression position, the feasibility of controlling the attitude of the dipole field coils with multiple TSR coils has been validated. However, this process may also cause drift in the background magnetic field and the formation of open magnetic field lines, which could drive plasma instabilities and lead to particle loss.

BackgroundLead-free double perovskite Cs2AgBiBr6 has garnered significant attention in the field of nuclear radiation detection as an environmentally friendly material. Experimental observations have revealed that doping Cs2AgBiBr6 with Cu+ significantly enhances the material's stability and photoelectric conversion efficiency.PurposeThis study aims to investigate the impact of Cu+ doping on the crystal structure and electrical properties of Cs2AgBiBr6.MethodsBased on density functional theory, first-principles calculations were applied to exploring the effects of Cu+ doping on the structure and electrical properties of Cs2AgBiBr6. Structural properties, such as the stability, doping formation energy, lattice parameters, elastic constants, of Cu+ doping on the lead-free double perovskite Cs2AgBiBr6 were investigated by simulation whilst the band analysis and density of states analysis were employed to study the impact of electrical performance.ResultsThe results indicate that Cu+ doping enhances the stability of Cs2AgBiBr6. The Cs2Ag1-xCuxBiBr6 compounds formed by doping, as well as the original Cs2AgBiBr6 material, exhibit indirect bandgap semiconductor behavior. The bandgap significantly narrows with an increase in the Cu+ doping ratio. Based on an analysis of the density of states (DOS), the bandgap narrowing can be attributed to the downward shift of the conduction band minimum dominated by Bi6p orbitals due to Cu+ doping.ConclusionsCs2Ag1-xCuxBiBr6 exhibits greater stability and superior electrical properties compared to Cs2AgBiBr6, making it a promising candidate material for semiconductor radiation detectors.

BackgroundExposure to high-intensity X-ray irradiation induces numerous defects in the thin films of optical components applied to X-ray free-electron laser (XFEL) facilities. These defects result in structural damage to the material and degradation of macroscopic properties, thereby affecting its service life and significantly compromising the reliability and stability of XFEL facilities.PurposeThis paper aims to design and implement an automated irradiation simulation software based on Python called automatic irradiation simulation based on LAMMPS (AISL) to support the simulation research of XFEL radiation damage to materials using molecular dynamics methods.MethodsAISL was designed to facilitate high-throughput automated studies on XFEL radiation damage by exploring the process of radiation-induced defects and promoting the accumulation of data for radiation-resistant materials. This software package established an automated workflow pipeline for simulation tasks, encompassing aspects such as batch submission management, scheduling of simulation tasks, reliable storage of computational data, and post-processing of thermodynamic information. Additionally, the metadata was automatically recorded in a radiation damage simulation database based on MongoDB, so were the workflow information, and calculation result files of the simulation. Finally, the utilization of AISL in simulating radiation damage in metal thin films of XFEL optical components was demonstrated to verify the validation of AISL.Results & ConclusionsThe demonstration results indicate that AISL is an effective, user-friendly software for conducting high-throughput automated irradiation simulation studies. It significantly enhances the efficiency of material irradiation damage simulation calculations using LAMMPS.

BackgroundThe accurate determination of oil saturation is an essential part of reservoir petrophysical evaluation. Both C/H and C/O logging methods are primarily used to determine oil saturation in low-salinity reservoirs.PurposeThe study aims to compare the logging response effects of the C/H and C/O methods.MethodsFirstly, a spherical model and the model set by the C/O logging instrument RPM (Halliburton Company) were applied to Monte Carlo simulations, and the window counts of the atoms formed under different geological parameters and pore structures were calculated by MCNP5. Then, the relationship between C/O and C/H with saturation and the sensitivity to saturation were analyzed, and the response effects of C/O and C/H were compared. Finally, the mechanisms underlying the different response effects of C/H under different models were clarified.Results & ConclusionThe analysis results show that C/H and C/O exhibit the same relationship with porosity and saturation in casing wells, and both can be used to calculate saturation. However, compared to C/O, C/H is more affected by the borehole environment and is less sensitive to saturation; therefore, it is not suitable for saturation calculations in complex boreholes.

BackgroundSolid-polymer electrolyte (SPE) water electrolysis has been widely used to concentrate low-level environmental tritium for measurements. It has the advantages of electrolyte elimination, infinite reduction ratio, high current, safety, and environmental friendliness. However, the use of high-activity tritium samples for the calibration of its tritium enrichment factor Et causes a tritium memory effect in the system.PurposeThis study aims to reduce the tritium contamination caused by high-activity tritium samples during calibration.MethodsFirst, the deuterium method (DM) was used to calibrate the prepared SPE-based tritium enrichment system. The strong correlation between the tritium-to-protium (β) and deuterium-to-protium separation factors (α) during the electrolysis was determined to obtain the electrolytic cell constant, k = αβ-1/βα-1. Then, based on the deuterium enrichment factor Ed, the tritium enrichment factor (Ed)k of the SPE-based tritium enrichment system was determined. Finally, the influence factors of the k-value were analyzed, and the tritium enrichment factors derived via low-level tritium measurements using the DM and spike-proxy method (SPM) were compared.ResultsThe value of k is approximately 1.088 9, with a slight variation of 0.89% among cells. The k-value depends on the electrode material, but it is almost independent of the initial volume of water.ConclusionsThe k-value can be used as a constant for calibration of the tritium enrichment factor for the same types of electrolyzer with high accuracy and precision.

BackgroundNeutrons/Gamma (n/γ) discrimination is critical for neutron detection in the presence of γ radiation and traditional pulse shape discrimination methods suffer from unstable discrimination accuracy.PurposeThis study aims to implement a machine-learning method that combines the kernel principal component analysis (KPCA), marine predator algorithm (MPA), and extreme learning machine (ELM) is proposed to improve the n/γ discrimination efficiency and accuracy against the traditional pulse shape discrimination methods.MethodsThe KPCA was used to reduce the dimensionality of the pulse signal characteristics of neutrons and gamma rays. Owing to the randomness in the ELM input layer weight and hidden layer bias, the MPA was employed to optimize the foregoing factors to improve the n/γ discrimination accuracy of the ELM. Finally, experimental data of Pu-C neutron source using BC-501A liquid scintillator detector were applied to effectiveness comparison of training and test with and without KPCA dimensionality reduction.ResultsComparison results reveal that the average discrimination accuracy of the KPCA-MPA-ELM is as high as 99.07%, which is 12.19%, 2.52%, and 1.56% higher than those of the ELM, MPA-ELM, and KPCA-ELM models, respectively. Compared with the charge comparison method and pulse gradient analysis method, the accuracy is improved by 1.80% and 5.91%, respectively.ConclusionsThe proposed model has a simple structure, exhibits good stability, hence be applied to handling high-dimensional data with good discrimination and generalization ability.

BackgroundLaser simulation technology is widely used in the research of transient ionizing radiation effects in semiconductor devices. Fully dielectrically isolated silicon-on-insulator (SOI) devices exhibit different responses to dose-rate gamma irradiation compared to bulk-Si devices.PurposeThis study aims to examine the photocurrents of both Si-based and SOI NMOS transistors, and investigate the performance of a SOI MCU with varying dose rates.MethodAn irradiation experiment was conducted on three types of transistors by using a 1 064 nm/12 ns laser device, and the photocurrent was tested under various laser energies. A pulsed γ-ray source was employed to perform the transient γ dose rate radiation test on an SOI-integrated circuit. The function, electrical parameters, and flipflop chain status of the SOI-integrated circuit under different dose rates were measured. Based on theoretical model for the generation of photocurrent in SOI transistor, the dose-rate threshold for logic flipping and corresponding critical charge were estimated based on theoretical model for the generation of photocurrent in SOI transistor.ResultsThe results indicate that the peak photocurrent of the SOI transistor is approximately 20 times lower than that of the bulk silicon transistor with the same feature size under identical irradiation conditions. This reduction is attributed to the decreased charge collection sensitive area of the SOI transistor. Within a dose rate range from 1.0×109 rad(Si)·s-1 to 4.2×1011 rad(Si)·s-1, the SOI-integrated circuit exhibites no latch-up effect. However, irradiation-induced upsets are observed in the SOI-integrated circuit.ConclusionsThese upsets caused by transient radiation effects manifest as transient functional interruptions, variations in operating current and voltage, and erroneous flip-flop statuses. These irradiation-induced upsets in the SOI-integrated circuit are likely attributable to, among other factors, transistor upsets and circuit-level voltage fluctuations on printed circuit board.

BackgroundWhen an X-ray tube operates, a large number of backscattered electrons are generated inside the tube. Under the influence of the electric field inside the X-ray tube, these backscattered electrons are pulled back to the anode target by the electric field, generating additional X-rays, which ultimately have a negative impact on the quality of the emitted beam.PurposeThis study aims at the distribution rules of backscattered electrons in transmission X-ray tubes, and the changes in backscattered electrons caused by target materials and tube voltages as well as their impact on the outgoing beam quality.MethodsFirstly, Geant4 was used to establish the physical model of the transmission X-ray tube, and the theoretical calculations were compared with experimental data to verify the accuracy of the calculation program. Then, C, Si, Cu, Ag, and W were selected as anode target materials, and the influences of different electric field strengths and these target materials on the distribution of backscattered electrons in the X-ray tube were analyzed in details. Finally, the influence of the backscattered electrons on the outgoing beam quality under the effect of tube voltage was investigated.ResultsThe full width at half maximum, photon yield, and characteristic peak yield of the outgoing beam of the X-ray tube increase when considering the electric field, indicating that the backscattered electrons have effect on the outgoing spectrum quality. Simultanously, the backscattered electrons pulled back by the electric field in the first and second generations predominantly affect the outgoing beam quality. A change in the tube voltage alters the extent of influence of the backscattered electrons on the outgoing beam quality, and a larger tube voltage leads to a more obvious influence of the backscattered electrons on the outgoing beam quality of the X-ray tube.ConclusionsThe distribution of inside backscattered electron is closely related to the characteristics of the X-ray tube, including the tube voltage and target material. The results of this study may provide a reference for numerical simulation calculations of miniature transmission X-ray tubes.

BackgroundIn the process of accelerated development of nuclear medicine department in recent years, the construction of decay tanks and the storage time of radioactive wastewater containing 131I have become issues of great concern for environmental regulatory agencies and hospitals. Regulations and standards, such as "Basic Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources" (GB 18871-2002), "Radiation Protection and Safety Requirements for Nuclear Medicine" (HJ 1188-2021), and "Reply to Consultation on Several Clauses of the Nuclear Medicine Standard", have stipulated the compliant discharge methods for radioactive wastewater containing 131I from hospitals.PurposeThis study aims to demonstrate the methods for calculating the optimal volume of the decay tank and the minimum storage time so as to minimize the unnecessary construction costs and land use while ensuring compliance with regulations.MethodsFirstly, according to theoretical induction, a formula for the upper bound of total activity of 131I was presented for a full decay tank in a hospital. Then, the RJ (Radioactivity Judgement) equation group was put forward to address the calculation of minimum decay time and volume of the decay tank. Finally, the actual monitoring data from four hospitals were applied to verification the recommendation of this study.ResultsVerification results demonstrate that when the temporary storage period for radioactive wastewater containing 131I reaches the minimum time calculated by the RJ equations, the total discharge activity of 131I complies with the national environmental protection standards. Among the three compliant discharge methods for radioactive wastewater containing 131I in the decay tank, the method specified in GB 18871-2002 is advantageous for the operation of the nuclear medicine department in the hospital.ConclusionsWith the methodologies detailed in this paper, it is possible to significantly reduce the required volume of the decay tank and the minimum storage time. These findings provide clear and specific guidance for the construction of decay tanks in nuclear medicine departments and for the supervision and inspection conducted by regulatory authorities.

BackgroundWithin GEN-IV reactors, nuclear graphite plays a crucial role as both a moderator and reflector in an environment characterized by high temperatures and intense fast neutron irradiation. The exposure to fast neutron irradiation induces the formation of numerous Frankel defects in the nuclear graphite. These defects undergo processes of annihilation and diffusion, ultimately giving rise to larger defect clusters. This transformation in the microstructure of nuclear graphite directly impacts its macroscopic properties, necessitating a thorough investigation.PurposeThis study aims to comprehensively explore the evolution of defects in nuclear graphite under conditions of high-temperature irradiation which is essential for advancing reactor safety.MethodsFirstly, the 30 MeV 107Ag5+ ion source was employe to irradiate IG-110 nuclear graphite at 420 ℃, simulating the defect evolution behavior during fast neutron irradiation of nuclear graphite. Then, the energy loss, defect distribution, and ion implantation profiles of 30 MeV 58Ni5+ and 107Ag5+ ion beams bombarding standard nuclear graphite ICRU-906 (density of 2.26 g?cm-3, displacement energy of 28 eV) were calculated using the full cascade damage model in the SRIM (Stopping and Range of Ions in Matter) software. The cross-sectional structure of IG-110 nuclear graphite was characterized using micro-Raman spectroscopy. Finally, the relationship between the Raman spectroscopic features at various depths of IG-110 nuclear graphite and the irradiation damage dose was compared to investigate the evolution of IG-110 nuclear graphite microstructure with increasing irradiation damage dose (Displacements Per Atom, DPA).ResultsWith the increase in particle fluence, the characteristic parameters of the Raman spectra of nuclear graphite, including the ID/IG ratio (the ratio of the D peak height to the G peak height), the Full Width at Half Maximum of the G peak (FWHM(G)), and the shift of the G peak, all show significant increments. Compared to samples irradiated with 58Ni5+ at the same irradiation damage dose, the graphite Raman spectra irradiated with 107Ag5+ demonstrate higher ID/IG ratios and FWHM(G). At the same FWHM(G) level, the ID/IG ratio of the graphite Raman spectra irradiated with 107Ag5+ is greater than that of the samples irradiated with 58Ni5+.ConclusionsThe results of this study suggest that irradiation with heavier ions induces a higher rate of defect accumulation in nuclear graphite, leading to a more rapid reduction in graphite grain size and promoting the progression towards nanocrystallization.

BackgroundBoron neutron capture therapy (BNCT) is a promising tumor therapy method that irradiates 10B attached tumors with thermal or epithermal neutrons. Specifically, 7Li(p,n)7Be is one of the main methods for producing thermal or epithermal neutrons. The Li target is a critical component for BNCT device.PurposeThis study aims to design a composite material Li target with a semi-tirelike surface structure that rotates around a central axis.MethodsBased on a 2.5 MeV, 20 mA proton beam, the TOPAS Monte Carlo software was used to simulate the influence of curved lithium target with different radial axial ratios on the angular distribution, energy spectrum, and flux of neutrons. The steady-state temperature distribution of the neutron target was simulated by using ANSYS software to investigate the impact of radial axial ratios on heat dissipation.ResultsThe simulated results show that the semi-tirelike curved surface enhances the concentration of the output neutron beam. When the radial axis ratio is 1.5, the number of neutrons with an exit angle of 0°~45° is 2.59 times that of the same plane target whereas the number of neutrons with an exit angle of greater than 90° is only 0.29 times that of a planar target in the same situation. The heat dissipation performance is improved by the fold channel design and curved surface structure. The maximum temperature of the lithium layer is only 100 ℃ for a 50-kW proton beam incident, which satisfies the heat dissipation requirement.ConclusionThe innovative curved neutron target of this study significantly improves the forward performance of the outgoing neutron and heat dissipation performance when compared with the conventional plane target. These types of advancements will lead to good prospects in the field of BNCT.

BackgroundThe volume of aviation gamma spectrum data is immense. If only a central processing unit (CPU) is used for data post-processing, it would be constrained by computational efficiency.PurposeThis study aims to propose a CUDA-based graphics processing unit (GPU) parallel solution that optimally accelerates the denoising of airborne gamma-ray spectral data using wavelet transformation.MethodsFirst, the impact of different block sizes on computational time was tested to determine the optimal block size for processing airborne gamma-ray spectral data. Subsequently, a GPU, instead of a CPU, was used to calculate the acceleration ratio for handling airborne gamma-ray spectral data of different volumes, and wavelet basis functions were used for those with the same data volume. Finally, by introducing white noise to the experimentally measured airborne gamma-ray spectral data, the signal-to-noise ratio of denoised data was calculated to optimize the threshold denoising method suitable for parallel acceleration of the GPU.ResultsThe optimal two-dimensional block sizes for denoising airborne gamma-ray spectral data are 64×64 and 128×128. Among the wavelet basis functions, those that achieved a total time acceleration ratio exceeding 100 compared to CPU processing account for 80%, while those that reached an acceleration ratio exceeding 90 constitute 91%. The coif5 function achieves an acceleration ratio of 353 times whilst the acceleration ratio of the threshold denoising function approaches 570.ConclusionsAll wavelet functions exhibit insufficient denoising effects at low signal-to-noise ratios and excessive denoising effects at high signal-to-noise ratios. Significant denoising can be achieved using hard thresholding of coif5, soft thresholding of coif1, and improved thresholding of bior3.7.

Transition metal compounds with partially filled 3d electron shells exhibit a variety of physical and chemical properties. They are widely used in sensors, magnetic memory, photoelectronic devices, photocatalysis, and electrochemistry. Therefore, determining and understanding their electronic structure is extremely important. The quantitative analysis of soft X-ray absorption spectroscopy depends on theoretical calculation method for the full multiplet based on crystal field theory and hybridization theory. By incorporating these elements, full multiplet theory can provide an accurate model of the fine electronic structure of transition metals and their ligands. A brief introduction of the basic principles, key parameters, and software relevant to soft X-ray absorption spectroscopy is firstly provided in this review. Subsequently, the applications of theoretical calculation in analyzing the electronic structures of 3d transition metals are discussed with an emphasis on its application in the field of electrochemistry. Finally, some prospects for future development in this field are proposed.

BackgroundSmall rod-controlled pressurized water reactors have foregone the use of soluble boron and heavily rely on control rods and burnable poison rods for reactor control.PurposeThis study aims to explore the influence of control rods on the key performance metrics of a long-term small rod-controlled pressurized water reactor.MethodsFirst, a KLT-40 reactor used for nuclear icebreakers was taken as research object, and a critical rod position-search burnup code was developed based on OpenMC. Then, the core lifetime and other indicators such as the axial power offset, fuel utilization, and radial power peak factor, were compared between with and without the designed control rods for the design and analysis of the control rod layout of KLT-40 reactor. Finally, the influence of different move-in/out strategies on axial power offset was analyzed.ResultsThe core lifetime can be extended from 590 effective full power days (EFPDs) to 650~698 EFPDs with the designed control rods. Adopting a strategy of prioritizing the movement of low-value rod groups effectively reduces the axial power offset, with values of -0.69 and +0.8 decreasing to -0.29 and +0.52, respectively.ConclusionsThe control rod burnup calculation strategy adopted to accurately calculate the core lifetime of small rod-controlled pressurized water reactors can effectively reduce the axial power offset by using a reasonable move-in/out strategy.

BackgroundLiF-UF4 and LiF-ThF4 are used as addition salts in thorium-based molten salt nuclear reactors. Monitoring the product quality during the preparation of molten salts necessitates the analysis of the contents of the main metal elements in LiF-UF4 and LiF-ThF4, namely, lithium, uranium, and thorium.PurposeThis study aims to establish effective, rapid analysis methods for determination of primary metal elements in LiF-UF4 and LiF-ThF4 molten salts.MethodsFirstly, the LiF-UF4 samples were treated with nitric acid and hydrogen peroxide whilst the LiF-ThF4 samples were dissolved in aluminum nitrate. Then, inductively coupled plasma atomic emission spectrometry was used as rapid analysis methods for measuring these two molten salts, and manganese was employed as the internal standard element to reduce the effect of signal drift.ResultsIn the analytical method of LiF-UF4, the recoveries of Li and U are detected to be 99.6%~102.4% and 99.6%~101.8%, respectively. The relative standard deviations (RSDs) of Li and U are 0.2%~0.7%, and 1.1%~2.0%, respectively. In the analytical method of LiF-ThF4, the recoveries of Li and Th are detected to be 99.6%~102.3% and 99.6%~102.4%, respectively. The relative standard deviations of Li and Th are 1.9%~2.0%, and 0.3%~0.4%, respectively.ConclusionsThe proposed methods are simple, time-efficient, accurate, and suitable for rapid analyses of large numbers of samples.

BackgroundPassive residual heat-removal system (PRHRS) based on natural circulation has been widely used in small reactors connected on secondary side loop.PurposeThis study aims to develop and analyze a passive residual heat-removal system on the secondary side of a small reactor to improve reactor safety.MethodsBased on the completed heat-transfer experiment of the intermediate circuit of a small integrated nuclear power plant, Reactor Excursion and Leak Analysis Program (RELAP) was employed to determine and analyze the natural circulation characteristics of the intermediate loop were determined and analyzed.ResultsThe heat transfer rate results of the program are in good agreement with the experimental data, and the natural cycle characteristics of PRHRS can be characterized. The pressure of the system loop is determined by the average temperature of the primary side of the steam generator (SG), and the difference in the inlet temperature of the SG primary side, mass flow rate, and height of the cooling and heating sources significantly influence the heat-transfer performance of the SG system. The heat-transfer performance of the waste-heat-removal system is more sensitive to the resistance of the system loop when the temperature of the primary-side inlet of the SG is higher.ConclusionsThese results provide a valuable application for further investigations of passive systems in small reactors.

BackgroundThe reactor period will dramatically decrease at the beginning of reactivity insertion in the nuclear reactor, which may trigger the protection system of the reactor period and lead to unnecessary shutdown. The instantaneous short reactor period is influenced greatly by the inserting rate of reactivity, but also related with the present delayed neutron precursors, which is difficult to quantify.PurposeThis study aims to explore the relationship between the instantaneous short reactor period and the inserting rate of reactivity from a theoretical perspective.MethodsA point reactor model was used to deduce the inserting rate of the reactivity function using the variable factors of reactivity, reactor period, and reactor dynamic parameters, with some conservative assumptions to omit the effect of delayed neutron precursors. The relationship between reactivity insertion rate and transient period of reactor was derived after analysis on the short period phenomenon. Then, the formula of relationship was verified for several transient cases.ResultsThe results show that the reactor periods are all larger than the aim reactor periods for all transient cases when using the rate constraint of reactivity insertion in the aforementioned formula. According to the proposed theoretical framework, unnecessary shutdown during an instantaneous short reactor period can be avoided.ConclusionsA theoretical framework proposed in this study can be applied to the control of rod withdrawal rate during the operation of nuclear reactors.

BackgroundNuclear isomers are crucial in cosmic element synthesis and have potential applications in controlling nuclear energy release. Specifically, Europium (Eu) is significant in fundamental studies. For instance, 152Eu is used as a reference source for radioactive experiments, and its isomeric state 152m1Eu has a probability of 73% to produce cosmological p-nuclei 152Gd with β- decay. Therefore, 152m1Eu is a crucial nuclide in the nuclearsynthesis of p-nuclei 152Gd.PurposeThis study aims to realize the efficient excitation of 152m1Eu with a bremsstrahlung source generated by laser plasma.MethodsFirstly, the laser-plasma bremsstrahlung source was utilized to achieve the efficient excitation of 152m1Eu (45.6 keV, T1/2=9.31 h) in the experiment with yields of 8×104 particles/shot by this isotope. Then, numerical simulations of the yield of 152m1, m2Eu were performed using the Geant4-GENBOD program to get generation time, and peak excitation efficiency evolution with electron temperature.ResultsThe results demonstrate that when the electron temperature reaches 15 MeV, the yield of 152m1, m2Eu approaches saturation. When the incident electron charge is 17.6 nC, the yield of 152m1Eu is approximately 8×106 particles/shot, and that of 152m2Eu is approximately 2×105 particles/shot. The generation time of 152m1, m2Eu in the target is approximately 32 ps. When the electron temperature reaches 15 MeV, the peak excitation efficiency of 152m1Eu is expected to be ~1017 particles/s, and that of 152m2Eu is expected to be ~1016 particles/s.ConclusionsThe ultrashort ultrahigh intensity laser technology can significantly enhance the excitation efficiency of isotopes of the same nucleus, and this will provide an important research avenue for the study of cosmic element synthesis and nuclear energy release control applications.

BackgroundSelf-powered neutron detectors (SPNDs) are critical devices in the monitoring and protection systems of nuclear reactors, and their signal current directly reflects the value and distribution of the core power. Insulators play an essential role in the design of SPNDs and are the main factor affecting the calculation accuracy of the signal current.PurposeThis study aims to improve the accuracy of the calculation method of the SPND signal current, ensuring that the measured currents accurately reflect the reactor conditions and meet the highest industrial standards.MethodsFirstly, the signal generation mechanism of the SPND was thoroughly discussed, and three independent calculation methods of the current based on the inherent physical characteristics of the space electric field of an insulator were proposed. Then, high-fidelity simulations of the SPND were performed using the Monte Carlo code, and the three methods were validated based on the simulation results. In addition to the current caused by the neutrons, the current caused by the photons inside the reactor was quantitatively analyzed. Meanwhile, extensive radiation experiments on the various reactors have been performed to verify these three current calculation methods.ResultsThe difference between the results obtained by using the three methods is less than 1%, demonstrating a considerable accuracy. In addition, the current of the rhodium SPND is primarily owing to the neutrons, whereas the photon-induced current is generally less than 5%. Experimental verification results on the several operating reactors show that the difference between the theoretical and experimental results is less than 3%, which also proves its effectiveness and accuracy.ConclusionsThis method has been applied to the large Chinese Gen-III advanced pressurized water reactor (HPR1000) and is universal. It can be used for the signal analysis of different types of SPNDs, as well as for providing valuable references for core monitoring systems in other reactors, such as the Gen-IV fast reactor as well as future fusion reactors.

BackgroundThe use of neutron-based methods to locate radioactive sources within sealed containers holds great practical significance.PurposeThis study aims to locate an AmLi source within the detection space of four position-sensitive neutron detectors.MethodsFirst of all, the Monte Carlo method was applied to designing the moderating shield of the detector. Then, a delay circuit was added to one end of the detector hence the position coordinate of the source y-axis (detector axis) was determined according to the time difference between the two ends of the detector to detect the source neutron signal, and the axial position function of each detector was calibrated. Finally, the detector was used to build a measurement space around sealed container, and the position coordinates of the x-axis and z-axis (the other two directions) were determined by the ratio of the neutron count rate of the two adjacent detectors, so was the function calibration. During the measurement of neutron point-source 3D coordinates, the detector with the largest count rate was first selected to determine the axial coordinates of the source, and then the coordinates of the other two directions of the source were determined according to the ratio of the count rate of the two adjacent detectors to the detector, so as to realize the location of the source.ResultsMeasurement results of five different positions of the point source in the detection space show that the positioning deviation on each coordinate axis is within 1.5 cm, and the relative standard deviation of neutrons measured in this detection space is within 1%.ConclusionsThis methodology effectively demonstrated the feasibility of employing neutron position-sensitive detectors to precisely locate radioactive sources, establishing a strong foundation for future endeavors aimed at accurately determining the positions of nuclear materials within processing equipment in nuclear facilities.

BackgroundThe silicon photomultiplier (SiPM) is sensitive to environmental noise, its performance is greatly affected by ambient noise.PurposeThis study aims to design a negative-feedback selective amplifier circuit for SiPM coupled plastic scintillator detector to reduce the noise and improve the overall performance of SiPM.MethodsIt is compared with the traditional OPA657 transimpedance feedback amplifier circuit. The circuit consisted of an RC filter input and an integrated operational amplifier AD8014 with advantages of a high gain and low input noise. The CR high-pass filter circuit was used as the comparator signal input to further filter out signal noise and prevent signal reflection from interfering with the amplifier. The experimental circuit is simulated using Micro-cap12 to obtain the relevant circuit parameters. Finally, the dark noise level and signal consistency of a 137Cs source at room temperature was record, and performance of this amplifier circuit for SiPM coupled plastic scintillator detector was compared with that of the traditional OPA657 transimpedance feedback amplifier circuit. [Results and Conclusions] Comparison results show that the proposed circuit effectively filters out ambient noise and exhibits a fast rise time. The output pulse rise time is found to be less than 12 ns, and the dark noise level is observed to be less than 30 mV, which are better than those of the transimpedance amplifier circuit.

BackgroundDetectors based on the third-generation semiconductor material silicon carbide (SiC) offer several important advantages, such as compactness, faster charge-collection times, and easier n/γ identification, and they are widely used in reactor core dose monitoring.PurposeIn this study, the n/γ signal amplitude, neutron fluence rate, and linear response calibration performances were tested systematically for a self-developed third-generation SiC semiconductor detector.MethodsFirstly, the neutron conversion layer material 6LiF (with a 95% abundance of 6Li) was sprayed onto a SiC substrate using electron beam evaporation vacuum coating technology to achieve the optimized thickness of 25 μm for the self-developed third-generation SiC semiconductor detector. Then, 241Am α radioactive source (activity 9.37×103 Bq) was used to observe α particle response signal amplitude, and γ response testing of radiation was conducted in the 137Cs γ source (activity 6.23×107 Bq) environment. In addition, the SiC detector's neutron flux response linearity, γ dose rate response linearity and calibration of neutron fluence rate response linearity were measured in the standard radiation field systems.ResultsThe measurement results show that the SiC semiconductor detector has a linear fit of R2 = 0.996 9 in the neutron fluence rate range of 1×103~1×106 cm-2?s-1, with a good linear response, and the response range of the neutron/gamma dose is 0.005~20 Gy?h-1.ConclusionThe SiC detectors with such good n/γ performance can be used for real-time and accurate monitoring of neutron and gamma doses in nuclear power field reactors.

BackgroundDuring the data acquisition of an extreme ultraviolet (EUV) spectroscopic diagnostic system in experimental advanced superconducting Tokamak (EAST), significant amounts of one-pixel noises are consistently observed. This is attributed to the influence of hard X-ray on the charge coupling device (CCD) detector.PurposeThis study aims to detect and denoise spectral images in the EUV spectral image processing system based on the field programmable gate array (FPGA).MethodsFirst of all, based on limiting filtering algorithm, spectral image processing was optimized by replacing fixed limiting thresholds and sample deviations using parameters such as standard deviation and deviation from the mean, and Andor Solis software to was applied to converting the SIF format spectra to BMP format spectral images. Then, data discrimination method was combined with limiting filtering algorithm to process spectral image data in stages according to the setting working area of the algorithm. Finally, a simulation test module was designed to process the video data converted from EUV spectral images in AX7Z100 ZYNQ FPGA platform, simulating the actual acquisition process for functional testing of the system. To assess the system's capacity to protect effective spectral data, the image data was transferred to one-dimensional spectral data for comparative analysis.ResultsBased on the experimental results, the EUV spectral image processing system of this study can effectively eliminate noise data points in the spectral image while essentially maintaining the integrity of the spectral data.ConclusionsThis study enhances the processing of the spectral data and provides a new technological path for EUV spectral image processing.

BackgroundExtraction of carbon from water is a crucial preprocessing step for measuring 14C in environmental waters using liquid scintillation spectrometry.PurposeThis study aims to explore the optimal technological conditions for extracting carbon from water using wet oxidation method.MethodsA wet oxidation system combining sodium persulfate and Fenton's reagent, along with phosphoric acid acidification and nitrogen bubbling, were employed for the wet oxidation carbon extraction experiments on two types of water samples with known (deionized water + sucrose) and unknown carbon components, each with a volume of 10 L. Simultaneously, carbon extraction experiments were conducted on the water samples having unknown carbon component, using a combination of wet oxidation and 185 nm ultraviolet (UV) oxidation so as to determine the optimal timing and sequence of reagent addition, as well as the optimized reagent dosage and ratio. Further experiments under optimized conditions were conducted to obtain more results for deep analysis.ResultsUnder the optimized conditions, after a 3-h reaction at 90 °C, the organic carbon extraction rate for the known carbon component (deionized water + sucrose) exceeds 96%. The total carbon extraction rate from the unknown carbon component water is (96.8±0.3)%, with an inorganic carbon extraction rate >98.5%, and an organic carbon extraction rate of (93.4±0.2)%, while the oxidation rate of tannic acid-type organic compounds is only (88±0.2)%. After the combination of wet oxidation and 185 nm UV oxidation, the total carbon extraction rate for the unknown carbon component increases to (98.3±0.5)%, with an inorganic carbon extraction rate ≥99% and an organic carbon extraction rate that can reach (95.6±1.4)%.ConclusionsResults of this study indicate that wet oxidation alone cannot represent the carbon recovery rate in actual water samples using typical organic compound carbon recovery rates. The combination of wet oxidation and 185 nm UV oxidation proves to be a more effective method for carbon extraction from water.

BackgroundAn electrolyte waste salt containing LiCl and various products is generated during the pyroprocessing of spent nuclear fuel in metal fast reactors. Separating metal impurities from waste salt can purify molten salt, facilitate salt recycling, and reduce the amount of waste salt, achieving waste minimization.PurposeThis study aims to investigate the effects of key factors on the application of the cold finger crystallization method used for removal of Sr and Ba from molten LiCl salt.MethodsA homemade cold finger experimental apparatus was applied to the experimental removal of two alkaline earth metals, Sr and Ba, from molten LiCl salt, and Fluent software was employed to simulate the application of cold finger crystallization equipment during dry reprocessing. The effects of crystal growth time, initial crystallization temperature, and initial SrCl2/BaCl2 concentrations on the removal ratio of the crystalline salt during the process were analyzed.ResultsThe initial temperature of molten salt is a critical factor that influences cold finger separation efficiency. When the initial temperature reaches 660 ℃, the removal efficiency improves. Moreover, when the impurity contents of Sr and Ba in molten salt are lower than 0.55%(w/w), the removal efficiency of the cold finger crystallization method can exceed 80%. Further analysis shows that the removal effects of different parts of molten salt crystals differ. The solvent salt at the top of the molten salt crystal is better, and the removal ratio of the bottom and inner salts is lower. Therefore, the optimal conditions for removing Sr and Ba from LiCl crystalline salt require an initial temperatures of 660~670 ℃, an airflow intensity of 10 L·min-1, and a growth time of 20 min. Under these optimal conditions, the removal ratio can reach 90%.ConclusionsThe proposed approach is feasible for purifying solvent salts from electrolyte waste molten salt via cold finger crystallization. This study provides a reference for purifying waste salt and reusing molten salt.

BackgroundPlutonium is an important element for nuclear energy production and radioactive waste disposal, and evaluating its species distribution in natural water systems is essential for investigating its migration behavior. Recently, with the advancements in related research, some new plutonium species have been discovered and confirmed, hence previous related research is now inadequate for accurately describing the species distribution of plutonium in solutions.PurposeThis study aims to understand the speciation distribution of plutonium in different water composition systems.MethodsThe geochemical calculation software PHEEQC was employed to systematically evaluate the effects of the pH value and coexisting ion concentration on the species distribution of plutonium. The plutonium species proportion in different natural water systems was estimated in according with the latest thermodynamic data embedded into PHEEQC software.ResultsThe results reveal that hexavalent plutonium in oxidized groundwater with low hardness mainly exists as PuO22+ or PuO2CO3 under acidic conditions, whereas PuO2(CO3)22- or PuO2(CO3)34- dominate under neutral or alkaline conditions. Although PuO22+ and PuO2CO3 are the main species of plutonium in acidic environments, CaPuO2(CO3)32- is the dominant species under neutral or basic conditions whereas the calcium concentration is high in the solution.ConclusionsTherefore, the hardness of water (particularly, the calcium concentration) can be concluded to be among the important factors affecting the species of plutonium and must be carefully considered in the geological disposal of plutonium.

BackgroundFull-field transmission X-ray microscopy (TXM)–X-ray absorption near-edge structural (XANES) (TXM-XANES) is an imaging method that combines TXM and XANES. By measuring the TXM images of multiple energy points before and after the K-edge of the element of interest, the distribution of elemental chemical states in the sample can be determined. Conventional TXM-XANES data requires the acquisition of images and background images at each energy point, which results in a large data volume and extended acquisition time. At the nanoscale, the instability of the mechanical structure and the movement of the sample may impact the TXM-XANES data analysis.PurposeThis study aims to use machine learning methods to achieve background-image sequence prediction modeling using only two spectral background images to reduce the data volume and shorten the acquisition time.MethodsMachine learning, polynomial regression, and linear interpolation were used to generate background image sequences. A prediction model of the complex linear relationship between image grayscale values, pixel points, energy, and other related features based on the known data was established. Subsequently, the entire background image sequence could be predicted using only two spectral background images. Finally, 2D energy distribution maps obtained by conventional TXM-XANES method and this improved TXM-XANES method for standard powder samples and lithium battery cathode material sample were compared and analyzed in details.ResultsThe proposed method achieves complete background-image sequence prediction modeling using only two spectral background images. The comparison results show that the proposed method requires a lower data volume and shorter acquisition time than the conventional TXM-XANES methods, which can significantly improve the experimental efficiency of TXM-XANES.ConclusionsThis study addresses the issues of prolonged data gathering time and poor experimental efficiency in TXM-XANES by developing a machine learning model that builds complex linear relationships between pixel values and related features, such as location and color, using machine learning. Using the two TXM-XANES background images for full-sequence background prediction achieves rapid prediction of the entire background image sequence.

BackgroundThe Shanghai Synchrotron Radiation Facility (SSRF) is a user facility that requires continuous addition of new beamlines to meet the needs of an increasing number of users. Adding new insertion devices (IDs) to the synchrotron radiation light source is a complex and critical task.PurposeThis study aims to highlight the important considerations of dual IDs installation and commissioning when adding new beamlines to synchrotron light sources.MethodsThe newly added dual IDs 04IVU (undulator) and 04Wigger for new beamline added to SSRF were taken as an example, key tasks that need to be addressed in the accelerator aspect of engineering were presented in details. Firstly, the design parameters and magnetic measurement results of the IDs were introduced, and the SPECTRA program was employed to calculate the spectral brightness of dual IDs. Then, problems encountered during the installation process and their solutions were discussed, and an orbit feedforward compensation system equipped with corrector coils for mitigating orbit distortions caused by the IDs was implemented. Finally, the significance, methods, and results of vacuum cleaning work in the front-end area, together with a summary of the issues encountered during the trial operation and corresponding solutions were outlined.ResultsThe maximum spectral brightness of 04IVU and 04Wiggler are 5.86×1019 Photon·s-1·mm-2·mrad-2·(0.1% B.W.)-1 and 2.75×1015 Photon·s-1·(0.1% B.W.)-1 at photon energy of 3 keV, respectively. After compensation, the distortions of 04IVU and 04Wiggler are reduced to below 2 μm.ConclusionThe dual IDs have been successfully installed and commissioned. This study provides reference value for installation and commissioning of accelerators associated with the new beamline in other synchrotron radiation light sources.

As a passive nondestructive nuclear technique, gamma ray spectrometry is used in many radioactivity laboratories. Gamma-ray spectrometry enables the identification of radionuclides in a sample from their emitted photon energy and calculation of their activities from the number of photons collected for each energy. However coincidence summing effects will influence the reliability of both radionuclide identification and calculation of activity. Coincidence summing effects appear when sources emitting coincident gamma rays are measured via gamma-ray spectrometry. Those effects that result in losses from the full energy peaks and enhancement of sum peaks influence the accuracy of the spectral analysis. To eliminate this influence, many correction methods have been established. Research on the coincidence summing effect (CSE) originated in the 1960s. Subsequently, many algorithm-based generation mechanisms of CSE have been built together with the development of corresponding correction software. Massive amounts of technological information on and achievements about coincidence summing correction have been reported by researchers from different countries, hence several intercomparisons of these methods, and self-consistency testing of cascaded additive effect correction algorithm were organized by the International Committee for Radionuclide Metrology (ICRM). Based on a detailed summary of the development history of correction methods, the CSE mechanisms, correction algorithms, correction software, and application of correction techniques were reviewed in this paper. Cobalt-60 was taken as an example to illustrate the influence of the summing-in and summing-out effects, with correction equations based on different measurement geometries and considering the impact of angular correlations. Meanwhile, the performance of different algorithms and software were compared and analyzed. Combined with the current research status, some suggestions are presented for future research for domestic researchers on the coincidence summing effect correction. First, the efficiency must be accurately established for the geometrical conditions of the measurement. Second, the number of cascades in the algorithm must be taken into account owing to its influence on the results. Third, correction software with a user-friendly interface and database of accurate decay schemes should be developed.

BackgroundLiquid molten salt reactors that use chloride salts as fuel are characterized by the high solubility of heavy metals and a hard energy spectrum, hence are ideal for transmuting transuranic nuclides (TRU). A small modular reactor exhibits the characteristics of a modular design and construction, which is one of the future development directions for nuclear energy.PurposeThis study aims to investigate the TRU incineration characteristics of a small modular chloride fast reactor (sm-MCFR) that can be refueled online and applied to the disposal of TRU in nuclear waste produced by pressurized-water reactors.MethodsFirstly, a 50-MW sm-MCFR scheme was proposed, and its neutron properties, as well as the performance of TRU incineration, were explored using the Monca program TMCBurnup (TRITON MODEC Coupled Burnup Code), a combination of the SCALE 6.1 (Standardized Computer Analyses for Licensing Evaluation) and the high-precision point burn-up program MODEC (Molten Salt Reactor Specific DEpletion Code). Then, the analysis of critical parameters, burn-up evolution, and the transmutation efficiency of both TRU mixed with Depleted Uranium (DU) and TRU combined with 232Th were investigated, using a straightforward post-processing approach.ResultsThe findings of this study indicate that using TRU as fission fuel in the sm-MCFR requires the online addition of TRU. When the heavy metal balance is maintained, the effective multiplication factor (keff) is less than 1. Conversely, when the balance is not maintained, keff > 1, allowing continuous operation. When operating at full power for 40 years, the core's residual TRU content will be significantly higher than the initial fuel load, with 657 kg remaining for the TRU+Th mix and 725 kg for the TRU+DU mix. Notably, the sm-MCFR demonstrates efficient transmutation when TRU is added online without maintaining the heavy metal balance. Over 40 years at full power, the transmutation rates will be 41% for TRU+DU and 49% for TRU+Th, effectively reducing the production of long-lived small-actinide elements.ConclusionsThe sm-MCFR can effectively incinerate TRU and provide a feasible scheme for minimizing spent fuel.

BackgroundU3Si2 is regarded as one of the most promising accident-tolerant nuclear fuels for light water reactors and is expected to replace the UO2 nuclear fuel in the future. Currently, spark plasma sintering (SPS) is an advanced technique for preparing U3Si2 pellets; however, the influence of SPS parameters on the performance of the pellets is unclear.PurposeThis study aims to investigate the effects of different sintering parameters (temperature and pressure) on the mechanical and thermal properties of the U3Si2 pellets prepared using SPS technology.MethodsThe thermal diffusivity of U3Si2 pellets was measured using a laser flash apparatus, and the thermal conductivity of the pellets was calculated. The mechanical properties of the pellets, including hardness, Young's modulus, and fracture toughness, were measured using nanoindenter. Thereafter, the influence of different sintering temperatures in the range of 1 000~1 300 ℃ and pressures in the range of 30~90 MPa on the mechanical and thermal properties of U3Si2 pellets were carefully examined.ResultsThe measurement results show that the thermal conductivity of the as-synthesized pellets increases linearly with temperature in the range 27~700 ℃. Moreover, increasing the sintering temperature and pressure improves the thermal conductivity of the U3Si2 pellets. The hardness and Young's modulus of the pellets increase with an increase in sintering temperature. They also exhibit a trend of first increasing and then stabilizing with increasing pressure, and tend to fully stabilize at 60 MPa. Moreover, the fracture toughness of the pellets decreases with the increase of sintering temperature and increases with increasing pressure.ConclusionsBased on the above results, optimized SPS parameters for the U3Si2 pellets are proposed, and this study provides a reference for the preparation of high-performance U3Si2 pellets.

BackgroundHigh-order harmonics of neutron diffusion equations can be used to reconstruct the neutron flux distribution in a reactor core, but traditional source iteration methods or modified source iteration methods have low solving efficiency.PurposeThis study aims to provide a reliable and efficient method for reconstructing the neutron flux distribution in reactor cores.MethodsFirstly, the neutron diffusion equation was discretized using the finite difference method. Then, the implicitly restarted Arnoldi method (IRAM) was employed to solve the eigenvalue problem of the neutron diffusion equation and obtain high-order harmonic samples for different macroscopic cross-section states. Subsequently, a low-order model for the neutron diffusion equation was constructed by using these samples and a combination of proper orthogonal decomposition (POD) and Galerkin projection, and an error model was developed to characterize the accuracy of eigenvalue and harmonic calculations. Finally, relevant programs were developed to reconstruct the neutron flux distribution in the two-dimensional steady-state TWIGL benchmark problem and validate the accuracy of the model.ResultsThe computation results show that the IRAM exhibits high accuracy in solving the high-order eigenvalues and harmonic problems of the neutron diffusion equation, with an error on the order of 10-14. The reconstruction of the neutron flux distribution based on the POD-Galerkin low-order model also maintains a high level of accuracy. The solution error increases with the order of the eigenvalues, with an error magnitude less than or equal to 10-12. The reconstructed neutron flux distribution closely matches the reference solution in the reactor core, and the error in the effective multiplication factor is only 8.7×10-5. Additionally, the computation time for the low-order model is only 10.18% of the full-order model.ConclusionsThis study provides a reliable and efficient method for reconstructing the neutron flux distribution in reactor cores. The method can be used not only to reconstruct the steady-state neutron flux distribution but also has the potential to predict the transient neutron flux distribution, which is expected to be further expanded in future applications.

BackgroundWith the continuous development of society and economy, nuclear energy has emerged as a crucial solution to address global energy shortages. However, uranium reserves on land have diminished after nearly half a century of extraction, resulting in irreversible harm to the natural environment. Consequently, seawater uranium extraction has become a more eco-friendly and abundant source of uranium resources, compared to terrestrial uranium mining. And extracting uranium from seawater emerges as a prudent strategy.PurposeThis study aims to use kinetic chromatography to extract uranium from seawater and investigate the separation mechanism.MethodsLeveraging the concepts of dynamic chromatography, a pulsed-injection energy-endowed kinetic chromatography column was developed. The column utilized spherical SiO2 with a diameter of 0.2 mm and a length of 5 m as the filler. After filling, it encompassed approximately 30 600 chromatographic separation units. A peristaltic pump for pulse injection was applied to the kinetic chromatography system, and a switching pulse injection device was equipped to enable time control splitting. Subsequently, a self-developed on-line spectrophotometric detector was utilized to monitor the concentration of target components in the solution in real-time so as to avoid human control errors during multistage separation experiments. Additionally, the behavior of uranyl ions in kinetic chromatography under varying conditions and determine the best separation conditions was investigated by experiments using different mobile phase carriers, pH values, injection flow rates, energy-endowed modes, and energy-endowed series, and various energy-endowed methods, including water bath heating, ultrasonic, and external magnetic fields, were employed to achieve optimal separation conditions. Finally, the separation factor of uranium and sodium ions in actual seawater uranium extraction was calculated to explore the separation mechanism by separate studies of uranium, europium, and sodium ions, as well as an actual seawater uranium extraction.ResultResults of this series of studies demonstrates that the best separation effect is achieved when using hydrochloric acid as the mobile phase carrier in dynamic chromatography, with a pH of 2, a sample flow rate of 4.109 mL·min-1, water bath heating as the energy-endowed mode, a heating temperature of 50 ℃, and a heating series of 4. Under these optimal conditions, the separation factor between uranium and sodium ions can reach 1.1854 in the separation studies of uranium, europium and sodium ions. In the real seawater uranium extraction study, the separation factor between uranium and sodium ions can reach 1.575. After simulation and calculation, the theoretical separation of uranium and sodium ions in seawater requires a minimum of 20 levels.ConclusionsThe efficient and rapid extraction and separation of uranium from seawater is facilitated in this study using a pulsed-injection energy-endowed kinetic chromatography column. This separation strategy allows for the efficient separation of light and heavy particles without interaction of the mobile and stationary, resulting in a high sample recovery rate and no need for column regeneration. This technique has potential for the separation of other nuclides, making it a versatile tool for nuclear chemistry research.

BackgroundThe direct simulation of the γ radiation field in a large space has a very low calculation efficiency.PurposeThis study aims to apply the global variance reduction (GVR) method to the calculation of the γ radiation field in a large space.MethodsFirstly, the volume correction factor was introduced for modifying the lower limit of the weight window wth to address the over-splitting problem caused by the volume difference between the counting cells/grids. The global quality factor (FOMG factor) calculated by the flux-based GVR method using the volume correction was 39 times higher than that obtained by direct simulation. Then, a non-counting area correction method was proposed to address the time-consuming problem encountered in non-counting area calculation while the FOMG factor was further improved by 40%. Finally, based on both the volume correction and non-counting area correction, the calculation of the γ radiation field were compared with that of seven GVR methods based on the particle error, weight, track, number, energy, collision and flux, respectively. The smoothing factor SI was introduced into the flux-based GVR method for results further analysis. [Results and Conclusions] The results show that the FOMG factor calculated by the seven GVR methods is about 2~3 orders of magnitude higher than that obtained by direct simulation, and the standard deviation σ is reduced by 2~3 orders of magnitude. The FOMG factor calculated by the weight-based GVR method is 2 304 times higher than that obtained by direct simulation; this value yields the best variance reduction effect among all GVR methods. As SI increases, the lower limit of the weight window wth of the simulation decreases, and the FOMG factor first increases and then decreases. When SI=0.8, the calculated FOMG factor has the largest value, which is 3 246 times higher than that obtained by direct simulation.

BackgroundA megawatt-class nuclear power system has been developed by coupling a heat pipe reactor with a supercritical carbon dioxide (S-CO2) Brayton cycle. This system offers advantages in terms of high safety, power density, and compactness.PurposeThis study aims at the operation characteristics of this power system with high efficiency and compactness.MethodsThe coupling code of a self-developed heat pipe reactor transient analysis code, Transient Analysis code for heat Pipe and AMTEC power conversion space Reactor power System (TAPIRS), and supercritical carbon dioxide Brayton cycle transient analysis code (SCTRAN/CO2) were utilized to analyze the open-loop dynamic characteristics under conditions of reactivity disturbance, load disturbance, cooling water temperature disturbance, and cooling water mass flowrate disturbance. Then, the control system was designed. On this basis, three load variation operation conditions, i.e., linear load variation, stepped load variation, and load rejection, were simulated and analyzed.ResultsThe simulation results show that the rotational speed of the new nuclear power system is sensitive to the disturbances and needs to be controlled. The bypass flowrate increases under low load conditions, hence the flowrate of the compressor needs to be controlled as well. The system can adjust the load from 0% to 100% at a rate of 6% FP (full power)·min-1. It is capable of implementing stepped load changes, although it experiences slightly more pronounced fluctuations. Under load rejection conditions, the stabilization time might be prolonged, but it will eventually stabilize with all parameters remaining within safe limits.ConclusionsThis study provides a reference for the conceptual design of new nuclear power systems with high efficiency and compactness.

BackgroundThe total dose effect of ionization on satellites in Geostationary Earth Orbit (GEO) is caused by space particle radiation in GEO.PurposeThis study aims to combine high-performance cerium-doped yttrium-lutetium silicate (LYSO:Ce) crystals with an aluminum layer to shield the effect of proton irradiation for effective electron radiation dose detection.MethodsFirstly, the detector model was established using Geant4, the shielding effects of different materials for LYSO:Ce detector were compared. Then, the response characteristics of the detector were analyzed, and the factors affecting the output response of the detector at different shielding layer thicknesses were investigated.ResultsThe results showed that the effect of proton irradiation can be eliminated by using 0.022 mm aluminum as the shielding layer. The LYSO:Ce crystal-based detector has a good linear response, and the secondary electrons and photons generated when electrons pass through the shield improve the response sensitivity of the detector. The ionization stopping power of the electron is approximately inversely proportional to the square of the incident electron velocity, and the detector response to radiation is enhanced when the transmission distance between the electron and the detector was appropriately increased. The highest electron detection efficiency of the detector occurs in the energy range of 0.04~1 MeV.ConclusionsThis study on the characteristics of electron radiation dose detector on the basis of LYSO:Ce crystals combined with an aluminum layer provides a technical reference and theoretical support for designing new scintillator space radiation detectors.

BackgroundEffective measurement using dual Geiger-Müller (GM) counter tubes hinges on the range switch control technology. This technology facilitates the selection of the appropriate counter tube for measurement. Nonetheless, the performance disparities between the two types of GM counter tubes imply that the conventional method of bifurcating the measurement range into two sections results in reduced linearity for the overlapping measurement ranges.PurposeThis study aims to propose a new control method for range switching to enhance the linearity of the overlapping ranges in the measurement using dual GM counter tubes.MethodsA dual GM counter detector was consisted of a low range GM counter tube with measurement range of 0.1 μ Sv· h-1~10 mSv·h-1, and a high range GM counter tube with measurement range of 1 mSv·h-1~100 Sv·h-1. The measurement range of 0.1 μSv·h-1~100 Sv·h-1 was segmented into three categories: low, medium, and high. Rapid and automatic transitions between these three ranges were facilitated by high-voltage control circuit, measurement range control circuit and dead time regulation circuit. During the medium range of 1~10 mSv·h-1 measurement, two range switching threshold points were set within the overlapping area, and data from the two GM counters were weighting processed respectively in the single-chip processor, hence appropriate weighting factors that maximize the linear fit of the measurement results of the dual GM counter were obtained. Finally, the 241Am source and 60Co ource were employed to test the dual GM counter detector circuits.ResultsPreliminary test results indicate that the proposed dual GM counter detector facilitates fast automatic transitions among the three measurement ranges, and the linear fit of the counter tube in the overlapping area from 1 000 μGy·h-1 to 10 000 μGy·h-1 is enhanced, making the linear fit of the dual GM counter reach up to 0.999 1 within the measurement range of 251~25 130 μGy·h-1.ConclusionsThe overall measurement linearity of the dual GM counter is effectively improved by proposed control method of this study for range switching.

BackgroundThe very small-angle neutron scattering (VSANS) spectrometer is currently being constructed at the China spallation neutron source (CSNS). The neutron detector plays a key role in meeting the detection requirements of the spectrometer.PurposeThis study aims to achieve accurate measurement of neutron diffraction in the small angle scattering mode which requires the position resolution of the neutron detector to be ≤2 mm, and the detection efficiency to be ≥60%@0.4 nm.MethodsA large-area position-sensitive neutron scintillator detector mainly composed of 6LiF/ZnS(Ag) scintillation screen, wavelength shift fiber (WLSF), and a silicon photomultiplier (SiPM) was designed and developed to achieve high efficiency and high-resolution real-time detection of thermal neutrons. The optical transmission performance of the 0.5-mm-diameter wavelength shift fiber was investigated in detail. The gain and the thermal noise characteristics of various SiPMs were examined. Finally, a 300 mm×300 mm engineering prototype of the detector was developed and tested by experiments. The detection efficiency was calculated with comparison of the incident neutron counts of the stander 3He tube whilst the detector's position resolution was tested using a boron-doped aluminum plate with a narrow slit bearing the label 'CSNS'.ResultsThe neutron beam test results show that the position resolution of the detector is 1.2 mm×1.2 mm and that the neutron detection efficiency is (61.8±0.2)%@0.4 nm.ConclusionsThe high-resolution neutron scintillator detector developed in this study satisfies the engineering design target, and meets the neutron scintillator detector meets the neutron diffraction measurement needs for the VSANS spectrometer of the CSNS.

BackgroundIonizing radiation can cause damage to animal's intestinal tissue. Citrulline is produced in the intestinal epithelial cell and has been proven to possess a protective effect on the gastrointestinal tract.PurposeThis study aims to investigate the protective effects and the underlying mechanisms of citrulline in the context of radiation-induced intestinal injuries.MethodsFirstly, a mouse model of an acute radiation-induced intestinal injury was established, incorporating a normal control, a simple irradiation, and an irradiation plus citrulline group. Then, these groups were employed to scrutinize the protective effects and mechanisms associated with citrulline. Subsequently, hematoxylin-eosin staining was used to examine the morphology of the mice's intestinal tissue, and the Elisa kit was employed to quantify endotoxin levels in plasma, as well as nitric oxide and inducible nitric oxide synthase in the intestinal tissue. Finally, focal adhesion kinase and Occludin levels in the intestinal tissue were assessed using western blotting.ResultsThe experimental results demonstrate that intraperitoneal injection of 1 g?kg-1?d-1 citrulline for one week following irradiation significantly extend the median survival time of irradiated mice and increase their body weight. Moreover, it markedly reduces plasma endotoxin levels, elevate the expression levels of focal adhesion kinase (FAK) and intestinal tight junction protein (Occludin), and decreases the expression levels of nitric oxide (NO) and inducible nitric oxide synthase (iNOS) in the intestinal tissue.ConclusionsCitrulline enhances the integrity of the intestinal barrier in irradiated mice, improves barrier function, mitigates nitrosative stress, and demonstrates a protective impact on radiation-induced intestinal damage in mice.

BackgroundAccurately quantifying the uranium in uranium yellow cake material is the key to selecting the subsequent processing technology. As an essential nondestructive testing method for uranium-containing materials, the active multiplicity method is proposed to quantify uranium by recording and analyzing 238U fission information induced by neutron sources. However, the quantitative results are biased owing to the neutron self-shielding of the uranium yellow cake material itself and differences in water content between samples.PurposeThis study aims to rapidly measure the uranium content of uranium yellow cake material using active multiplicity method and further improve of measurement accuracy.MethodsFirst of all, following the comparison of the excitation effects of different neutron sources on a sample using the MCNP (Monte Carlo N-Particle Transport) program, a 241Am-Be source was selected to simulate the sample measurement process and optimized using MATLAB programming combined with MCNP. Then, the curve of multiplication factor M versus the uranium mass was obtained by simulation, and an appropriate M was selected according to the net content of the sample. Finally, the quantitative error caused by the difference between neutron absorption and water content in the process was investigated, and the double rate was corrected using the relationship between S0/Si and D0/Di and then calculated.ResultsThe simulation results of a series of samples with different masses and water contents show that a large gap is found between M and the leakage multiplication factor ML caused by neutron self-shielding. The error in uranium quantification is less than 5%; neutron self-shielding due to the change in water content affects the single, double, and triple counting rates (S/D/T). The relative error of uranium quantification can be controlled at around 10%.ConclusionsThis study has significance and important reference value for further research on the application of the active multiplicity method in the production and measurement of uranium yellow cake.

BackgroundSextupoles are an important component of storage ring octonal unit in high energy synchrotron radiation sources. They require complex technological processes and precise center extraction. Therefore, it is necessary to find out the most reasonable calibration scheme for the mechanical center.PurposeThis study aims at the calibration scheme for the mechanical center extraction of high energy photon Ssurce (HEPS) sextupoles, and obtaining the corrected mechanical central coordinate system.MethodsThe method of directly measuring the reference plane of the sextupole was adopted for the mechanical center calibration of magnets. By rotating the conventional calibration coordinate system with a given pole seam deviation angle, the three polar seam surfaces were brought closer to the theoretical positions to decrease the main diagonal component of the sextupoles. The mechanical center calibrations were performed twice for each hexacode iron to further reduce the impact of the polar seam error.ResultsThe calibration results show that the calibration repeat accuracy is 0.005 mm for the sextupole. The standard deviation between the measured value and the design value of the pole seam spacing is 0.015 mm. Additionally, the standard deviation of the reference point before and after the rotation of the coordinate system is 0.09 mm, with a maximum rotation angle of 0.6 mrad.ConclusionsThe calibration scheme of this study can be used to improve the calibration accuracy and provide reference for the calibration of similar equipment. It ensures smooth installation of accelerator devices and is of significant importance for accelerator collimation measurements.

BackgroundWhen conducting experiments at a synchrotron radiation facility, it is necessary to scan the position of the sample or the beam energy within a certain range. The scanning modes are categorized as step scanning and on-the-fly. In the step-scanning mode, there is a significant amount of dead time in the motion mechanism, which leads to long data acquisition times and low experimental efficiency. By contrast, the on-the-fly mode allows for continuous motion while triggering the detectors and sensors, avoiding the dead time of the motion mechanism and greatly improving the efficiency of the experiments.PurposeThis study aims to propose and implement a hardware-triggered on-the-fly system to achieve rapid and continuous scanning during experimental processes for Hefei light source (HLS).MethodsThe proposed on-the-fly system mainly consisted of a synchronous signal acquisition module, a synchronous motion control module, and a software control module. The design of the hardware synchronous trigger module in the synchronous signal acquisition module was based on the field-programmable gate array (FPGA) development board Zynq7020. The encoder signals were decoded into position signals and used to trigger other devices based on position or time. This made it the central component for the on-the-fly mode. The design of the synchronous motion control module was based on a highly synchronous EtherCAT bus. The motion controller supported multi-axis coordination and programmable trajectory motion to meet the requirements of different experiments. The software control module utilized the experimental physics and industrial control system (EPICS) architecture for device control, and experiment flow control and data acquisition were implemented using Bluesky platform. Finally, the proposed on-the-fly system has been deployed and tested at the Soft X-ray Magnetic Circular Dichroism (XMCD) experimental station of HLS.ResultsThe experimental results demonstrate that the on-the-fly mode system not only meets the spectral performance requirements, but also reduces the single acquisition time from tens of minutes in the step-scanning mode to approximately 1 min.ConclusionsTherefore, the on-the-fly system designed and implemented in this study significantly improves the experimental efficiency and user experience.

BackgroundThe beam measurement group of Shanghai Synchrotron Radiation Facility (SSRF) has developed a new software package, HOTCAP, for high-speed oscilloscope-based three-dimensional bunch charge and position measurements to investigate the transient process of injection and beam instability in a high-energy electron storage ring. However, the software package does not specifically optimize the algorithm efficiency for data-processing speed.PurposeThis study aims to optimize real-time performance of the HOTCAP software so that the time required to complete the processing and analysis of single-measurement data fully satisfies the requirements of real-time measurements.MethodsAn operational efficiency test and algorithm optimization were conducted for each functional module of the HOTCAP software package to improve the overall performance. The specific time consumption data of each module in the processing flow were calculated, and the most time-consuming algorithm for extracting the three-dimensional position of charges was specially optimized to reduce duplicate calculations by using cached variables.ResultsAfter optimization, the processing time of the single-measurement data is reduced by more than 10 times.ConclusionsThe optimized HOTCAP software by this study satisfies the real-time monitoring and online data release requirements of the high-energy electron storage ring status.

BackgroundThe high-precision magnetoelectric velocity sensor can measure the vibration of the superconducting cavity in a low-temperature environment.PurposeThis study aims to quantitatively investigate the effects of temperature on the vibration characteristics of a superconducting cavity and provide recommendations for designing mechanical and cryogenic systems cryomodule.MethodsThe superconducting cavity in a 1.3 GHz cryomodule of the Shanghai HIgh repetitioN rate XFEL and Extreme light facility (SHINE) was taken as research object. The mechanical vibration in frequency range of 1~100 Hz, with a direction perpendicular to the beam direction was concerned, and six vibration sensors were arranged at two measuring points to monitor the velocity signals of different components. Subsequently, the displacement power spectral density, displacement root mean square, and frequency response function of superconducting cavity at the temperature of 300.0 K, 125.0 K, and 2.0 K were quantitatively analyzed using the spectrum analysis method.ResultsVibrations of the superconducting cavity caused by the flow at 2.0 K is 9.4% and 4.5% in the vertical and transverse directions, respectively, of that caused by ground source at the Shanghai Synchrotron Radiation Facility. In a cryogenic environment, the new vibration source is the cold flow, and the different fluid states have different effects on the vibration of the superconducting cavity in the vertical and lateral directions.ConclusionsThe study is valuable for guiding the tests and optimal design of cryomodules. The vibration of superconducting cavities at low temperatures can be measured using high-precision magnetoelectric velocity sensors. It is necessary to measure and analyze the potential source and its impact to satisfy the sub-micron beam stability requirements of superconducting linac and suppress the cavity frequency shift caused by mechanical vibration.

Generally, non-destructive testing methods based on the fast neutron technology, such as the associated alpha particle detection technology, can detect hidden hazardous materials without affecting the detected object. The effective neutrons are labeled by recoil alpha particles produced in a D-T reaction, which can significantly improve the signal-to-noise ratio of neutron detection. Meanwhile, the spatial information of the detected object can be obtained based on the spatial resolution of alpha particles. Therefore, neutron detection methods based on accompanying alpha particles have important application prospects in the field of security inspection. The principle and system composition of the neutron detection method based on associated alpha particles are briefly introduced herein. Then, key components of the detection system, such as the neutron tube, alpha particle detector, and gamma detector, are described. Next, the associated alpha particle neutron detection system, which is being studied worldwide, and its progress are presented. Finally, the prospects of this detection method are discussed.

BackgroundThe conversion of UF6, which is a primary nuclear product, to UF4 in fluoride molten salt phase is expected to be used in the preparation or reconstitution of nuclear fuel salt for molten salt reactors, thus simplifying the process of molten salt reactor fuel production. Determination of the concentration of the key intermediate UF3 plays an important role in obtaining the reaction parameters.PurposeThis study aims to establish a method for measuring UF3 concentration in solid fluoride molten salts.MethodsThe X-ray diffraction (XRD) was employed to test the homemade standards and obtain the internal standard curve of UF3. Firstly, the α-Al2O3 was taken as the internal standard to obtain the XRD peak height internal standard curve (R=0.986) and peak area internal standard curve of LiF-BeF2-UF3 molten salt. Then, these two internal standard curves were applied to measuring the known content of LiUF5 and UF3 solid mixed samples to compare their accuracies. Finally, measurements were conducted on rapidly cooled LiF-BeF2-UF3 solid molten salt samples and naturally cooled LiF-BeF2-UF3-LiUF5 solid molten salt samples to evaluate the stability and accuracy of the curve, and the relative error was obtained.ResultsIn the UF3 concentration range of 1.00~10.00 wt%, the correlation coefficient of the internal standard curve based on the peak area determined for of LiF-BeF2-UF3 molten salt is 0.995. Measuring results of solid mixed samples of LiUF5 and UF3 with known concentrations indicate that the peak area internal standard curve achieves better accuracy with a relative measurement error of no more than 8.7%. In addition, the results of the same content samples with different cooling methods confirm the good stability and accuracy of the proposed method with less than 5.4% relative standard deviation.ConclusionsThe established method can be used for the quantitative analysis of solid LiF-BeF2-UF3 and LiF-BeF2-UF3-LiUF5 molten salts with good measurement accuracy and repeatability.

BackgroundIn thermal pipelines of nuclear power systems, thermal stratification is a common phenomenon that can cause stress concentration and deformation of pipeline structures, thereby leading to safety hazards. A stagnant branch pipe is connected to the main coolant pipe, and a large temperature difference exists between the fluid in the pipe and the coolant in the main pipe of the primary circuit. Due to factors such as turbulent flow penetration and valve leakage, thermal stratification is prone to occur in the branch pipe.PurposeThis study aims to analyze the temperature change characteristics and flow characteristics of thermal stratification in stagnant branch pipes and provide a theoretical basis for subsequent experimental research and stress analysis.MethodsFirstly, a stagnant branch pipe model was established, and numerical simulation of thermal stratification phenomenon in stagnant branch pipes was conducted using FLUENT 2022 to analyze the temperature variation characteristics of the pipe wall and the distribution characteristics of the flow field inside the pipe. Then, the SST k-ω model was used to perform three-dimensional numerical simulation of the thermal stratification of stagnant branch pipes, with a leakage flow rate of 0.062 kg·s-1, leakage temperature of 488.15 K, and leakage pressure of 6 MPa.ResultsThermal stratification is prone to occur in horizontal pipe sections. Without insulation measures and a large pipe diameter, thermal stratification can be exacerbated, while the curved section can effectively reduce the temperature difference of the cross-section. A backflow phenomenon occurs in the horizontal section of the stagnant branch pipe, while the structure of the large and small end pipe sections causes secondary backflow in the flow field inside the pipe. The backflow phenomenon is not conducive to the mixing of cold and hot fluids in the pipe; consequently, the influence time of thermal stratification is longer.ConclusionsA significant difference in the thermal stratification phenomenon exists between the stagnant branch pipe and equal cross-section pipes.

BackgroundThe accuracy of transient thermal hydraulic parameter prediction of reactor cores under various working conditions directly affects reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which are often modeled as time-series prediction problems.PurposeThis study aims to solve the accuracy problem of continuous prediction of core thermal hydraulic parameters under instantaneous conditions and to test the feasibility of a gated cycle unit based on the attention mechanism in core parameter prediction.MethodsThe 1/2 full core model of China Experimental Fast Reactor (CEFR) core was taken as the research object. The subchannel SUBCHANFLOW program was employed to generate the time series of transient core thermal hydraulic parameters. The gated recurrent unit (GRU) model based on soft attention was used to predict the mass flow and temperature time series of the core.ResultsThe results show that, compared with the adaptive radial basis function (RBF) neural network, the GRU network model with soft attention offers better prediction results. The average relative error of temperature is <0.5% when the step size is 3, and the prediction effect is quite good within 15 s. The average relative error of mass flow rate is <5% when the step size is 10, and fairly good prediction effect is achieved in the subsequent 12 s.ConclusionsThe model constructed in this study not only exhibits higher prediction accuracy in the continuous prediction process but also captures the trend characteristics in the dynamic time series, which is of considerable value for maintaining reactor safety and effectively preventing nuclear power plant accidents. The GRU model based on soft attention can provide continuous prediction for a period of time under transient reactor conditions, providing a reference value in engineering applications and improving reactor safety.

BackgroundA heat pipe reactor is ideal for underwater unmanned vehicles (UUV) because it is simple, is compact, and has high inherent safety.PurposeA passive residual heat removal system that uses natural circulation to cool the adiabatic section of heat pipes was designed based on the characteristics of a new type of megawatt compact nuclear power plant with a heat pipe reactor.MethodsFirstly, based on the characteristics of 3.5 megawatt compact nuclear power plant for UUV, natural circulation of water was utilized to cool the adiabatic section of heat pipes. Then, the computational fluid dynamics software STAR-CCM+ was used to simulate and analyze the heat removal capacity of the passive residual heat removal system with different geometric parameters, made it conservatively meeting the demand of maximum residual heat removal power.Results & ConclusionsThe results show that a baffle around the adiabatic section of heat pipe bundle is beneficial to reduce the maximum temperature of the fluid. The widths of the inlet and outlet of the baffle have almost no effect on the heat removal capacity, while extending the lower part of the baffle is unfavorable to natural circulation. When the axial length of the emergency cooling chamber is 160 mm, it can conservatively meet the maximum residual heat power of 0.14 MW. The maximum fluid temperature is 288 ℃, which is lower than the boiling point under working pressure, and normal operation is possible in ambient temperatures ranging from 5 ℃ to 25 ℃.

BackgroundThe physical quantity concerned in the clinical application of 125I particle source brachytherapy is the 1 cm water absorbed dose rate D˙w,1 cm. However, there is no corresponding standard for this physical quantity in China. It is planned to develop an ionization chamber for the absolute measurement of the water absorbed dose rate of the 125I particle source as a standard device, hence the internal electric field of the ionization chamber must first be analyzed to obtain the most reasonable design scheme to satisfy the condition that the electric field intensity is uniformly distributed in the ionization chamber.PurposeThis study aims to simulate the internal electric field of the ionization chamber for the design of particle source water absorbed dose absolute measurement device.MethodsFirstly, the internal model of the ionization chamber was established by Maxwell software to simulate the distribution of electric field intensity under the six variables of the ionization chamber: with or without a protective electrode, different protective electrode ring width, different insulation ring width, different grid number, different grid shape and different grid thickness. Then, qualitative and quantitative analysis were carried out using finite element method. Finally, the influence of different variables on the distribution of electric field intensity in the ionization chamber of the absolute measurement of water absorbed dose of 125I particle source was obtained.ResultsAnalysis results show the ratio of the width of the guard electrode to the radius of the collector electrode must be not less than 2 for design of the ionization chamber. The edge effect at the edge of the collector is increased with width increment of the insulator ring, hence the width of the insulator ring should be reduced as much as possible. When the number of grid electrodes is 15, the variation of the electric field intensity can be reduced to about 1%. When the grid section is rectangular, the variation in electric field intensity is small compared with the circular and triangular grid sections. The larger the thickness of the grid, the more severe the edge effect at the edge of the grid, and the thickness of the grid should be reduced as much as possible.ConclusionsThe uniformity of the electric field can be effectively improved by increasing the number of grid electrodes. Results of this study is helpful to optimal design of the standard device for measuring the absolute water absorbed dose of 125I particle source.

BackgroundGaussian signals have symmetry and completeness, hence, the Gaussian filtering method is widely used in nuclear signal processing and radiation energy spectrum analysis. The mathematical description of Gaussian signals is relatively complex, which makes it difficult to construct digital Gaussian pulse shaping filters for nuclear pulse signal real-time processing. The commonly used digital quasi-Gaussian shaping algorithm in radiation measurement systems is derived from the differential equations of Sallen-Key and CR-(RC)n filters in analog nuclear electronics. However, its output shaping pulse signals have poor symmetry and problems such as undershoot occur when used in a single stage.PurposeThis study aims to explore the convolutional quasi-Gaussian pulse shaping filter algorithm and apply it to the processing of X-ray fluorescence measurement system experimental platform to obtain measured nuclear pulse data.MethodsFirstly, a convolutional quasi-Gaussian pulse shaping algorithm was implemented based on trapezoidal pulse signals, and bipolar pulse shaping was achieved after the initial convolution. Then, the second convolution was accumulated and summed to obtain left and right symmetric quasi-Gaussian shaping pulses. The digital recursive formula of the quasi-Gaussian pulse shaping algorithm was obtained using the Z-transform method, and the effectiveness of the algorithm and the influence of shaping parameters on the amplitude frequency characteristics were studied through simulation. Finally, quasi-gaussian shaping algorithm and trapezoidal shaping algorithm were applied separately to the offline process of measured nuclear pulse data from the X-ray fluorescence measurement system experimental platform.ResultsThe quasi-Gaussian pulse shaping filter has better high-frequency noise suppression performance compared to trapezoidal pulse shaping filters. With the increase of the values of shaping parameters na and nc, the filter passband decreases whilst the low-frequency amplitude relatively increases, and the high-frequency noise suppression effect is enhanced. However, this also leads to shaping pulse broadening and increases the probability of pulse pile-up.ConclusionsThe experimental results demonstrate that the quasi-Gaussian pulse shaping algorithm has better pile-up pulse separation ability. Under identical X-ray tube voltage and current conditions and peaking time, the energy resolution of the energy spectra obtained by both algorithms is fundamentally equivalent. However, the energy spectrum with quasi-Gaussian pulse shaping has a higher characteristic peak area.

BackgroundWith the rapid development of nuclear energy and the wide application of nuclear technology, the radioactivity level of bodies of water has become a highly concerning issue for the public and governments, especially after the Fukushima nuclear accident in Japan.PurposeThis study aims to develop an online γ radioactivitymonitoring system based on 4G Remote Terminal Unit (4G-RTU) to meet the needs of online and emergency monitoring of water radioactivity.MethodsFirst, the γ-ray monitoring device based on sodium iodide detector, 4G-RTU, an integrated power supply waterproof and compression resistant floating device and corresponding software were employed to compose an online monitoring. Second, Qt programming control software was used to realize the remote control of the system, real-time radioactivity monitoring, and data upload. Finally, the original data obtained by the system were used to test the performance indicators, applicability, accuracy, and software functions to verify the practicability of the system.ResultsWithin the coverage of the 4G network, the system realizes remote control of equipment, real-time online monitoring, and data upload throughout the day. The detectable energy range of the system is 30~3 000 keV, and the energy resolution of the system for 137Cs at 662 keV is 7.3% with minimum detectable activity of 0.75 Bq?L-1. The spectral drift for 208Tl at 2 614 keV is 0.33%, and the linearity of the spectral energy is 0.999 970. The maximum value of energy spectrum stability is 2.28% for 7 h continuous operating, and the minimum value is -2.36%. The operating temperature range of the system is in the range of -5 ℃ to +50 ℃.ConclusionsThe on-line monitoring system meets the application demand and achieves the expected function. It has important popularization value and application prospects in the field of online radioactive monitoring of bodies of water such as oceans, lakes, and rivers.

BackgroundThe unfolding-synthesis technique is commonly used in digital pulse processing systems for nuclear radiation measurements.PurposeThis study aims to propose a novel flat-topped widened peak-shaping algorithm based on the pulse unfolding-synthesis technique is proposed.MethodsFirstly, the repetition and polynomials were utilizes to shape the digital pulses, and the nuclear pulse signal was unfolded into unit pulses series. Then, an impulse response system was employed to synthesize these unit pulses series to achieve the desired peak shape. Finally, the improved flat-topped peak-shaping algorithm was compared and analyzed against traditional filter shaping algorithms in terms of accuracy in amplitude extraction, interference resistance, and pile-up recognition.ResultsExperimental results demonstrate that, under the same shaping time, the energy resolution of the flat-topped peak-shaping algorithm for the γ characteristic peak of 137Cs is 7.2%, outperforming the trapezoidal, triangular, and Gaussian shaping algorithms. Additionally, it exhibits high counting rate performance.ConclusionsThe flat-topped peak-shaping algorithm of this study can effectively replace traditional pulse-shaping methods and be utilized for high-precision, high-count-rate γ spectroscopy measurements.

BackgroundThe separation and removal of fission products and the recovery of carrier molten salt during nuclear fuel reprocessing in molten salt reactors reduces waste and facilitates useful substance recycling.PurposeThis study aims to separate and remove the rare earth (RE) fission products and recovery of carrier molten salts.MethodsFirstly, high temperature precipitation reactions of rare earth (RE = Ce, Nd, Sm, Eu, Y, Yb) and thorium fluorides were studied in LiF-BeF2 melt using sodium sulfide hydate (Na2S·5H2O) as the precipitant. Their removal ratios were subsequently compared under different conditions. Then, a combined precipitation distillation method was employed to further heat the precipitated mixed salt to 950 °C and distilled under vacuum conditions at a pressure of 10 Pa for 20 min. The content and removal ratio of rare earth Nd, as well as contents of oxygen and sulfur, in condensed collected salts were investigated after using above improved precipitation-distillation processing. Finally, further analysis of the composition of sediment was conducted using X-ray Diffraction (XRD), X-ray Photoelectron Spectroscopy (XPS) and Energy Dispersive Spectrometer (EDS).ResultsThe results demonstrate that RE removal ratios are less than 90% when the RE to precipitant ratio is 1:2 at 600 ℃ whilst the RE Nd content in the salt collected by condensation is reduced to 1.39×10-4 g?g-1, and its removal ratio is increased to 99.6% with further improved precipitation-distillation processing. Simultaneously, the oxygen and sulfur contents are 8.5×10-5 g?g-1 and 1.50×10-4 g?g-1, respectively. Analysis results of XRD, XPS and EDS indicate that the sediment mainly consists of RE sulfide and sulfur oxide.ConclusionsThis study confirms the feasibility of separating RE from waste salt using the sulfide precipitation method and that over 99% RE separation efficiency can be achieved using precipitation-distillation combined treatment. Therefore, this provides a reference method for purifying waste salt and realizing molten salt reuse.

BackgroundIn thorium-based molten salt reactors (TMSRs), 233Pa is an important intermediate nuclide in the conversion chain of 232Th to 233U, and 233PaF5 can be effectively separated from carrier salt by low-pressure distillation. However, some of the metal fluoride is vaporized along with 233PaF5 simultaneously, and the evaporated fluoride may condense at different temperatures.PurposeThis study aims to investigate the condensation behavior of 233PaF5 and other key metal fluorides in the FLiBeZr molten salt during the low-pressure distillation process.MethodThe FLiBeZr molten salt containing PaF5 and various metal fluorides was low-pressure distillated to examine the condensation behavior of 233PaF5 and key metal fluorides at different temperatures. Then, the optimal temperature of key metal fluorides was evaluated and the separation characteristics between them and 233PaF5 were figured out. Finally, the separation factors of 233PaF5 and the other metal fluorides were calculated and compared with those of the entire low-pressure distillation process at the optimum condensation temperature of 233PaF5.ResultsThe results show that the optimal condensation temperature for 233PaF5 and 95NbF5 is within the range of 600~700 ℃, whilst it is within the range of 400~500 ℃ for 237UF4 and 95ZrF4. The comparison results show that there is no significant difference in the separation factors of 95NbF5 and 233PaF5, but the separation factors of 233UF4 and 95ZrF4 are increased by a factor of two to 20 times.ConclusionIt is confirmed that the separation coefficient is determined primarily by volatilization, but can be further improved by varying the condensation temperature.

BackgroundLead-free/lead-composite nuclear radiation protection materials are becoming increasingly prevalent in military applications to safeguard the health and safety of military personnel. Currently, accurate measurement and tracing of the shielding performance of military nuclear radiation protective clothing is difficult.PurposeThis study aims to measure lead equivalent in nuclear radiation protection materials for satisfying the radiation performance evaluation requirements of international mainstream military radiation protective clothing.MethodsFirstly, a device to measure lead equivalence in protective clothing was developed. Then, the reference radiation mass for lead equivalence at a photon energy of 130 keV was determined using a combination of Monte Carlo simulation and experimental measurements. Finally, the shielding performances of a standard lead sheet were assessed using developed device under narrow beam conditions, and shielding performances of military nuclear radiation protective clothing materials produced by a few manufacturers were tested in the same conditions.ResultsEvaluation results show the relative expanded uncertainty of lead equivalent measurement results for standard lead sheet and military nuclear radiation protective clothing samples correspond to 4.2% (k=2).ConclusionsThis study identifies measurement conditions for subsequent performance tests of additional military nuclear radiation protective materials prior to delivery.

BackgroundCultural relics are precious and nonrenewable resources. Scientific dating is the key to research on cultural relics, and luminescence technology is an important method for dating ceramic cultural relics. Currently, the methods for dating ceramic cultural relics include the conventional thermoluminescence (TL) method and the TL predose method. Few reports on the use of optically stimulated luminescence (OSL) to determine the age of cultural relics are available.PurposeThis study aims to verify the reliability of the dating of porcelain cultural relics by luminescence method, and expand the method of dating porcelain cultural relics.MethodsThe conventional TL, TL predose, and OSL single-aliquot regenerative-dose (SAR) methods were employed to date an celadon glazed porcelain piece unearthed from the Qinglong Town ruins in Shanghai. Then, the dating results of three method were compared for applicability analysis.ResultsThe results indicate that the signal for the porcelain obtained using the conventional TL method is almost zero when the temperature is between 300 ℃ and 450 ℃, making it difficult to accurately calculate the equivalent dose. The ages of the TL predose and SAR methods are (1.16±0.05) ka and (1.35±0.05) ka, respectively, which are consistent within the error range.ConclusionsThe results of this study indicate that the OSL technique can be used to date porcelain cultural relics. For the celadon glazed porcelain piece examined in this study, the dating results of TL pre-dose and SAR methods are kept consistent.

BackgroundLuminescence dating technology has made significant advancements in determining the chronology of archaeological materials subjected to low firing temperatures. However, the luminescence dating of archaeological materials subjected to high firing temperatures remains challenging.PurposeThis study aims to explore the luminescence emission spectrum characteristics and luminescence properties of high-firing temperature quartz to verify the feasibility of thermoluminescence (TL) signals from different bands in luminescence dating.MethodsFirstly, the high-firing temperature (about 950 °C) quartz extracted from pottery unearthed at the Lingjiatan archaeological site was taken as a case study, spectral measurement platform was established using a Ris? DA-20 luminescence dating instrument coupled with an Andor spectrometer and a charge-coupled device camera to analyze the luminescence spectral properties of archaeological quartz with high firing temperatures. Then, five filter combinations and two photomultiplier tubes (PMTs) were used to compare the TL and isothermal thermoluminescence (ITL) sensitivities of blue and red emissions. Kinetic parameters for Blue TL and Red TL were determined by deconvolving the glow curves with the general-order equation. Finally, exposure experiments were conducted on the Blue and Red TL using a solar simulator. The single aliquot regenerative dose (SAR) protocol was implemented to assess the applicability of the Blue TL-SAR, Blue ITL-SAR, Red TL-SAR, Red ITL-SAR, and optically stimulated luminescence (OSL)-SAR methods for dating archaeological quartz exposed to high temperatures during production or use.ConclusionsThe spectral analysis reveals that the archaeological quartz subjected to high firing temperature exhibits significant Red TL emissions at approximately 620 nm, which is correlated with the TL peak at 375 °C. This Red TL at 375 °C exhibits a marked insensitivity to light. The multi-wavelength TL, multiwavelength ITL, and conventional OSL dating results are consistent with the known radiocarbon age within the error range. This study demonstrates the potential feasibility of using luminescence signals of different wavelengths for chronological studies of archaeological materials subjected to high firing temperatures.

BackgroundRock surface luminescence dating has developed rapidly in the past decade. It has been widely used to obtain exposure ages and erosion rates of various rocks in archaeology, geology, and geography, such as stone artifacts, glacier bedrock, gravel, and bedrock fault surfaces. However, these is little study on the effect of parameters related to this method, such as attenuation coefficient μ, decay rate of the trapped charge at the rock surface σφ0ˉ, environmental dose rate D˙, and characteristic saturation dose D0 on exposure age and erosion rate results.PurposeThis study aims to quantitatively investigate the impact of relevant parameters on exposure age and erosion rate, and examine the limits of exposure age and erosion rate obtained by the method under different parameters.MethodsFirst, parameters that might have an impact on age and erosion rate results were determined through theoretical analyses. Then, the relationship among the parameters, depth of half saturation and exposure age (erosion rate) were studied using numerical simulations. The impact of each relevant parameter on exposure age and erosion rate was observed for varied parameter values. Finally, the dating limit was determined from the inflection point in the simulated profiles.ResultsThe smaller μ value is, the greater the rate of change in the depth of half saturation would be when increasing the same exposure time. However, the rate of change in the depth of half saturation remains constant for different σφ0ˉ values when the same erosion rate is increased.ConclusionsAmong the parameters, σφ0ˉ and, more remarkably, μ significantly affect the dating result. In general, D˙ and D0 have little effect on the exposure age and erosion rate; therefore, the differences in D˙ between the surface and interior of rocks may be ignored. The growth rate of depth of half saturation of granite gneiss is significantly higher than that of sandstone for the same exposure time increment. Therefore, light-colored rocks such as granite should be prioritized for collection during field work. The ranges of dating and obtaining erosion rate using this method are 10-3~102 ka and 10-2~103 mm?ka-1, respectively.

BackgroundGamma source irradiation is the most commonly used way at present for the equivalent dose (DE) determination of fossil samples in electron spin resonance (ESR) dating. However, it is facility-limited and time-consuming in many cases.PurposeThis study aims to establish the standardised growth curves (SGCs) of fossil enamel samples for determining the DE by ESR without gamma irradiation.MethodsFirst of all, we analyzed 20 tooth samples from the Late Pleistocene sites, and they exhibited similar dose response characteristics. Then, based on our preliminary work, we attempted to establish the SGCs of these Late Pleistocene fossil teeth using three different methods: (1) a simple method (fitting the natural dose points of fossil samples from Middle to Upper Pleistocene sites with exponential functions), (2) an average method (fitting the dose points with averaged ESR signal intensities), and (3) a representative sample method (establishing a SGC by using a representative sample). Finally, dose values obtained by each method were compared with those determined by the additive dose method (ADM).ResultsThe results of DE determined by the simple and average methods are close, with a deviation of less than 32% from the ADM results. The dose values obtained by SGC using the representative sample method generally agree with those of the ADM, with a deviation within 26%, which is the smallest among the three methods.ConclusionAlthough the uncertainties of the dose values obtained for the SGCs are not very close to those obtained using the ADM, it indicates the potential to quickly determine a more reasonable Dmax for irradiation, identify the possible intrusion of fossil samples, and analyze small or precious fossil samples.

BackgroundThe neutronics and thermal-hydraulic characteristics of lead-bismuth cooled reactors are significantly affected by the geometric configuration of fuel assembly and lattice parameters. Reactor cores loaded with different geometry type fuel assemblies have different critical dimensions and fuel loadings under the same refueling cycle and thermal-hydraulic constraints.PurposeThis study aims to analyze these key factors and select a geometric structure of fuel assembly that is conducive to miniaturization and lightweight of lead-bismuth reactor.MethodsFirst of all, the core model of a 4 MWt small lead-bismuth reactor was established, and simulation analysis of reactor physical characteristics was conducted using the RMC Monte Carlo program developed independently by the Reactor Engineering Calculation and Analysis Laboratory of Tsinghua University and the nuclear database ADS-2.0 released by the International Atomic Energy Agency (IAEA) in 2008. Then, three fuel assembly schemes of rod bundle type, annular type and honeycomb coal type were selected for comparison and analysis in term of fuel consumption characteristics, reactivity coefficient and steady-state thermal parameters under the same core size, fuel loading, coolant flow area, cladding and air gap volume, 10-year refueling cycle and basically consistent steady-state thermal safety margin.Results & ConclusionsThe results show that compared with the rod bundle fuel assembly and the annular fuel assembly, the honeycomb coal fuel assembly has good steady-state thermal characteristics and hard neutron spectrum. The core of the honeycomb coal fuel assembly can realize smaller core size and fuel loading, and has obvious expansion negative feedback, and can effectively flatten the power distribution and reduce the core pressure drop. It is a fuel assembly solution that is conducive to the miniaturization and light weight of lead-bismuth reactors.

BackgroundThe medical isotope production aqueous reactor (MIPR) has advantages of small size, low power, and high inherent safety, hence is one of better candidate reactor types for the production of 99Mo and other medical isotopes.PurposeThis study aims at the effects of extraction methods and reprocessing capacity on the production efficiency of 99Mo based on low-enriched uranium MIPR designed with neutronic optimization.MethodsFirst, the calculation method was verified according to existing experimental data, and the neutronic optimization of the MIPR was performed for core design by using SCALE6.1 code and ENDF/B-VII database with 238 groups. Then, the 99Mo production efficiency under different extraction methods as well as processing capacities was investigated based on the optimized core structure. The range of achievable critical uranium concentration and enrichment was determined.ResultsThere is a minimum critical mass under different enrichment, and with the increase of 235U enrichment, the uranium concentration at the minimum critical uranium mass decreases. The effective multiplication factor decreases linearly with an increase in nitric acid concentration, and the corresponding nitric acid reactivity coefficient is approximately -1.400×10-2 L·mol-1. With an increase in uranium concentration, the void and temperature reactivity coefficients decrease, and the corresponding reactivity coefficients are approximately (-100~-250)×10-3 ℃-1 and (-18~-30)×10-5 ℃-1.ConclusionThe production efficiency of the MIPR production of the 99Mo online extraction method is slightly greater than that of the offline batch processing method with an increment of about 16% under a five-day production cycle. The reprocessing capability has a greater impact on the production efficiency of the online extraction method. If the reprocessing rate is increased five times, the increment of production efficiency is about 113% under a five-day production cycle.

BackgroundBecause of the excellent properties of lead-based materials as reactor coolants, lead-based fast reactors have become a key type of fourth-generation advanced nuclear energy systems. A small passive long-life Lead–bismuth -cooled fast Reactor (SPALLER) is designed by the University of South China for profound research.PurposeThis study aims to improve the inherent safety and cost-effectiveness of lead–bismuth-cooled fast reactors, and determine the maximum core power of this kind of reactor.MethodsFirstly, the SPALLER was taken as research object, and five steady-state limitations and two accident limitations were proposed to meet the transportation size, material durability, and long-term operational stability of the reactor core and ensure safety under accident conditions. Then, a neutronics maximum power calculation platform was built through Latin hypercube sampling and a Kriging proxy model whilst the steady-state limitations were considered as multi-objective optimization problems with complex multidimensional nonlinear constraints. Meanwhile, the neutronics maximum power and natural circulation power of SPALLER-100 at different core heights were calculated by taking the natural circulation ability of SPALLER-100 into account. Finally, a design scheme was obtained to meet the requirements of neutronic and thermal-hydraulic assessments of this reactor while producing maximum power. Consequently, during the full life-cycle of SPALLER-100, a safety analysis of three typical accident scenarios (loss of heat sink, transient over power, and coolant inlet temperature undercooling) was performed using a quasi-static reactivity balance approach.ResultsThe results show that the maximum core power can be increased from 100 MW to 120 MW, and the neutronics maximum power calculation platform has high accuracy with safe and economical maximum power scheme.ConclusionsThis study can provide reference for other types of natural circulation reactors to maximize power output.

BackgroundSilicon carbide (SiC) composite claddings are candidate solutions for accident resistant fuel claddings in light water reactors.PurposeThis study aims to estimate the failure probability of a double-layer structured SiC cladding under a loss-of-coolant accident (LOCA).MethodsBased on a failure probability calculation method for SiC composite cladding, a quasi-steady state method was used to simulate and calculate the SiC composite cladding failure probability under transient conditions. Sensitivity analysis of the two characteristic parameters of Weibull distribution was performed by analyzing the proportion of various stresses under accident conditions. The effects of different burn-up conditions on the failure probability were investigated, and the failure probability of the cladding under different layer thickness ratios was simulated.Results & ConclusionsSimulation results indicate that the transient failure probability of SiC composite claddings is significantly affected by changes in the proportion of the composite layer and Weibull parameter, as well as the occurrence of LOCAs under different burn-up conditions. This study makes contribution to the development and design of accident resistant fuel claddings, providing reference for further investigations on the failure probability of SiC composite claddings.

BackgroundDuring reactor operation, zirconium (Zr) alloy cladding is continuously oxidized as it gets in contact with fuel, and combines with the fuel to form a firm chemical interaction layer. This affects the thermal conductivity of the fuel gap, the mechanical properties of the cladding, and the mechanical interaction of the fuel cladding.PurposeThis work aims to obtain relevant analytical data on the chemical interaction layer between the irradiation Zr-alloy cladding and uranium oxide (UO2) pellets in a pressurized water reactor (PWR).MethodsFirst of all, the D13 intact fuel rod with a burnup of 45 GWd·tU-1 for PWRs in a nuclear power plant was chosen as UO2 pellets with a pellet enrichment of 4.45 wt%, and M5 Zr-alloy was used as the cladding materials. Then, a series of operations (cutting, pellet separation, inlaying, secondary cutting, inlaying, and polishing of cladding tubes) was conducted in the hot cell. The polished sample was transferred to the lead chamber and the UO2 fuel pellet was removed using 4 mol?L-1 of nitric acid solution. The cladding tube was separated from the chemical interaction layer, and a low-speed cutter was used to cut the cladding tube to a width of 2~3 mm in the glove box. Finally, the morphology and structure of the chemical interaction layer were analyzed using metallographic microscopy, scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), and hot cell Raman spectroscopy.ResultsThe analysis results show that the gap between the fuel pellet and cladding is approximately 14~19 μm after the fuel running to a burnup of 45 GWd·tU-1. In the chemical interaction layer, the time sequence of mechanical contact at different locations is different, resulting in discontinuity of the interaction layer. The SEM-EDS results show that the chemical interaction layer is in the shape of "worms" composed of U, Zr, and O to form a mixed phase (U,Zr)Ox compound.ConclusionsThe results of this study indicate that the chemical interaction layer is mainly composed of tetragonal zirconia (t-ZrO2) and monoclinic zirconia (m-ZrO2).

BackgroundThe annular fuel has a closely arranged structure, and the coolant flow at both the gap between the stringers and the near wall surface is small, which is unfavorable to the coolant mixing between the subchannels and the uniform circumferential temperature distribution.PurposeThis study aims to explore the effect of the ratio of gate spacing to gate diameter on the distribution of temperature along the circumference direction.MethodsBased on the software code ANSYS FLUENT, a computational fluid dynamics (CFD) analysis model for annular fuel assemblies was established. Then, the calculations in hydromechanics and the numerical simulation using operating parameters of typical pressurized water reactor (PWR) were performed to analyze the coolant flow and heat transfer characteristic when the annular fuels in square or hexagonal arrangement under different grid ratios. The circumferential non-uniformity of annular fuel outer temperature distribution was investigated under circumstances of various pitch-to-diameter ratio.ResultsComputational results show that an appropriate increase of grid ratio is beneficial to the uniform circumferential temperature distribution of stringers. The appropriate grid ratio of square component is between 1.07 and 1.09, and the non-uniformity of circumferential temperature distribution of triangle component is slightly lower than that of square component. Therefore, the appropriate grid ratio is between 1.06 and 1.09.ConclusionsThe temperature distribution at the bar gap is improved most obviously by increasing the grid ratio and the improvement in the near surface takes the second. The results of this study provide a reference for the subsequent optimization design of the grid ratio of annular fuel.

BackgroundThe calculation error of the stacked pulse amplitude generated by traditional pulse shaping methods leads to distortion in the X-ray fluorescence spectrum; thus, it is difficult to accurately analyze the spectrum measured in a high-stacking rate background.PurposeThis study aims to propose a transformer model based on deep learning for the pulse amplitude estimation of radiation measurements using high-performance silicon drift detectors.MethodsFirstly, multi-head attention was applied to the transformer model, and an encoder-decoder structure with embedded positional encoding was employed to estimate the amplitude of stacked pulses. Then, a predefined mathematical model was used to simulate the pulse signal output by the detector for model training, and Gaussian noise corresponding to thermal noise and shot noise was added to the signal to simulate real nuclear pulses. Finally, experimental verifications were carried out on powdered iron ore samples and powdered rock samples, and relative error, corresponding to the accuracy of pulse amplitude estimation, was used as a model performance evaluation indicator.Results & ConclusionsExperimental verification results show that the average relative error obtained for eight offline pulse sequences of powdered iron ore samples and powdered rock samples is 0.89%, which means that the model can accurately estimate the amplitude of stacked pulses.

BackgroundDue to the complex structure of the ventilation ducts in nuclear facilities, the concentration distribution of radionuclides such as aerosols in the ducts is uneven. The inhomogeneity of aerosol distribution brings great challenges to the sampling representativeness of radiation monitoring. In chemical processes, static stirring devices are commonly used to enhance the homogeneity of the product mix. However, this device has not been applied in the nuclear power field.PurposeAccordingly, this study aims to improve the mixing uniformity of aerosols in air supply pipelines by using static stirring devices, so as to provide a reference for the representativeness of radiation monitoring sampling.MethodsThe stirring effects of three different stirring devices were investigated through numerical simulations. The RNG k-ε model was used to simulate the gas phase flow field, and the discrete phase model (DPM) was employed in ANSYS CFX software to simulate the behavior of aerosol particles. Selection of the particle size of aerosols followed the recommendations in the sampling representative standards, with the specific size of 10 μm. The other boundary conditions in the simulation were based on the actual operating conditions of a nuclear power plant. As a result, the effects of different stirring devices on the flow field and aerosol concentration distribution were obtained.ResultsThe static stirring device can form strong swirls, thereby improving the uniformity of aerosol distribution. Increasing the twist angle of the blades and the proportion area of the inner blades strengthened the generated vortex field, further affecting the diffusion of aerosols. The static stirring device with an increased inner blade area exhibited a moderate swirl intensity, and better stirring effect than those of the other two structures. The coefficient of variation of the aerosol concentration decreased by 30.60%.ConclusionsInstalling a static stirring device in an air supply duct is a feasible method to improve the mixing uniformity of aerosols. Owing to the complex structure of ventilation ducts in nuclear facilities, the concentration distribution of radionuclides, such as aerosols, in the ducts is uneven. This inhomogeneous aerosol distribution poses significant challenges to the sampling representativeness of radiation monitoring. In chemical processes, static stirring devices are commonly used to enhance the homogeneity of the product mix. However, these devices have not yet been applied in the field of nuclear power.

BackgroundNuclear high-temperature resistant ceramic materials have been widely used in nuclear energy, military, and aerospace fields in recent years owing to their excellent thermal insulation and high-temperature oxidation resistance.PurposeThis study aims to investigate the mass and heat transfer process between plasma fluid and flying particles in supersonic plasma spraying during the preparation of yttrium-stabilized zirconia thermal barrier coatings, so as to reveal the process parameters of flying particles.MethodsFirstly, the computational fluid dynamics (CFD) approach was employed to simulate the interaction between flying particles in the plasma spraying process. Then, a three-dimensional mathematical model of the plasma spraying flow field was established, and the jet characteristics of different spraying parameters in the de Laval nozzle and the melting and stress state of flying particles were analyzed by using this model. Furthermore, the online monitoring device Spray Watch 2i (Osier, Finland) was used to compare the online measurement of the velocity and temperature of flying particles obtained with the simulation results.ResultsThe comparison results show that relative errors are within 15%, verifying the simulation results effectively by experimental results. When the spraying power is reduced from 71 kW to 36 kW (i.e., reduced by 49.2%), the maximum velocity of the plasma jet is reduced by 8.5%, and the maximum temperature is reduced by 22.2%.ConclusionsA correlation between plasma spraying parameters, jet characteristics, and melting of flying particles is revealed in this study, providing theoretical guidance for the precise control of high-performance thermal insulation coating structures required for accident resistant fuel cladding in nuclear reactions.

BackgroundLow-level radioactive wastewater (LLW) is generated during the operation of nuclear facilities. Usually, LLW is discharged directly into the ocean in the form of liquid effluent after purification under the discharge management limits. However, discharging LLW into inland water bodies is difficult for inland nuclear facilities because of the poor dispersion and the lack of public acceptance. Thus, LLW disposal has become one of the challenges limiting the development of inland nuclear facilities. Liquid-to-gas discharge, which is based on high-pressure spray evaporation technology, is an alternative solution for LLW disposal for inland nuclear facilities.PurposeThis study aims to develop and validate a model for simulating spray flow and evaporation to assess the design feasibility.MethodsA numerical method coupling a two-phase flow model, mass transfer model, and heat transfer model were established to describe droplet evaporation during the flow process. To validate this numerical method, a sample high-pressure spray evaporation system was developed which included three key subsystems: carrier gas generation, source term generation, and measurement systems. Finally, considering evaporation and deposition factors, three experimental cases were designed for experimental comparison of the droplet diameter, number, and deposition rate among these cases.ResultsThe comparison results show that the numerical method is highly consistent with the experimental results, with a maximum uncertainty of 15%.ConclusionsThe numerical model developed in this study can be used for the technological design of liquid-to-gas LLW discharge based on high-pressure spray evaporation technology.

BackgroundAccelerator-driven subcritical systems (ADS) are among the most promising options for next-generation nuclear power systems. Various radionuclides are produced during the process of protons bombarding the target in the ADS, and the cross-sections of various long-lived radionuclides have not been accurately measured. These long-lived nuclides are related to ADS radioactive waste treatment, therefore, accurate evaluation of long-lived radionuclides generated in ADS spallation targets is a key topic in applied research.PurposeThis study aims to determinate production cross-sections of natPb(p,x)207Bi and natPb(p,x)194Hg reactions according to measurement data, and compare them with existing experimental and theoretical results.MethodsThe proton activation method was employed to effectively estimate the cross section of long-lived nuclides produced by the interaction between protons and spallation target materials. Four proton-irradiated natural lead samples were irradiated with protons at energies of 40 MeV, 70 MeV, 100 MeV, and 400 MeV for 90 min, 75 min, 40 min, and 25 min, respectively. After cooling for approximately 20 a, the samples were measured using an ultralow background gamma spectrometer GeTHU in the China Jinping Underground Laboratory (CJPL), and the GeTHU detection efficiency was calculated using the Simulation and Analysis for Germanium Experiments (SAGE) simulation framework. Combined with the irradiation parameters of the samples, the total production cross-sections of the two nuclides were calculated using a cross-section calculation formula. Experimental results are evaluated and compared with those of existing studies.ResultsThe production cross-sections of natPb(p,x)207Bi reaction in the natural lead samples irradiated by protons with four different energies (40 MeV, 70 MeV, 100 MeV and 400 MeV) are calculated as (40.70±3.59) mb, (19.31±1.43) mb, (13.15±0.96) mb, and (2.90±0.22) mb, respectively. The calculated production cross-section of natPb(p,x)194Hg reaction in the natural lead sample irradiated by protons with an energy of 400 MeV is (57.07±7.83) mb. Based on the same samples, the measurement results of cross-section remain consistent within the error range. The cross-sections of natPb(p, x)207Bi are closer to TENDL's evaluated cross-sections. The cross-section of natPb(p, x)194Hg is consistent with the theoretical expectation of INCL++/ABLA. In addition, different sources that contribute to the total uncertainty of both reactions are explained in detail.ConclusionsThe production cross-sections of natPb(p,x)207Bi and natPb(p,x)194Hg reactions measured herein were calculated independently and showed good agreement with existing results. These results demonstrate that GeTHU is capable of measuring low-activity and long-lived radionuclides in the CJPL. Finally, the results of this study also provide the latest experimental evidence for the evaluation of radioactive waste in ADS.

BackgroundNuclear science and technology are closely related to the lives of people. However, nuclear radiation may harm the health of the general public; hence, nuclear radiation monitoring must be strengthened. A wired nuclear radiation monitoring system has the characteristics of complex wiring, a long construction period, high cost, poor mobility, and more difficult troubleshooting.PurposeThis study aims to address the demand for convenient measurement and monitoring of gamma radiation fields.MethodsBased on the LoRa wireless communication technology, a γ radiation monitoring system using silicon photomultiplier (SiPM) tube-scintillator detector was designed. The main functions of the system included data collection from the detector, data processing and transmission using a STM32 single-core processor. The collected data packaging and transmission were processed in STM32 microcontroller using the LoRa wireless communication module. Considering the possible channel congestion in the monitoring system and the frequent data transmission, a dynamic optimal path communication algorithm was designed to find the optimal reconnection path and realize the priority allocation of data transmission.Results & ConclusionsThe test results show that the data transmission stability of γ radiation monitoring system based on LoRa is higher than 99.57%, and has the advantages of flexible networking, a further distance of transmission, low cost, and substantial expansion, hence has a broad reference prospect.

BackgroundThe solar X-ray detector (SXD) is the main scientific instrument onboard the Macau Science Satellite-1B (MSS-1B). It consists of two parts—a soft X-ray detection unit and a hard X-ray detection unit—with a dual-channel design comprising a silicon drift detector (SDD) and a cadmium zinc telluride detector (CZT). Both the precise energy spectrum and intensity of the Sun can be simultaneously obtained by the SXD, hence to quantify the level of solar flares and study their evolutionary process.PurposeThis study aims to calibrate the detection efficiencies of the SDD and CZT, so as to invert the observed data for obtaining real solar X-ray data.MethodsThe Monte Carlo code MCNP5 (Monte Carlo N-Particle 5) was employed to calculate the SDD and CZT efficiencies by simulation. Soft and hard X-ray detection efficiency calibration experiments were performed using a monochromatic X-ray ground calibration facility via relative measurement methods.ResultsThe experimental results for the SDD-1 and CZT-1 efficiency calibration agree well with the predicted results of the simulation. In particular, the maximum relative error between the experimental and simulated efficiencies of SDD-1 dose not exceed 3.59%@16 keV, and the maximum relative error between the experimental and simulated efficiencies of CZT-1 dose not exceed 9.54%@120 keV. The relative expanded uncertainty of the monochromatic X-ray flow intensity measurement is 3.8% (k=2), and the uncertainty of the simulation results for the SXD is 0.12%.ConclusionsThis study provides not only data support for SXD onboard MSS-1B satellite, but also valuable guidance for the calibration of other astronomical satellites' detectors in the future.

BackgroundLost radioactive sources needs to be quickly retrieved, positioning of radioactive source in complex environment is the key to find the lost radioactive source. [Propose] This study aims to develope a novel approach for the rapid positioning of orphan sources using a NaI(Tl) array detection device.MethodFirst of all, by leveraging the shadow effect between array detectors, a response curve between gamma-ray incidence angles and counts was obtained through the use of Monte Carlo simulation software. Then, the support vector machine (SVM) method was employed to establish a predictive mathematical model for the counting rate of array detectors as a function of gamma-ray incidence angle, utilizing. Finally, a radioactive source localization physical experiment platform was constructed, and a series of incidence angle response experiments were conducted for the validation of this approach applied to radioactive source localization under varying conditions.ResultsEexperimental results demonstrate that, through the use of the SVM regression prediction model, the maximum average deviation of the angle is 9.21° whilst the minimum is 1.77° for the angle prediction of an orphan 137Cs point source.ConclusionsThis method can achieve rapid and accurate localization of an orphan radioactive source.

BackgroundIn the Shanghai High Repetition rate XFEL (X-ray free electron laser) and Extreme Light (SHINE) facility, the vertical linear polarization laser is generated by using 40 planar superconducting undulators (SCUs) with a period length of 16 mm, length of 4 m, and a gap of 4 mm. At present, the Hall probes are the most reliable method for measuring the undulator magnetic field whilst the positioning accuracy of the sensitive center of the Hall probe is one of the main factors affecting the accuracy of magnetic field measurement.PurposeThis study aims to calibrate the position of the Hall probes' sensitive region for magnetic field measurements of SCU with high-precision.MethodsThe experimental platform for magnetic field point measurement of SCUs was introduced in details, a sledge with three mounted Hall probes and a retro-reflector were applied for magnetic field measurement. By flipping the sledge, the lateral distance between the sensitive centers of the Hall probe and each other were obtained, so did the lateral distance between the sensitive centers of the Hall probe and the apex of the pyramid prism. Therefore, the position of the Hall probes' sensitive region and center of the retro-reflector were calibrated.Results & ConclusionsThe positional calibration of the Hall probes has an accuracy higher than ±10 μm, which meets the requirements for magnetic field measurement.

Silicon carbide (SiC) crystal can be used as a passive monitor to measure the neutron irradiation temperature in nuclear reactors, which has significant application prospects for advanced reactors operating in high-temperature intense irradiation environments. Since the SiC temperature measurement technique was proposed in the 1960s, various temperature measurement methods have been developed on basis of neutron irradiation effects in the structural, thermal and electrical properties of SiC. These methods involve measuring changes in macro-size, density, thermal diffusivity or the electrical resistivity of SiC. This study summarizes the fundamental principles and characteristics of these methods firstly, then the research progress on SiC temperature measurement system required for advanced nuclear reactors at the China Institute of Atomic Energy (CIAE) is emphatically reported, and the measurement accuracy of SiC monitor is analyzed by calculating the lattice swelling rate of the neutron-irradiated SiC using a theoretical model, which verified the reliability of the temperature measurement results of the system. Finally, experimental methods for further improving measurement efficiency of SiC monitor are discussed.

The year 2023 marks the 35th anniversary of the establishment of the Beijing Tandem Accelerator Nuclear Physics National Laboratory. Accelerators and nuclear reactors are the two main tools for studying nuclear science. In 1988, the Tandem Laboratory was officially founded, serving as a significant research hub for nuclear physics in our country. It has consistently played a leading role in nuclear science innovation, achieving 140 000 h of stable operation. The laboratory has undertaken research in nuclear physics research, nuclear data measurement, nuclear physics applications, and interdisciplinary studies. This has resulted in a series of internationally recognized basic and technological achievements that meet national major demands, fostering a group of outstanding talents. It has provided solid support for the continuous development of nuclear physics research and nuclear technology strategy in our country. This article provides a comprehensive overview of the 35 years of development of the Tandem Laboratory.

Supernovae are the most gorgeous fireworks that people can observe in the universe. Their explosion can produce a maximum luminosity 10 billion times that of the Sun, helping scientists see farther. Type Ia supernovae can be used as a standard candle to facilitate measurement of the distance between galaxies in the universe. A supernova explosion will also propel a large number of heavy elements into interstellar space, which is a major driving force for the chemical evolution of galaxies. In addition, supernovae are crucial to the origin of elements in the Milky Way, the formation of the structure of the solar system, and the evolution of life on the Earth. The study of supernovae will further enrich our understanding of the universe and help us solve the mysteries of the expansion of the universe, the generation of heavy elements, and the origin of life. At present, scientists predict that the next supernova will explode at any time, and preparations are in progress for observing the coming supernova.

The nucleus is a quantum many-body complex system governed by the nuclear force, and it is prone to global changes such as deformation, rotation, vibration, fission, and clustering. In the past >30 a, we have witnessed the rapid expansion of the experimentally attainable nuclear chart and new discoveries and breakthroughs in studies on unstable nuclei. Examples include the halo nuclei and the associated exotic structural phenomena, the shell evolution observed using in-beam γ spectroscopy through the application of the achromatic magnetic spectrometer, the measurement of the basic properties of unstable nuclei, and the discovery of new magic numbers and rich phenomena in multi-nucleon correlations along with the formation of clusters and molecules. In the coming years, the expanded area of the nuclear chart—particularly the medium-heavy-mass neutron-rich region—will be the host of extreme exotic structures, the astrophysical r-process, and the reaction pathways to reach the superheavy island. Therefore, many new-generation radioactive ion-beam facilities are under development worldwide, and essential breakthroughs are foreseen.

BackgroundIsoscalar pairing plays an important role in the spin-isospin excitation of nuclei. The discovery of super Gamow-Teller (GT) states in N≈Z nuclei has motivated researchers to explore the effects of isoscalar pairing on spin-isospin excitations.PurposeThis study aims to investigate the effects of the isoscalar pairing interaction on GT and spin-dipole (SD) transitions in 42Ca.MethodsBy solving the relativistic Hartree-Bogoliubov equation, we obtained the canonical single-nucleon basis and occupation amplitudes, which were used as inputs for the quasiparticle phase-random approximation (QRPA) calculation. Using the QRPA model, the GT and SD transitions in 42Ca were calculated, where the Gaussian isoscalar pairing force was adopted, with its strength being a free parameter.ResultsFor GT states, the isoscalar pairing mixed the spin-flip transition configuration into the low-lying GT state, enhancing the collectivity of the low-energy GT state and significantly increasing its transition strength. Meanwhile, the isoscalar pairing force induced a shift of the low-energy GT state toward lower energies owing to the attractive properties of the isoscalar pairing force. For SD states, the isoscalar pairing force hardly affected the strengths and energies of SD states in 42Ca.ConclusionsIsoscalar pairing force was essential for restoring the SU(4) symmetry and hence reproducing the low-energy super GT state of 42Ca in the experiment, whereas it hardly affected the SD states.

China Jinping Underground Laboratory (CJPL) has the deepest rock overburden in the world, which considerably shields the detectors from muons. Thus, it has ultra-low radiation background level and is useful for experiments investigating rare physical events. Previously, experiments including the CDEX (China Dark matter EXperiment), PandaX (Particle and Astrophysical Xenon Experiments), JUNA (Jinping Underground Nuclear Astrophysics Experiment), and neutrino experiment have been carried out at CJPL and have given good results in dark matter detection, neutrinoless double beta decay, and more. This review introduces the construction process of CJPL, and introduces the facilities, results, and future plans of the aforementioned experiments. The CDEX used a high-purity germanium detector array for the dark matter detection and neutrinoless double-beta decay searches; whereas, for the same searches, PandaX used a dual-phase liquid xenon time projection chamber detector. A proton and helium accelerator was used by JUNA to simulate four nuclear reactions that occur in the Universe. A 103-kg prototype was constructed for feasibility verification by the neutrino experiment. The CDEX, PandaX, and JUNA collaboration groups give their latest results, all of which have approached or replaced the best results in the world. These experiments verify the extraordinary experimental conditions at CJPL. With the construction of CJPL-II, we expect an increase in the number of experiments based in Jinping and for further significant results to be achieved.

The fundamental properties of unstable nuclei are highly related to the nuclear structure and effective nucleon-nucleon interaction, and they can be used to study various exotic structures of unstable nuclei. Laser spectroscopy is a powerful tool used to study nuclear properties and structure by probing the hyperfine structure and isotope shift of the corresponding atoms or ions, from which the nuclear properties can be extracted in a nuclear model-independent manner. Multi-step laser resonance ionization spectroscopy (RIS) can be used to measure the atomic or ionic hyperfine structure. Based on this approach, various experimental techniques have been developed at radioactive ion beam (RIB) facilities worldwide to study the nuclear properties and structure of atomic nuclei. In this paper, the RIS approaches and relevant RIS experimental techniques are first introduced. Subsequently, the recently-developed collinear resonance ionization spectroscopy experimental technique, which can be used to measure the atomic or ionic hyperfine structure spectrum with a high-resolution and high sensitivity and plays an important role in the study of the nuclear properties and structure of unstable nuclei in the large mass regions of nuclear charts, is discussed in detail. Finally, the development status of RIS and its application in domestic RIB facilities are discussed.

The physics of radioactive nuclear beams is one of the frontiers of nuclear physics. New phenomena and physics appear in exotic nuclei far from the β-stability line. The neutron skin is an exotic phenomena in unstable nuclei and is closely correlated with the properties of the equation of state (EOS) of asymmetric nuclear matter and neutron stars. This study sought to examine previous studies on the effect of the neutron skin on nuclei-nuclei collisions to identify good observables for determining the neutron skin thickness, which could in turn be used to investigate the EOS of asymmetric nuclear matter. Various theoretical models are used to study the effect of neutron skin in nuclei-nuclei collisions. The statistical abrasion-ablation (SAA) and isospin-dependent quantum molecular dynamics (IQMD) models are used to study the neutron abrasion cross-section, neutron/proton ratio, and t/3He ratios. A nuclear structure model is used to investigate the relation between the neutron skin and α-cluster formation, α decay, nuclear surface, and nuclear temperature. Strong correlations have been found between the neutron skin thickness and neutron abrasion cross-section, neutron/proton ratio, and t/3He ratios, photo production, and other quantities. By measuring quantities that have a strong correlation with the neutron skin, the skin thickness can be obtained. The EOS of asymmetric nuclear matter and properties of neutron stars can be studied or constrained by using the obtained neutron skin data. Further investigations are necessary for determining observables that are useful for determining the neutron skin thickness from experimental measurements.

With the rapid development of radioactive-ion-beam facilities worldwide, many exotic nuclear phenomena have been observed or predicted in nuclei far from the β-stability line or close to the neutron (proton) drip lines, such as halos in atomic nuclei and shape decoupling in deformed halo nuclei. The study of exotic nuclear phenomena, including halos, is at the frontier of current nuclear physics research. The covariant density functional theory (CDFT) is one of the most successful models in nuclear physics. The CDFT has been widely used to study structures and properties of exotic nuclei. The deformed relativistic Hartree-Bogoliubov theory in continuum (DRHBc) has been developed and achieved a self-consistent description of deformed halo nuclei by including deformation and continuum effects, with the deformed relativistic Hartree-Bogoliubov equations solved in the Dirac Woods-Saxon basis. The DRHBc theory has been used to predict the deformed halo structure in 44Mg and the shape decoupling between the core and halo. The theory has also been used to address unresolved problems concerning the radius and configuration of valence neutrons in 22C, deformed halos in carbon and boron isotopes, particles in the classically forbidden regions in magnesium isotopes, and other similar phenomena. The rotational excitation of deformed halos has been investigated by implementing an angular momentum projection based on the DRHBc theory. This investigation has shown that the effects of deformed halos and shape decoupling are also present in the low-lying rotational excitation states of deformed halo nuclei.

With experimental facilities being developed globally, producing superheavy nuclei using heavy-ion collision has become feasible, which is essential for exploring charge and mass limits of nuclei and understanding the r-process in nuclear astrophysics. Fusion reactions are crucial for the synthesis of superheavy nuclei, yet only neutron-deficient superheavy nuclei get produced due to the limited neutron number of stable beams. Recent experiments suggest that multinucleon transfer reactions are promising for producing new neutron-rich superheavy nuclei. As a result, transport models are required for extracting physics information from these experiments and making predictions about incident energies and projectile-target combinations, to synthesize new super-heavy nuclei. In this article, we introduce the development of transport models such as the dinuclear system (DNS) model, quantum molecular dynamics (QMD) type model, Boltzmann type model, and Time-dependent Hatree-Fock (TDHF) type model, and conclude with their latest applications in the synthesis of superheavy nuclei, especially in fusion reactions and multinucleon transfer reactions. In addition, various international large-scale scientific facilities, as well as their scientific objectives, and future plans, are also summarized.

Backgroundβ-decay half-life is one of the fundamental physical properties of unstable nuclei and plays an important role in nuclear physics and astrophysics.PurposeThis study aimed to provide accurate nuclear β-decay half-life predictions and reasonable uncertainties associated with the predictions.MethodsNuclear β-decay half-lives were studied based on the Bayesian neural network (BNN) approach. Three types of neural networks with x = (Z, N), x = (Z, N, Qβ), and x = (Z, N, δ, Qβ) were constructed as inputs to explore the influence of the input on the prediction. The posterior distributions were sampled using the Markov chain Monte Carlo algorithm. The mathematical expectations and standard deviations of the neural network predictions on the posterior distributions were used as the predicted values and errors of the BNN approach.ResultsThe learning accuracy can be significantly improved by incorporating the β-decay energy and physical quantity related to the nuclear pair effect into the neural network input layer and then using the logarithm of β-decay half-life as the network output. For nuclei with half-lives of less than 1 s, the prediction accuracy is approximately 0.2 orders of magnitude, which is similar to that afforded by the BNN method by learning the differences between the logarithms of the experimental half-lives and theoretical results.ConclusionsThe Bayesian neural network can accurately predict β-decay half-lives. When extrapolated to the unknown nuclear region, the predicted β-decay half-lives agree with the results of other theoretical models within errors, especially for nuclei with Z ? 50.

We developed a Gamow shell model based on first principles and successfully applied it to the nuclei around driplines. Herein, we review the theoretical and technical developments of this method. Starting from the realistic nuclear forces, the model uses the Berggren basis, which contains bound, resonant, and scattering continuum states. Therefore, the Gamow shell model can handle the coupling to the continuum. In the complex-momentum plane, we used many-body perturbation theory (i.e., so-called Q^-box folded diagrams) to derive the Hamiltonian for the valence space. Subsequently, the shell-model calculations, which included the resonance and continuum effects, were performed. Therefore, such ab initio calculations can describe the weakly bound properties of nuclei near driplines and unbound resonance properties of nuclei beyond driplines. In this study, the symmetry breaking between oxygen isotopes and their mirror nuclei is discussed, and the important continuum effects on the excitation spectra of neutron-rich carbon isotopes are analyzed.

BackgroundTo date, various nuclides up to Z = 118 have been discovered and synthesized, raising the challenge of synthesizing nuclides with Z ≥ 119. Recently, the fusion-evaporation reactions 243Am54Cr, xn119297-x and 243Am55Mn, xn120298-x have been suggested as methods for synthesizing new elements with Z = 119 and 120. As α-decay is a powerful tool for the identification of new elements or nuclides, accurate predictions of the α-decay properties of the reaction products could be a useful reference for future experiments.PurposeThis study aims to provide quantitative predictions of the α-decay, spontaneous fission, and β-decay half-lives for the α-decay chains of 293, 294119 and 294, 295120 and to demonstrate the competition between the decay modes for these nuclei.MethodsAn improved density-dependent cluster model (DDCM+) is used to calculate the α-decay half-lives, taking the anisotropy of the surface diffuseness into account. The spontaneous fission half-lives are calculated using the Karpov formula, which is related to the fissility parameter and fission barrier height of the potential energy surface. The β-decay half-lives are determined using a finite-range droplet model (FRDM).ResultsThe predictive α-decay half-lives for the α-decay chains of 293, 294119 and 294, 295120 are obtained using the DDCM+ model, and the theoretical half-lives of the spontaneous fission and β-decay for these nuclides are also presented.ConclusionsFor the α-decay chains of 293, 294119 and 294, 295120, α-decay is predicted to be the dominant decay mode for most of the nuclei, while the half-lives of spontaneous fission and β-decay are predicted to be comparable to those of the α-decay near the region of A = 261. We expect that these results will serve as a useful reference for the synthesis of new isotopes in the future.

We reviewed the recent progress on strange particle production and hypernuclear physics both in experiments and in theories. The temporal evolutions of nucleons and resonances are described by the Skyrme energy density functional and relativistic covariant density functional theory, in which the meson-nucleon and hyperon-nucleon interactions are considered. Calculations are performed for the reactions of 12C+12C, 40Ca+40Ca, 112Sn+112Sn, and 197Au+197Au. The in-medium effects and high-density symmetry energy from the production of kaon, antikaon, and hyperon (Λ, Σ, Ξ) are investigated systematically. A quantum coalescence method is used to construct the hypernucleus, and the phase-space distribution is investigated in terms of the mass, charge, kinetic energy, rapidity distribution, collective flows, etc. Pre-equilibrium cluster emission in heavy-ion collisions is analyzed by implementing 2-, 3-, and 4-body nucleon collisions. The relativistic quantum molecular dynamics model is introduced by including ρ and δ coupling for nucleon transportation, and the collective flows are calculated for protons and neutrons.

BackgroundMachine learning, which has been widely applied to scientific research in recent years, can be used to investigate the inherent correlations within a large number of complex data.PurposeWe evaluate the performances of two types of machine-learning algorithms for correcting nuclear mass models, reconstructing the impact parameter in heavy-ion collisions, and extracting the symmetry energy slope parameter. We also discuss the extrapolation and generalization ability of the machine-learning models.MethodFor correcting the nuclear mass models, 10 characteristic quantities are fed into the LightGBM to mimic the residual between the experimental and the theoretical binding energies. For impact parameter or symmetry energy, two types of observables constructed based on the particle information simulated by using the UrQMD transport model for setting up the different impact parameters or symmetry energy slope parameters are used as inputs to a conventional neural network and the LightGBM to extract the original information.ResultAnalysis of these nuclear physics problems reveals the potential applicability of machine-learning methods.ConclusionsMachine-learning methods can be used to investigate new physical problems, thereby promoting the development of both theory and experiment.

BackgroundThe space environment contains numerous high-energy particles, and a single high-energy particle passing through a spacecraft shell bombards the electronic devices within, triggering single-particle effects such as device logic state upset and function failures, which, in turn, affect spacecraft operation reliability and mission accomplishment.PurposeNotably, ground accelerator irradiation tests provide an important and effective means for simulating space single event effects and for predicting the risks of single event effect rates for electronic devices in space applications. Generally, electronic devices can be used in spacecraft only if their resistance radiation indicators meet astronautical application requirements.MethodsSpacecraft are typically exposed to space radiation particles, primarily heavy ions and protons; therefore, single event effect simulation testing for electronic devices relies predominantly on heavy ion and proton accelerators. To address the requirements of single event effect testing, technologies such as large-area beam expansion and homogenization, high-precision beam current diagnosis, and efficient test terminals have been developed to fulfill the requirements of various test tasks.ResultsParticular focus is placed on the CIAE's (China Institute of Atomic Energy) heavy ion single event and proton single event effect simulation test techniques and the heavy ion microbeam technique for radiation sensitive area identification for electronic devices. Subsequently, the aforementioned techniques are applied to a single event effect risk evaluation for astronautical electronic devices.ConclusionsIn the future, the demand for radiation-resistant devices is expected to continue to increase in the aerospace, nuclear industry, and other radiation application fields. It is, therefore, necessary to further exploit the irradiation potential of existing domestic single event effect simulation equipment and establish new accelerator platforms with improved capacity for single event effect simulation testing.

Aerospace integrated circuits represent core components of space electronic systems, and anti-radiation hardening is a key technology to ensure the reliable operation of aerospace integrated circuits in the space domain. As the feature sizes of integrated circuits shrink to the nanometer scale, the single-event effect gradually becomes the most critical factor limiting the radiation-hardened performance level of aerospace integrated circuits. In this study, radiation hardened by design is utilized as a method to develop radiation-hardened performance. Based on single-event radiation tests on a heavy ion accelerator, new methods are proposed for the single-event test evaluation of new processes and devices. Consequently, new technique development and radiation effect law research are also undertaken. The effectiveness of the design hardening technology is evaluated, and a single-event radiation damage mechanism is discovered. The proposed technology provides key support for the production of high-reliability and long-lifetime aerospace integrated circuit products.

Through the use of the accelerator facilities at home and abroad, the nuclear reaction group of the China Institute of Atomic Energy has made many remarkable achievements in the study of fusion-fission dynamics, fusion-enhancement mechanisms at sub-barrier energies, reaction dynamics induced by exotic nuclei, and the related exotic nuclear structure and proton decay. In this study, some representative achievements are reviewed briefly. (1) The fusion mechanisms at near-barrier energies were investigated systematically, and a self-consistent method to evaluate the coupled-channel effects was proposed. (2) Nuclear deformation parameters were extracted from backward quasi-elastic scattering, which offered evidence for hexadecapole shapes. (3) A surrogate capture method was developed, based on which the first 239Pu(n,2n) excitation function developed in China was derived. (4) Systematic studies of exotic decay spectroscopies for proton-rich nuclei in the sd-shell were performed, following which a β2p decay of 22Si and a large isospin-asymmetry decay were discovered, and a strongly isospin-mixed doublet in 26Si was revealed. (5) Systematic studies of reaction mechanisms induced by exotic nuclei at energies close to the Coulomb barrier were performed, providing evidence for the failure of the dispersion relation in the optical potential of 6He+209Bi, and the reaction dynamics of proton drip-line nuclei of 8B and 17F were investigated. Future research based on the new HiTOF and BRIF facilities is discussed as well.

The HI-13 tandem accelerator, located at the Beijing Tandem Accelerator National Laboratory, has been in operation for 35 years. To ensure the continued performance of the accelerator, the operation and maintenance team has prioritized focus on various aspects. The operation team conducted research that involved developing key components, cultivating a high-quality operational team, improving the machine time efficiency, and increasing the participation of users outside the China Institute of Atomic Energy (CIAE). The primary emphasis has been on developing key components and upgrading subsystems. These efforts have successfully maintained and improved the accelerator's performance, ensuring its safe and stable operation. Finally, the paper alse discusses the challenges faced by tandem accelerators and presents future development plans.

The atomic nucleus, governed by short-range nuclear force, is a quantum many-body system that plays a vital role in the visible energy-mass dynamics of the universe and significantly influences the sustenance, development of society, and the security of nations. There have been numerous discoveries in the past decades concerning exotic structures and properties of short-lived nuclei. These findings have sparked breakthroughs in our understanding of nuclear structures and have given rise to a new field called radioactive ion beam physics, which focuses on the study of unstable nuclei. For more than 30 years, the Beijing Tandem-Accelerator Nuclear-Physics National Laboratory has provided a basic research platform for low-energy nuclear physics experiments. The experimental nuclear physics team at Peking University has continuously developed a dedicated experimental apparatus, conducted a series of physics experiments at the Beijing HI-13 tandem accelerator, and achieved important results related to exotic nuclear structures. In this article, we present several notable experimental achievements of our team at the HI-13 accelerator. These include the investigation of the shape evolution of germanium isotopes (around A=70) using in-beam γ-spectroscopy, the exploration of cluster structures in light neutron-rich nuclei through direct nuclear reactions, and the development and commissioning of collinear laser spectroscopy experiments at the Beijing Radioactive Ion-beam Facility.

Nuclear data, especially neutron nuclear data, forms the foundation of national defense, nuclear energy development, and the applications of nuclear technology. It also plays a critical role in fundamental nuclear physics research. The quality of nuclear data directly impacts the effectiveness, safety, reliability, and economy of related devices and products. Experimental data serves as the foundation for developing theoretical models and nuclear data libraries. Therefore, experimental research holds a paramount position in nuclear data research. The experimental research on nuclear data in China commenced in the mid-1950s and has achieved fruitful results after decades of development. In this article, we provide a brief overview of the progress made in experimental research on nuclear data in China and outline potential future advancements.

The first radioactive ion beam line, GIRAFFE, has been built at the CIAE HI-13 tandem accelerator in China. A total of eleven types of radioactive ion beam, including 6He, 7Be, and 8Li, have been generated. Several significant reactions in nuclear astrophysics have been indirectly measured via transfer reactions, and research on nuclear structure, relevant to nuclear astrophysics, has been performed using charge exchange reactions and thick-target experimental methods. A series of single nucleon or α cluster transfer reactions have been measured using a Q3D magnetic spectrometer, and the astrophysical S-factors and reaction rates for essential reactions have been obtained. The obtained results serve as a crucial experimental foundation for research involving element abundance and celestial body models.

Cluster structures can be stable in the interior of atomic nuclei. The study of α-cluster structure of atomic nuclei and its effects are important topics in nuclear physics as well as astrophysics. In the past few decades, cluster structure effects in atomic nuclei have been much studied for heavy-ion nuclear reactions. This paper summarizes the authors' studies on the α-cluster structure effects on nuclei in nuclear reactions and relativistic heavy-ion collisions. For example, the cluster structure of atomic nuclei has been studied through giant resonances of atomic nuclei. The cluster structure of the nucleus is studied through the emission and correlation of particles (including neutrons, protons, and photons) in nuclear reactions and through collective flows. We extend the cluster effect of atomic nuclei to relativistic heavy-ion collisions, e.g., to the study of collective flows and their rise and fall, the HBT (Hanbury Brown and Twiss) correlation, multiplicity correlations, the dihadron azimuthal correlation, and electromagnetic fields.

Academician ZHANG Huanqiao is an outstanding scientist cultivated by the new China in the 1950s. The impoverished and weak war environment of the old China and the hardship era of the new China have nurtured his patriotic determination to serve his motherland. He adhered to the frontline of scientific research and devoted himself wholeheartedly to it. His research fields spanned neutron physics, fission physics and heavy-ion nuclear physics, and he has achieved excellent results under difficult conditions. He committed himself to the country and measured the urgently needed nuclear data to cooperate with nuclear weapon development. He is rigorous and realistic, and repeatedly verifies the experimental results to ensure accuracy. He dares to take the lead and constantly delves into new research fields. All his actions reflect a silent loyalty and love for his motherland and science. He is an inheritor and speaker of the older generation scientists' spirit, and a microcosm of the scientist's spirit and the development of science and technology in the new China. We write this article for sharing on the occasion of academician ZHANG Huanqiao's 90th birthday.

BackgroundCurrently, the destructive puncture manometry method is used to measure helium pressure inside fuel rods. However, this method is expensive and does not guarantee 100% coverage. Hence the non-destructive testing (NDT) equipment is introduced for non-destructive measurement of helium pressure inside fuel rods.PurposeThis study aims to analyze the reliability of NDT testing method for the measurement of helium pressure inside fuel rods.MethodsThree standard rods with helium pressure values of 0.98 MPa, 1.76 MPa, 2.45 MPa, respectively, were selected for experimental test. The experimental fuel rods were first used to obtain the results comparison of heat transfer method and puncture manometry, then the control variates were employed to control the fuel rod temperature, the time interval of a single measurement, and the ambient temperature respectively, so as to determine influencing factors in the NDT method. Finally, reliability analysis of NDT method was performed according to experimental results.ResultsThe results of the NDT method are consistent with that of the puncture manometry method at a temperature range of 24~30 oC with less than 0.05 MPa deviation. Minimum repeat measurement time interval for NDT measuring helium pressure of the same standard rod or fuel rod is 2 min.ConclusionsThe NDT method for measuring the helium pressure of fuel rods is reliable, and the measurement results are stable in different environmental conditions.

BackgroundIn a pressurized water reactor (PWR) loss-of-coolant accident (LOCA), high temperature and high internal pressure of the fuel rod can lead to ballooning of fuel rod cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow and thus affect the core heat transfer during reflood phase and subsequent severe accidents. However, the commonly used integrated severe accident analysis codes use simple parametric models to simulate these aspects and therefore cannot consider the influence of multiple coupled factors. This results in a lack of accuracy of the simulation results.PurposeThis study aims to analyze the key phenomena in core degradation, and develop a thermal-mechanical (TM) behavior module for assessing the failure of cladding and analyzing the flow blockage.MethodsFirst of all, the fuel rod thermal–mechanical behavior (FRTMB) module developed for analyzing the TM behavior of fuel rods was integrated into the integrated severe accident analysis code (ISAA). Then, on the basis of the FRTMB module, the flow blockage model of the ISAA-FRTMB code was improved to suit for simulating changes in coolant flow rate caused by fuel rod deformation. Finally, the QUENCH-LOCA-0 experiment was simulated by using improved ISAA-FRTMB code to verify the correctness and effectiveness of the model, and the peak cladding temperatures were compared in order to verify the validity of the flow blockage model.ResultsThe results including cladding failure time, circumferential strain, flow blockage rate and cladding temperature predicted by the code are in good agreement with the experimental data. The maximum circumferential strain of the simulated cladding, as indicated by the experimental results, is in the range of 25%?50%, and the errors of the predicted cladding rupture time and temperature are within 4%.ConclusionUnder the stress caused by internal pressure, the cladding deforms outward owing to thermal creep with the increase of temperature. Rapid thermal creep and swelling lead to cladding failure. The maximum circumferential strain of the simulated cladding, as indicated by the experimental results, is in the range of 25%?50%, and the errors of the predicted cladding rupture time and temperature are within 4%. The correctness and effectiveness of FRTMB module are thus verified.

BackgroundFLiBe is commonly used as the coolant and carrier salt in liquid molten salt reactors (MSRs). Its certain moderating properties and thermal neutron scattering attributes affect the neutronic performance of the MSR, and this in turn influences the physical design and safe operation of the reactor. Consequently, studying FLiBe's thermal neutron scattering data is essential for MSRs.PurposeThis study aims to analyze the influence of of FLiBe thermal neutron scattering on neutronic performances of a 65-MW MSR.MethodsFirst, according to the requirements, a core model of a 65-MW MSR was established by using the general Monte Carlo procedure. Then, the neutronics performance of the MSR was calculated by considering the scattering cross-section of the free gas model and FLiBe thermal neutron scattering data (e.g., the neutron spectrum, effective multiplication factor, and nuclide reactivity rate). Finally, the changes in the influence of FLiBe thermal neutron scattering effect on neutronic properties under different energy spectra were compared.ResultsThe computation results show that, by considering the thermal scattering effect of FLiBe molten salt, the neutron energy spectrum in the core of the MSR becomes harder, 235U fission rate decreases, the keff value of the reactor decreases, but the density coefficient in the temperature reaction coefficient of the fuel keeps almost unchanged, and the Doppler coefficient decreases by 0.28×10-5 K-1. With the hardening of the energy spectrum, the variation in the 235U fission rate reduction decreases, and the decrease in keff caused by thermal neutron scattering changs from 9.2×10-4 to 2×10-4.ConclusionsTherefore, it is necessary to incorporate FliBe's thermal neutron scattering data into the physical calculations for the MSR core.

BackgroundThe core corium may melt through the reactor pressure vessel wall then lead to the failure of the second barrier during a serious accident. Core catcher can collect and cool the corium and prevent the development of severe accident.PurposeThis study aims to establish a computational model to explore the cooling process of crucible core catcher adopted by VVER (Vodo-Vodyanoi Energetichesky Reactor) designed by Russia.MethodsAccording to the derived parameters based on VVER core catcher design data, non-isothermal flow calculation module of COMSOL was established to simulate the flow field, temperature field, and crust distribution of corium pool. The solidus temperature and liquidus temperature and the exponential form change of corium viscosity were referred to the research results of VULCANO item.ResultsFor the double layered structure of the corium pool in core catcher, the metal layer solidifies quickly after a core meltdown accident. Constantly changing natural convection flows are formed in the upper and middle part of the oxide layer and the temperature distribution is relatively uniform. No strong convection exists in the lower part of the oxide layer lead to obvious thermal stratification. Most of the corium cooled in the upper part of the oxide layer will transfer to the lower part by gravity and natural convection before full solidification, resulting in a slow increase in the thickness of the upper crust and a rapid increase in the bottom crust of the oxide layer.ConclusionsThe safety margin of crucible core catcher of VVER is sufficient, however the relevant equipment, support and auxiliary system are required to remain operational for a long time to realize the design function.

BackgroundIn the international fourth-generation nuclear power system, the lead-bismuth fast reactor is one of the most concerned technologies. However, insoluble particulate matter generated in the flow of liquid lead-bismuth alloys will collect locally in the flow channel and affect the operation of lead-bismuth fast reactors.PurposeThis study aims to find the motion deposition of particulate matter in the flow channel, understand its influence on the safe operation of small lead-bismuth fast reactors, and provide a reference for the safe design of lead-bismuth reactors.MethodsFirstly, based on the design scheme of 100 MWth small natural circulation lead cooled fast reactor SNCLFR-100, the particle deposition in the rod bundle channels that were divided into three types according to the relative position and wall conditions: triangle like channels, pentagon like channels and trapezoid like channels, was numerically simulated using ANSYS software, and the particle deposition movement was obtained. Then, the effects of particle type, particle size and particle velocity on particle deposition were obtained on the basis of grey correlation degree theory. Finally, the correlation degree of various factors affecting particle deposition rate was analyzed.ResultsThe results show that the particle deposition mainly occurs at the inlet stage,the surface of the inlet section is large area adhesion deposition,and the surface of the middle and rear sections is point-like deposition. With the increase of axial distance, the magnitude of turbulent kinetic energy is the main factor affecting the radial distribution of particulate matter. The increase of particle density and particle size will strengthen the deposition of particulate matter. The increase of particle velocity will reduce the particle deposition. The degree of influence on particle deposition is particle size>type> particle velocity.ConclusionsDuring the operation of lead-bismuth fast reactor, attention should be paid to the deposition of particles in the inlet section and to remove the particles with larger particle size.

BackgroundThe metallic materials utilized for nuclear reactors undergo corosion due to the inherient high-temperature and high-pressure environment. Consequently, the corrosion products may be deposited in the core, called crud, and impact the fuel operation, core reactivity, and primary radioactivity, such as crud-induced localized corrosion or crud-induced power shift.PurposeThus, this study aimed to establish a model that can quantitatively analyze these corrosion products, the results of which can then be used to evaluate the impact of these products.MethodsBased on the corrosion and release dynamic theory, combined with the assumption of metallic oxide volume ratio (Pilling-Bedworth Ratio), a corrosion and release model of metallic materials was developed. The model was validated based on experimental data from Inconel 690.ResultsThe verification result indicates that the proposed model is reasonable and scientific, and hence can be used to quantify the amount of main corrosion and release products of metallic materials for nuclear reactors.ConclusionsThis study provides a model of the main elements of corrosion products including Ni and Fe ferrite for PWR plants, which can be used for evaluating the impact of corrosion products. However, some of the microelements of corrosion cannot be quantified by using this model as the corresponding equations were over-determined. Hence this aspect requires further research in the future.

BackgroundMolten salt reactors, one of the important types of fourth-generation advanced reactors, use high-boiling-point molten salt as a nuclear fuel carrier after melting, hence have the characteristics of high-temperature output and normal-pressure operation. A heat-pipe molten salt reactor based on thermoelectric power generation has the advantages of its components, that is, high output temperature, high thermoelectric conversion efficiency, simple structure, safety, and reliability. Therefore, the reactor of heat-pipe molten salt has significant advantages in the field of energy systems as it is an ideal energy source for outer space and deep-sea exploration missions. However, because of the low thermal conductivity of the molten salt in the core, the dense arrangement of heat pipes complicates the heat transfer design of the thermal power generator in the condensing section of the heat pipes.PurposeThis study aims to design a heat-pipe–thermal power generation coupling system structure suitable for molten salt reactors, and analyze its heat transfer characteristics on the basis of design requirements of the reactor.MethodsFirstly, the condensing section of the core heat pipe was designed using a tower thermoelectric power generation system. A thermoelectric generator was placed between the outer wall of the hot-side tower and the inner wall of the cold-side tower, and the gap between the generators was made of an insulating material to reduce heat leakage. Then, a heat transfer simulation of a four-layer tower thermoelectric power generation system suitable for a heat-pipe molten salt reactor was performed using the ANSYS Workbench. Finally, temperature distribution and variation under different power values at each layer of the thermoelectric generator and every thermoelectric generator, etc., were analyzed.ResultsThe analysis results reveal that, when the system is running with maximum heat-pipe temperature of 696 ℃, the temperature distribution in the overall tower is uniform, the effective heat utilization rate is >96%, the system leakage heat is <4%, and the temperature difference between the two sides of the generator is >490 ℃, which is conducive for improving the thermoelectric conversion efficiency.ConclusionsThe structural design of this study is feasible and conducive for promoting the application of thermoelectric power generation in a heat-pipe molten salt reactor.

BackgroundTraditional X-ray fluorescence spectrum analysis has the limitations of poor accuracy of the characteristic peak counting rate and shadow peak.PurposeThis study aims to propose a long and short term memory (LSTM) neural network model based on deep learning for the loss correction of the characteristic peak count rate and shadow peak.MethodsFirstly, a LSTM neural network model based on deep learning was proposed to estimate accurately the amplitudes of nuclear pulse signals by learning samples. Then, a convolutional neural network (CNN) with unique convolutional kernel structure was introduced to deal with the challenges of large sample size of the nuclear pulse signal and the low training efficiency of the model by extracting the sample features layer by layer, thereby effectively reducing the number of samples and the complexity of model training. Finally, a series of offline nuclear pulse sequences of powdered iron ore samples were used to generate the dataset required for model training. Among the 64 000 entries in this dataset, 44 800 were used as training sets, 12 800 were used as validation sets, and the remaining 6 400 were used as testing sets.ResultsThe trained CNN-LSTM model saves considerable training time, overcomes the defects of local convergence of traditional methods, and accurately estimates the parameters of input pulse under different degrees of distortion. Results show that the accuracy rate of the training and verification sets is greater than 99%. An analysis of the count repair results reveals that the average value of the correction ratio of the three shadow peaks, that is, the correction ratio of the depth learning model trained in this study to the count loss derived from the distorted pulses, is 91.52%.ConclusionsThe CNN-LSTM model can effectively correct the shadow peaks derived from the amplitude loss of distorted pulses and improve the accuracy of the characteristic peak count rate in X-ray fluorescence spectra. The model is shown to have high application value for the field of X-ray fluorescence spectroscopy.

BackgroundElliptic flow (v2) is one of the most important observations for exploring the properties of nuclear matter using heavy-ion collisions. v2 is not only affected by dynamic processes but is also related to the Fermi momentum of the initial nucleus.PurposeThis study aims to quantitatively determine the effect of the initial Fermi momentum on the time evolution of v2.MethodsFirst, based on the Ultrarelativistic Quantum Molecular Dynamics (UrQMD) model, gold-gold (Au+Au) collisions at beam energies of 0.4A GeV and 0.8A GeV with impact parameter b = 6 fm were simulated. In the initial stage, three cases were considered: without Fermi momentum, with Fermi momentum, and with half-Fermi momentum. Then, by reverse tracing the nucleons that were emitted at mid-rapidity (|y0|<0.1) throughout the reaction process, the time evolution of v2 for these traced nucleons was investigated in detail. Finally, the influence of the initial Fermi momentum on v2 of the nucleons in the mid-rapidity region in heavy-ion collisions at intermediate energies was examined.ResultsThe yield of free nucleons calculated by considering the Fermi momentum was much larger than that obtained without the Fermi momentum, owing to the reduction in nucleon-nucleon collisions. However, v2 shows the opposite effect; it is obtained by considering that the Fermi momentum is much smaller than that in the latter case because of the stronger blocking effect of the spectator nucleons.ConclusionsOur results indicate that the initialization of the nucleon momentum must be carefully considered in the transport model.

BackgroundDiamond material demonstrates excellent temperature and radiation resistance properties, and detectors made from diamond exhibit good potential for use under harsh environments.PurposeThis study aims to analyze the structure and working principle of diamond thermal neutron detectors, and establish a physical model of such a detector applied to 2 MW thorium molten salt experimental reactor-liquid fueled (TMSR-LF1) radiation field by using MCNP program.MethodsFirst of all, 6Li and 10B were selected as neutron conversion materials considering the neutrons of TMSR-LF1 mainly concentrated in the 10-8~10-6 MeV energy range, and the Stopping and the Range of the Ions in Matter (SRIM) program was employed to calculate the range of secondary charged particles generated by the reaction in the neutron conversion layer and diamond layer. Then, the MCNP program was used to establish a physical model of diamond neutron detector applied to 2 MW TMSR-LF1 radiation field. Finally, the effects of the neutron conversion layer thickness (6LiF, 10B), diamond thickness, and γ screening threshold on the neutron detection efficiency, γ detection efficiency, and n/γ suppression ratio of the detector were determined through simulation results.ResultsThe results reveals that 6LiF is more suitable than 10B for use in the neutron conversion layer in neutron and γ mixed fields. With the increase of the 6LiF thickness, the neutron detection efficiency first increases and then decreases, and the optimal thickness of 6LiF is 25 μm. The n/γ discrimination performance of the detector deteriorates with the increase of diamond thickness, but the diamond thickness must be greater than 20 μm to ensure insensitivity of the detector to γ, hence a γ screening threshold is needed to prevent excessive γ interference for thick diamond layers.ConclusionThe influence of detector structural parameters on detector performance obtained by this study has guiding significance for the subsequent fabrication of and research on such detectors.

BackgroundRecently, global concerns regarding the illicit transportation and trafficking of nuclear materials and other radioactive sources have increased, leading to increased demands for efficient and rapid security and non-proliferation technologies. The International Atomic Energy Agency's Incident and Trafficking Database has reported 3 235 confirmed incidents involving nuclear and other radioactive materials out of regulatory control from 1993 to 2017. Of these incidents, 278 are associated with trafficking or malicious use of materials such as highly enriched uranium, plutonium, and plutonium-beryllium neutron sources. Therefore, developing depth-of-interaction detector for neutrons and gamma rays is important for effective control of nuclear and radiation materials at national and international cross points such as borders, ports, and airports.PurposeThis study aims to design a depth-of-interaction detector for neutrons and gamma rays and characterize its performance.MethodsHereby, an EJ276 plastic scintillator (Φ3 cm× 15 cm) coupled with two silicon photomultipliers (SiPMs) in both sides was designed as a depth-of-interaction detector for neutrons and gamma rays. The short gate time was optimized to achieve better neutron/gamma-ray discrimination, and the reaction position was determined based on the amplitude ratio and time of flight (TOF) difference between signals from two sides. Finally, Am-Be neutron source and 137Cs γ source were applied to detector parameter optimization and resolution calibration for performance characterization.ResultsExperimental results demonstrate that good consistency in the detection efficiency of the detector at different incident positions, where the resolution of the one-dimensional reaction position is approximately 4.4 cm.ConclusionsThe designed depth-of-interaction detector can be used toreplace detector arrays in neutron scatter cameras and coded-aperture imagers to reduce costs and system complexity.

BackgroundComplete kinematic measurements in the medium or high-energy region is a common experimental method to study the structure and properties of exotic nuclides on the neutron-rich side. The experiment setup in the Cooling Storage Ring - Radioactive Ion Beam Line in Lanzhou (CSR-RIBLLII), a typical nuclear external target facility, comprises many detectors with different requirements. The anticoincidence (Veto) detector is an essential part of the external target facility for eliminating the interference of charged particles and measuring medium or high-energy neutrons with high reliability and performance by combining them with a neutron wall detector. The original Veto detector with photomultiplier (PMT) readouts has many disadvantages, such as low detection efficiency and poor uniformity, resulting in significant differences or contradictions between experimental and calculation results.PurposeThis study aims to upgrade the original Veto detector using wave length shifter fiber (WLS) and silicon photomultiplier (SiPM) to improve the detection efficiency of charged particles.MethodsFirstly, a new configuration for the anticoincidence Veto detector unit was designed and the detector thickness was increased by 5 mm compared to the previous Veto detector, resulting in a final thickness of 1 cm. The Veto detector was embedded with 15, 7, and 3 WLS fibers from both ends, and read using SiPM. Furthermore, to systematically explore the performance of the detector unit, a linear relationship was calibrated between the number of photons of the SiPMs and the number of Analog-to-Digital Converter (ADC) channels. This relationship was used to accurately calculate the threshold value, laying a foundation for calculating detection efficiency. Then, based on Multi-Wire Proportional Chamber (MWPC), a detection efficiency test platform was established, and time position conversion and track selection data analysis methods were developed as test methods. Finally, a detailed test on the whole and each part of the anticoincidence Veto detector unit was carried out on the MWPC test platform.ResultsTest results show the highest anticoincidence efficiencies of SiPMs at both ends for the Veto detector embedded with 15, 7, and 3 WLS fibers are 99.99%, 99.94%, and 99.82%, respectively; increased by over 22.74% compared with the original Veto detector.ConclusionsThe new Veto detector based on WLS fiber and SiPM readout meets the needs of the CSR-RIBLLII external target facility.

BackgroundThe extraction of uranium (U) and its alternative resources, such as thorium (Th) and plutonium (Pu), from seawater is essential to address the scarcity of terrestrial U resources. The development of a separation material with high adsorption properties is the key to solving this problem.PurposeThis study aims to reveal the adsorption behavior of actinides (U, Th, and Pu) on the surface of a two-dimensional metal material, antimonene.MethodsThe Hubbard U values, Ueff, were determined for the on-site Coulomb interactions of 5f electrons of U and Pu atoms using the linear response method. Furthermore, the adsorption energy, adsorption configuration, electronic structures, charge transfer, and highest occupied molecular orbital wavefunction of a U, Th, or Pu atom adsorbed on the surface of monolayer antimonene were analyzed using the DFT+U approximation. The variation of the adsorption rate with temperature was further calculated by the equilibrium adsorption rate equation.ResultsThe calculated Ueff values of U and Pu atoms are 2.24 eV and 2.84 eV, respectively. The Pu atom is energetically unfavorable to be adsorbed on antimonene (with a negative adsorption energy for each adsorption site), whereas the U and Th atoms exhibit strong chemical adsorption on its surface. Antimonene also offers abundant surficial stable adsorption sites for the U and Th adatoms. The most energetically stable sites for the U and Th adatoms are the B (Bridge)-H (Hollow) site and H (Hollow) site, with adsorption energies of 4.40 eV and 3.62 eV, respectively. The impurity states are generated in the band gap of antimonene upon the adsorption of the U or Th atom, and the strong p-d coupling between the U or Th adatom and antimonene in the impurity states contributes to the strong adsorption of the adatoms. The desorption temperatures of U and Th on the surface of antimonene reach 837 K and 660 K, respectively.ConclusionsThe results indicate that antimonene is an excellent two-dimensional adsorbent material for U and Th and has potential for several applications such as in the extraction of actinides from seawater.

BackgroundShanghai High Repetition rate XFEL and Extreme light facility (SHINE) employs a White Rabbit (WR)-based timing system. This timing system operates via the utilization of beamline–endstation division, which receives external reference timing signals and distributes them to each beamline and endstation via WR timing network devices, including master nodes, WR switches and slave nodes.PurposeThis study aims to develop a timing equipment control system (TECS) to address the requirements of remote monitoring and control of distributed timing equipment.MethodsBased on Experiment Physics and Industrial Control System (EPICS) and Simple Network Management Protocol (SNMP), an approach for acquiring timing equipment parameters was employed. These parameters was stored in the resident memory database though EPICS Input/Output Controller (IOC) and accessed via a user interface developed with PyDM (Python Display Manager). Archive and retrieval of timing equipment parameters were implemented in the Archiver Appliance historical archiving system. Finally, test environment was set up in laboratory to verify the validity and reliability of this TECS.Results & ConclusionsThis control system underwent testing exhibits its effective functionalities, including real-time monitoring equipment parameters, as well as remote control of equipment signal delay and pulse width. These capabilities are essential in meeting the requirements of SHINE beamlines and endstations.

BackgroundDynamic micro-computed tomography (micro-CT) using monochromatic X-ray offers higher density resolution and lower radiation damage compared to that using white X-ray, however balancing its imaging spatial and temporal resolution is challenging. Currently, the reported highest temporal resolution of monochromatic X-ray dynamic micro-CT is 13.3 Hz with a detector effective pixel size of 5 μm.PurposeThis study aims to develop a monochromatic X-ray dynamic micro-CT system with a higher spatial and temporal resolution to meet the experimental needs of the fast X-ray imaging beamline (BL16U2) users at Shanghai Synchrotron Radiation Facility (SSRF).MethodsFirstly, an experimental system of dynamic micro-CT with the high flux density monochromatic X-ray from an undulator source was established by combination of a high-speed rotary stage and a large numerical aperture triple-lens fast X-ray imaging detection system on the BL16U2 beamline at SSRF. Then, a demonstration experiment with a fast-foaming polyurethane material as a sample was performed to examine the spatial-temporal resolution of this experimental system, moreover a quantitative analysis of the bubble motion during foaming process was performed.ResultsExperimental results of foaming process of the fast-foaming polyurethane material based on the monochromatic X-ray dynamic micro-CT system show that a temporal resolution of 20 Hz of the dynamic micro-CT was achieved with 15 keV monochromatic X-ray and an effective detector pixel size of 2.2 μm.ConclusionsThe developed monochromatic X-ray dynamic micro-CT system has a high spatial-temporal resolution and can perform four-dimensional quantitative analysis of complex motion systems, providing a powerful experimental research platform for users of BL16U2 beamline at SSRF.

Neutron depth profiling (NDP) offers unique advantages in the measurement of element depth distributions, characterized by its high sensitivity and non-destructive nature. This article presents an overview of the principles and data processing methods employed in NDP technology, followed by a comprehensive comparison of various NDP devices and their corresponding parameters on a global scale. Furthermore, potential avenues for upgrades of NDP devices are explored. Given the remarkable sensitivity and non-destructive attributes of NDP technology in detecting 6Li, it proves particularly well-suited for in-situ measurements in lithium batteries, rendering it an invaluable tool for research in this field. The article underscores the application of NDP in lithium battery research whilst its utilization in high-temperature alloys, semiconductor materials, and nuclear materials is introduced as well.

BackgroundMany existing studies have shown that the use of suitable surface modification methods can enhance the boiling heat transfer effect of metal components, making it have a broad potential application prospect in the pressurized water reactor. However, for the weak alkaline environment of high temperature and high pressure in the reactor, little literature is reported on whether this enhanced effect can be maintained for a long time.PurposeThis study aims to explore the effect of corrosion on boiling heat transfer characteristics of metal specimens with micro-structure surface.MethodsFirst of all, three micro-structures of micro-groove, micro-porous and micro-columns were processed on the surface of stainless steel plate specimens by laser processing. Then the specimens were placed in the high-temperature and high-pressure environment simulating the actual reactor conditions to carry out corrosion experiments for up to 200 d. Finally, the pool boiling experiment and visualization study of the specimens before and after corrosion were carried out for comparison.ResultsThe results show that the surface critical heat flux (CHF) of the three micro-structured metal specimens increases and then decreases with the increase of corrosion time, among which the micro-pores specimens have the largest bubble generation rate at the beginning of nuclear boiling, and the micro-groove specimens have the highest CHF.ConclusionsThe influence law and mechanism of long-term corrosion in pressurized water reactor on the enhanced heat transfer effect of different micro-structure surfaces are partially revealed by this study.

BackgroundThe propagation of pressure waves in nuclear energy systems will cause hydraulic load effects, and it is particularly important for the analysis of structural loads to accurately simulate the propagation process of pressure waves. System analysis codes such as RELAP5, TRACE, etc. are widely used in the simulation and analysis of reactor pressure wave propagation. But system analysis codes can only simulate one-dimensional pressure wave propagation behavior.PurposeTo cope with the multi-directional and multi-dimensional pressure wave propagation issue, corresponding model and algorithm study is carried out in this paper to investigate the two-dimensional pressure wave propagation behavior in two-phase steam-water flow condition.MethodsBy employing a four-step algorithm of time-step separation, and a non-equilibrium phase transition heat transfer model, a two- dimensional two-phase flow pressure wave propagation code (TPFPWPC-2D) is developed based on 2D axisymmetric cylindrical coordinate system. The code verification is carried out by using a typical benchmark of steam-water two-phase shock tube. Finally, in order to verify the ability of TPFPWPC-2D code to simulate the two-dimensional propagation of pressure waves, numerical simulations of the pressure wave propagation behavior in a cylindrical space region were conducted.ResultsThe results of code verification show that the new code proposed here agrees well with the two system analysis codes RELAP5 and WAHA. The 2D simulation application shows that the new code can capture the 2D propagation processes of pressure wave reasonably, especially the reflection and superposition characteristics.ConclusionsFrom the results mentioned above, conclusions can be drawn that the new code developed in this paper can simulate the two-dimensional axisymmetric propagation characteristics reasonably in both quantitative and qualitative levels.

BackgroundIn order to accurately predict the friction pressure drop characteristics of liquid lead bismuth in the cross-section of the fuel assembly rod bundle, a suitable friction pressure drop model should be selected.PurposeThis study aims to investigate Friction pressure drop model for wire-wrapped rod bundles in full flow.MethodsEight different frictional pressure drop models within wire-wrapped rod bundles were evaluated their applicability by using statistical analysis. The prediction accuracy of experimental data from different models in different flow regimes was explored corresponding to laminar flow, transitional flow, and turbulence.ResultsThe analysis results show that the friction coefficient is not only related to the number of rod bundles (Nr) and the pitch-to-diameter ratio (P/D), but also related to the wire lead length-to-diameter ratio (H/D). The modified BDD model in the laminar flow range and this work model are more consistent with the experimental data. The modified BDD model, CTD model and this work model are relatively consistent with the experimental data in the transition flow range. The Rehme model, the UCTD model and this work model in the turbulent range are more consistent with the experimental data.ConclusionsTherefore, the model presented in this study is suitable for predicting friction pressure drop in the cross-section of the fuel assembly bundle in the full flow state.

BackgroundWith the increase of complexity of reactor core design, the core modeling and calculation have brought challenges.PurposeThis study aims to implement the accurate modeling and calculation of unstructured geometry core.MethodsBased on discrete ordinate nodal method for arbitrary triangular-z geometry, the precise modeling and mesh generation of unstructured core were established by constructive solid geometry (CSG), and Block-Jacobi parallel algorithm was employed to reduce calculation time of reactor core. Finally, based on the developed SARAX program, core physics calculations for new complex geometries of a space reactor and a heat pipe reactor were performed for accuracy verification by using Block-Jacobi parallel algorithm combining with established precise model and mesh.ResultsThe verification results show that the effective multiplication factor and radial power distribution agree with that of multi-group Monte-Carlo calculation. The calculation deviation of eigenvalues is less than 3.00×10-3, and the relative deviation of radial power distribution is less than 1.5%.ConclusionsResults of this study show that SARAX code has the ability of modeling and higher accuracy in the calculation of unstructured geometry core.

BackgroundAs an innovative technology of nuclear power, magnetically suspended high-temperature molten salt canned motor pump (referred to as molten salt canned motor pump) can be used in the fourth generation molten salt reactor (MSR). Improving pump performance via hydraulic optimization design is significant to fourth-generation nuclear power technology.PurposeThis study aims to investigate the influence of different working fluids on the hydraulic optimization design of magnetically suspended high-temperature molten salt-canned motor pumps and provide suggestions for the optimal design of magnetically suspended high-temperature molten salt-canned motor pumps.MethodsFirstly, ANSYS CFX software was employed to perform a numerical simulation of a magnetically suspended high-temperature molten salt canned motor pump. Based on response surface methodology (RSM), approximate models between significant parameters and optimization objectives were established. Then, taking the efficiency and head as optimization objectives, a non-dominated sorting genetic algorithm II (NSGA-II) was used to design the magnetically suspended high-temperature molten salt canned motor pump under molten salt and water.ResultsCompared with water working fluid, the optimization space of the pump under molten salt working fluid is larger. When the efficiency of the optimization model under the two working fluids is the same, the impeller inlet diameter and the blade outlet placement angle of the molten salt optimization model are reduced, whereas the impeller outlet width and the diffuser throat plane width are increased. The efficiency of the finally determined molten salt optimization model is increased by 0.75% and the head is raised by 0.078 2 m whilst the efficiency of the water optimization model is increased by 0.55%, and the head is reduced by 0.035 9 m.ConclusionsThe research results of this study can be used to guide the hydraulic structure design of a magnetically suspended high-temperature molten salt-canned motor pump.

BackgroundThe passive residual heat removal (PRHR) system is an important innovative design of the advanced pressurized water reactor technology. Under accident conditions, PRHR system can transport decay heat in the form of natural circulation to ensure core cooling. However, the heat exchange function of PRHR system will be lost when the PRHR pipeline breaks. With development of the accident process, the coupling effect between different safety equipments of the passive core cooling system (PXS) will be affected. Besides, thermal hydraulic state of the reactor coolant system (RCS) will also be affected via complex interaction mechanism. As a result, the new thermal hydraulic phenomena occur, and thus ultimately affecting the accident mitigation capacity of the PXS.PurposeThis study aims to confirm the safety characteristics of PXS and identify the new thermal hydraulic phenomena of advanced passive nuclear power plant during accident with multiple failures.MethodsA series of integral effect tests of loss of coolant accident (LOCA) were conducted on the advanced core-cooling mechanism experiment (ACME) facility. The influence of failure of PRHR HX flow and heat exchanging function on LOCA accident process were investigated on the basis of the test cases including PRHR pipeline break and cold leg (CL) break. The unique thermal hydraulic phenomena occurred during PRHR LOCA were explored, and their influence laws on the coupling effect among PXS safety equipments, and the influence laws on thermal hydraulic state of RCS were obtained.ResultsThe results show that there is a momentary reverse flow and heat transfer process in PRHR HX at the beginning of PRHR LOCA compared with typical CL LOCA. Besides, the natural circulation process between the core and steam generators (SGs) plays a critical role in cooling and depressurization of RCS, and its corresponding time-averaged heat transfer power is increased by about 30%. Besides, the asymmetric arrangement of PXS leads to a significant difference of transient thermal hydraulic state between the RCS branches, namely the PRHR cools the coolant via one RCS loop while two core makeup tanks (CMTs) inject the cold water to the core via the other RCS loop, and the pipeline resistance distribution shows a significant impact on the injection performance of safety equipment with low driven head such as CMTs.ConclusionsThe unique and important thermal hydraulic phenomena in the early stage of the accident, namely reverse flow and heat transfer process in PRHR HX and natural circulation process between core and SGs, are identified. The asymmetric arrangement effect will be more noticeable when the break occurs in PRHR pipeline.

BackgroundThe triple-to-double coincidence ratio-?erenkov (TDCR-?erenkov) method can be applied to the activity measurement of radionuclides by detecting the ?erenkov photons produced in a non-scintillation solution. The computation of the detection efficiency of this method is based on the premise that the energy of emitted β is completely deposited in the solution. However, this precondition is ideal and does not apply to the actual measurement because of the counting loss caused by the restrictions of a finite solution and the wall of the counting vial (i.e., wall effect).PurposeThis study aims to analyze the influence of the wall effect on the computation of detection efficiency.MethodsThe transport process of emitted β from the solution to the vial wall was analyzed in sections. Thereby the relationship between the number of ?erenkov photons and the deposition energy spectra of emitted β with different energies in different matrices was obtained. This relationship was used to further improve the calculation model of the TDCR-?erenkov method. Subsequently, the calculation model was simplified to reduce the required time. Geant4 calculated the deposition spectra of emitted β in different matrices, subsequently, the efficiency of different nuclides was calculated using curves of the number of ?erenkov photons vs. the energy of the emitted β. To verify the accuracy of the improved calculation model, measurements were carried out on a variety of pure β-emitters.ResultsThe results derived from the improved TDCR-?erenkov method are in good agreement with those of the TDCR-LS method. Especially for high energy β-emitters, the relative deviation of the results between the TDCR-?erenkov and TDCR-LS methods is reduced from 0.47% for the original method to 0.02% (90Y), and 0.64% to -0.16% (32P).ConclusionsThe TDCR-?erenkov method is more accurate when considering the wall effect in the activity measurement of high-energy β-emitters.

BackgroundThe China Spallation Neutron Source (CSNS) is a multidisciplinary research platform. Its high-energy 1.6 GeV proton beam serves various applications in aerospace devices and particle detector testing. However, certain irradiation applications and high-performance detectors require different beam energies. A degrader was designed to adjust the proton energy to a desired range.PurposeThis study presents a reasonable degrader scheme for the 1.6 GeV proton test beam at the CSNS.MethodsThe physical process of the 1.6 GeV high-energy proton beam passing through a degrader made of either of three different materials (iron, copper, and tungsten) was simulated using FLUKA, a Monte Carlo particle transport code. Parameters such as the degrader thickness, the energy deposition, the outgoing proton beam intensity, and the irradiation dose were determined through simulations. The optimal degrader material was identified. In addition, a continuously adjustable structure of the degrader was given.ResultsIron displays slight advantages in terms of energy deposition and radiation dose distribution, compared to copper and tungsten. Furthermore, the phase-space distribution of the outgoing proton beam and the secondary pion beam were also given, providing important references for future beam-line design.ConclusionAn optimal degrader structure made of iron for the CSNS high-energy proton beam is proposed. The secondary pion test beam is also feasible at the proton test end station. This is a significant development for future engineering design.

BackgroundThe performance of solid oxide fuel cells (SOFCs) can be promoted by optimizing cathode materials.PurposeThis study aims to boost the electrochemical performances of cathodes for SOFCs by doping transition metal at the B-site of double perovskite.MethodsFirstly, a series of B-site doped PrBa0.8Ca0.2Co2O5+δ(PBCC) oxides as cathodes for SOFCs were prepared by sol-gel. The effects of B-site doped content and doped elements on the crystalline structure of the cathodes were analyzed by X-ray diffraction (XRD) and scanning electron microscope (SEM). Then, the trends of conductivity and thermal expansion coefficient with B-site doped PBCC oxides were investigated. Finally, the electrochemical performances of cathodes with different B-site doped PBCC oxides were tested to find optimal doping element type and content.ResultsTest results show that polarization is reduced and the electrochemical catalytic activity is improved when 5 mol% of Fe is doped on the B-site of the PBCC cathode. Compared to the PBCC cathode, the max power density of the full cell with a 5-mol% Fe-doped cathode increases from 988 mW?cm-2 to 1 259 mW?cm-2 at 700 ℃.ConclusionsThe electrochemical performances of SOFCs can be boosted by modifying the B-site of double perovskite using transition metal.

BackgroundDouble perovskites have become a research hotspot in recent years due to their flexible structure, easy doping, and good thermal stability. Photoluminescence (PL) of rare-earth-doped double perovskite materials has been frequently reported, but few studies on thermoluminescence (TL) have been conducted.PurposeThis study aims to investigate the TL characteristics of Y2-x-yBixEuyMgTiO6 (0≤x<1, 0≤y<1) phosphors.MethodBi3+ and Eu3+ co-doped Y2MgTiO6 samples were synthesized by a high-temperature solid phase method, and the X-ray diffraction (XRD), PL, and TL of the samples were measured.ResultsXRD analysis results show that the crystal structures of all samples are monoclinic P21/n, and Bi3+ and Eu3+ are doped into Y2MgTiO6 by substituting Y3+. The PL results show that Y1.79Bi0.01Eu0.20MgTiO6 has a strong red emission near 620 nm (corresponding to the 5D0→7F2 transition of Eu3+), which is accompanied by a long afterglow. The TL curves of the samples doped with different concentrations of Bi and Eu ions show that Y1.79Bi0.01Eu0.20MgTiO6 has the highest TL sensitivity, and the samples exhibits two significant TL peaks near 510 K and 610 K. The TL spectrum is more abundant than the fluorescence spectrum, and the 5D0→7FJ (J =1,2,3,4) transition of Eu3+ can be observed. The TL intensity of the sample has a good linear relationship with the irradiation dose in the range of 2~1 000 Gy. The TL kinetic parameters of the samples are analyzed using two methods under different preheating temperatures (Tm and Tstop) and glow curve deconvolution. The analysis results show that the depth of the TL trap in the sample extends from 0.80 eV to 1.40 eV.ConclusionsThe results of this study indicate that the TL spectrum is richer than the PL spectrum and that Y1.79Bi0.01Eu0.20MgTiO6 may be used as TL dosimeter material for large dose detection.

BackgroundThe accurate acquisition and correlation coincidence of positron annihilation signal form the basis of the lifetime spectrum sensitive characterization of microscopic defects in materials. The complex radiation background interferes with the acquisition of positron annihilation signals, particularly in the study of neutron radiation damage of nuclear structural materials. The γ ray background generated by radionuclides induced by neutron activation affects the measurement results of positron lifetime spectrometer.PurposeThis study aims to investigate the effect of γ background on positron annihilation lifetime measurement.MethodsFirst of all, the positron lifetime measurement system is built in a "fast-fast coincidence" manner, and radiation background simulation experiments are designed by selecting two typical nuclides, 60Co and 137Cs sources, with nearby feature γ photon energy for measuring positron annihilation lifetime. Then, the spectra under two typical activity ratios are compared with the activated neutron-irradiated samples.ResultsThe simulation results indicate that the double high energy γ rays generated by 60Co sources are the primary factors affecting the spectrum shape and lifetime components. When the 60Co/22Na activity ratio is relatively low, 1.9, the peak-to-valley ratio of the spectrum significantly degrades, with the increase of random coincidence probability caused by radiation background. Further, at high activity ratio of 3.3, besides random coincidence, the false coincidence probability increases sharply, and the spectral shape is evidently distorted. For neutron-irradiated RPV steel, the lifetime value is reduced by 17% and 46% at low and high activity, respectively, compared with the non-irradiated samples.ConclusionsUsing the simulation method of radiation background sources and the influence rules of interference γ in this study, new techniques for eliminating γ background could be further explored in positron annihilation lifetime measurement.

During the neutron detection process, owing to the effects of inelastic scattering and slow neutron capture, a neutron-gamma mixed radiation field is formed, which increases the complexity of neutron detection. Organic scintillators are widely used in neutron detection because of their high flashing efficiency, short decay time, and high detection efficiency. Pulse shape discrimination (PSD) is a key technology for discriminating neutrons and gamma rays according to the difference in pulse shape caused by the difference in particle decay time in organic scintillators. Traditional PSD methods include time-domain and frequency-domain discrimination methods. In recent years, various machine-learning techniques applied to neutron-gamma discrimination have achieved better results. To better use organic scintillators and the corresponding neutron-gamma discrimination methods in neutron detection, we conducted a comprehensive analysis of the glowing mechanism of organic scintillators, PSD principle, organic scintillator types, and neutron-gamma discrimination methods and investigated the performance evaluation indexes of organic scintillators and neutron-gamma discrimination methods. Finally, the future development directions of organic scintillators and neutron-gamma discrimination methods were examined.

Large volumes of water containing tritium are generated during the operation, decommissioning, and incident-management of nuclear installations and related facilities, and are expected to increase with the ongoing expansion of nuclear power generation. If released into the environment, this water could pose a substantial environmental threat to living organisms. However, conventional isotope separation techniques, such as cryogenic distillation and catalytic exchange, are inadequate for efficiently isolating significant quantities of low-level tritiated water because of the complex machinery and excessive energy required, and the potential for hydrogen-gas detonation. In contrast, water distillation (WD), as a traditional technology, has the unique advantages of simple operation, no corrosive and toxic substances, and no hydrogen explosion risk, but problems of small separation coefficient and low efficiency always exist in this technique. However, the separation effect of WD can be effectively improved by improving the process variables in the process of WD, such as temperature, distillation column diameter, and packing dimensions, so as to adapt WD to the separation of industrial tritiated water. This study provides a thorough exposition of the basic principles and distinctive features of water distillation and examines the effect of various operational parameters on separation efficiency to adapt the process for the industrial separation of tritiated water. The impact of various process variables on the separation efficiency of tritiated water via distillation was investigated, and results show that optimizing these variables can markedly improve the separation efficiency of water distillation. In particular, decreasing the dimensions of the packing material or altering its properties can lead to higher separation factors and lower residual tritiated water concentration. These findings suggest that water distillation can be used for the separation of tritiated water. By optimizing its operational parameters, water distillation can become a viable method for the industrial separation of tritiated water and is expected to play a significant role in this field in the future.

BackgroundFeCrAl alloy cladding, as an accident tolerant fuel (ATF) mid-term commercial technology approach, has received extensive attention.PurposeThis study aims to investigate the effect of trace Y on the internal pressure burst and oxidation properties of FeCrAl alloy cladding.MethodsFirstly, the crystalline grain size and micro-morphologies of FeCrAl and FeCrAlY alloy cladding samples were observed by optical microscope. Internal pressure burst and high temperature oxidation tests were carried out by burst test equipment and thermo-gravimetric analyzer with a moisture generator. Then, X-ray diffractometry (XRD), scanning electron microscope (SEM) and energy dispersive spectrometer (EDS) were employed to analyze the composition of oxidation products, surface and cross-sectional micro-morphologies of FeCrAl and FeCrAlY alloy cladding samples before and after high-temperature stream oxidation and the distribution of elements on the surface oxidation products.ResultsThe results show that trace Y is mainly dissolved in the FeCrAl alloy matrix, and no Fe-Y phase is formed. The inclusion of Y do not change the burst strength and the rupture opening morphology at room temperature (RT) to 1 000 ℃, and the high-temperature steam oxidation resistance of FeCrAl alloy cladding is significantly improved by the trace Y. Under the condition of steam oxidation at 800 ℃, 1 000 ℃ and 1 200 °C for 8 h, the oxidation weight gain of FeCrAlY alloy cladding decrease by 65.1%, 60.0% and 31.5%, respectively. Compared with the single Al2O3 oxide film on the surface of FeCrAl alloy cladding, the Y-containing composite oxide film with lower internal stress, higher compactness and better adhesion with the substrate is formed on the surface of FeCrAlY alloy cladding.ConclusionsTherefore, the addition of trace Y do not change the burst properties of FeCrAl alloy cladding, however, the high-temperature steam oxidation resistance of FeCrAl alloy cladding is significantly improved.

BackgroundThe trend towards increasingly narrow apertures in multipole magnets poses a challenge to many conventional measurement methods. Consequently, these methods' applicability in small aperture multipole magnets is limited. However, the single stretched wire measurement technique has emerged as a promising alternative due to its minimal space requirements within the measurement domain. Therefore, this technique is well-suited for accurately measuring magnetic fields in small aperture magnets.PurposeThis study aims to introduce a novel approach for analyzing the gradient integral and multipole errors of the quadrupole magnet, to address the limitations associated with the current single stretched wire method (SSWM).MethodsFirstly, a magnetic measurement system based on the single stretched wire method was constructed with two boasted key advantages: minimal space occupation within the measurement domain, and flexible motion modes. Then, leveraging these features, measurements of the four poles of a quadrupole magnet by employing a hyperbolic trajectory was acquired, and a new technique for analyzing both the gradient integral and multipole errors associated with the quadrupole magnet was developed. Finally, the feasibility of this SSWM was verified by comparing the results obtained from our system to those derived from the rotating coil method.ResultsMeasurement results of a quadrupole magnet with the inscribed radius of 11 mm and gradient of 100 T?m-1 by SSWM show that the repeatability of three measurements is better than ±1.5×10-4 which is less than one-third of the maximum value of multipole error of 5×10-4, so it can meet the measurement requirements.ConclusionsThe methodology outlined in this study for constructing the measurement system and analyzing the resultant data offers a practical and effective solution for the future magnetic field measurements of small aperture magnets.

BackgroundShanghai High repetition rate XFEL (X-ray free electron laser) and Extreme light facility (SHINE) is a high-repetition-rate X-ray-free electron laser. The timing system of the beamlines and endstations must provide high-precision bunch IDs and a timing trigger for the equipment that works in single-pulse mode.PurposeThis study aims to design a data acquisition (DAQ) testing system to simultaneously acquire X-ray bunch IDs with their corresponding detector data package for subsequent data processing.MethodsThis DAQ testing system was developed on the Zynq UltraScale+ system-on-chip (SOC), and the White Rabbit protocol was employed for the timing system environment. A Bunch ID obtained from the FPGA mezzanine card (FMC) of the embedded White Rabbit node (WRN) was transferred to the server using a TCP protocol stack built on LwIP (light weight internet protocol). Finally, a Basler camera was employed to test this DAQ system, in which the pypylon library was applied to raw data acquisition software for camera snapshot whilst two channels of data were collected by an upper computer and saved to a database for comparison.Results & ConclusionsThe number of bunch IDs obtained by this acquisition test system is the same as that of image frames taken by Basler camera, which demonstrates that the testing system can satisfy the requirements of bunch ID acquisition in SHINE beamlines and endstations.

BackgroundNarrow rectangular channels are widely used in various fields because of their compact structure and other advantages.PurposeThis study aims to improve the prediction method of critical heat flux (CHF) in narrow rectangular channels for reactor safety and economy by conducting CHF visualization tests in narrow rectangular channels with different gap size to explore the CHF triggering mechanism.MethodsFirstly, a high-temperature and high-pressure experimental loop with narrow rectangular channels was built, and the visualisation video and thermal-hydraulic data were collected simultaneously. It was found that the flow patterns correspond to bubble flow, slug flow, churn flow and annular flow when CHF occurs with the gap size of 5 mm, 3 mm, 2 mm and 1 mm, respectively.ResultsBefore the occurrence of CHF, bubble flow, slug flow and churn flow experience temperature fluctuations. In the annular flow, the CHF involves a gradual expansion of the area from the initial dry spot; in the churn flow, the CHF covers a smaller area; while the slug flow affected the widest area; in the bubble flow, the temperature fluctuations at the heating wall are the most frequent. Furthermore, when the system pressure is in the range of 1?4 MPa and the gap size is 1 mm, there is a non-linear relationship between the system pressure and the CHF, while in the other channels the CHF increases as the system pressure increases.ConclusionsThe narrow gap size has a very important effect on CHF in narrow rectangular channels, and the findings of this paper can lay the foundation for the establishment of a CHF mechanism model in narrow rectangular channel.

BackgroundAs an innovative nuclear fuel assembly, the helical cruciform fuel (HCF) assembly has the characteristics of large specific heat transfer area, short heat conduction path, strong inter-channel mixing and free from the grid spacers. Compared with the traditional cylindrical fuel assembly, the HCF assembly can raise the core power density with compromise on the safety margin. However, the concentrated stress might take place at the location of self-support points, resulting in the plastic deformation and even rupture.PurposeThis study aims to analyze the thermal-mechanical behaviors of HCF bundle under steady conditions and accident transitions, so as to obtain the stress and strain of HCF rods, based on which, the integrity of fuel cladding was assessed.MethodsFirstly, a 3×3 typical HCF geometrical assembly model without four rods in corners was constructed and discretized by hexahedral mesh. Then, the steady and transient convective conditions were applied to the outer surfaces of rods to simulate the various working conditions, including single phase, boiling, reactivity insertion accident and loss of coolant accident. Finally, the governing equations for mechanics and heat transfer were established and solved in ANSYS using the thermal and mechanical modules.ResultsThe results show that, the maximum von Mises stress and plastic deformation take place at the location where adjacent rods contact, where the stress and strain are determined by both the contact constrain condition and the temperature difference between cladding inner and outer surfaces. However, at the elbow of the blades, the stress and strain are mainly affected by the radial temperature gradient in the cladding material. For the cladding, the plastic deformation is larger while the von Mises stress is smaller under the flow boiling condition compared with these under the single-phase cooling condition. Furthermore, the integrity of fuel cladding can be maintained under the conditions of reactivity insertion and loss of coolant accidents, where the stress and the temperature are lower than the break limit and the zirconium-water reaction temperature, respectively.ConclusionsFrom the thermal-mechanical analysis on the HCF assembly, this kind of innovative fuel assembly shows good mechanical performance under normal and accidental conditions.

PurposeThis study aims to improve the low accuracy of the aerosol model in the ISAA code by developing high-precision natural deposition model of aerosol in the containment.MethodsFirstly, the aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Then, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved respectively. Finally, AHMED (Aerosol and Heat Transfer Measurement Device), ABCOVE (Aerosol Behavior Code Validation and Evaluation) and LACE (Light Water Reactor Aerosoal Containment Experiments) experiments were employed to validate and evaluate the improved ISAA code.ResultsCalculation results show that the improved model is applicable to more accurate simulation of the peak aerosol mass and responding to the influence of the containment pressure and temperature on the natural deposition rate of aerosols, and the calculation accuracy of the residual mass of aerosols in the containment is significantly improved simultaneously.ConclusionsThe performance of improved ISAA with high-precision aerosol models of this study meets the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.BackgroudNuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle to the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity.

BackgroundThe numerical results from a neutron source, as the important input parameter for the transport calculation, directly affect the accuracy of shielding calculations for reactor. The apparent differences between core sources are related to their geometric model, burnup, and power distribution.PurposeThis study aims to improve the calculation accuracy of the core neutron source for nuclear reactor shielding.MethodsFirstly, the geometric weight of each component in the core was generated by analyzing the characteristics of the radial source distribution on the basis of neutron importance, and the fine source mesh calculation was conducted for the peripheral components with high geometric weight and the region with a large power gradient. Then, a layered approach for different axial height positions was employed to reduce the influence of the axial power peak factor to achieve stable transport calculation results, and the source and geometry meshes were mapped according to the volume weight method to ensure the conservation of total source. Finally, the NUREG/CR-6115 core as a benchmark model was used for numerical verification.ResultsNumerical verification results indicate that, compared with the average source calculation, the multi-weight source mesh mapping algorithm reduces the root mean square of the relative error in the fast neutron fluence by 18.46% between the transport calculation results and the reference value.ConclusionsThe multi-weight source mesh mapping algorithm can be employed to obtain an accurate source distribution, improve the accuracy of the shielding calculation, and satisfy the requirements of engineering applications.

BackgroundLiquid molten salt reactor has many features such as high economy, safety and on-line fuel processing. The emergency draining salt passive residual heat removal system (EDS-PRHRS) is a unique residual heat removal system design for liquid fuel molten salt reactor, in which safely export residual heat of the molten salt in the salt draining tank is the first requirement for EDS-PRHRS design.PurposeThis study aims to analyze the transient characteristics of EDS-PRHRS salt discharge tank during operation by simulation.MethodsFirstly, the accident analysis of the passive residual heat removal system was carried out. The peak temperature of the molten salt was mainly found in the full heat discharge phase of the salt discharge tank. Then, a computational model of the molten salt coupled to the heat exchanger element was established for this stage of the discharge tank and numerical simulations were carried out by using computational fluid dynamics (CFD) analysis software Fluent. The Mixture model was used to simulate the boiling heat exchange of water in the heat exchanger element. Finally, different parameter sensitivity analysis scenarios were designed to investigate the effect on the transient.ResultsThe analysis results show that the heat exchange power of the heat exchange element gradually decreases with time, and the temperatures of the outer wall and the hot spot of molten salt have a peak with time.ConclusionsBy increasing the axial height of the thimble and enhancing the emissivity of the air gap layer, the temperature peak can be significantly reduced, and the peak value can be slightly reduced by delaying the salt discharge time. In addition, the triangular arrangement of can delay the local solidification time. The study results can provide some reference for EDS-PRHRS design.

BackgroundThe gas-cooled fast reactor (GFR) has great advantages of finance and sustainability which combines the features of high temperature gas-cooled reactor and fast reactor. However, safety issue has become the main challenge in the development of GFR due to the high temperature and high neutron flux in the GFR core. Coated particle fuel (CPF) has been widely used in high temperature reactor (HTR) due to the excellent high temperature tolerance.PurposeIn order to strengthen the safety property in GFR, this paper puts forward a block-type fuel assembly (FA) model based on CPF. Based on the FA model, neutronics analysis and thermal hydraulics validation is carried out to verify the rationality of the design.MethodsMonte Carlo method is used in the calculation. Physical parameters including plutonium fraction in the U-Pu mixture fuel, diameter of fuel pins/coolant channels, the number of coolant channels, pitch-to-diameter ratio, thickness of cladding and thickness of assembly wrapper were selected and sensitivity analysis were conducted on the FA property to these parameters.ResultsAnalysis results show that among the above six parameters, plutonium fraction and pitch-to-diameter ration have the most obvious effect on the neutronic property and the number of the coolant channels mainly influences the power distribution of the GFR FA. Finally, temperature distribution of the FA is calculated using single channel model under a low coolant fraction and requirements in terms of thermal-hydraulic property are put forward for the FA parameters.ConclusionsThe block-type FA model put forward in this paper meets the design requirements well. The research conclusion of this paper provides reference for the future study on GFR nuclear design.

BackgroundTo measure the beam positions of High Energy Photon Source (HEPS), different types of beam position monitors (BPMs) have been developed. The position sensitivity coefficient is an important parameter of BPMs by which the position of the beam can be calculated.PurposeThis study aims to establish a method for calculating the position sensitivity coefficient of BPMs.MethodsThe position sensitivity coefficients of various types of BPMs, such as round, elliptical, and octagonal pipes, were determined by using the boundary element method (BEM). The azimuth button angles in the elliptical BPM of the HEPS booster and the button distances in an octagonal BPM on the Beijing Electron Positron Collider II (BEPCII) storage ring were derivated by the application of BEM. Furthermore, the position sensitivity mappings of the BPMs was calculated.ResultsThe difference in sensitivity results of the round BPM calculated by the BEM and the analytical value is approximately 1%. The error between the calculated and experimental measurement results of the position sensitivity coefficients of the elliptical and octagonal sections is approximately 2%.ConclusionsThe BEM is a reliable method for calculating the position sensitivity coefficient of BPMs, which can be used in BPM design.

BackgroundThe Gaussian pulse shaping algorithm has the advantages of high signal-to-noise ratio and low ballistic deficit. Therefore, the radiation detector output signal is often shaped to a Gaussian waveform in the actual nuclear radiation measurement system even if the signal is more likely to be a dual exponential signal.PurposeThis study aims at gaussian pulse shaping algorithm for dual exponential nuclear signals based on wavelet transform.MethodsBased on the simulated nuclear pulse signal, the influence of the shaping parameters on the pulse shape and the filtering performance of the shaped pulse was investigated. A FAST-SDD detector was used to acquire the X-ray fluorescence signals emitted by a standard manganese sample. The measured nuclear signals were processed by Gaussian pulse shaping and trapezoidal pulse shaping algorithms respectively before generating energy spectrum. The performance of the two shaping algorithms on filtering and pile-up pulse separation were compared by using the full width at half maximum and the area of the 5.89 keV peak.Results & ConclusionThe comparison results show that the best energy resolutions corresponding to Gaussian and trapezoidal pulse shaping algorithms are achieved when the peaking time ranges from 3.2 μs to 6.4 μs, and the difference between two algorithms is less than 5 eV. Besides, the Gaussian pulse shaping algorithm performs better than trapezoidal pulse shaping algorithm on pile-up pulse separation with the same peaking times.

BackgroundA CdZnTe (CZT) detector is a compound semiconductor detector with a high atomic number and high detection efficiency, it can be used at room temperatures to detect short wavelength radiation such as X-ray and γ ray.PurposeThis study aims to investigate the factors affecting the energy spectrum characteristics of the CZT detector.MethodsThe geometric model of the detector was established by using Geant4 software, and the intrinsic detection efficiency and absorption rate of CZT crystal in the planar size of 10 cm×10 cm were simulated. The charge collection efficiency of the crystal was calculated using the Hecht formula and the γ-ray energy spectrum was obtained by collecting the deposition energy and position information in the crystal. By analyzing the physical properties of crystals, the impact of physical properties on detector performance was explored.ResultsSimulation results show that incomplete charge collection significantly influences the spectral performance of the detector. When the γ ray energy is less than 50 keV, the spectrum is not affected by hole wake whilst the influence of hole wake is more obvious when the energy is between 50 keV and 100 keV. The energy spectrum is gradually aggravated by the influence of hole wake when the γ ray energy is above 100 keV.ConclusionsThis effect of hole wake for CZT detector can be reduced by increasing the bias voltage, but the increased bias voltage shifts the spectrum's peak, and the shift amount is determined by the maximum charge collection efficiency of the crystal.

BackgroundAerosol particle size of radon progeny is the key parameter of the radiation dose conversion coefficient in radon progeny. It is necessary to develop a measuring device for the aerosol particle size of radioactive aerosol to measure the aerosol particle size distribution of environmental radon progeny. Inertial impactor is a kind of widely used particulate classification sampler.PurposeThis paper aims to design and implement an impactor applicable to radon progeny aerosol with a cutting size of 1 μm.MethodsFirst of all, several kinds of inertial impactors were analyzed on the basis of aerodynamic theory, the design parameters of the impaction sampler structure, such as the diameter of the collecting plate, the distance between the collecting plate and the inner wall, the distance between the nozzle and the collecting plate, the height of the nozzle, were simulated by using computational fluid dynamics (CFD) analysis software Fluent and discrete phase model. Then, based on simulation results, a set of optimized design parameters were obtained and a porous impingent sampler was implemented for radon progeny aerosol. Finally, this impactor was calibrated by a GRIMM11-D aerodynamics particle size analyzer in a laboratory.ResultsThe optimized design parameters show that the nozzle distance D, the nozzle height T, the distance S from the nozzle to the collecting plate, and the nozzle diameter W have relationship of D/W=1.5~3.5, T/W=1~5, S/W=1. The experimental calibration results of designed porous impingent sampler are basically consistent with that of CFD numerical simulation with dp50=(1±0.07) μm, σg1=1.33, σg2=1.35, and the cutting particle size of the impactors meets the practical application requirements.ConclusionsThis paper focuses on the design of an impactor sampler. Through simulation and comparison tests with ELPI+ instrument, the effective cutting of 1 μm particle size is realized, which provides convenience and ideas for the further optimization design and online particle size fractional measurement of radioactive aerosol.

BackgroundUnder high-temperature operating conditions, the tritium would be generated inside the core of thorium-based molten salt reactor (TMSR) and probably diffuse through the structural material into the environment. Establishing an Al2O3/Ni-Al composite tritium permeation barrier coating may help address this issue.PurposeThis study aims to explore the optimal preparation process, especially the in-situ oxidation process.MethodsThe Al2O3/Ni-Al composite coating was prepared on the surface of GH3535 alloy by pack cementation aluminizing (PCA) followed by vacuum in-situ oxidation, and the effects of oxidation temperature and vacuum on the microstructure of Al2O3 films were analyzed by experiments. Grazing incidence X-ray diffraction (GIXRD), scanning electron microscopy (SEM), and transmission electron microscope (TEM), X-ray energy dispersive spectroscopy (EDS) were used to characterize the phase composition and crystal structure of the alumina film, as well as morphologies of the surface and cross-section.ResultsThe experimental results show that the low oxygen partial pressure can increase the forming temperature of alumina film, but can form a more compact film with flat surface. Higher oxidation temperature is conducive to the formation of thicker alumina films, but also greatly increases the surface defects.ConclusionsBy in-situ oxidation process at 1.2 Pa-850 ℃-72 h, alumina thin films with good properties can be obtained on the surface of GH3535 alloy: The phase of film contains γ and α, the thickness is about 0.8 μm, and the surface is compact without defects.

BackgroundThe SXFEL (Soft X-ray Free-Electron Laser facility) is the first X-ray coherent light source in China. To monitor the beam loss in its undulator beamline, a quartz fiber beam loss monitoring system based on the Cherenkov radiation principle was designed and installed. The quartz fiber is insensitive to high-energy gamma rays, making it suitable for a strong SXFEL radiation field environment.PurposeThis study aims to apply quartz fiber beam loss monitoring (BLM) system to the undulator beamline of SXFEL, and carry out position calibration experiment to measure the fiber attenuation coefficient, and performance of the system in the beam tuning period.MethodsFirst, two pure quartz composition fibers with 400 μm inner diameter of core and high concentration of hydroxide ions were employed. The beam loss signal was generated by falling YAG (Ce:Y3Al5O12, target film) of the beam profile monitor at a fixed position and adjusting the trigger time delay to make the position of the beam loss signal the same as that of the beam profile monitor for position calibration experiment. Second, to measure the fiber attenuation coefficient, the coefficient was fitted by bringing the peak value of the beam loss signal generated by the falling YAG at different positions of the SBP (Shanghai-XFEL Beamline Project) beamline and the corresponding fiber position into the signal attenuation formula.ResultsThe fiber BLM can accurately reflect the position of the beam loss with upstream position resolution of approximate 0.2 m in the experiment test, as well as in the period of beam tuning. The refractive index of quazrtz firber core is approximately 1.5, hence the relationship between the beam loss position and signal arrival to upstream PMT time interval is 0.12 m·ns-1. The measured fiber attenuation coefficient is around 74 dB·km-1, which is consistent with the calculation result and similar to the measurement result of SPring-8 Angstrom Compact Free Electron Laser (SACLA) using the same type of optical fiber.ConclusionsThe fiber beam loss monitoring system has a good position resolution and has the potential to meet the requirements of SXFEL beam tuning.

BackgroundSilver nanoclusters, being a novel variety of nanomaterial, have garnered significant attention owing to their exceedingly minute dimensions, and distinct physical and chemical characteristics.PurposeThis study aims to present a straightforward and efficient method for fabricating silver nanoclusters composites using radiation technology.MethodsFirstly, silver nanoclusters in aqueous solution were directly synthesized through radiation reduction. By means of radiation grafting technique, polyacrylic acid templates was grafted onto an array of matrix materials, thereby producing solid templates. Subsequently, these solid templates were employed to achieve in situ synthesis of silver nanoclusters composites, obviating the need for water-soluble template materials. Finally, the fluorescence detection performance and catalytic performance of silver nanoclusters were tested by fluorescence spectrometer and the UV visible spectrum.ResultsThe silver nanoclusters and composites prepared in this study have retained the photoluminescence and catalytic activity characteristic of silver nanoclusters, thereby presenting potential applications in metal ion detection and catalytic degradation of 4-nitrophenol. Furthermore, it is noteworthy that the combination of the base material and silver nanoclusters is capable of manifesting a synergistic effect, thereby enhancing the overall performance of silver nanoclusters.ConclusionsThe utilization of radiation technique has enabled a simplified route of silver nanocluster composites. In addition, the versatility of this synthesis route extends across a variety of matrix materials, thereby broadening the scope of potential applications for silver nanocluster composites.

BackgroundThe annihilation radiation exhibits conspicuous features in the low-energy segment of the orbital gamma spectrum, which contains a substantial amount of geological information about the lunar surface. The fluence rate can reflect the element composition, density, maturity, and other characteristics directly.PurposeIn order to further clarify the primary source and influencing mechanism of annihilation radiation on the lunar surface.MethodsA quantitative model for the annihilation radiation characteristic peak of orbital gamma spectrometers was established. The gamma rays induced by protons of varying energies were simulated using GEANT4 to further understand the primary source and mechanism of annihilation radiation on the lunar surface. The data from the "Chang'e-1" high-energy particle detector (CE1-HPD) was used as the input term, and the annihilation radiation characteristic information induced by 4~400 MeV protons galactic cosmic rays interacting with five typical rocks was calculated. After subtracting the 0.511 MeV characteristic peak collected by the "Chang'e-1" gamma spectrometer (CE1-GRS) from natural radioactivity, the results were compared with the annihilation radiation fluence rate induced by 4~400 MeV protons galactic cosmic rays.ResultsThe results indicate that the rock's composition have a negligible effect on the annihilation radiation. The probability of cascading shower generating annihilation radiation is directly proportional to the incident proton energy. Additionally, the contribution of 4~400 MeV protons to the annihilation radiation present in the orbital gamma spectrum is relatively low, only (1.97±0.66)×10-4 .ConclusionsThe established model has proven to be accurate in reflecting the related characteristics of the gamma radiation field on the lunar surface and can be used for quantitative analysis of annihilation radiation. The results indicate that the contribution of 4~400 MeV protons galactic cosmic rays to the annihilation radiation present in the orbital gamma spectrum is minimal.

BackgroundHEPS (High Energy Photon Source) needs to control the beam orbit change within 10% of the cluster size within a certain frequency. In order to meet the beam orbit stability requirements in HEPS, it is necessary to establish a fast orbit feedback (FOFB) system.PurposeThis study aims to design and implement an effective feedback bandwidth of FOFB system that is greater than 500 Hz, and the delay of the whole system is less than 160 μs.MethodsBased on this requirement, a two-layer communication of loop centralized computing system network topology was designed and implemented for FOFB system of HEPS. And on this basis, the FPGA (field-programmable gate array) firmware algorithm of the signaling pathway of FOFB system was realized, including beam position acquisition, loop data transmission, FOFB algorithm, power control interface and the testing logic.ResultsThe measurement and analysis results show that each module in the data transmission link of the FOFB system can be used normally, and the total delay time of the system is about 140.46 μs, which has reached the intended design target.ConclusionsThe FOFB design of this study lays a foundation for the future construction, optimization and debugging of FOFB system on HEPS storage ring with good flexibility and scalability, providing a feasible solution for the future establishment of fast orbit feedback system in other storage rings.

BackgroundEthylene is an important raw material in petrochemical industry. Semi-hydrogenation of acetylene in an ethylene is an industrially important process. Conventional supported monometallic Pd catalysts offer high acetylene conversion, but they suffer from very low selectivity to ethylene due to over-hydrogenation.PurposeThis study aims to prepare a catalyst with high acetylene conversion and simultaneous selectivity to ethylene, surpassing conventional Pd catalysts, and explore the structure activity relationship of palladium-bismuth catalyst in acetylene hydrogenation.MethodsFirstly, PdBi/SiO2 catalyst was synthesized via a deposition-precipitation method for industrial hydrogenation of acetylene to ethylene. Then, comparison of catalytic activity and selectivity with traditional catalysts in the semi hydrogenation reaction of acetylene was conducted. Finally, the X-ray Absorption Fine Structure (XAFS), High-Angle Annular Dark-Field Scanning Transmission Electron Microscopy (HAADF-STEM), and X-ray Energy Dispersive Spectroscopy (EDS) were employed to explore the reaction mechanism.ResultsCompared with the Pd catalyst, PdBi/SiO2 catalyst exhibits increased reactivity at a lower temperature, with 100% acetylene conversion and 90% selectivity.ConclusionsThe Pd-Bi alloys structure is confirmed to effectively inhibit the formation of PdHx, weaken the cracking rate of hydrogen and the adsorption of ethylene on palladium surface, and inhibit the excessive hydrogenation of ethylene to produce by-product ethane. The simple synthesis PdBi structure provides new ideas and insights for industrial catalysts.

The use of the relativistic heavy ion collision experiment has extended our insights into the diverse possibilities available to a truly strongly-interacting system. The main goal of this experiment is to describe the properties of the different phases of quantum chromodynamics (QCD) and to chart the QCD phase diagram on the T-mu plane. For the phase diagram, apart from the general phase boundary lines, some specific characteristics such as the possible critical endpoint (CEP), associated coexistence region, and strongly-coupling quark-gluon plasma (sQGP) have to be identified. Here, the CEP separates the first-order phase transition from the second-order transition (or crossover) when the case beyond the chiral limit is considered. However, convincing signals have not yet been obtained using the relativistic heavy ion collider (RHIC) experiment. Theoretically, strong interaction systems hold significant features: asymptotic freedom in the ultraviolet region, dynamical chiral symmetry breaking, and confinement in the infrared region. Such features can be uniformly displayed in the phase structure of the matter in the temperature T and chemical potential planes. Consequently, several investigations have been experimentally and theoretically performed. However, the strong coupling feature in the low-energy region prevents the use of perturbative calculation methods, which creates the need for the development of nonperturbative approaches. Additionally, lattice QCD simulations have been widely implemented; however, the "sign problem" delays the progress in the large chemical potential region. Therefore, the Dyson-Schwinger equation (DSE) equation method and functional renormalization group approach, which inherently include both dynamical chiral symmetry breaking (DCSB) and confinement, play an important role. The QCD DSE approach is a method based on the continuum quantum field theory. The new criteria were proposed based on the DSE and studied using the deconfinement and Chiral symmetry restoration phase transition of QCD. Currently, functional methods can be used to provide a reliable estimation of the CEP location. First, reliability is achieved using a thorough investigation of the truncation of the DSE, state of the art truncation is then performed causing a converging result between the different methods, and the predication of the lattice simulation at low chemical potential is confirmed. The results show a fast convergence of the truncation owing to the infrared fixed point of the QCD coupling, which allows the capturing of the QCD running behavior using a finite set of two- and three-point Green functions. The estimated location of the CEP based on the current computation is μB at 600~650 MeV and T at 100~110 MeV. The existing functional QCD methods are non-perturbative continuum methods that are capable of simultaneously describing both the DCSB and confinement. Although they are limited by the truncations, the use of functional QCD approaches has resulted in progress in the study of the QCD phase structure and thermal properties, where a complete phase diagram and related thermal properties have been obtained in a large chemical potential range, which can provide a reference for the exploration of the QCD features. Most of the theoretical studies using effective models or certain truncations have observed the existence of the CEP; however, the determination of its location is still a work in progress because it varies based on the computation. Moreover, searching for QCD phase transition signals, particularly the CEP, is the main goals of current and future experimental programs on the relativistic heavy ion collider.

In high-energy heavy ion collisions, quarks and gluons are released from the colliding nucleus to form a new state of nuclear matter called deconfined quark gluon plasma (QGP). To study the transition from normal nuclear matter or hadron resonance gas to QGP, non-perturbative quantum chromodynamics (QCD) must be solved on supercomputers using the lattice numerical method (lattice Quantum Chromodynamics, lattice QCD). However, lattice QCD only works for zero and small baryon chemical potential regions that can be described by the Taylor expansion and provides the nuclear equation of state (EoS) and QCD transition in these regions. For large baryon chemical potential regions that cannot be described by the Taylor expansion, lattice QCD fails to provide the nuclear EoS and QCD transition owing to the famous sign problem. Machine learning helps to study the nuclear EoS and QCD phase transition. First, machine learning can determine the nuclear EoS and QCD transition using the momentum distribution of final state hadrons in heavy-ion collisions, with data from both heavy-ion collision experiments and relativistic hydrodynamic simulations. Second, it can contribute to the direct solution of the sign problem in lattice QCD. The present paper reviews the applications of machine learning to the study of the QCD phase transition in heavy-ion collisions. This study (1) introduces nuclear EoS and QCD transition as well as the difficulty of the lattice QCD method, (2) analyzes the nuclear EoS using Bayesian analysis, (3) identifies the nuclear EoS and QCD phase transition using different types of deep neural networks (e.g., convolutional neural network, point cloud network, and many-event averaging), (4) searches for critical self-similarity using a dynamical edge convolution-based graph neural network, (5) learns the quasi-particle mass using a physically informed network and auto-differentiation, (6) discards unphysical regions in the nuclear EoS with a critical endpoint using active learning, (7) discusses unsupervised learning for the nuclear liquid-gas phase transition, (8) determines the nuclear symmetry energy in heavy-ion collisions, (9) investigates Mach cones using deep learning assisted jet tomography, and (10) accelerates the sampling of lattice QCD configurations using a physically constrained neural network while solving the sign problem in lattice QCD using deep learning.

The RHIC-STAR (Relativistic Heavy Ion Collider-Solenoid Tracker at RHIC) experiments have measured the cumulants of net-proton (a proxy for net-baryon), net-charge, and net-kaon (proxy of net-strangeness) multiplicity distributions in Au+Au collisions at different centers of mass with energies ranging from 7.7 GeV to 200 GeV. Recent results have shown that the ratio of the fourth-order net-proton cumulant over the second-order one (κσ2) exhibits a nonmonotonic energy dependence. In relativistic heavy-ion collision experiments, only information about the final state particles can be measured. Therefore, we investigated the fluctuations of the conserved charges (baryon, electric charge, and strangeness) in Au+Au collisions using a multiphase transport (AMPT) model. This model can basically describe the results measured by the RHIC-STAR experiment. More importantly, the AMPT model is used to understand the key impacts of the dynamical evolution of relativistic heavy-ion collisions on fluctuations and correlation functions, including the creation and diffusion of conserved charges, hadronization, hadronic rescatterings, and weak decays. It was discovered that the correlation between positive and negative charges may originate from the string melting mechanism. Baryon (proton) correlation functions are consistent with the expectation of baryon number conservation. Net-strangeness (net-kaon) originates from pair production. We studied the correspondence between representative quantities and their conserved charges and found that their behaviors are qualitatively consistent yet quantitatively different. Although the physics of quantum chromodynamics (QCD) critical fluctuations is not included in the AMPT model, our results are expected to provide a baseline for the search of possible critical behavior at the QCD critical end point in relativistic heavy-ion collisions. We incorporated critical density fluctuations into the model and found that they play a role.

The searching for potential quantum chromodynamics (QCD) phase transition signals is a fundamental goal of on-going experiments on heavy-ion collisions, which is critical to understanding the properties of strongly interacting matter under extreme conditions, the inner structure of compact stars, gravitational waves emitted from neutron star mergers, etc. In particular, the beam energy scan program carried out at the Relativistic Heavy-Ion Collider (RHIC) provides a unique tool that enables studies into the QCD phase diagram and the conjectured QCD critical point. Besides the event-by-event fluctuation of conserved charges, which has been widely accepted as a useful study of the QCD critical point, the production of light nuclei can serve as a sensitive observable to the QCD phase transitions in high-energy heavy-ion collisions. The density fluctuation and correlation among nucleons are automatically encoded in the production of light nuclei in heavy-ion collisions. This study aims to demonstrate how to probe QCD phase transition with light nuclei production in heavy-ion collisions. The progress of studies in this area over the last few years is reviewed. The nucleon coalescence model provides a suitable tool for the study of the effects of density fluctuation/correlation on light nuclei production. A transport model based on the Nambu-Jona-Lasinio (NJL) model is developed to simulate the occurrence of the first-order chiral phase transition in heavy-ion collisions. Within the coalescence model for light nuclei production, the yield ratio NtNp/Nd2 of protons (p), deuterons(d), and tritons (t) is shown to be sensitive to the nucleon density fluctuation and correlation and can function as a good probe to the non-smooth QCD phase transition. The production of light nuclei in heavy-ion collisions encodes the information about baryon density fluctuations and correlations, and enhancements of the yield ratioNtNp/Nd2 could serve as an indicator for the occurrence of a first-order or second-order QCD phase transition.

Understanding the phase structures of strong interaction matter is an active frontier in nuclear physics research currently, and it will provide crucial insights into heavy-ion collision experiments as well as neutron star observations. Most studies in this area focus on the influence of extremely high temperatures and baryon densities on matter properties, especially pertaining to phase transitions such as the chiral symmetry breaking and color superconductivity. Recent experimental and theoretical studies reported that in non-central heavy-ion collisions, systems carry a large initial angular momentum that becomes very strong vorticity fields in the bulk fluid. This has thus introduced several new questions regarding the properties of strong interaction matter under vorticity fields, and have led to many novel results. Thermal field theory calculations based on rotating frame and mean-field approximation have been developed to study various phase transitions under rotation, such as chiral symmetry breaking, color superconductivity and superfluidity at high isospin asymmetry. The results have demonstrated important impacts of vorticity fields on the phase boundaries of these transitions, and have also revealed nontrivial new phase structures of strong interaction matter under rotation. A new dimension of the usual QCD phase diagram has been unveiled. The study of rotation-induced phase transition extends phase transition research to a broader space. There remain more unexplored issues that merit further study.

Exploring the quantum chromodynamics (QCD) phase diagram at finite bayron density regime through the beam energy scan (BES) program at the relativistic heavy-ion collider (RHIC) is one of the key frontiers in high energy nuclear physics. The high precision data anticipated from the second phase of the BES program would potentially enable the discovery of the conjectured QCD critical point, a landmark point on the phase diagram. In this paper, the progress made by the beam energy scan theory (BEST) collaboration, which was formed with the goal of providing a theoretical framework for analyzing data from BESII, is reviewed. In addition, the challenge of investigating the QCD phase diagram with future facilities is discussed.

The quantum chromodynamics (QCD) phase diagram is of great interest to researchers in the field of high energy nuclear physics. We review the present research status of several aspects of this topic. This review includes the search for the phase transition mechanism resulting in high-order baryon number fluctuations, how chiral imbalance, finite volume, and under rotations affect the QCD diagram, and the applications of the equation of states of dense QCD matter in the study of compact stars. The Nambu-Jona-Lasinio model and Dyson-Schwinger equations approach are the most commonly used methods described in this review. It is found that the theoretical results of high-order baryon number fluctuations are in good agreement with the experimental data. The chiral imbalance, finite volume, and rotation of quark-gluon plasma (QGP) have a quantitative impact on the chiral condensate and the QCD phase structure. In the study of compact stars, the theoretical results from equation of states of dense QCD matter agree well with pulsar observations. Further research will be required to form a complete understanding of the QCD phase diagram, particularly given the abundance of QGP.

We review the current status of quantum chromodynamics (QCD) properties in strong magnetic fields from lattice QCD. After a general introduction, we briefly present the implementation of a background magnetic field onto a lattice and discuss the recent progress on QCD properties at zero temperature, QCD transition temperature and inverse magnetic catalysis, and QCD phase structure in strong magnetic fields. Finally, we summarize this study.

We aim to study the effects of chemical potential and angular velocity on the critical endpoint of quantum chromodynamics (QCD). We used several probes (drag force, jet quenching parameter, heavy vector meson spectral function) to characterize the phase transition and studied gravitational waves from the holographic QCD phase transition in the early universe. We used different holographic QCD models to discuss the QCD phase transition, energy loss, spectral function, and gravitational waves. We found that the chemical potential and angular velocity changed the location of the critical endpoint, and the drag force and jet quenching parameter were temperature dependent and enhanced near the phase transition temperature. The magnetic field had a nontrivial effect on the spectral function. We conclude that the chemical potential decreases ωc, and the angular velocity decreases μc and the phase transition temperature. The jet quenching parameter and drag force can characterize the phase transition, and the magnetic field promotes the dissociation of heavy vector mesons. Moreover, the energy density of gravitational waves decreases as the gluon condensate increases, and the peak frequency shifts downward with increasing gluon condensate.Exploring the phase structure of QCD is an important task in high-energy heavy ion collision physics, and recently, there has been considerable interest in the QCD phase transition for rotating backgrounds.

The goal of relativistic heavy-ion collisions is to determine the phase boundary of quantum chromodynamics (QCD) phase transitions. Critically sensitive observables are suggested to be higher-order cumulants of conserved charges. The non-monotonous behavior of higher cumulants was observed at the relativistic heavy-ion collider (RHIC). However, it remains unclear whether these non-monotonous behaviors are critically related. We studied the influences of non-critical fluctuations, finite system size, and limited evolution time to determine if they cause non-monotonous behavior. First, we examined the minimum statistics required for measuring the fourth cumulant. The minimum statistic obtained using the centrality bin width correction (CBWC) method was 25 M. We suggest using a 0.1% centrality bin in the CBWC method instead of each Nch. With a 0.1 centrality bin width, 1 M statistics are sufficient. We then pointed out the statistical fluctuations from the limited number of final particles. By assuming the independent emission of each positive (or negative) charged particle, the statistical fluctuations of positive (or negative) charged particles were presented by a Poisson distribution, and the statistical fluctuations of net-charged particles were their evolution. The obtained statistical fluctuations for net protons, net electronic charges, and net baryons were consistent with those from the Hadron Resonance Gas model. In addition, the measured cumulants at RHIC/STAR are dominated by these Poisson-like statistical fluctuations. At the end of this section, we suggest the pooling method of mixed events and demonstrate that the sample of mixed events accurately presents the contributions of the background. Dynamic cumulants were defined as the cumulant of the original sample minus that of the mixed sample. Dynamical cumulants were shown to simultaneously reduce the influence of the statistical fluctuations, centrality bin width effects, and detector efficiency. Second, because the system is finite, the correlation length at the critical point is not developed to infinity in contrast to the system at thermal limits. Using a Monte Carlo simulation of the three-dimensional three-state Potts model, we demonstrated the fluctuations of the second- and fourth-order generalized susceptibilities near the temperatures of the external fields of the first-, second-, and crossover regions. Both the second- and fourth-order susceptibilities showed similar peak-like and oscillation-like fluctuations in the three regions. Therefore, non-monotonic fluctuations are associated with the second-order phase transition and the first-order phase and crossover in a finite-size system. The exponent of finite-size scaling (FSS) characterizes the order of transitions or crossover. To determine the parameters of the phase transition using the FSS, we studied the behavior of a fixed point in the FSS. To quantify the behavior of the fixed point, we define the width of the scaled observables of different sizes at a given temperature and scaling exponent ratio. The minimum width reveals the position of the fixed point in the plane of the temperature and scaling exponent ratio. The value of this ratio indicates the nature of the fixed point, which can be a critical, first-order phase transition line point, or crossover region point. To demonstrate the effectiveness of this method, we applied it to three typical samples produced by a three-dimensional three-state Potts model. The results show that the method is more precise and effective than conventional methods. Possible applications of the proposed method are also discussed. Finally, because of the limited evolution time, some processes in relativistic heavy-ion collisions may not reach thermal equilibrium. To estimate the influence of the nonequilibrium evolution, we used the three-dimensional Ising model with the Metropolis algorithm to study the evolution from nonequilibrium to equilibrium on the phase boundary. The order parameter exponentially approaches its equilibrium value, as suggested by the Langevin equation. The average relaxation time is defined. The relaxation time is well represented by the average relaxation time, which diverges as the zth power of the system size at a critical temperature, similar to the relaxation time in dynamical equations. During nonequilibrium evolution, the third and fourth cumulants of the order parameter could be positive or negative depending on the observation time, which is consistent with the calculations of dynamical models at the crossover side. The nonequilibrium evolution at the crossover side lasts briefly, and its influence is weaker than that at the first-order phase transition line. These qualitative features are instructive for experimentally determining the critical point and phase boundary in quantum chromodynamics.

One of the main goals of relativistic heavy-ion collision (HIC) is to search for the critical end point (CEP) of quantum chromodynamics (QCD), and distribution of the net-proton number from experimental measurements shows non-monotonic behavior, which indicates the existence of a CEP. The purpose of this work is to investigate the relationship between the net-proton number fluctuation and collision energy, and to explain the experimentally measured behavior. This study investigates the three-flavor Polyakov-loop Nambu-Jona-Lasinio (PNJL) model, which contains quark degrees from the NJL (Nambu-Jona-Lasinio) model and effective gluon contributions from Polyakov-loop, based on the equilibrium assumption and mean-field approximation. In addition, we study the phase diagram and C4/C2 of baryon number fluctuation as a function of collision energy along the freeze-out lines fitted from experimental data. With an appropriate form of freeze-out line, the collision energy decreases in the region of 7.7~200 GeV, and C4/C2 decreases slightly then increases, which is in agreement with the experimental data. Additionally, these results indicate that the equilibrium assumption is appropriate for the exploration of the system evolution after HIC, and the relationship between the freeze-out and phase transition lines is highly sensitive for observables.

The exploration of the critical point on the QCD (Quantum Chromodynamics) phase diagram is one of the most important goals of the beam energy scan program in relativistic heavy-ion collisions (RHIC-BES). Preliminary experimental measurement observed the non-monotonic behavior of net-proton fluctuations as a function of collision energy, which qualitatively agrees with the prediction of the static theoretical models and this hints the existence of the QCD critical point. The system created in heavy-ion collision is highly expanding system with which the dynamical effects dramatically modify the critical fluctuations near the QCD critical point. To confirm the existence of QCD critical point and study the phase structure of QCD system at finite temperature and finite density region, a series of dynamical models near the QCD critical point has been developed. This paper reviews the recent developments related to the exploration of the QCD critical point from experimental and theoretical viewpoints. In particular, we emphasize on the developments and challenges of the dynamical model near the QCD critical point and the first-order phase transition.

Several experiments are being conducted at heavy-ion colliders around the world to determine the location of the proposed critical end point of quantum chromodynamics (QCD) in the T-μB phase diagram. As the presence of a very strong magnetic field is relevant to peripheral heavy-ion collisions, magnetars, and the early Universe, it is important to investigate the effect of a high magnetic field strength on QCD phase diagrams. We summarize the recent status and new developments in studies investigating QCD phase transitions under an extremely strong magnetic field. By doing so, we believe that this work will promote both theoretical and experimental research in this field. TheT-B phase diagrams are produced by Lattice QCD simulations. Other phase diagrams (E-B, μB-B,μI-B, andΩ-B) are mainly studied by using the chiral effective Nambu Jona-Lasinio model. A rotating magnetic field is adopted for the study of color superconductivity. The Ginzburg-Landau approximation is used to studyπ-superfluidity andρ-superconductivity in a very strong magnetic field. Physical effects, besides a magnetic fieldB, can also be measured when sketching a QCD phase diagram, such as temperatureT, strong electric fieldE, chemical potentialsμ, and rotational angular velocityΩ. We present five QCD phase diagrams: T-B,E-B, μB-B,μI-B, andΩ-B. The following phases are present in many (if not all) of the five QCD phase diagrams: chiral symmetry breaking, chiral symmetry restoration, inhomogeneous chiral phase, π0-condensation,π-superfluidity,ρ-superconductivity, and color superconductivity. The running of the coupling constant with magnetic field is consistent with the decrease of the pseudo-critical deconfinement temperature, providing a natural explanation for the inverse magnetic catalysis effect. We also found that a chiral anomaly induces pseudoscalar condensation in a parallel electromagnetic field, and that there appears to be a chiral-symmetry restoration phase in theE-B phase diagram. Without consideration of confinement, color superconductivity is typically favored for large baryon chemical potential; however, chiral density wave is also possible in the largeB and relatively smallμB region of the phase diagram. In an external magnetic field, theπ-superfluid with finite isospin chemical potential acts similarly to a Type-II superconductor with finite electric chemical potential. Bothπ-superfluidity andρ-superconductivity are possible in a parallel magnetic field and rotation, but the latter is more favored for largerΩ particles.

Recent progress in studies on quantum chromodynamics (QCD) phase transition and related critical phenomena within the functional renormalization group (fRG) approach were reviewed, including the nonperturbative critical exponents and baryon number fluctuations, which are pertinent to the critical end point (CEP) in the QCD phase diagram. The fRG is a nonperturbative continuum field approach, in which quantum thermal fluctuations are successively integrated with the evolution of the renormalization group (RG) scale. Different methods of finding solutions to the flow or fixed-point equations of a nonperturbative effective potential have been discussed, for example, the Taylor expansion, expansion of the spatial dimension ε=4-d, and the recently proposed direct solution of the global potential. Furthermore, the baryon number of fluctuations is relevant to the critical phenomena of the CEP. Both have been discussed, and one explores the underlying reasons for the observed non-monotonic dependence of the kurtosis of the net proton number of distributions on collision energy in experiments.

Experimental evidences at the relativistic heavy ion collisions (RHIC) and large hadron collider (LHC) have demonstrated the formation of quark gluon plasma (QGP) in ultra-relativistic heavy-ion collisions at a small baryon chemical potential, where the phase transition from hadronic matter to QGP is suggested to be a crossover from state-of-the-art lattice quantum chromodynamics (QCD) calculations. It has been conjectured that there is a first-order phase transition and a critical point at a finite μB region in the QCD phase diagram. This study reviewed recent progress in searching for the QCD critical point from RHIC-STAR experiments.

BackgroundPack cementation aluminizing technology is a common method for preparing tritium barrier coatings, and its relative parameters during the preparation process have an important influence on the microstructure of the aluminide layer and the tritium barrier properties of the in-situ oxidized Al2O3 coating.PurposeThis study aims to investigate the effects of pack aluminizing conditions on the microstructure of the Fe-Al layer and analyze the related kinetic analysis of the aluminizing process.MethodsFirst of all, a pack aluminizing process activated by 1 wt% AlCl3 was used to fabricate aluminide coatings on the substrate of 316L stainless steel in the 923 K to 1 173 K range. Then, scanning electron microscope (SEM), energy dispersive spectrometer (EDS), and X-ray diffraction (XRD) were employed to characterize the cross-sectional microstructure and composition of the aluminized layer. Finally, the effects of aluminizing temperature and time on the microstructure and composition of the aluminized layer were analyzed, and the kinetic parameters of the formation of the Fe-Al layer and the relationship between aluminizing time and the thickness of the aluminized layer were further obtained.Results & ConclusionsThe experimental results show that the main phases of the aluminized layer are Fe2Al5 and FeAl with a certain amount of FeAl(Cr,Ni) precipitates. The high aluminizing temperature would accelerate the growth of aluminized layer and lead to the formation of a thick intermediate layer between the substrate and outer aluminized layer above 1 023 K. Simultaneously, extending the aluminizing time could increase the thickness of the Fe-Al layer, but has no effect on the phase composition. The relation between aluminizing temperature and the growth velocity of the Fe-Al layer is in accord with Arrhenius' equation, and the relative activation energy of the aluminizing process is about 79.23 kJ·mol-1. During the process of pack aluminizing, the relationship between the aluminizing time and the Fe-Al coating thickness is h=14.585t1/2+19.514.

BackgroundMolten salt systems have been extensively applied in the generation of nuclear and solar energy owing to their excellent heat transfer and storage performance. Therefore, it is essential to explore their physical and chemical properties, which are largely determined by their composition and ionic structure. In this regard, high-temperature (HT) nuclear magnetic resonance (NMR) has been demonstrated as an effective solution for qualitative analysis. Due to the existence of vent holes on the top of the commercialized standard NMR sample cell, it is not suitable for the study of some volatile, toxic or radioactive molten salt systems.PurposeThis study aims to implement ceramic NMR sample cells that are suitable for different molten salt systems.MethodsFirstly, a novel sealed sample cell that meets the requirements of molten salt systems was designed and produced by using an inner tube comprising AlN, BN, and Al2O3 ceramic materials, and an outer tube composed of ZrO2 ceramic materials. Then, with KBr as the standard sample, temperature calibration for this sample cell was conducted on the basis of 79Br NMR method. Finally, LiCl-KCl two-component molten salt (59.2 mol% LiCl-40.8 mol% KCl, eutectic melting point 353 ℃) was selected for HT-NMR experiment to verify the accuracy of temperature calibration, and check the performance of the HT-NMR method.ResultsExperimental results show that the sample cells are applicable in molten salt systems at a maximum temperature of 700 ℃, which meets the detection requirements of most molten salt systems. Additionally, the measurement results of the 79Br chemical shift of the KBr samples and captured HT 35Cl NMR spectra of the LiCl-KCl molten salt verify the reliability of the sample cell.ConclusionsBased on the results, the HT-NMR (High Temperature NMR) sample cells proposed in this study can be widely applied in various fields when the molten salt system requirements are met.

BackgroundCore fuel salt emergency drain system designed for fuel salt drain and afterheat removal, provides a safe shutdown mode for a molten salt reactor (MSR). It has important significance to evaluate reliability of the system for the safety of MSRs.PurposeThis study aims to quantitatively analyze the failure probabilty of the system and identify the pivotal factors that affecting the system failures, and provide suggestions for optimization of the system in engineering application.MethodsFirst of all, the fault tree analysis was employed to model the reliability of the core fuel salt drain system of MSRE through RiskSpectrum software. Then, the minimum cut sets and importance analysis was adopted to identify the most important basic event in fault tree of the system. Finally, two optimization methods, i.e., reduce the use of welds in bayonet cooling thimbles, and use different types of valves to isolate cooling gas flow of freeze valve, were proposed.ResultsThe results show that failure probability of the system is 5.62×10-4, and the identified pivotal factors affecting the system failures are welds leakage failures of thimbles and common cause failure of two groups valves of freeze valve. The optimization methods based on results of fault tree analysis can significantly reduce the system failure probability.ConclusionsThis study provides reference value for design and engineering application of the core fuel salt emergency drain system for MSRs.

BackgroundUnder the condition of sub-cooled nucleate boiling (SNB), corrosion products in primary coolant of nuclear reactor will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Previous literature shown, zinc injection in primary coolant is an important method to inhibit the fuel crud deposition on the fuel cladding surface.PurposeThis study aims to investigate the influence of zinc concentration on the behavior of fuel crud deposition, and eventually provide guidance for zinc injection in primary coolant of nuclear power plant.MethodsThe fuel crud deposition tests of domestic zirconium alloy fuel cladding in different zinc concentrations were carried out by using a self-made fuel crud deposition device. Tubular crud deposition test specimen with built-in heating unit was designed and prepared for simulation study. After the tests, stereo microscope (SM) and scanning electron microscope (SEM) were employed to observe the macro and micro morphology of fuel crud whilst the composition of of fuel crud was observed and analyzed by the energy dispersive spectroscopy (EDS) with SEM, and X-ray photoelectron spectroscopy (XPS) was used to analyze the contents of Zn and B elements in the crud phase and inside the crud.ResultsObservation results show that the chimney-like crud formed on the fuel cladding surface becomes less obvious with increasing the zinc concentration in the coolant and the crud surface becomes flatter. Simutabeously, the crud thickness, the ratio of Ni/Fe and the boron precipitation mass within the crud are decreasing with increase of the zinc concentration. When the zinc concentration increases to 100 μg?L-1, new Zn-containing phases precipitate within the crud.ConclusionsWithin the zinc concentration of 0~100 μg?L-1, zinc injection in primary coolant of reactor can significantly inhibit the crud deposition on the fuel cladding surface.

BackgroundIn the neutron activation calculation, the inherent uncertainty of the input nuclear data will cause a certain impact on the calculation results. The uncertainty of the calculation results plays an important role in the source term analysis and radiation shielding design of nuclear facilities.PurposeThis study aims to analyze the uncertainty of neutron activation calculation based on direct derivative method.MethodsFirstly, the direct derivation method and Gear algorithm for uncertainty analysis were investigated, and the activation coefficient matrix and sensitivity coefficient matrix were constructed. Then, the Gear algorithm was employed to solve the activation equation and sensitivity equation simultaneously, and the sensitivity coefficient of nuclide inventory to nuclear data was obtained. The relative uncertainty of nuclide inventory was obtained by combining the relative uncertainty of nuclear data. Finally, This method was integrated into the neutron activation program ABURN, and typical examples was selected to test and verify its performance.ResultsThe calculation results of the nuclide inventory and its sensitivity coefficient and relative uncertainty by the ABURN program have little deviation from the analytical solution and the numerical solution of the European activation program FISPACT, most of the deviations do not exceed 0.2%, and the maximum deviation does not exceed 1%.ConclusionsVerification results show that the method and procedure developed in this paper have the ability to analyze the sensitivity and uncertainty of nuclide inventory with high precision, and can provide tools and data support for the radiation protection of nuclear facilities and the source term analysis.

BackgroundThere is usually a strong coupling of neutronics-thermal hydraulics (N-TH) fields inside nuclear reactors.PurposeThis study aims to accurately simulate the multi-physics fields in nuclear reactors by developing a three-dimensional N-TH coupling code MORPHY tailored to advanced complex reactors.MethodsFirst of all, a three-dimensional triangular-z nodal variational nodal method (VNM) was employed for neutronics calculation. and the stiffness confinement method (SCM) was used to solve the neutron temporal-spatial equation; thermal hydraulic calculations were based on the one-dimensional multi-channel model and the one-dimensional cylindrical thermal conductivity model. Then, the accuracy of neutron dynamics was verified by TWIGL benchmark, Dodds benchmark, and the typical pressurized water reactor (PWR) benchmark NEACRP. Finally, the effects of different coupling methods and angle discrete orders on the results were analyzed and compared against reference solutions by PARCS.ResultsVerification results of TWIGL benchmark show that the deviation of relative power from the reference results is less than 0.5%. Compared with the results of Dodds benchmark, it verifies the MORPHY code's ability to describe unstructured meshes. The transient coupling calculation capability of MORPHY is verified by NEACRP benchmark.ConclusionsNumerical solutions by MORPHY are in good agreement with reference results of the TWIGL, Dodds and NEACRP benchmark problems. It is concluded that MORPHY can adapt to the transient N-TH coupling analysis of nuclear reactor cores.

BackgroundThe pre-equilibrium cluster manifests the nuclear structure and reaction dynamics of collision system. The systematical investigation of cluster emission in transfer reactions is of significance in deep understanding of the synthesis of superheavy nuclei, shell evolution, new isotope production, etc.PurposeThe dynamics of pre-equilibrium cluster in a few of nucleon transfer reaction has been described by theoretical model, such as exciton model, cluster model. However, the cluster emission in massive transfer is very complicated because of the emission mechanism associated with the structure properties and also the dynamical process.MethodsIn this work, the pre-equilibrium cluster emission in massive transfer reaction has been systematically investigated within the dinuclear system model. The model has been successfully used for describing the massive fusion reaction and multi-nucleon transfer dynamics. The nucleon exchange and energy dissipation take place once the dinuclear system is formed. The nucleon transfer between the binary fragments is governed by the single-particle Hamiltonian and proceeds around the Fermi surface formed by the dinuclear system. The master equation is used for the nucleon transfer dynamics and the relative motion energy dissipation is taken into account. The dynamics of neutron, proton, deuteron, triton, 3He, 4He, 6,7Li and 8,9Be in collisions of 12C+209Bi, 40Ca+208Pb and 48Ca+238U near Coulomb barrier energies is analyzed, i.e., temporal evolution of production rate, kinetic energy spectra and angular distribution.ResultsIt is found that the emission probability of 4He is the same magnitude of proton emission and several orders larger than the one of 3He. Both the nuclear structure and dynamical effects influence the pre-equilibrium cluster production.ConclusionsThe pre-equilibrium clusters are emitted from the 'projectile-like' and 'target-like' fragments and the angular distributions manifest the similar trends. The kinetic energy spectra of clusters are shown as the Boltzmann distribution. The method is also extended to the cluster emission in weakly bound nuclei induced reactions by considering the preformation factor for the cluster construction.

BackgroundPlastic scintillators have potential for application in neutron detection. Two sizes (?2.54 cm×2.54 cm, ?5.08 cm×5.08 cm) of plastic scintillators are self-developed by scientific research team in the school of physics, Sichuan University.PurposeThis study aims to experimental test the neutron/gamma (n-) discrimination performance for two self-developed plastic scintillators.MethodsA photomultiplier tube (PMT) was used to build detection systems, and high speed oscilloscope (LECROY HDO6104A) was employed to sample signal of detector for the energy calibration of the self-developed plastic scintillator. The pulse amplitude spectrum of 137Cs γ radiation source was measured and compared with the MCNP5 simulation spectrum to obtain the position information of the Compton edge and accurately calibrate the energies of γ rays. The data obtained from a 241Am-Be neutron source were analyzed using the charge integration method, and parameters such as the figure of merit (FOM), peak-to-valley ratio for neutrons, and the proportion of leaked neutrons over all neutron events were used to quantify the n-γ discrimination in different energy zones. The detection efficiencies of two self-developed plastic scintillators relative to the Commercial off-the-Shelf (COTS) EJ-299-33A were determined.ResultsThe results show that the FOM of ?2.54 cm×2.54 cm self-developed plastic scintillator is higher that of ?5.08 cm×5.08 cm self-developed plastic scintillator, and the detection efficiency of two self developed plastic scintillators relative to EJ-299-33A is about 0.49 and 1.0, respectively.ConclusionsThe performance of the ?5.08 cm×5.08 cm self-developed plastic scintillator is comparable to that of the COTS plastic scintillator EJ-299-33A with near the same discrimination ability.

BackgroundThe miniature fission ionization chamber is a widely used neutron detector for the in-core neutron flux monitoring of a nuclear reactor. Typically, the in-core neutron flux rate measurement system of a domestic CPR1000 nuclear power unit adopts a mobile miniature fission ionization chamber as the neutron probe to measure the neutron flux of the reactor and provide an in-core neutron flux distribution map during the operation. Therefore, it is an important piece of safety and control equipment for nuclear power plants.PurposeThis study aims to develope a mobile miniature fission ionization chamber neutron detector according to the service conditions and technical requirements of current foreign products.MethodsThe nuclear properties of self-made fission ionization chamber neutron detector was developed strictly following the national standard GB/T 7164-2022 and the industry standard NB/T 20215-2013. The gamma sensitivity was tested and compared with a reference commercial fission detector using a 60Co gamma source. The thermal neutron detection characteristic, including the length and slope of plateau, the thermal neutron sensitivity and linearity were tested in one test channel of the China Mianyang Research Reactor (CMRR) with neutron flux from 1×109 n?cm-2?s-1 to 4×1013 n?cm-2?s-1.Results & ConclusionsThe test results indicate that "domestic substitution" of this in-core safety product can be achieved, and the nuclear characteristics of self-developed prototypes are comparable to those of foreign products.

BackgroundCompton imaging technology is a new radiation hotspot location technology that does not require collimation and has a wide field of view, high efficiency, and broad application prospects. With the development of nuclear technology, Compton cameras with the above-mentioned advantages have a wide range of applications not only in the nuclear industry but also in the field of nuclear medicine, hence recently become a popular research field worldwide.PurposeThis study aims to develop a double-layer separated Compton camera for far-field imaging of specific radiation scenes in nuclear facilities.MethodsFirst of all, two pixel-type cadmium zinc telluride (CZT) detectors and an application-specific integrated circuit (ASIC)-based readout electronics system were adopted for the development of a double-layer separated Compton camera. A list-mode maximum likelihood expectation maximization (LM-MLEM) image reconstruction algorithm was implemented in the host computer software. Then, 137Cs point source was used for experimental test of imaging performance of the system, and the parameters affecting the imaging performance, such as the detector layer spacing and area of the absorption layer, were optimized. Finally, far-field three-dimensional imaging of the radiation source was performed by moving the measuring position of the detector.ResultsThe test results show that the energy resolution of the CZT detectors is approximately 3% (FWHM@662 keV), which can determine the location of the point source at a distance of 5 m, and the angular resolution for θ and φ directions of the optimized system is approximately 10°.ConclusionsDouble-layer separated Compton camera of this study has advantages of adjustable structure, low detector cost, relatively simple readout electronics, and wide imaging field of view. The angular resolution of this double-layer separated Compton camera can be improved by proper adjustment of the imaging influence parameters (such as layer spacing and the area of the absorption layer).

Background131I-BEV-PTX-SPIONs is a type of nanoparticle used in cancer therapy. It is composed of four components: a radioactive isotope of iodine (131I), a chemotherapy drug called paclitaxel (PTX), a type of nanoparticle called superparamagnetic iron oxide nanoparticles (SPIONs), and a molecule called bevacizumab (BEV) which is an antibody that targets and blocks the growth of blood vessels that supply tumors.PurposeThis study aims to investigate the preparation and biological distribution of 131I-BEV-PTX-SPIONs.MethodsFirst of all, 131I-BEV-PTX-SPIONs were prepared, synthetized and identified. The transmission electron microscope (TEM) were employed to observe the particle characteristics. Then, 30 tumor-burdened nude mice were divided into the single targeting group and the dual targeting group for evaluation of 131I-BEV-PTX-SPIONs distribution in these nude mice, each group was divided into five sub-groups based on time points of 2 h, 6 h, 12 h, 24 h and 48 h, 3 in each sub-group. Finally, 131I-BEV-PTX-SPIONs were injected into the caudal vein of these mice, and experiments of biological distribution in vivo and SPECT imaging were carried out, and results were analyzed using GraphPad Prism 8.3 software.ResultsThe nanospheres in prepared 131I-BEV-PTX-SPIONs are obtained in good mono dispersion with a diameter of approximately 220 nm by TEM observation. 131I-BEV-PTX-SPIONs obtained in a high radiolabeling yield is about 81.4% with the radiochemical purity of over 99% and good stability shown in the 0.2 mol·L-1 PB buffer. And it could attain sustained PTX release in vitro. Comparing with the cellular uptake of 131I, a higher uptake and sustained PTX release in vitro are shown for 131I-BEV-PTX-SPIONs. Biodistribution experimental results show: After the injection of 131I-BEV-PTX-SPIONs, with the extension of time, the radiation count of the tumor is relatively higher, at 12 h reaching the peak. And the T/NT ratio increased gradually, and it reaches 7.8±0.50 at 48 h. The counts and the ratios at 6 h, 12 h, 24 h and 48 h are notably higher in the dual targeting group than the single targeting group (P131I-BEV-PTX-SPIONs, the tumor site has a radioactive build-up. With the extension of time, the accumulation of radioactivity increased and remained stable, and the T/NT ratio rises steadily.ConclusionsThese results demonstrated the potential of 131I-BEV-PTX-SPIONs in the diagnosis and treatment of lung cancer, and it was worthy of further study.

BackgroundHard X-ray Imager (HXI) is one of the three scientific payloads onboard of the advanced space-based solar observatory (ASO-S). The calorimeter of HXI consists of 99 LaBr3 crystal and photomultiplier tube (PMT) detection units. A highly integrated charge-measurement application-specific integrated circuit (ASIC) with model IDE3381 is adopted in the front-end electronics of the calorimeter to process the signals from the 99 detection units on the space-limited and power-limited satellite platform.PurposeThis study aims to evaluate the radiation tolerance of model IDE3381 ASIC in a space radiation environment.MethodsA test bench with a flexible structure was designed by separating the device undergoing testing from the data acquisition (DAQ) system, hence shielding DAQ from the radiation environment. The performance of ASIC was automatically tested and monitored in the test bench during radiation tests. Both single-event effect (SEE), including single-event upset and single-event latch-up, and total ionizing dose (TID) tests were carried out by using a heavy ion beam and 60Co gamma-ray, respectively.ResultsThe test results show that the SEE threshold of model IDE3381 ASIC is greater than 75 MeV?cm2?mg-1, and the TID capacity is greater than 30 krad(Si).ConclusionsThe radiation tolerance of the charge measurement ASIC (model IDE 3381) meets the requirements of ASO-S HXI flight model.

BackgroundStable isotopes play a crucial role in a variety of fields such as energy, military, semiconductor, agriculture, medicine, pharmacology, biology, food industry, and chemistry. With the rapid growth of nuclear science and technology applications in China, there has been an increasing demand for isotopes that cannot be met by current production capacities. Thus, the development of electromagnetic isotope separators capable of producing high yields and high isotopic purity has become necessary.PurposeThis study aims to develop an electromagnetic isotope separator based on a 2.45 GHz microwave ion source and isotopic magnet for studying a number of important heavy isotopes, such as xenon and molybdenum isotopes.MethodsFirstly, adjustable axial magnetic field in the source was designed by a double-solenoids to obtain high density plasma, and a high coupling efficiency matching waveguide was optimized by CST microwave computing module. Then. a crucible built in the discharge chamber was used to melt metal oxide for generating heavy metal ion beams. Finally, the discharge chamber, microwave coupling waveguides and heating oven of the ion source were simulated and designed for the generation of heavy ions.ResultsSimulation result shows that the temperature around the crucible is 917 ℃ when the current of heating wire is set to 70 A, and 100 mA hydrogen beam is generated during commissioning. The designed crucible in the discharge chamber can generate metal vapor efficiently for ionization, and achieve producing 20 emA Xe+ and 5 emA Mo+ respectively at the energy of 40 keV.ConclusionsThe feasible scheme of the magnetic field and microwave coupling design of this study are verified. The design of the 2.45 GHz electron cyclotron resonance (ECR) ion source provides a feasible and effective solution for the high yields isotope ions.

BackgroundNeutron radiography (NR) is an important nondestructive testing method. NR is particularly useful for detection of light materials in medium and large heavy samples. Especially, the fast neutrons can penetrate the heavy materials and reveal the structure of the light materials. Compared to accelerator neutron sources, the fission neutrons elicited from a reactor are stable and of high quality. The fission neutron imaging is a useful complementary testing technology, especially for industrial applications that require high throughput and large-scale testing.PurposeThis study aims to investigate the super field of view neutron imaging by fission neutrons elicited from research reactor.MethodsBased on theoretical analysis and Monte-Carlo simulation, one filter combination was employed to improve the proportion of fission neutrons in the thermal neutron beamline at China Mianyang Research Reactor (CMRR). A fission neutron imaging system was constructed by employing a large field fast neutron fluorescent screen, short focus distance lens, and scientific charge coupled device (CCD) camera. Finally, some samples were tested using fission neutron tomography.ResultsThe fission neutron flux reaches up to 3×105 cm-2·s-1 when the L/D ratio is about 260. The field of view of NR is up to 400 mm×400 mm with resolution was better than 0.5 mm. Using super field of view method, samples less than 600 mm can be tested with this new system.ConclusionsCombination of theoretical calculation and experimental methods, fission neutron imaging can be improved to overcome some of the limitations of traditional neutron radiography techniques, and meet the needs of large sample detection in the future.

BackgroundSerial X-ray crystallography has developed rapidly due to its advantages of data collection at room temperature, low radiation damage and time resolution. To solve protein structures by using the serial X-ray crystallography, a large amount of produced diffraction data needs to be screened for finding the effective diffraction patterns. The use of convolutional neural networks (CNN) can not only automate the data screening process, but also improve the accuracy of data classification comparing with the traditional "point finding method".PurposeThis study aims to explore five types of popular convolutional neural networks, i.e., AlexNet, GoogleNet, MobileNets, Vgg16, ResNet, for screening crystallographic diffraction patterns, and compare the accuracy and efficiency of them to build up a fast and accurate convolutional neural network tool for screening the diffraction patterns of different protein crystal samples.MethodsFirstly, the primitive data for model training extracted from the coherent X-ray image database, collected by Linac Coherent Light Source (LCLS) and Spring-8 Angstrom Compact free electron laser (SACLA), were pre-processed by gray level equalization and data enhancement. The deep learning models were trained by iteration of the preprocessed data. Then, the selected convolutional neural network through the comparison of accuracy and efficiency was used to process further the experimental data of protein crystals diffractions.ResultsThe results show that MobileNets not only has the accuracy similar to large networks such as ResNet, GoogleNet-Inception, but also runs faster.ConclusionsMobileNets provides an effective and convenient screening tool for serial X-ray crystallography experimental data.

BackgroundAccording to the requirements of unmanned underwater vehicles for high reliability, high power, and long-life power, the design scheme of the megawatt heat pipe nuclear reactor silent unmanned portable reactor (UPR-s) is proposed by Xi'an Jiaotong University.PurposeThis study aims to design the shielding scheme for UPR-s to ensure the radiation safety of the cabin.MethodsFirst of all, according to the UPS-s scheme applied to the underwater unmanned vehicle (UVV), the layout of the nuclear system and shielding was designed, and the source terms of the reactor core under both full power and shutdown status were calculated by using NECP-SARAX code. Then, initial shielding model was established with consideration of several alternative shielding materials. The deterministic neutron-photon shielding calculation code NECP-hydra was employed to analyze several shielding schemes: the initial model layout, composite shielding layout, and shadow shielding layout. Finally, the accumulated fast neutron fluence, photon dose and source intensity at the safety plane were analyzed, and a shielding optimization scheme meeting the requirements was proposed on the basis of the numerical analysis results.ResultsCalculation results of shielding optimization scheme show that the maximum accumulated fast neutron fluence and photon dose of the safety plane at full power are 9.48×1011 n·cm-2 and 7.29×105 rad, respectively. Under shutdown conditions, the maximum safe plane dose rate is 0.004 49 mSv·h-1, and the total weight of core plus shielding is 296.35 kg.ConclusionsThe key parameters of optimized shielding scheme, including the cumulative fast neutron fluence, photon dose, and total shielding weight, satisfy the given design requirements.

BackgroundTritium can be released into the environment in a loss of vacuum (LOVA) scenario in a fusion reactor. The simulation of the atmospheric dispersion behaviour of tritium is one of the core components of the assessment of the radioactive consequences.PurposeThis study aims to analyse the behaviour of tritium dispersion in the atmosphere after a fusion reactor accident.MethodsBased on the Gauss model and the Pasquill stability classification method, an analytical model of tritium dispersion was developed for transient cases considering the effects of gravitational settling, smoke lifting, and wind speed, etc. The calculation of the model for dry settling at the ground boundary was improved by adding ground reflection coefficients to the Gauss model. Finally, the Canadian tritium release experiment and the tritium release accident at the Savannah River plant in the United States were used to verify the applicability of the model.ResultsVerification results show that the accuracy of the developed model is the same as that of UFOTRI and the HotSpot 3.0 code. For the LOVA scenario of International Thermonuclear Experimental Reactor (ITER), the atmospheric dispersion behaviour of tritium is obtained for multiple release heights, different wind velocities and tritium phased releases.ConclusionsThe phased release of tritium results in two highly radioactive regions along the downwind direction, and the increase in release height and wind speed will enhance the atmospheric diffusion behaviour of tritium and thus reduce the accumulation of radioactivity in the near field.

BackgroundThe nuclear transmutation is the only way to reduce the radioactive hazard of the high level long-lived radioactive minor actinides (MA). The majority of commercial reactors in operation in the world are pressurized water reactor (PWRs), hence the transmutation efficiency of minor actinide nuclide (MA) in PWR are crucial problem in the area of the nuclear waste disposal.PurposeThis study aims to improve the transmutation efficiency of MA and flatten the core power distribution by using MA nuclide for PWR.MethodsFirst of all, the HPR1000 (Hualong #1) model 177 core structure was taken as reference PWR, thermal-fast neutron convertible material 6LiD was introduced to design coated axially non-uniform MA/6LiD transmutation rods which structurally applicable to the PWR. The internal component of the transmutation rods was UO2, and the external component was the transmutation coating material composed of MA and 6LiD nuclides. The layout of the coating material on the transmutation rods was axially three, five and seven segments structure, and the coating thickness gradually decreased from the middle to both ends. Then, the Monte Carlo program RMC2.0 developed by the Reactor Engineering Calculation and Analysis Laboratory of Tsinghua University was employed to establish the core and calculate the effect of transmutation coating material composition on core keff.ResultsThe results show the best transmutation effect up to 23.25% is realized when the mass ratio of 6LiD to MA in the transmutation coating material is 2∶8. Among the coated axially nonuniform transmutation rods, the seven-segment transmutation rod has the best transmutation effect, and the transmutation rate is 25.43%. The best fission effect of three-segment transmutation rod has the fission rate of 4.48% for MA nuclide. At the same time, the transmutation rod with axial non-uniform structure can reduce the axial power peak factor of the core from 1.778 to 1.375.ConclusionsCompared with axially uniform rods, these axially non-uniform MA/6LiD transmutation rods have good transmutation efficiency, especially the fission rate, and good performance on flatten axial power distribution is achieved, simutanously.

BackgroundThe activation method is taken to measure the in-core distribution of neutron spectrum for the designed 10 MW solid-fuel thorium molten-salt reactor (TMSR-SF1). The neutron activation foil sample is loaded outside the reactor and quickly transported to the measurement position in the reactor through the transmission device for irradiation, and then is transferred outside of reactor to the energy spectrum samples for de-spectrographic analysis.PurposeIn order to realize the rapid entry and exit of the neutron activation foil sample into and out of the reactor, a pneumatic conveying system with double-layer casing is designed in this research.MethodsThe principle of the conveying system and the structure of the double-casing tube were adopted in this study. ANSYS Fluent software and 6DOF dynamic grid technology were used to analyze the movement and stress of the sample under different pipe gaps, so as to determine the sample pipe gap value. Then the flow parameters and gas-solid two-phase flow resistance of the conveying system were calculated in detail using the pneumatic conveying theory. Finally, a prototype was developed for experiments to verify the principle of the conveying system.ResultsThe analysis results showed that the sample speed is decreased with the increase of the pipe gap. The experiments results show that the velocity of the sample and the pressure loss of the gas-solid two-phase flow increase with the increase of the gas flow rate. Under the same flow rate, the experimental speed of the sample movement and the pressure loss of the gas-solid two-phase flow are in good agreement with the theoretical calculations.ConclusionsThe pneumatic convey system with double-casing tube can be applied to transport the sample into and out of the reactor, and the theoretical calculations values of pneumatic conveying parameters in this study are reliable.

BackgroundThe high-fidelity neutron transport calculation requires refined geometric modeling whilst the unstructured meshes have strong adaptability to copy with the changes bring by complex geometry structure, and overcome the deficiencies of structured meshes in modeling capability.PurposeThis study aims to develop and validate a two-dimensional shielding calculation code ThorSNIPE which can be used to improve the modeling ability for analysis complex problems.MethodsFirst of all, problem solving model was established with discrete ordinates method and finite element method on the basis of the first order Boltzmann transport equation. The computational performance of continuous finite element method and discontinuous finite element method were compared and analyzed. Mass-matrix lumping technique was further applied to improve the reliability of solving model. Then, a two-dimensional discrete ordinate-finite element shielding calculation program ThorSNIPE was developed on the basis of above model. Finally, the code was validated by BWR cell critical benchmark, Argonne-5-A1 fixed source benchmark and Dog leg duct benchmark.Results & ConclusionsThe numerical results show that calculation value provided by ThorSNIPE is in good agreement with reference value, indicating that ThorSNIPE is suitable for complex shielding calculation, and Mass-matrix lumping technique can effectively suppress the non-physical spatial oscillations without reducing the calculation accuracy.

BackgroundCompared with the traditional pressurized water reactor (PWR), the core design of large advanced PWR CAP1400 has significant changes, such as the increase in the number of fuel assemblies, the increase in reactor power, the increase in the average temperature of core coolant, etc. These changes have an important impact on the results of rod ejection accident, and then affect the safety of reactor.PurposeThis study aims to verify the safety of large advanced PWR under rod ejection accident condition and the influence of key input parameters on accident analysis results.MethodsBased on the neutron dynamics software TWINKLE and fuel performance analysis program FACTRAN, the typical four types of operating conditions, including the beginning of life the hot full power and the hot zero power, the end of life the hot full power and the hot zero power, were selected to carry out the simulation calculation of the control rod ejection accident analysis for large advanced PWR, and the sensitivity analysis of key input parameters of rod ejection accident conditions was performed by using the direct numerical perturbation method.Results & ConclusionsSimulation results show that the power peak is the most sensitive to the worth of rod ejection, but less sensitive to shutdown reactivity. The consequences of the control rod ejection accident designed for CAP1400 can meet the requirements of acceptance criteria and the reactor is in the safe and controllable state.

BackgroundElliptical orbiting satellites passing through the inner radiation belt are exposed to high-energy and high-flux protons and electrons. Therefore, electronic devices of satellites need to resist ultra-high cumulative radiation doses.PurposeThis study aims to propose a composite material structure for shielding space protons and electrons, instead of the traditional aluminum structure.MethodsThe interaction of elliptical orbital protons and electrons with four shielding materials (polyethylene/polypropylene, tantalum and aluminum) was simulated by the MULASSIS (Multi-Layered Shielding Simulation Software). The radiation particle energy spectrum calculated by SPENVIS software was used as the input of MULASSIS particle energy spectrum. The changes of total dose and displacement dose after shielding with areal density of the four shielding materials were compared and analyzed. Finally, the most suitable composite shielding structure was selected by considering the proton and electron shielding effects and mechanical properties of the four materials.ResultsThe results show that under same areal density, the order of the shield effectiveness from large to small, of four different materials on orbital proton and electron is polypropylene, polyethylene, aluminum and tantalum, among which polypropylene and polyethylene have almost the same shielding effect. The polyethylene-aluminum composite shielding structure is selected for construction design. The shielding targets of total dose and displacement dose to ensure the reliable operation of elliptical satellites are 50 krad(Si) and 2×1010 p?cm-2 (equivalent to 10 MeV protons) respectively.ConclusionsCompared with the single aluminum shield, at least 27.8% shielding mass is saved by using the polyethylene-aluminum composite protective structure in the above ratio.

BackgroundThe comprehensive research facility for fusion technology magnet performance research platform (MPRP) is a large-scale experimental platform established for advanced superconducting magnet experiments. The retrieval speed of MPRP historical data is slow due to massive storage.PurposeThe study aims to develop a MPRP data archiving system (MPDAS) and increase its retrieval speed.MethodsFirst of all, the experimental physics and industrial control system (EPICS) data archiving plug-in was designed for MPDAS. Both MongoDB Sharding and Replica Set mechanism were employed to build a highly scalable data storage architecture. Then, the core ideas of three traditional cache replacement algorithms, LRU (least recently used), LFU (least frequently used) and FIFO (first in first out) were drawn by MPDAS to establish a data temperature model based on Newton's law of cooling. A multi-dimensional feature data partitioning algorithm was implemented to integrate access time, access frequency and storage order, hence the hot and cold historical data were identified to realize data tiered storage. Finally, the retrieval speed of MPDAS was improved by preferentially accessing Redis when querying historical data, and selecting different retrieval strategies based on hit results and data integrity.ResultsThe system test results show that the functional characteristics of MPDAS meet the design requirements. Compared with FIFO, LRU, and LFU, the Redis hit rate of the MPDAS when the hot database stores 1% of the historical data is increased by 38.05%, 26.91%, and 11.06% respectively.ConclusionsBy increasing the hit rate of hot data, the average response time of data retrieval can be directly reduced. The retrieval response speed of MPDAS is effectively improved by quantifying the heat of historical data and dividing the heat and cold.

BackgroundSilicon carbide junction barrier Schottky (SiC JBS) diode is a kind of power device based on wide bandgap semiconductor. SiC JBS diode is expected to become an important part of electric propulsion systems in the radiation application field in the future space exploration due to its excellent high-voltage, high-frequency and high-power characteristics. However, there are a large number of protons in the typical orbit of spacecraft, which always threaten the stable operation of spacecraft, including its key components.PurposeThis study aims to explore the resist ability of SiC JBSs to the degradation of medium energy proton irradiation, and clarify the mechanism of radiation effect of SiC JBSs from medium energy proton.MethodsBased the proton equivalent displacement damage dose in low Earth orbit for ten years, the SiC JBSs were firstly irradiated using 10 MeV protons at fluences ranging from 3×109 cm-2 to 3×1010 cm-2 at room temperature and without bias voltage. And the macro electrical characteristics of the SiC JBSs both before and after irradiation, including the forward current-voltage (I-V), reverse I-V and capacitance-voltage (C-V) characteristics, were tested. Then the irradiation-induced defects characteristics were tested by deep level transient spectrum (DLTS). Further, the related degradation mechanism that was associated with this phenomenon was also investigated using based on the test data and mathematical calculation. Finally, irradiation experiments of accelerator protons were carried out for commercial SiC JBSs.ResultsThe results show that the forward electrical characteristic of the SiC JBSs is stable, and the leakage current decreases at low reverse safety voltage. But the rated breakdown voltage is seriously degraded with the increase of irradiation fluence. The main contribution to the change of SiC JBSs characteristics originates from the increase of interface charge, deep level defects and Schottky barrier height, and the decrease of carrier density and carrier diffusion length in the drift region.ConclusionsAnalysis of the radiation damage process and mechanism of SiC JBSs in this study provides a research basis for its evaluation and verification before applied to medium energy proton environment.

BackgroundLaser-induced breakdown spectroscopy (LIBS) is a new technology for testing and analyzing the composition and content of elements in materials.PurposeThis study aim to determine the content and distribution of hydrogen isotopes in hydrogen storage materials by using LIBS quantitative analysis technology.MethodsThrough independent design, construction and integration, a LIBS system was established for in-situ measurement of hydrogen isotopes in a vacuum chamber. Titanium sheets were used to prepare titanium-hydrogen samples with different concentrations of hydrogen and deuterium atoms to investigate the content and distribution of hydrogen isotopes in titanium using LIBS technology. The plasma parameters were calculated from the emission spectrum of titanium, and quantitative analysis on the content of hydrogen and deuterium atoms in the titanium sheet was carried out. Finally, the internal calibration method was employed to draw the calibration curves of hydrogen and deuterium, respectively, so as to determine the accuracy of this technology.ResultsThe plasma temperature calculated from the Boltzmann diagram is (16 000±1 000) K. Test results show that the linearity of calibration curves is increased by 4% by integrated intensity calibration, and the error of hydrogen isotope quantitative analysis is reduced by 2.8%. Based on the fitted curve, the concentration is consistent with the concentration determined by the pressure drop method during the sample preparation process. The average measurement error of hydrogen and deuterium is 3.19% and 1.94% respectively.ConclusionsProvided that the plasma state conforms to the local thermal equilibrium and the plasma temperature of different elements is consistent, LIBS quantitative analysis can accurately measure the contents of hydrogen isotopes by using the internal standard method. Signal enhancement effect and data accuracy of LIBS meet the requirements of quantitative analysis.

BackgroundThe electron beams produced by laser plasma acceleration have excellent quality for pulse lengths of the order of fs. Due to the existence of a strong laser field, there are difficulties in direct applications, and more applications need to transmit the electron beams to the application terminal. The energy spread leads to the generation of energy chirp of the electron beam in the transmission.PurposeThis study aims to explore the design of the beam optics to compress the pulse length and keep it on the fs scale.MethodsAn achromatic beamline consisting of bending magnets and quadrupole magnets was designed to compress the pulse length of electron beams. Critical parameters of an achromatic beamline were given by a derived formula. Transformation matrix was employed to investigate the differences of the pulse lengths in achromatic transmission and non-achromatic transmission. The pulse lengths of electron beams with different energies were scanned with different deflection angles (0.3 rad, 0.6 rad, 0.9 rad) and deflection radii (0.15 m, 0.25 m, 0.35 m) to study the influence of beamline parameters. Finally, the magnetic field gradients of the quadrupole lens were adjusted to realize the compression of electron beams with different energies in a beamline.ResultsComparing to non-achromatic transmission, the pulse lengths of electrons with the same energy and different initial divergence angles can be compressed effectively in the achromatic beamline. The larger the deflection angle or the deflection radius, the longer the pulse duration of the electron beam with higher energy (>25 MeV). By adjusting the magnetic field gradients of the quadrupole lens, the pulse lengths can be reduced from more than 100 fs to around 20 fs at higher energies.ConclusionsUsing a fixed-size achromatic beamline, combined with magnetic field strength adjustment, the pulse lengths of electron beams with different energies can be kept on the order of fs after transmission.

BackgroundSynchrotron radiation experimental methods have unique advantages in studying the structure and physical properties of materials, but it is a challenge for many experimental methods to achieve synchrotron radiation in situ high temperature conditions, especially above 2 000 K. Laser heating methods can achieve rapid, micro-region extreme high temperature conditions, and have become an important tool for the study of high temperature physical properties.PurposeThis study aims to develop a portable laser heating device for Shanghai Synchrotron Radiation Facility (SSRF) in situ experiments in the field of extreme high temperature research, such as high entropy alloys, turbine blades, aviation materials, etc.MethodsA 100 W continuously tunable near-infrared fiber laser was used as the heating souce, the sample was heated up by laser through the focusing lens and generated thermal radiation. The radiation spectrum was collected through the spectral collection focusing lens and measured by spectrometers. The temperature gradient and temperature stability of the sample were fitted by the blackbody radiation method. Finally, the melting experiment of pure tungsten sheets in vacuum was conducted to verify its maximum heating temperature, and the temperature gradient and stability measurement of the device were calibrated with platinum samples.ResultsWe Experimental results show that melting point of about 3 695 K for tungsten sheets in vacuum is achieved using this device, and X-ray diffraction patterns of MoS2 and CTAB-MoS2 materials under 1 608 K in situ are obtained at the surface diffraction beamline station of SSRF.ConclusionsThe laser heating method developed expands the extreme experimental conditions in SSRF, and provides an important means to study high temperature physics for materials.

The nuclear fusion reaction using deuterium and tritium fuel produces a large number of neutrons, γ rays and activation products, which have an impact on the radiation safety of people and the environment. In order to reduce the impact of ionizing radiation, it is necessary to know accurately well the time and space distribution information of nuclear radiation field intensity in fusion device. The magnetic confinement fusion devices built in the world have established a complete nuclear radiation monitoring system according to their own operating conditions to deal with the potential impact of ionizing radiation. By monitoring the radiation dose during the operation and maintenance of the magnetic confinement fusion device, the ionizing radiation and radionuclide data of the experimental site and the surrounding environment are obtained, which provides data support for radiation safety protection management. Based on the investigation of radiation monitoring systems of magnetic confinement fusion device at home and abroad, the main ionizing radiation source terms and monitoring system architecture are reviewed in this paper, and the measuring methods and common detectors of neutron and γ radiation dose in magnetic confinement fusion are introduced. Finally, the research status of radiation monitoring system for nuclear fusion devices at home and abroad is summarized, and the development trend and goal of nuclear radiation monitoring system in the future are prospected.

BackgroundNECP-SARAX is a neutronics analysis code system for advanced reactors developed by the Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In the past few years, a considerable amount of verification and validation work has been done based on CEFR, PHENIX, SUPERPHENIX, JOYO MK-I, ZPR, and ZPPR reactors. The results indicate that NECP-SARAX offers high performance for fast spectrum reactor analysis. Meanwhile, the fuel and control rod assemblies of these reactors are used for verification of the cross-section generation code TULIP. While TULIP has demonstrated promising preliminary results in fast spectrum system analysis, a comprehensive systematic verification and validation process remains essential.PurposeThis study aims to validate the applicability of TULIP code for various fast spectrum systems.MethodFirstly, a total of 147 critical experiment benchmarks were selected from ICSBEP and used for analysis. The initial results demonstrated that the keff bias between TULIP and Monte Carlo codes exceeded 10-2 for an experimental benchmark with a thick reflector. Then, a homogeneous two-nuclide problem simplified from the HMF021-002 benchmark was subsequently used to analyze this phenomenon, and the intermediate-weight nuclides had resonance-like fluctuating scattering cross sections above the resonance energy. Finally, to address this phenomenon, the TULIP code was undergone enhancements, mainly focusing on optimizing the resonance calculation strategy and method using ultra fine group to deal with thee self-shielding effect of resonance-like cross sections in the non resonant region under high loading of intermediate-weight nuclides.ResultsIn a fast spectrum system with a large amount of structural material, the self-shielding effect of the resonance-like cross section of the intermediate-weight nuclides above the resonance range becomes non-negligible. The optimized TULIP method reduces the keff bias to within 3×10-3 for these benchmarks with a thick reflector.ConclusionsNew numerical results indicate that the enhanced TULIP code has good performance for various fast spectrum system analyses.

BackgroundThe sodium-cooled fast reactor adopts the three loops design with sodium-sodium-water. When a double-ended guillotine (DEG) break occurs in the steam generator (SG) tube, a large leakage sodium-water reaction (SWR) accident occurs, which threatens the safety and integrity of the secondary loop. A protection system is therefore designed to ensure secondary loop integrity.PurposeThis study aims to analyze the influence of protection system critical parameters on the large leakage SWR with paralleling SGs.MethodsFirst of all, a large leakage SWR model, including the water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and protection system models were established. Then, the large leakage SWR model was verified using the experimental data, and the 3-DEG large leakage SWR was simulated on the basis of the secondary loop structure. The integrity of the secondary loop and the protection system response were analyzed. Finally, a sensitivity analysis was performed for the critical parameters of the protection system, including the bursting pressure of liquid rupture disks, bursting delay time of rupture disks, location of liquid rupture disks, length of the release pipe, and volume of the primary accident discharged tank. Parameters with key influence on the integrity of the secondary loop and the protection system response were determined.ResultsThe 3-DEG large leakage sodium water reaction accident results in a peak pressure of 2.003 MPa in the reaction zone and 1.329 MPa in the critical equipment of the secondary circuit except for the reaction zone. The smaller bursting pressure and delay time, and the location of liquid bursting disks at the bottom chamber and the release pipe shorter length are more conducive to the integrity of the secondary loop and the protection system response.ConclusionsThis study provides a reference value for the design requirements of large leakage SWR protection systems with paralleling SGs.

BackgroundTraditional safety analysis methods rely on expert advice and user self-evaluation, lacking the ability to quantify output uncertainty. In contrast, the best estimation plus uncertainty (BEPU) methodology can quantify the uncertainty of the output, thereby avoiding unnecessary conservative assumptions and improving the economic viability of nuclear power. It is now widely used in the design and safety analysis of nuclear reactors. However, owing to the cognitive limitations of science and numerical approximation in programs, most thermal-hydraulic programs lack sufficient input uncertainty information related to internal models, often relying on expert advice.PurposeThis study aims to investigate the uncertainty quantification methodology for model parameters in sub-channel codes using Markov Chain Monte Carlo (MCMC) sampling.MethodsFirstly, the PSBT void fraction distribution experiments were employed to evaluate the prediction ability of the subchannel program COBRA-IV, and a Python-based uncertainty analysis methodology was developed to quantitatively analyze the model parameter uncertainties that affect the void fraction. Then, the model parameters were assumed to be independent, with their uncertainties following a normal distribution. Based on the Bayesian principle, the most likely maximum a posteriori probability function (PDF) of the model parameters were obtained by combining the prior and observed information, despite the limited actual uncertainty information. Finally, an MCMC sampling methodology was adopted to solve the Bayesian relation, and the statistical uncertainty information of the model parameters were obtained using a stable a posteriori Markov chain, which requires at least 104 magnitudes to achieve convergence and the corresponding forward program runs. Therefore, to reduce the calculation cost and improve the calculation efficiency, a high-precision adaptive BPNN surrogate model was constructed to replace the complex and time-consuming forward program code. Furthermore, a set of uncertainty quantification methods with Python was developed to simultaneously quantify the uncertainty of the model parameters using a statistical method. During the selection of a slip model we discovered that both the slip ratio and turbulence mixing coefficient significantly affected the void fraction. Therefore, we developed.ResultsThe results indicate that after obtaining the uncertainty of the model parameters, the 95% confidence interval of the results generated by the forward propagation of input uncertainty enveloped the experimental values well. Furthermore, by incorporating the mean value of the model parameter uncertainties, obtained via uncertainty quantification, the modified model output exhibited a closer agreement with the experimental values than with the reference values.ConclusionsThe uncertainty quantification analysis methodology established in this study can be applied to the uncertainty analysis of subchannel program model parameters.

BackgroundScintillator-based fast-ion loss detector (FILD) can measure the velocity-space distribution of fast-ion loss and are key in studying the control mechanism of fast-ion loss in nuclear fusion devices.PurposeThis study aims to obtain the velocity-space distribution of fast-ion loss from corresponding FILD data and evaluate the capacity of current FILD probe for the further improvement of diagnostic design.MethodsThe FILDSIM code was employed to establish the linkage between the fast-ion image digitized by FILD and the velocity-space distribution of fast-ion loss. The detection coverage of fast ions on the scintillator was assessed through reverse tracing of the lost fast ions, considering the geometry of the FILD probe as well as the pitch angle and gyroradius of fast ions entering the pinhole of the FILD probe.ResultsThe obtained velocity-space distribution of fast-ion loss under ICRH indicates that the energy of lost fast hydrogen minority ions is above 200 keV. Moreover, analysis shows that the geometry of the probe, particularly the shell behind the scintillator, obstructs the diagnostic detection range, creating a null region on the scintillator.ConclusionsThe acquisition of the velocity-space distribution of fast-ion loss lays the foundation for further evaluation and control of fast-ion loss under ion cyclotron resonance heating. In addition, the investigation of the probe detection range provides a basis for upgrading diagnostic systems.

Energy loss during interactions between high-energy particles and target materials mainly consists of nuclear and electronic energy losses. Electronic stopping and electron-phonon coupling effects are two different mechanisms that reflect electronic energy loss effects. To accurately simulate the irradiation damage process of high-energy particles, it is necessary to solve the key scientific problem of the influence of electronic energy loss on irradiation damage. This paper reviews the most recent progress on the irradiation damage behavior study of several key structural materials under the influence of electronic energy loss effects, elaborates the effects of electronic stopping, electron-phonon coupling, and electronic thermal conductivity on irradiation defects. The influence laws of electronic energy loss effects on the irradiation damage of target materialsare summarized and the existing problems in the research of high-energy particle irradiation of target materials are highlighted. Finally, the prospectives are outlined for future research directions.

BackgroundA general-purpose heat source (GPHS) is the most established heat source module for radioisotope thermoelectric generators (RTGs) and a key reference for the equivalence of electrically heated analog heat sources during the development and testing of radioisotope power supplies.PurposeThe study aims to develop a highly efficient electric heating simulation heat source to meet the requirements for equivalent testing and verification of non-nuclear units of radioisotope power systems.MethodsFirstly, an electrically heated analog heat source was designed and imitated based on various actual GPHS performance parameters. Based on the simulation calculations, the thermodynamic equivalent substitutability between the simulated heat source and real GPHS in terms of material and dimensional differences was evaluated. Then, the performance of its operation in different application scenarios was analyzed, and an optimized application environment was proposed. Finally, based on the experimental test results, the uniformity of the thermal output characteristics of the imitation GPHS-simulated heat source was compared with that of a real GPHS, and the practical application characteristics of the imitation GPHS in RTGs were also evaluated.ResultsAt an input power of 250 W, the average surface temperature of the GPHS-simulated heat source reaches 515 K. The temperature variation trend with power is consistent with that of the simulation results. Experimental test results show that the energy conversion efficiency of the RTG module is increased to ~6% with a 250-W heat source power supply.ConclusionsThe proposed and constructed good equivalent simulated GPHS heat source with reference to the thermal properties of a real GPHS can be applied to RTGs and provides an effective and unified reference standard for the performance evaluation of radioisotope power sources.

BackgroundDuring the operation of nuclear power plants, a large amount of low and intermediate level waste (LILW) is generated, which is usually prepared into 200-L and 400-L waste drums. To ensure the safe disposal of these waste drums, they must be analyzed to determine the type and activity of the nuclides contained within them. Non-destructive assay (NDA) has been widely used in the detection of waste drums in nuclear power plants, along with segmented gamma scanning (SGS) and tomographic gamma scanning (TGS). However, the low measurement accuracy of SGS and the long measurement time of TGS limit the practical application of these methods.PurposeThis sudy aims to shorten the measurement time while maintaining high measurement accuracy by proposing a new neural network-based method for measuring the activity of waste drum.MethodsWhen the waste drum was filled with a uniform distribution of medium and rotated at a constant speed during measurement, the point source was equivalent to a ring source. The equivalent ring source in the waste drum possessed an activity equal to the total activity of all sources. The neural network model is established, the count rate of the detector at different positions is used as input, and the radius of the equivalent ring source is used as output. Finally, the total activity of the waste drum is calculated. The simulated measurement is carried out in a 400-L waste drum, the medium is concrete, the radioactive source is Co-60, and 50 groups of single-source and 10 groups of multi-source are generated randomly. Different methods are used to reconstruct the activity of the waste drum.ResultsWhen there is only one radioactive source in the waste drum, the mean relative error (MRE) of activity reconstruction by the new method is 4.26%, which is much lower than that of SGS (68.15%) and close to that of TGS with 60 grids (3.97%). When there are multiple radioactive sources in the waste drum, the MRE of activity reconstruction by the new method is 24.27%, which is lower than that of SGS (48.02%) and close to that of TGS with 60 grids (28.61%). This new method achieves the equal measurement accuracy of TGS but reduce the measurement time to 1/20 of TGS.ConclusionCompared to traditional measurement methods, the new method greatly shortens the measurement time while maintaining high precision, thereby providing technical support for the measurement of LILW.

BackgroundLarge-area plastic scintillators are widely used in radiation monitoring. They are typically employed as gamma radiation counters through signal over-threshold detection. The value of the threshold voltage affects the detector efficiency and minimum detectable activity. When rays are incident at different positions of the large-area plastic, the photon collection efficiencies generated through collecting end pair differ, leading to differences in the detector efficiency.PurposeThis study aims to explore appropriate threshold voltage to reduce the differences and maintain a low minimum detectable activity (MDA), hence achieving high detection efficiency for large-area plastic scintillator.MethodsFirstly, an energy spectra acquisition system was designed using STM32F429 produced by STM semiconductor as the main control chip and a counter for detector efficiency testing of large-area plastic scintillator in size of 400 mm×300 mm×50 mm. Then, the reasonable threshold voltage was determined according to the background energy spectra of the large-area plastic scintillator, the energy spectrum of the 60Co and 137Cs sources at different positions of the large-area plastic scintillator with acquisition time of 3 min, and signal-amplification relationship.ResultsThe determined threshold voltage using above mentioned method for 60Co source is 93.7 mV. To achieve high detector efficiency and low MDA for the large-area plastic scintillator, the best determined threshold voltage for 60Co source is 95 mV. At this voltage, the detector efficiency for 60Co source is 22%, the MDA is 78 Bq.ConclusionsThe proposed method has reference value for large-area plastic scintillation designing and manufacturing detectors.

BackgroundStudying the transport behavior of impurities in plasma and developing effective impurity control methods are important for achieving high-performance plasma discharge and ensuring the safe operation of the device.PurposeThis study aims to design a control system for the experimental advanced superconducting Tokamak (EAST) laser blow-off (LBO) impurity system.MethodsA new automatic control process was adopted to enable to inject tracer particles of different elements into the EAST plasma repeatedly, quantitatively, and controllably. An accurate control of the focusing lens displacement and laser triggering time were achieved through the STM32 microcontroller system and its output PWM waves for stepper motor driving, hence the diameter of the laser spot was adjustable to change the amount of impurities injected. Finally, the designed control system for LBO was tested in staging environment to verify its practicability and accuracy.ResultsThe test results show that the system can rapidly detect the external trigger signal and achieve precise timing, with less than 0.4 mm position error for laser spot focusing.ConclusionsThe design of control system meets the requirements of the laser blowing impurity injection experiment. This study is of considerable significance for research on EAST plasma impurity transport.

BackgroundThe targeted acquisition of the radioactive element content or radioactivity in a radioactivity is an important task in geological exploration and radioactive pollution investigation. During the process of targeted gamma radiation sampling, gamma rays from non-target areas significantly interfere with the measurement results.PurposeThis study aims to design a dual-energy targeted gamma radiation sampling probe that uses a cerium bromide scintillation detector on the basis of the difference in the linear attenuation coefficients of the high- and low-energy gamma rays from the same radioactive decay series in the lead shielding layer of the detector.MethodsFirstly, Monte Carlo (MC) numerical simulations were employed to determine the optimal lead shielding layer thickness for the dual-energy targeted gamma radiation sampling probe detecting targets of the 0.609 MeV and 1.764 MeV gamma rays emitted by 214Bi. Then, the directional proportionality coefficients were calculated and applied to obtaining 0.609 MeV gamma ray counts of dual energy γ radiation probe within lead shielded angle. Finally, MC numerical simulations with four types of interfering radiation sources and physical experiments with two radium sources were conducted to validate the results that calculated using the directional proportionality coefficients.ResultsThe simulation result of optimal lead shielding layer thickness for the dual-energy targeted gamma radiation sampling probe is 6 mm, and the calculated directional proportionality coefficients of a and A are 0.268 and 0.451, respectively. Validation results show that the relative error between the counts for the 0.609 MeV gamma rays within the shielded angle and the net peak area counts measured with the dual-energy targeted gamma radiation sampling probe for two radium sources is within ±2.52% with average relative error of 0.63%. The relative errors between the measured uranium content and recommended values in the tested models are all less than 5%.ConclusionsThe dual-energy targeted gamma radiation sampling results for a radioactive mixed standard model and three radioactive models indicated that the designed dual-energy targeted gamma radiation sampling probe is capable of targeted gamma radiation sampling.

BackgroundNickel-, iron- and tungsten-based alloys are commonly used as structural materials of reactors. During their operational life, these alloys undergo intense neutron irradiation.PurposeThis study aims to analyze the post-irradiation defect evolution and its mechanisms in these materials for comprehending the effects of irradiation on them.MethodsThe displacement cascades in nickel, iron, and tungsten were examined at various temperatures (300?500 K), primary knock-on atom (PKA) energies (<20 keV), and directions (<135>, <122> and <100>) by using molecular dynamics (MD) simulations. Firstly, the model was initially relaxed at each specified temperature under a canonical ensemble for 10 ps, applying periodic boundary conditions in every direction. Then, an atom was randomly chosen as a PKA and assigned kinetic energy to initiate the cascade collision simulation in the micro-canonical ensemble. Finally, the Open Visualization Tool package was employed for visualization and data analysis of the irradiation cascade processes.ResultsThe simulation results reveal that nickel and iron exhibit similar steady-state defects. At lower PKA energies (<5 keV), nickel exhibits marginally fewer defects than iron. However, as the PKA energy surpasses 5 keV, the number of defects in nickel becomes slightly more than that in iron. Furthermore, under identical irradiation conditions, tungsten demonstrates superior radiation resistance, with fewer steady-state defects when compared with both nickel and iron.ConclusionsThe defect evolution during various cascade displacement phases in three metals and their defect recombination rates are crucial to understanding the disparities in radiation damage resilience. The derived results help to comprehend the radiation characteristics of these metals. Additionally, the primary radiation damage dataset compiled for these metals lays a foundation for further larger-scale simulations of their radiation attributes using rate theory or cluster dynamics methods.

BackgroundDue to the specific imaging structure of neutron imaging system, the geometric unsharpness is unavoidable in neutron photographic images after imaging.PurposeThis study aims to design a geometric unsharpness correction method for neutron photographic image based on improved Richardson–Lucy (RL) algorithm.MethodsFirst, according to the geometric unsharpness of the neutron photographic image, the mathematical model of the point spread function (PSF) was established. Then, the Laplace operator and median filter were used to remove the PSF related gamma spots noise (GSN) in the image, and the neutron photographic image was restored by RL algorithm. Finally, the restoration quality of neutron photographic images of the line-pair sample was evaluated using average gradient (AG) and spatial frequency (SF).ResultsThe results of the line-pair sample demonstrate that compared with four existing correction algorithms of geometric unsharpness, the proposed method improves AG and SF by 60.23% and 29.90%, respectively.ConclusionsThe proposed method of this study can effectively correct the geometric unsharpness caused by the amplification of gamma spots noise in the process of neutron photographic image restoration, providing an important technical support for performing high-resolution neutron radiography.

BackgroundThe barrel-sampled electromagnetic calorimeter (Ecal) is an important part of the Multi-Purpose Detector (MPD) in Nuclotron-based Ion Collider fAcility (NICA) that is being built in Russia. It is primarily used to detect energy, time, and position information of electrons and photons in the energy domain from 10 MeV to a few GeV. MPD-Ecal is comprised of 2 400 modules with 16 towers per module. Each tower is made up of alternating layers of 211 scintillator sheets and 210 lead sheets, as well as 16 wave-length shift fibers.PurposeThis study aims to evaluate the performance, such as energy resolution, time resolution, and coordinate resolution, etc., of the Ecal by simulation.MethodsThe Geant4 software was employed to simulate single-energy electron incident on Ecal to examine the effects of several parameters on the performance of Ecal. Influences of the position of the particle incidence point, the number and thickness of the scintillator and lead layers, the polish of the optical fibre end-face, and the energy and type of incident particles on the energy deposition and resolution, time distribution and time resolution, and coordinate resolution were investigated in details. Finally, the time resolution of a single tower was simulated using the natural cosmic ray package, and the tower's expected time resolution in the cosmic ray test was obtained.ResultsAs the electron incidence position moves from the edge to the center of the module, the energy deposition within the scintillator rises from 718 MeV to 758 MeV. With a limited tower length of 415.5 mm, increasing the number of scintillator layers decreases the energy resolution of the module, improves the time resolution of the tower, and worsens the coordinate resolution of the Ecal. Taking into account the performance of the Ecal gauge, the optimal number of scintillator layers in the tower is 211. SiPM detects 42% more photoelectrons at a polish of 0.6 when the fiber end is coated with a reflective material than when it is not. As the polish of the fiber end-face increases, so does the number of photoelectrons detected by the SiPM, and the time resolution of the tower improves. When the fibre end-face polish is 1.0, the time resolution of the tower is less than 103 ps while time resolution of the tower (with 211 layers) in the cosmic ray is 185 ps.ConclusionsThe time and coordinate resolutions of Ecal improve with increasing electron energy under the same circumstances.

BackgroundWater is the primary resource required for the exploitation of lunar resources. Investigating the distribution of water on the surface of the moon has become a focal point in the lunar exploration plans of several nations.PurposeThis study aims to quantify the presence of hydrogen and analyze its spatial distribution on the lunar surface using data from the Chang'E-2 gamma-ray spectrometer (CE2-GRS).MethodsFirstly, an analytical method combining branch specific stripping with the nonlinear least-square fitting Gaussian function was proposed to subtract the characteristic γ rays of interfering nuclides (214Bi@2.204 MeV, 27Al@2.210 MeV, 49Ca@2.371 MeV) ranging from 2.1 MeV to 2.5 MeV. Then, a characteristic function between the abundance and counts of aluminum γ rays around the moon was established to subtract the counts of aluminum in the mixed peak. Finally, the spatial distribution of hydrogen γ rays for counts per 3 s on the whole lunar surface was obtained.ResultsThe analysis results show that high-value characteristics exhibited in some areas, including the Aiken basin, Mare Ingenii, Mare Imbrium, and Oceanus Procellarum, are approximately 2.6 times the average value of hydrogen counts among the 14 major maria. Comparison between the distribution characteristics of hydrogen elements on the lunar surface and the data of epithermal neutrons from Lunar Prospector (LP) reveals a highly negative correlation between the distribution characteristics of the two in these regions.ConclusionsThe distribution characteristics of hydrogen elements on the lunar surface further predict that there may be a large amount of structural water in the Mare Ingenii, Mare Imbrium, and Oceanus Procellarum, formed by the combination of hydroxyl groups or molecular hydrogen (H2), achieving a better understanding of the orbit γ deep mining and scientific application of energy spectrum data.

BackgroundThe in-vessel retention (IVR) strategy is an important measure for mitigating severe reactor accidents. It has been successfully applied to the severe-accident management of advanced pressurized water reactors such as the AP1000, HPR1000, and CAP1400. In the implementation of the IVR strategy, the lower head is deformed by the heat load of the high-temperature melt. This affects the heat-removal capacity of the pressure-vessel external cooling and the successful implementation of the IVR strategy. It is necessary to examine the stress, failure, and deformation of the lower head.PurposeThis study aims to develop a large deformation model for the pressure-vessel lower head and analysis of its application in the FOREVER experiment.MethodsA mechanistic model called the lower-head large-deformation model was developed to address the limitations of the simplified film stress model of the Integrated Severe Accident Analysis (ISAA) program Lower Head Thermal Creep Module (LHTCM), which is very simple, and the absence of a deformation calculation module in the LHTCM model. This model was based on Timoshenko plate and shell theory, the Norton creep law, and large-deformation plasticity theory. Then, the model was integrated into the ISAA program to calculate the FOREVER-EC2 experiment.ResultsThe overall deformation result predicted by the large-deformation model exhibits the characteristic egg-like shape, with maximum displacement occurring at the bottom position of the lower head. The failure time predicted by the large-deformation model is 394.33 min, with an error of only 1.9% relative to the experiment. The predicted bottom elongation is consistent with the experimentally measured value. Additionally, the predicted location of the breach is consistent with the experiment, occurring between 75o and 85°.ConclusionsThe lower-head large-deformation model of this study can accurately predict the stress, failure time, overall deformation, and location of the breach of the lower head in a severe core melting accident.

BackgroundInstitute of Modern Physics, Chinese Academy of Sciences is developing a high-precision calibration device for lead-bismuth flowmeters based on the static mass method. In static mass method-based calibration devices, the uncertainty component of the diverter is the main component of the uncertainty of the flowmeter calibration device.PurposeThis study aims to predict the system error of a cylinder-driven diverter due to the asymmetry of cylinder forward and reverse strokes in the calibration process of lead-bismuth flowmeters, realize the a priori analysis of class B uncertainty components of a cylinder-driven commutator, and investigate whether the class B uncertainty can accurately reflect the magnitude of the relative system error introduced by the diverter.MethodsFirstly, a two-way coupling calculation method based on computational fluid dynamics (CFD) was developed for the diverter, and the SST k-ω turbulence model and Euler multiphase flow model were employed for calculation. Then, the stroke difference method was employed to obtain the class B uncertainty components for the diverter under different flow conditions and torque drives, and the calibration relative system errors under different timing methods and velocity distributions at the commutator nozzle outlet were calculated. Finally, the operating characteristics of the cylinder-driven diverter were analyzed.Results & ConclusionsComputation results show that the greater the thrust of the cylinder, the smaller is the type B uncertainty of the diverter. Compared to using the start or end of stroke as the timing moment, the calibration error caused by the diverter is the lowest when the diverter baffle is moved to the midpoint of the stroke as the timing moment, and the error is approximately 1/7?1/30 times those of the other two timing methods. The magnitude of diverter class B uncertainty reflects the envelope value of the relative systematic error caused by the diverter when using different timing moments.

BackgroundMolten salt reactors (MSRs) are fourth-generation advanced nuclear energy systems that exhibit characteristics such as high safety, high economy, nonproliferation, and sustainability. To ensure the safe operation of MSRs, identifying transient conditions promptly and accurately is crucial. However, current system transient identification methods rely on manual identification by operators, introducing significant human factors seriously affecting nuclear power safety.PurposeThis study aims to establish a transient identification model for an MSR system based on the K-nearest neighbor (KNN) method, so as to reduce human factors introduced during the traditional system transient identification process, and improve the operational safety of the MSR.MethodsDatasets for the system transient identification model were generated by using the RELAP5-TMSR code to simulate 11 operating conditions of the molten salt reactor experiment (MSRE) built and operated at Oak Ridge National Laboratory in the United States. Subsequently, a system transient identification model based on the KNN method was developed by training, optimizing, and validating these datasets. Four metrics, i.e., accuracy, precision, recall, and F1-score were applied to evaluating the system transient identification model. Finally, the robustness of the model was tested and optimized under noisy conditions.ResultsThe results demonstrate that the KNN-based transient identification model for the MSR system achieves a 99.99% F1-score on the test datasets. The system transient identification model also exhibits high robustness, with an F1-score of 94.32% under noisy conditions. The optimized system transient identification model achieves a 99.73% F1-score when identifying transient conditions under noise, accurately identifying the transient conditions of the MSRE.ConclusionsThe KNN-based transient identification model for the MSR system can satisfy the requirements of transient identification of the MSR system, hence be applied to intelligent MSR operations and maintenance, ensuring safe MSR operation.

BackgroundThe kilowatt reactor using stirling technology (KRUSTY) is a heat-pipe-cooled reactor experimental system that uses a Stirling engine to convert thermal energy to electricity, it is the only one published experimental data for heat-pipe-cooled reactor systems. The KRUSTY experimental data under different working scenarios include the cold startup and load change processes, heat pipe failure, reactivity insertion, and heat sink loss.PurposeThis study aims to validate the self-developed system transient analysis code named TAPIRS-D for the heat-pipe-cooled reactor concept using KRUSTY experimental data.MethodsFirstly, an in-house system code for a heat-pipe-cooled reactor named TAPIRS-D was introduced, with the main theoretical module briefly explained, including the reactor power calculation module, heat transfer module for fuel assembly, and heat pipes. Then, the TAPIRS-D was applied for the first time to the simulation of the key processes of the KRUSTY prototypic reactor test under normal operation and accident conditions. Finally, comparison between the simulation data and experimental data was conducted for the validation of this analysis code.ResultsComparison results demonstrate that the maximum relative prediction error for the fuel temperature is less than 2%, and the reactor power average prediction error is less than 10%.ConclusionsThe prediction trend of the numerical simulation by TAPIRS-D fits well with the experimental data on key parameters such as core power and the temperature of fuel and heat pipes, which indicates that TAPIRS-D is well developed and is capable of conducting safety analysis for heat pipe cooled reactor concepts. The validation of this system analysis code provides a good reference for other newly developed system codes for heat pipe reactors.

BackgroundThe annular fuel assembly of lead-bismuth cooled fast reactor has many safety advantages, but during its operation, due to the corrosive effect of lead-bismuth coolant, it is prone to blockage accidents, resulting in deterioration of heat transfer and jeopardizing the integrity of the first barrier. Therefore, it is urgent to research and analyze the blockage accident for the annular fuel assembly of the lead-bismuth-cooled fast reactor.MethodsA 5×5 single annular fuel assembly model was established, and numerical simulations for blockage of the inner and outer channel were carried out with different blockage areas, blockage thicknesses, and axial position of the blockage based on the computational fluid dynamics (CFD) software Fluent. The temperature distribution of the claddings, the flow field distribution near the blockage, the mass flow change of the channel, the radial temperature distribution, and the heat distribution of the fuel element at the blockage are compared with the result in no-blockage case.ResultsSimulation results indicate that the increase in the blockage area leads to a significant increase in the cladding temperature of the blockage area, an expansion of the scope of the recirculation area expands, the position of the highest temperature point of the fuel pellet shifts to the blockage side, and the heat flux density on the blockage side decreases. When the blockage fraction is large, the changes of parameters are similar to the above conclusions as the blockage thickness increases; when the blockage is located at the entrance, the local temperature rise of the cladding is smaller than that when the blockage is located at the center; with the increase of the blockage area and thickness and as the blockage position gets closer to the entrance of the active zone, the flow loss of the inner channel increases significantly, while the flow of the outer channel is almost unaffected.ConclusionsTherefore, the damage is more serious when the blockage accident occurs in the inner channel.

BackgroundNuclear power simulation technology has been widely applied in areas such as reactor design, safety analysis, independent safety evaluation, accident mitigation measures, design and optimization of control and protection system, verification of advanced main control room design, and operator training. This technology has effectively improved the safety and economy of nuclear power plants.PurposeThis study aims to develop a real-time modeling and simulation platform for a liquid-fueled molten salt reactor (LF-MSR) based on the open-source and open architecture of the experimental physics and industrial control system (EPICS).MethodsFirst, the real-time interaction function of the LF-MSR system code, RELAP5-TMSR, was improved, and the control and protection system and human-machine interaction interface were extended. Then, the ThorTypography simulation platform was preliminarily developed for LF-MSR by integrating the above three main functional modules. Finally, ThorTypography was validated using the pump start-up experiment, pump coast-down experiment, natural circulation experiment, and reactivity insertion experiment from the Molten Salt Reactor Experiment (MSRE) as the benchmark.ResultsThe test results of ThorTypography are consistent with the calculation results of RELAP5-TMSR and are in good agreement with the MSRE data. Moreover, the total simulation time is consistent with the total physical problem time.ConclusionsThorTypography is suitable for real-time modeling and simulation of LF-MSR systems and can provide effective support for LF-MSR design, operation, and safety analysis.

BackgroundGenerally, pulse truncation events caused by measurement systems often present challenges to pulse height analysis in the field of spectroscopy and radiometry, resulting in spectral distortion.PurposeThis study aims to propose a composite neural network model for accurately estimating the heights of truncated pulses.MethodsFirstly, a long and short-term memory (LSTM) network was embedded into the UNet structure to construct a composite neural network model (LSTM-UNet). Then, the model was trained for height estimation of truncated pulses output by silicon drift detectors using a simulated pulse dataset for which the pulse amplitude matrix superimposed with noise was taken as input signal while the output signal was a set of expanded pulse heights. Finally, the performance of the model using relative error indicators was evaluated by analyses of powder iron ore and powder rock samples.ResultsThe average relative error of the UNet-LSTM model for pulse height estimation analysis on simulated pulse sequences is approximately 2.31%, which is 1.91% lower than the average relative error of traditional trapezoidal shaping algorithms. Verification results of the UNet-LSTM model on measured pulse sequences with different degrees of truncation show that the average relative error obtained during the height estimation of two samples and eight sets of offline pulse sequences is 2.36%.ConclusionsThe results reveal that the proposed model can accurately estimate truncated pulse heights.

BackgroundAs a fundamental property of peptide molecules, chirality has been demonstrated to play an important role in controlling the structures and characteristics of peptide supramolecular systems. However, the mechanism through which chirality takes effect has not been clarified.PurposeThis study aims to examine the self- and co-assembled nanostructures and analyze the intermolecular interactions that drive the assembly by employing diphenylalanine (FF), along with the core recognition sequence of Amyloid-β protein (Aβ) and its enantiomer D-Phe-D-Phe (ff), in a model system.MethodsA series of structural and morphological analyses were conducted in the experiments. First, the scanning electron microscopy (SEM) and atomic force microscopy (AFM) images of the assembled nanostructures were obtained to observe the microscopic morphology and topological structure of the assembled FF and ff. Subsequently, circular dichroism (CD) and Fourier transform infrared spectroscopy (FTIR) were employed to characterize the secondary structures of peptides in the nanostructures. Finally, fluorescence emission spectrum and X-ray diffraction (XRD) analyses revealed the intermolecular interactions between peptide and solvent molecules.ResultsThe findings demonstrate that FF and ff self-assemble into similar fibrous nanostructures, and their chirality primarily affects the interactions between peptide molecules, as well as those between peptide and water molecules. Furthermore, the formation of new crystalline phases for the co-assembly of FF and ff was confirmed by XRD.ConclusionsOur results may facilitate the understanding of the formation mechanism of amyloid fibers and design of peptide supramolecular materials.

BackgroundAccurately calculating the borehole size and standoff between the tool and borehole wall is the premise for correcting the effect of the borehole environment and improving measurement precision in logging while drilling (LWD).PurposeThis study aims to obtain an accurate calculation method for the standoff size of LWD.MethodsFirstly, the theoretical relationship of the standoff was derived based on the broad-beam γ-ray attenuation model for azimuth density LWD. Then, Monte Carlo N-Particle transport code (MCNP) was employed for simulation, and results were benchmarked according to experimental tool data. By analyzing the influence of standoff, mud and formation density, and other factors on the detector response, an accurate standoff calculation formula was derived. Finally, logging curves were drawn using the CIFLog platform and the calculated standoff results were compared with the measured ultrasonic results, hence to verify the accuracy of the proposed standoff calculation method.ResultsThe calculated standoff results in the simulation are fundamentally consistent with the theoretical value. The calculation result using measured data processing is in approximate agreement with the ultrasonic standoff, and the calculated calipers have good correspondence with the wireline well diameter. The proposed method provides accurate measurement even when the standoff is small and ultrasonic standoff measurement is abnormal. The applicable range of standoff calculation based on azimuthal density LWD is 0~3.81 cm.ConclusionsThe broad-beam γ-ray attenuation based standoff calculation method for density LWD complements to larger range ultrasonic measurements to provide key parameters for borehole correction for LWD tools.

BackgroundGallium nitride (GaN) power devices have garnered attention in the anti-irradiation field owing to their excellent performance.PurposeThis study aims to explore the anti-γ-ray damage ability of gallium nitride power devices and clarify the mechanism of radiation degradation.MethodsFirstly, the domestically produced commercial NP20G65D6 P-GaN gate enhanced AlGaN/GaN High Electron Mobility Transistor (HEMT) device was taken as test sample. Then, 60Co γ-ray source with different irradiation doses of 0.3 Mrad (Si), 0.6 Mrad (Si), and 1.0 Mrad (Si), respectively, was employed to conduct total dose irradiation experiments under different bias (ON-state, OFF-state, and GND-state) conditions and annealing tests at different temperatures for enhanced AlGaN/GaN HEMT devices. Finally, the response law between the electrical performance of the device and the bias condition and annealing environment were analyzed to reveal the degradation mechanism of device sensitive parameters.ResultsThe results indicate that as the γ-ray irradiation dose increases, the device's threshold voltage exhibits a negative drift, and the transconductance peak, saturation leakage current, and reverse gate leakage current gradually increase. Simultaneously, the electrical characteristics of the device deteriorate more rapidly under the ON-state bias condition. Furthermore, annealing at high temperatures leads to a more apparent recovery of the electrical properties of devices. The analysis demonstrates that the higher the γ-ray irradiation dose, the more radiation defects are generated. The gate bias reduces the initial recombination rate of electron-hole pairs caused by irradiation, increases the number of holes escaping the initial recombination, and further increase the concentration of defect charge. The high-temperature environment causes tunneling annealing or thermal excitation annealing, which is conducive to the recovery of device performance.ConclusionsThe radiation damage process and mechanism of gallium nitride power devices of this study provides data support for evaluating and verifying its application in a space environment.

BackgroundNuclear reactions have been crucial in the evolution of the universe since the big bang. The cross section of nuclear reaction that occurs during the early evolution of the star is extremely low, so it cannot be accurately measured in a ground laboratory. The China Jinping Underground Laboratory (CJPL), which is the deepest operational underground laboratory in the world, offers unique ultra-low background conditions that facilitate the direct evaluation of the nuclear reactions occurring during the early evolution of stars.PurposeAsymptotic giant branch (AGB) stars are thought to be the major contributor to Galactic fluorine production. However, the astronomical fluorine overabundance cannot be explained by using the current standard AGB models. Direct measurements of 19F(p,αγ)16O reactions can help solve this problem.MethodsExperiments were conducted on a high-current 400-kV JUNA (Jinping Underground laboratory for Nuclear Astrophysics) accelerator at the CJPL. A 4π BGO detector array was specially designed for the JUNA project.ResultsThe astrophysical S factors in the energy region of 72.4~188.8 keV were experimentally derived for the first time, covering the astrophysical Gamow window. The thermonuclear 19F(p,αγ)16O rate was determined at a low temperature of about 0.05 GK for astrophysical modeling. The present low-energy S factors significantly deviated from previous theoretical predictions, and the associated uncertainties were considerably reduced.

BackgroundThe calibration of radon measuring instruments is the most important and key technical link to ensure the reliability and accuracy of radon observation data and plays a vital role in seismic monitoring and prediction.PurposeThis study aims to obtain a calibration method suitable for the calibration of newly connected seismic water radon monitor according to the National Metrological Verification Regulations for Radon Meters (JJG825-2013), the calibration experiment is carried out for the analog radon observatory (manual), which is newly adopted by the seismic network, and the calibration method suitable for the seismic water radon observation is obtained.MethodsFirstly, the FD-125 radon-thorium analyzer was used as the experimental instrument, the calibration experiments were carried out in the standard radon chamber by three calibration methods: circulation method, vacuum method, and flow-gas method. Then, the volume response coefficients of three methods were calculated, and the uncertainty sources and mathematical models were analyzed according to JJG825—2013 regulation. Measured values by internationally recognized standard AlphaGUARDPQ2000Pro radon meter were taken as theoretical references of radon concentration for radon chambers, hence the uncertainties of three calibration methods were obtained.ResultsThe results show that the circulation calibration method has the highest detection efficiency and fewer influencing factors, and is suitable for manual and digital radon detector calibration. The vacuum calibration method is affected by the standard radon gas diluted by pipeline gas and pressure balance, resulting in low negative pressure sampling efficiency and high volume activity response coefficient R. As for the flow-gas calibration method, the continuous injecting of radon gas into the scintillation chamber affects the dynamic stability of radon gas in the scintillation chamber, and radon gas is not discharged to the background during calibration, resulting in the high volume activity response coefficient R and high intrinsic error.ConclusionsThe cyclic method is more suitable for the calibration of new seismic water radon observatories, and this study provides a reference for the calibration of other new water radon observatories and gas radon observatories.

BackgroundThe Energy Resolved Neutron Imaging Spectrometer (ERNI) of China Spallation Neutron Source (CSNS) is in process of building, whose diffraction detector with ninety degree partition adopts 6LiF/ZnS neutron scintillator detector with a 90° partition as its detection equipment.PurposeThis study aims to develop the readout electronics for high position resolution neutron scintillator detector of ERNI.MethodsFirstly, a capacitive network combined with center-of-gravity of the induced charge distribution was adopted for the design of readout electronics. Then, a prototype of the readout electronics system consisted of three parts: capacitive network circuit, preamplifier board and digital readout board, was developed. After functional verification, the relevant performance parameters of the developed prototype were experimentally tested in the laboratory and in the No.20 beam line of CSNS.ResultsThe experimental results show that the integration nonlinearity of electronics is better than 0.95%, the time resolution is about 12 ns, the position resolution is 1mm, and the detection efficiency is 65%@1.6 ?.ConclusionsThe prototype meets the design specifications of the project. The successful development of the prototype provides a reliable technical support for the smooth development of the spectrometer experiment in ERNI of CSNS in the future.

BackgroundAlthough the neutron image conversion screen is a key component of thermal neutron radiograph, its parameters can severely affect both the spatial resolution and thermal neutron-photon conversion efficiency.PurposeThis study aims to design a neutron image conversion screen for a thermal neutron transmission imaging system based on a compact D-D neutron source.MethodsFirstly, the Geant4 (Geometry and Tracking) program was used to simulate the physical process of thermal neutron transmission imaging and two-dimensional images of transmitted photons, and establish a thermal neutron radiography model based on LiF(ZnS) and LiF(GOS) image conversion screens, and the Siemens star image indicator model. Then, the line spread function (LSF) was employed to calculate spatial position resolution of neutron transmission imaging, and the relationships between the thickness of thermal neutron image conversion screens and the spatial resolution, as well as that between the thickness of thermal neutron image conversion screens and neutron-photon conversion efficiency were evaluated and calculated. Finally, based on parameters of thermal neutron radiography imaging system based on compact D-D neutron source at Lanzhou University, recommended thicknesses for LiF(ZnS) and LiF(GOS) conversion screens were applied to the spatial resolution test experiments.ResultsThe recommended thicknesses for LiF(GOS) and LiF(ZnS) image conversion screens are 40 μm and 80 μm, respectively, the spatial resolution of the thermal neutron radiography reach 45 and 63 μm, respectively, and the neutron-photon conversion efficiencies are 136.34 and 126.81, respectively.ConclusionsThis study lays the technical basis for the development of a thermal neutron radiography based on compact D-D neutron sources. It may be also applicable to other thermal neutron imaging systems.

BackgroundRadio Frequency Quadrupoles (RFQs) are applied widely in proton accelerator facilities. The development of the high-frequency RFQs enables the construction of compact proton accelerator facilities, but faces with more tuning difficulties. The traditional tuning methods based on the theoretical characteristics of the RFQ cannot achieve good results with compact proton accelerators. The tuning method based on the response matrix and SVD method meets the problem of the solution being out of the range that tuners can reach.PurposeThis study aims to purpose a novel tuning method for efficient reduction of dipole components by a transform of the tuning method, so as to tune the compact RFQs better and limit the range of the solution.MethodsFirstly, a tuning method based on the response matrix and the least squares method was designed and implemented. The solution was limited, and a different weight was assigned to the dipole components for diminution during the tuning progress. Then, the tuning method was experimentally tested in the simulation environment on the aluminum RFQ prototype for verification. Both tuning with single tuner and multiple tuners were tested in simulation.ResultsVerification results show that high precision is achieved and the solution is within the expected range even without limitation. Experimental results of verification on the aluminum RFQ prototype show that the quadrupole and dipole component errors are 1.57% and 24.09%, respectively, in the initial state, but reduce to 1.39% and 2.33%, respectively, after five rounds of tuning.ConclusionsThe novel tuning method based on response matrix is verified by this study for its validity of limiting the range of the solution and reducing the dipole components efficiently, hence can be applied to RFQs operating at other frequencies as well. It can contribute to the development of compact proton accelerators and promote medical proton facilities in the future.

BackgroundUltra-high dose rate (UHDR) radiation of electron or proton beam has been shown to spare normal tissues surrounding the tumors while killing tumor cells effectively which is called the FLASH effect (FE). However, the internal mechanisms of FE has not yet been fully revealed, and the optimal parameter range for its use remains unknown.PurposeThis study aims to design an UHDR cell irradiation experimental platform that provides a stable, appropriate and wide range of adjustable dose and average dose rates for exploring the FE dependence on total dose.MethodsBased on a 7 MeV medical proton linear injector, a single scattering nozzle was designed and optimized using the Monte Carlo code FLUKA. A 40-μm-thick tantalum foil, acting as both a vacuum window and a scatterer, was comprised in the nozzle with a source-to-surface distance of 26 cm. Finally, a single pulsed shoot-through UHDR cell penetration irradiation experiment was conducted by simulation using optimized parameters for this platform.ResultsThe simulation results demonstrate that the experimental platform can provide a 2 cm diameter irradiation field with a dose homogeneity of 4.9%. By adjusting the beam intensities (0.1~1 mA) and pulse widths (20~200 μs) of proton beam pulses, the dose and corresponding average dose rate of this platform can be adjusted within the range of 6~667 Gy and 3.3×105~3.3×106 Gy·s-1, respectively. Results of simulated UHDR cell irradiation experiment show that the monolayer cells can be irradiated using a single pulsed shoot-through mode with a dose rate of 3.3×105 Gy·s-1 and doses ranging from 7~40 Gy.ConclusionThis platform enables UHDR experiments to explore the FE dependence on total dose, providing further experimental data for clarifying the FE mechanisms.

BackgroundLoss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal–hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.PurposeThis study aims to develop theoretical models of the reflooding stage in addition to providing benchmark data for evaluating the safety analysis code for LOCAs in a reactor and for the design of the residual heat removal system.MethodsA series of bottom reflooding tests were conducted on a 5×5 rod bundle in the film boiling test facility at the nuclear safety and operation laboratory (NUSOL) of Xi'an Jiaotong University using uniformly heated rods. The experimental results were analyzed in detail, and the surface parameters of the heated rod bundle were obtained by solving a one-dimensional transient inverse heat conduction problem. The effects of different experimental conditions on the velocity of the quench front propagation were investigated. Furthermore, the experimental results were compared and calculated using the thermal safety analysis code, and the problems with the thermal safety analysis code RELAP5 reflooding simulation are summarized.ResultsOur results indicate that a high inlet flow rate, high inlet subcooling degree, and low power density are favorable for the propagation of the cold front during the reflooding process. Additionally, the root mean square (RMS) error of the simulated quench time and peak cladding temperature (PCT) are 40.994 s and 61.465 K, respectively. However, the simulation results have a relatively large error compared with the experimental results in the post-critical heat flux (CHF) heat transfer stage, primarily owing to the issues with the boiling mode judgment and membrane boiling heat transfer model.ConclusionsThe experimental data of this study can serve as new verification data for flow and heat transfer prediction models during the reflooding process; it can also be used to evaluate and optimize the thermal-hydraulic safety analysis code. Loss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal-hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.

BackgroundDouble-layered heat exchanger tubes can effectively reduce the probability of heat exchanger tube rupture accidents from design; however, the thermal contact resistance between tubes may decrease the heat transfer efficiency of the heat exchanger tubes. This is not conducive to the smooth heat export of the first circuit system of lead-bismuth reactor; therefore, there is an urgent need to optimize the design of heat exchanger tubes.PurposeThis study aims to optimize the structure of the double-layered tubular main heat exchanger, so as to improve its heat transfer efficiency.MethodsFirstly, the double layer heat exchange tube type main heat exchanger of lead bismuth reactor was taken as the research object, the interstices of double-layered heat exchanger tubes was filled with gallium-based graphene nanofluid as the thermal interface material. Then, the influence of heat exchanger tube length, wall thickness, outer diameter, and spacing on the heat transfer performance were analyzed and results were compared with that of a double-layered tubular main heat exchanger without the thermal interface material. Finally, taking the JF factor and cost efficiency ratio (CER) as optimization objectives and using the aforementioned four parameters as optimization variables, the heat transfer performance of the main heat exchanger was optimized and evaluated on the basis of a genetic algorithm, hence to obtain a new double-layered heat exchanger design for a lead-bismuth reactor.ResultsThe results of comparative calculations indicate that the total heat transfer coefficient, first-loop pressure drop, JF factor, and CER factor of optimized main heat exchanger increase by 5.79%, 2.32%, 5%, and 24.62%, respectively.ConclusionsFilling the gap of double-layered heat exchanger tubes with a gallium-based graphene nanofluid can effectively improve the heat transfer performance of the double-layered tube exchanger while reducing the accident probability of steam generator tube rupture.

BackgroundInert matrix fuel (IMF) can efficiently convert plutonium and long-lived minor actinides used for preventing the proliferation of nuclear weapons and improving spent fuel disposal, hence has been becoming a hot research topic in recent years. The sol-gel method has the advantage of uniform elemental distribution of the products and the wet operation process is less likely to produce radioactive dust, therefore, it has been used to prepare zirconium-based IMF in the research.PurposeThis study aims to prepare a colloidal solution with good dispersive properties and to obtain IMF microspheres with good sphericity, uniform size, and homogeneous elemental distribution.MethodsFirst of all, ThxZr1-xO2 inert matrix fuel was prepared by an external gelation process, and the sol-gel viscosity was used as the main gelation index. Then, the variation tendency of sol viscosity with c(NH4+)/c(NO3-) was investigated for different metal ions concentrations and different temperatures. Finally, the statistical distributions of colloidal particle sizes were obtained for different metal ions and reaction temperatures by laser particle sizing tests, and the X-ray diffraction (XRD) was used to study the structure of IMF after heat treatment at different temperatures.ResultsThe results showed that the complex gelation parameters and properties can be categorized and quantified using gelation field diagrams. In addition, ThxZr1-xO2 IMF kernels with uniform element distribution, good sphericity, and integral appearance were obtained by optimizing the process parameters. Zirconia showed low solubility behavior in the thorium-oxide system, leading to the generation of a biphasic structure.ConclusionsThe results of this study indicate that zirconium-based spherical IMF microspheres with good performance can be prepared by external gelation method.

BackgroundThe magnitude of the thermal stress in the first wall system is one of the key factors affecting the safe operation of the fusion reactor.PurposeThis study aims to investigate the effect of a rough substrate on the thermal stress in the W/316L stainless steel first wall system.MethodThe finite-element analysis software Ansys Workbench was employ to analyze the distribution of thermal stree in a W/316L stainless steel first wall system with a rough substrate. Depth analysis was conducted on factors such as the temperature, coating thickness, and substrate thickness that affect the magnitude of thermal stress. Meanwhile, starting from the interface shear stress in the system, the influence of rough substrate on the bonding strength of coatings was simultaneously investigated.ResultsThe simulation results indicate that the thermal stress in the rough substrate system increases with the the increase of temperature and substrate thickness, but decreases with the increase of coating thickness. The maximum thermal stress and the adhesive strength between the coating and the substrate are raised by the introduction of the rough substrate.ConclusionsResults of this study can provide reference for the development of high-adhesive strength first wall coating systems.

BackgroundThe recoil release caused by the collision of a neutron with a target nucleus has a significant impact on the analysis of activated corrosion product sources in a reactor. In water cooled reactors, a recoil release in the irradiated area can cause the activation corrosion products to leave the wall surface and enter the coolant, which then migrates to the non-irradiated area with the flow of coolant, thereby making the non-irradiated equipment also radioactive.PurposeThis study aims to analyze the influence of recoil release on the source term of activated corrosion product in reactor.MethodsBased on the investigation of the mode of action of recoil release in a reactor, a calculation model and a program module for recoil release was established and integrated into the CATE program. Then, the effects of recoil release on the concentrations of activation corrosion products in the nuclear reactor were analyzed by using two approaches, one involved specifying a recoil release probability, while the other involved dynamically calculating the recoil release probability. Finally, values of the main activated corrosion products nuclides 58Co and 60Co in the core and steam generator before and after considering recoil release were calculated, and the impact of recoil release on the activation corrosion products and its implications for the actual operation of the reactor were explored.ResultsThe calculation results indicate that the recoil release probability decreases from 45% at the beginning of the simulation to 0.3% at the end of the simulation. However, the variation pattern of the activity ratio of 58Co and 60Co in the core and steam generator remains the same as that without recoil release. The activity ratio is 91% and 203% respectively, compared to the case without recoil release.ConclusionsThe total probability of recoil release is related to the thickness of the corrosion product layer and gradually decreases with the operating time of the reactor.

BackgroundThe primary coolant flow rate is essential in preventing departure from nucleate boiling. The implementation of a low-leakage core loading pattern in advanced passive (AP) technology-based nuclear power units has increased the temperature difference gradient at the core outlet, resulting in elevated uncertainty in the flow rate calculations when using the heat balance method.PurposeThis study aims to validate a measurement and calculation method based on the Bernoulli equation model for accurately determining the primary coolant flow rate in AP nuclear power units, hence meeting the design and regulatory requirements.MethodsFirst of all, measurements were conducted for the primary loop main equipment and bend pipe flowmeter pressure differentials during the commissioning phases. Calorimertic balance tests were performed at power levels of 50%, 75%, 90%, and 100%. Then, the bend pipe flowmeter coefficients were calibrated using the flow rate values obtained from the hot function test and 100% rated thermal power (RTP). Finally, based on weighted factors, the total flow rate values for the reactor coolant system (RCS) were calculated with emphasis on the minimization of uncertainties.ResultsThe proposed measurement and calculation method yields primary coolant flow rate values with a relative error of less than 4%. The total flow rate after loading is within the range of 95.8% to 104% of the expected optimum flow rate. The uncertainty of the volumetric flow rate calculated from NAPs is lower than 1.9%, demonstrating a novel approach for precise measurements in other units.ConclusionsThe method of this study offers an advanced perspective for reactor coolant precise measurements in other units, with primary coolant flow rate values exhibiting minimal relative error and volumetric flow rate values from NAPs demonstrating low uncertainty.

BackgroundThe experimental advanced superconducting tokamak (EAST) feeder system is an essential part of the device that connects the superconducting magnet and high-temperature superconducting current lead. It provides the magnet with feeding and energy release channels in the event of quenching. In recent years, owing to the increase in the inlet temperature of the toroidal field (TF) feeder system, the outlet temperature has occasionally exceeded the threshold of 6.1 K, resulting in the termination of the experiment.PurposeThis study aims to ensure the continuation of the EAST experiment by thermal stability analysis under new TF feeder outlet temperature threshold of 6.5 K.MethodsBased on the system structure and low-temperature operation data of the TF feeder and superconducting conductor, a mathematical model for temperature margin and current shunt temperature of superconducting conductor was established. Then, the mathematical model and GANDALF software were employed to calculate the temperature and stability margins of superconducting conductors during operation under different background magnetic fields and operating currents.ResultsThe calculations results indicate that under the new threshold of 6.5 K, the temperature margin of the conductor is greater than 1.5 K and that the stability energy margin is greater than 200mJ·cm-3.ConclusionsThe superconducting conductor remains safe for use after increasing the threshold.

BackgroundThe high-temperature liquid lead-bismuth metal has a scouring and wear effect on the head of the axial flow lead-bismuth pump impeller blades in a lead-cooled fast reactor system, causing the protective layer on the blade surface to break down and material corrosion rate to accelerate.PurposeThis study aims to reduce the scouring wear effect of the high temperature liquid metals on blade surfaces.MethodsFirst of all, three types of leaf top clearance structures, i.e., plane, chamfered right angle, and chamfered rounded angle, were designed. Then, the reliability of the simulation results was verified by the scaling conversion method, and the commercial CFD software ANSYS CFX with SST k-ω turbulence model was employed to analyze the variation of flow velocity, shear force, and flow pattern with scouring and wear characteristics under different leaf top clearance structures. Finally, the energy loss of the high temperature liquid lead-bismuth metal on the material surface was analyzed using the wall entropy yield.ResultsAnalysis results show that the head and efficiency of the chamfered right angle model are reduced by 1.02% and 0.64%, respectively, compared with those of the flat surface under standard operating conditions, and the chamfered angle model shows a 0.51% reduction in the head and 0.51% efficiency increase. The impeller scouring wear effect occurs predominantly near the inlet edge of the blade rim, and the effect of the high temperature liquid metal on the blade head scouring wear is improved by the chamfered and rounded designs.ConclusionsThe chamfered and rounded designs reduces the mechanical energy loss on the blade surface by reducing the flow velocity at the top clearance and reducing the scouring wear effect at this location. Therefore, the rounded design and the right angle design could improve the influence of high-temperature liquid lead-bismuth metal on the erosion wear of the blade head.

BackgroundThe use of controlled X-ray sources instead of 137Cs radioactive sources in density logging has become a new trend. The high voltage on the target substrate significantly affects the intensity of the X-ray source, and the density measurement uncertainty can be maintained at 0.01 g?cm-3 when the high voltage is 350 kV.PurposeThis study aims to examine the depth-of-investigation characteristics and influence of a 350-kV high-voltage X-ray density logging instrument.MethodsThe depth of investigation of various source distance detectors in 20% water-bearing limestone formation was studied using the Monte Carlo method. By comparing the investigation characteristics of 350-kV high-voltage X-ray source and 137Cs source density logging, the reasons for the differences in the depth of investigation among them were analyzed. Moreover, the contribution of mudcake and formation to the detector and the density deviations of various detectors were analyzed via simulation. Finally, the influence of mudcake on the density logging response of the well wall was explored.ResultsThe results indicate that the depth of investigation of X-ray density logging instrument increases with the augment of source distance. Compared to the 137Cs source density logging, the scattered particles of the X-ray density logging are mainly concentrated at 1~3 cm from the bore wall, resulting in the depth-of-investigation differences between the two techniques. Furthermore, the contributions of mudcake and formation to different source distance detectors are different, and the detector density deviation decreases with the increase in source distance.ConclusionsThis study affords a theoretical basis for the depth-of-investigation characteristics and influence of 350-kV high-voltage X-ray density logging.

BackgroundDirect-current radiofrequency (DC-RF) hybrid plasma has broad application prospects in the field of nuclear ultrafine powder material preparation owing to its characteristics of high temperature and high chemical activity.PurposeThis study aims to explore the flow and heat-transfer characteristics of DC-RF hybrid plasma, so as to provide references for the design and stable operation of the plasma generator device.MethodsFirst of all, the hybrid plasma generator was assumed to be a two-dimensional axisymmetric model, and the device was filled with pure argon plasma in a local thermodynamic equilibrium (LTE), steady, and turbulent flow state. Then, the ANSYS FLUENT software was employed to establish a two-dimensional model for the DC-RF hybrid plasma torch structure, and the spatial distributions of the temperature and flow field in DC-RF hybrid plasma torch were simulated using the k-ε turbulence model with SIMPLE algorithm based on velocity and pressure coupling solver. Finally, the effects of changes in the operating parameters were analyzed based on these results.Results & ConclusionsThe simulation results indicate that increases in the DC arc current, reaction gas flow rate, and cooling gas flow rate can reduce backflow effects at the entrance of hybrid plasma torch. The temperature and area of the plasma arc near the RF coil increase with the RF coil current. However, an excessive current and gas flow rate may adversely affect the operation of the device. Various requirements of material handling processes on the premise of stable operation of the device can be satisfied by adjusting working parameters for the control of the hybrid plasma flow field profiles.

BackgroundThere are no commercially available channel electron multipliers (CEMs) made of glass in domestic market of China; more complex CEMs with helix channels are scarcer.PurposeThis study aims to develop a CEM with a single helix channel, and test its performace for satisfying the requirements of high-end users of such products.MethodsFirst of all, a series of manufacturing process designs and improvements were made on the basis of the formula of microchannel sheet glass, resulting in the production of a single spiral channel electron multiplier with suitable performance. Then, a CEM analog mode test device with a disc-incense type tantalum filament as the input current and a CEM pulse-counting mode test device with an ultraviolet light-emitting diode combined with a gold cathode as the input signal were set up to conduct comprehensive testing of the device's performance parameters.ResultsThe newly developed CEM with single helix channel achieves gains of 1×104~1×106 in the analog mode and 1×107~1×108 in the pulse-counting mode. The gain value increases with the increase of the working voltage, and the rise time of the output pulse is 2~3 ns.ConclusionsThe overall performance of the developed CEM is close to that of foreign counterparts, and the CEM can be used in related instruments.

BackgroundCosine function shaping (or cos shaping) is used to digitally shape nuclear pulse signals, as the shaping method is simple and has high operability and flexibility.PurposeThis study aims to explore different cosine shaping methods of nuclear pulse signals, and evaluate their effect.MethodsFirstly, based on the single exponential decay signal and cosine pulse signal, transfer functions and cascade formulas of three different cosine shaping methods in the Z-domain were derived. The influence of the parameter selection on the shaping effect in the cosine shaping algorithm was analyzed. Then, the cosine shaping methods were developed for the simulated nuclear signals and the actual sampled nuclear signals, and the cosine shaping, amplitude extraction, and energy spectrum construction of the digital nuclear signals were implemented in the field programmable gate array (FPGA) system. Finally, the gamma (γ) energy spectrum of 137Cs (NaI(Tl) detector) was evaluated using the different cosine shaping methods.ResultsThe results of γ energy spectrum from 137Cs (NaI(Tl) detector) demonstrate the satisfactory performance of all three digital cosine shaping algorithms in terms of energy resolution and counting. The symmetric zero-area cosine shaping performance index is improved relative to conventional methods.ConclusionsThe three kinds of digital cosine shaping methods all achieve accurate cosine shaping for simulated and real nuclear signals. The three cosine shaping methods proposed in this study may be applied to shape functions in other research areas.

BackgroundIn the realm of cosmological ray studies employing plastic scintillation fibers, it is essential to conduct quantitative analyses of the photon number from the fiber's output pulse for the successful design of readout electronics.PurposeGiven the absence of a weak cursor setting device such as a single photon source. This study aims to quantitatively analyze the number of photons generated by photon incidents within the fiber calibrating without a weak photon source such as a single photon source.MethodsFirstly, photon numbers within weak optical pulses induced by muons in optical fibers with diameters of 1 mm and 2 mm were determinated by the calibration method that making use of inherent non-photogenerated carrier characteristics of the silicon photomultiplier tube (SiPM). Then, the Geant4 software was employed to simulate the theoretical photon yield of muons in these optical fibers, and the simulation results were compared with experimental data for validation.ResultsThe verification results indicate that the anticipated photon count in the optical pulses within fibers with diameters of the 1 mm and 2 mm fibers are 44 and 85, respectively. The deviation from the simulation results is 4.55% and 10.59%, respectively.ConclusionsThe results validate the efficiency of the low photon number measurement method, demonstrating its ability to accurately measure the photon count generated by the incident fiber without the need for additional calibration equipment. This method may extend to other scenarios that require the measurement of photon numbers in weak light pulse situations.

BackgroundTitanium and its alloys are widely utilized within the military, aerospace, shipping, nuclear energy, and biomedical fields because of the advantages of low density, high strength, good corrosion resistance, and high biocompatibility. Moreover, titanium films are important materials for surface protection due to their high hardness and good compactness. During the preparation of titanium films, the surface morphology and phase structure will be influenced by substrate properties (e.g., surface morphology and temperature), working gas pressure, and other factors. Substrate temperature mainly influences the growth process of thin films, which directly affects the grain structure of the films, and thus changes the corresponding mechanical properties.PurposeThis study aims to establish the relationship between substrate temperature and mechanical properties of titanium films.MethodsFirstly, titanium film samples were prepared at a substrate temperature range of 600~750 ℃ by using resistance evaporation coating on the surface of molybdenum substrate. Then, the structural characterization of the film was examined using X-ray diffraction (XRD) so as to obtain the preferred orientation of titanium film. Scanning electron microscopy (SEM) and atomic force microscopy (AFM) were employed to identify the surface morphologies of the titanium films, including grain size distribution and surface roughness. Finally, an AFM nano-indentation method was performed to examine the mechanical properties of the titanium films and obtain the elastic modulus of titanium film.ResultsThe results demonstrate that the substrate temperature significantly influences the microstructure and mechanical properties of titanium films. When the substrate temperature increases from 600 ℃ to 750 ℃, the preferred orientation of titanium films changes from (101) to (002) due to the competition between the minimization of strain energy and surface energy. With the increase of substrate temperature, the mobility of titanium atoms on the substrate increases, resulting in increased average grain size, surface roughness, and elastic modulus of the titanium films. The average grain size increases by 26% as the substrate temperature increased from 600 ℃ to 750 ℃.ConclusionsThe microstructure, surface morphology, and mechanical properties of titanium films are sensitive to substrate temperature. A high substrate temperature in the process of resistance evaporation is more desirable to obtain titanium films with high mechanical properties.

BackgroundThe electrostatic lens plays an important role in obtaining high quality focused electron/ion/slow positron beams with high spatial resolution and brightness. A novel electric lens, composed of simplified electrode structures with tube diameter gradually decreasing, is proposed for focusing charged particle beams with low energy and large spot size.PurposeThis study aims to investigate the beam dynamics of this designed electrostatic lens for validation.MethodsBased on overall structure of electric lens focusing system, the charged particle optical simulation software SIMION was employed to optimize the parameters of this focusing system. Then combined with electron beam experiments, the key technologies involved in the electronic lens were studied in detail, including influencing factors, their distributions, and the focusing performance of the lens.ResultsThe results show that the large transverse space of the initial electron beam can be compressed effectively by arranging the electrode structure and electric potential of the lens, with focusing efficiencies exceeding 80%.ConclusionsThe focusing method proposed in this study has significant lateral compression advantages with wide application prospects in many focusing scenarios of different charged particle beams, such as reactor positron sources.

BackgroundWith advantages of low system pressure, stable operation and good economic performance, molten salt heat exchanger has recently been widely applied to the field of energy as concentrating solar power, nuclear power engineering, high temperature hydrogen production, and so on.PurposeThis study aims to analyze the thermal stress generated in the main components of the U-tube heat exchanger due to the high operating temperature of the molten salt and the large temperature difference between the hot and cold fluids.MethodsFluid-thermal-solid coupling method was adopted in this study. First of all, the main thermal performance parameters of the heat exchanger were obtained by using computational fluid dynamics (CFD) computation, and compared with experimental results to verify the accuracy of the CFD fluid simulation results. On this basis, the heat transfer process was analyzed in details for the molten salt tube-shell heat exchange under the operating condition, and the flow field and temperature field of the heat exchanger were obtained. Finally, the stress field generated by the coupling of flow field, pressure field and temperature field was calculated by Ansys workbench finite element software, and the stress distribution of the tube sheet connected with the heat exchange tube and shell was emphatically analyzed to find the maximum stress value of the tube sheet and the stress change rule of some paths.ResultsThe result shows that the CFD fluid simulation method is feasible with a maximum deviation of 3.07%. The larger stress is found at the connection area between the tube plate and the non-tube, which is located near the inner tube wall on the shell-side with about 2 mm away from the lower surface of tube plate.ConclusionsResults of this study provides important reference for the actual operation and structural deign of molten salt heat exchanger.

BackgroundAmong the mitigating strategies for severe accidents, the in-vessel retention (IVR) is one of the useful remission measurements. The key point to evaluating IVR is to analyze that the final steady-state thermal load of the melt does not exceed the critical heat flux (CHF), which occurs during boiling heat transfer on the outer wall of the lower head, and the remaining wall thickness of the lower head can carry the melt to prevent the structural failure.PurposeThis study aims to analyze heat transfer of the reactor pressure vessel (RPV) lower head under severe accidents by using ASTEC code.MethodsFirst of all, the composition and mass of the molten substance were assumed to be UO2, 92 353.29 kg; Fe, 43 000 kg; Zr, 23 133.9 kg; Zr oxidation, 41.8%, for a large advanced pressurized water reactor (LAPWR). With the heavy metal oxide layer and metal layer of stable molten pool in the lower RPV of this LAPWR, the average value of core decay power and the physical properties of molten materials in RPV were input as the condition boundaries for ASTEC, the middle break accident sequence was selected for the calculation of the thermal parameters of the coolant, the outer wall CHF and the final thickness of the lower head. Then, the CHAWLA-CHAN heat transfer relationship was used to calculate the heat transfer coefficient between the melt and the inner wall of the lower head. The key safety related issues such as the heat transfer parameters of the outer wall of the lower head, the heat transfer through the lower head, and the wall thickness of the lower head were analyzed. Finally, the IVR effectiveness was estimated by the thermal properties and the structure of the lower head.ResultsWhen the decay power is 21 MW and the core molten pool is divided into two layers, the average thickness of the oxide layer is 1.6 m, and the metal layer is 0.8 m. The results show that the heat exchange is more intense in the upper part of the lower head, and the maximum value of the heat flux occurs at the junction of the two melt layers, which the corresponding surface angle is 77.5°~80°. The inner wall of the lower head will be melted by the molten metal layer in the location of the minimum thickness of the lower head, and the final remaining thickness is less than 2.0 cm.

BackgroundCompared with conventional rod-type nuclear fuel, annular fuel has higher power density and better heat transfer efficiency, which can significantly improve the safety and economy of the reactor.PurposeThis study aims to investigate the effect of ring fuel element geometry on the thermal performance and to correct the initial parameters.MethodsThe initial parameters of the ring fuel element were set and the thermal conductivity calculation program of the ring fuel element was prepared. The effects of the ring fuel flow distribution ratio, inner and outer cladding thickness, inner and outer air gap thickness and core block thickness on the thermal performance of the ring fuel element were investigated by three evaluation criteria developed and geometric corrections are made.ResultsAppropriately increasing the flow distribution ratio, decreasing the inner casing thickness, increasing the outer casing thickness, decreasing the inner and outer air gap spacing and decreasing the core block thickness can improve the thermal performance of the components; setting the flow distribution ratio to 1, the inner casing thickness 0.06 cm is amended to 0.04 cm, the outer casing thickness 0.06 cm is amended to 0.07 cm, the inner and outer air gap spacing 0.035 cm. The thickness of core block is amended to 0.5 cm.ConclusionsThermal performance of annular fuel elements is significantly improved after appropriate geometry correction is made.

BackgroundNuclear critical safety analysis is the key technology to ensure the safety of spent fuel reprocessing plant. However, the present critical safety analysis codes for solution system are either limited in the geometric scope of application, or have poor engineering practicability due to low computational efficiency.PurposeThis study aims to develop a method suitable for wide application range and high accuracy for nuclear critical safety analysis, so as to provide technical support for spent fuel reprocessing plant.MethodsAccording to the characteristics of spent fuel solution system, a set of methods, such as the zero-dimension cross-section calculation and whole system group condensation model, the three-dimensional space-time neutron dynamics model based on PCQS, and the R-Z two dimensional thermal and radiolysis gas simulation model, were combined to establish a paralleled 3D critical safety analysis code hydra-TD. In addition, some experiments of SILENE facility at France were modeled and calculated by using hydra-TD code to verify its effectiveness.ResultsThe verification results indicate that there are very small errors of key parameters such as the first fission power peak, multiplication time and total fission times.ConclusionThe code hydra-TD developed in this study can be applied to simulation of the multi-physics processes in the critical transients of the fuel solution, hence has the ability of critical safety analysis.

BackgroundSawtooth oscillations are macroscopic instabilities in plasma. To better control the sawtooth oscillation in an advanced experimental superconducting tokamak (EAST) device, it is necessary to develop sawtooth controlling methods that help improve the confined performance of plasma in the EAST device.PurposeThis study aims to analyze sawtooth behavior under the on-axis heating by ion cyclotron resonance frequency (ICRF) wave in the EAST device.MethodsFirst of all, the soft X-ray integrated signal intensity data was used to analyze the sawtooth period and amplitude. The radius of q=1 surface and the plasma pressure gradient at q=1 surface were calculated using a soft X-ray intensity profile. Then the neutron yield flux was obtained from the neutron yield flux diagnostic data. Finally, the equilibrium reconstruction results of the equilibrium fitting algorithm (EFIT) were combined with polarimeter-interferometer (POINT) diagnostic data to investigate the relationship between the variation of ICRF and plasma current density.ResultsExperimental results show that the sawtooth period is positively correlated with the ICRF power, and the variation in sawtooth period is roughly same as that in sawtooth amplitude and plasma pressure gradient at q=1 surface. The ICRF power needs to exceed 0.8 MW to change the radius of q=1 surface. The sawtooth period and q=1 surface with ICRF power change are more sensitive under solely ICRF heating than under ICRF+lower hybrid wave (LHW)+electron cyclotron resonance heating (ECRH).ConclusionsSawtooth behavior of EAST plasma is affected by the fast ions produced by ICRF and the radius change of q=1 surface.

Background55Fe is a low-energy radionuclide that is difficult to measure and decays to a ground state of 55Mn through pure electron capture (EC), accompanied by the emission of Auger electrons and low-energy X-ray. As iron is the main component of nuclear reactor building materials, significant amounts of 55Fe have been produced in nuclear reactors and other neutron-producing nuclear facilities.PurposeThis study aims to develop an 55Fe nuclide standard through the absolute measurement of 55Fe activity and provides activity traceability services for 55Fe measuring instruments to ensure the accuracy and consistency of the measurement results of calibration instruments.MethodsThe liquid scintillation triple-to-double coincidence ratio (TDCR) method was applied to determining the activity of 55Fe. First, based on nuclear and atomic data of 55Fe, the electron deposition spectrum of 55Fe in a scintillator was calculated using a random atomic rearrangement model. Second, the counting efficiency of single-energy electron was computed based on the free parameter model. The total efficiency curve of 55Fe was then obtained by summing the efficiency of all deposited electrons. Finally, the experimental counting efficiency was derived by measuring the TDCR value and combining it with the total efficiency curve to realize an absolute measurement of 55Fe activity.ResultsThe experimental results show that correction factors for the asymmetric effect of photomultiplier tube (PMT) quantum efficiency obtained on test samples are between 1.001 and 1.005. The measured specific activity of 55Fe is 94.15 kBq?g-1 with a relative standard uncertainty of 0.45%. Experimental efficiency is better than 63% for double coincidence logic sum of liquid scintillation counter.ConclusionsThis study demonstrates that low relative standard uncertainty of 55Fe activity could be achieved using the liquid scintillation TDCR method with high detection efficiency, and more consistent measurement results can be obtained after applying the asymmetry correction of PMT quantum efficiency.

BackgroundWith the development of negative ion based neutral beam injection system (NNBI) for China fusion engineering test reactor (CFETR), the output power control system of its supporting the radio frequency (RF) power source is one of the key technologies to realize the improvement of its performance.PurposeThe study aims to design an improved output power control system of RF power source to solve the problems of output power stability and insufficient control accuracy in the use of existing RF power sources.MethodsThe software and hardware separation control structure designed by ARM+CPLD dual-core were employed to ensure the operation efficiency of the output power control algorithm of the RF power source and the communication stability of the peripheral equipment. Multi-stage progressive power control method and 12-bit digital signal were adopted to control the opening and closing of RF power amplifier, so as to realize high-precision control of output power. The capacitive voltage divider method and current transformer method were combined to accurately sample the actual output power of the RF power source for implementing high-stability control of output power with a closed-loop power control method. Meanwhile, the upper computer software design of man-machine interaction based on serial communication of self-defined protocol was adopted to complete the man-machine interaction function of output power control.ResultsThe control system has perfect human-computer interaction software function, and test results of the prototype RF power output power control system with simulation load show that the control accuracy of the output power is higher than 0.1% when the rated output power is 50 kW, and the stability fluctuation is less than 0.5%.ConclusionsThis scheme with impedance matching networks is expected to meet the performance requirements of CFETR NNBI RF power supply for output power control.

BackgroundIn the long-term uninterrupted work of the real-time on-line monitoring system of water radioactivity, the spectrum drift, line broadening and shift of peak position are caused by the temperature change of the detector and various electronic components and the aging of components, which leads to the difficulty of spectral line analysis and the error of analytical results.PurposeThis study aims to develop a calibration device for real-time on-line monitoring system of water radioactivity based on cerium bromide detector.MethodsThe device was designed to consists of 137Cs standard source (exemption source), lead block, lead chamber with calibration hole and linear motor. The optimum opening radius of the calibration hole and the optimum thickness of the lead block were obtained by Monte Carlo simulation. The standard 137Cs source was used as the standard reference peak, and the calibration of peak position and peak area, the peak position drift and peak area of 137Cs full-energy peak was analyzed by software with real-time gain calculation and parameters adjustment. Finally, the device was applied to the field application verification.Results & ConclusionsResults of Monte Carlo simulation indicate that the optimum radius of the calibration hole is 2.2 cm and the optimum thickness of the lead block is 5 cm. The verification results shows that the device can limit the change of peak position and peak area to ±1% and ±5%, respectively.

As a wide variety of radioactive materials with different characteristics and potential environmental risks existed, the domestic classification of radioactive items management was made, and The Classification and List of radioactive materials (Trial) was formulated. In order to optimize the classification of radioactive materials in China, the current situation of the classification of radioactive materials and classification basis at home and abroad were investigated. The main differences of classification basis were summarized and followed by analysis of problems existing in the classification of radioactive materials in China. New ideas and suggestions to solving the problems was put forward. First, the enumeration items such as radioactive sources and radioactive wastes in the classification of radioactive materials needs to be deleted. Second, the Classification and List is suggested to be revised by adopting the United Nation (UN) Number corresponding to different classifications of radioactive material which is unique and consistent with the international. Finally, a reference for the revision of The Regulations on the Safety Management of The Transportation of Radioactive Materials and The Classification and List of radioactive materials (Trial) is provided.

The associated radioactive waste residue has characteristics of long half-life of nuclides, wide range of radioactive levels, and so on. There are many industries involved in associated radioactive waste in China with wide sources, large types and quantities, hence restrict the healthy development of the industry. This study aims to propose effective safety disposal measures suitable for the complex physical and chemical properties of such waste residues, large differences in regional distribution and disposal status. Through comparison and analysis with hazardous waste, medium and low level radioactive waste, uranium mining and smelting waste, the technical support for the regulatory concept of associated radioactive waste residue was provided. The current disposal methods, design requirements, applicable scenarios, advantages-disadvantages and application examples of this kind of waste residue were systematically introduced, and the disposal strategies of associated radioactive waste residue were suggested. Meanwhile, a typical case of a regional disposal site project was taken as example, the characteristics of the project and waste residue were deeply analyzed, and the countermeasures were given to provide reference for relevant disposal work. Finally, based on above work, some suggestions are put forward to promote the final disposal of associated radioactive waste residue in China.

BackgroundThe current radiation dose calculation technology can only give 3D static dose results of upright human body, which cannot meet the needs of future accurate protection.PurposeThis study aims to achieve 4D dose calculation by establishing a complete method.MethodsFirstly, three algorithms, namely, the rotation matrix method, the volume graph Laplace operator method and the As-Rigid-As-Possible (ARAP) method, were employed to realize the deformation of mesh phantom. Then the phantom deformation guided by motion capture was investigated, and the tetrahedral cutting technology based on Delaunay algorithm was applied to the high speed Monte Carlo calculation. Finally, the 4D dose calculation application system was implemented and used for field test of nuclear power plant (NPP).ResultsThe comparison between the calculated and measured individual 4D dose values shows that the deviation of Hp(10) is less than 10%, and the deviations of Hp(3) and Hp(0.07) are less than 15% are verified.ConclusionsThe reliability and practicability of 4D radiation dose calculation of human body proposed in this study are verified by application results in specific radiation operation process in NPP, which is expected to achieve precise protection of personnel in the future scenarios such as NPP operation and maintenance, nuclear facility decommissioning and medical interventional treatment.

BackgroundIn the aspect of long-distance transport and deposition of airborne nuclear pollutants, Eulerian-Lagrangian method can combine the theoretical advantages of the Lagrange method and Euler method, but and there are few studies in China.PurposeThis study aims to verify the effectiveness of this method for long distance transport and deposition of airborne nuclear pollutants by simulating a nuclear leakage accident of one nuclear power plant in China.MethodsAssuming that a nuclear power plant in the eastern coastal area of China has a leakage similar to the Fukushima nuclear accident, the numerical simulation of the long-distance transport process of nuclear pollutants in the atmosphere was carried out by using the Euler-Lagrangian method of MATCH (Multi-scale Atmospheric Transport and Chemistry) module in JRODOS (Java Real-time On-line Decision Support) system. The results of surface deposition and the distribution of dose rate field were combined with the actual weather map to verify the trend.ResultsThe simulation and verification results show that the wet deposition plays an important role in the removal of nuclear pollutants, and the Eulerian-Lagrangian method can give the main characteristics of long-distance transport and deposition of nuclear pollutants in the atmosphere.ConclusionsThe simulation results are in good agreement with the actual weather trend, and this method can provide auxiliary reference for China's nuclear accident consequence assessment and emergency decision-making.

BackgroundRadioactive oil is one of the organic "difficult wastes" produced by the operation of nuclear power plants.PurposeThis study aims to develop a nuclide separation and treatment technology for radioactive waste oil and explore engineering application.MethodsFirst of all, a set of radioactive oil nuclide separation and purification treatment engineering equipments was developed and applied to obtain test samples collected from one NPP. Then a nuclide separation and purification process based on the oxidative aging method were developed. Finally, the high-purity germanium for nuclide γ spectrometer was employed to measure 58Co, 60Co, 54Mn, 110mAg, 137Cs, 134Cs, 124Sb, 125Sb, 59Fe, 95Zr, 95N, etc., γ nuclides, low background liquid scintillation counter was applied to the measurement of 3H and 14C, and high sensitive automatic liquid scintillation counter was used to measure 55Fe and 63Ni. Total α and total β were obtained by using low background α/β measuring instrument. [Results and Conclusions] The results show that the decontamination coefficient of the oxidative aging process reaches more than 2 orders of magnitude; the radioactive oil treated by the engineering device amounts to the clearance level, hence can be managed as ordinary hazardous waste. The additional dose that may be caused in the process of incineration (including transportation) for the treated radioactive oil is far below the dose limits, which meets the dosage guidelines for reuse, and is in line with waste minimization principles.

Associated radioactive mines are widely distributed in China, with many industries, complex processes and technologies, close relationship with the public, and great impact on the radiation environment. They are relatively weak links in nuclear and radiation safety supervision. In the process of implementing the upper level document, some specific problems have to be faced, such as the definition of associated radioactive ore, the scope of associated radioactive ore development and utilization, etc. In the field of radiation environment supervision, there are issues such as effluent and radiation environment monitoring specifications that need to be solved. In particular, the treatment and disposal of associated radioactive solid waste is the key problem that currently restricts the sustainable development of associated radioactive ore development and utilization. The second national census of pollution sources reported the cumulative storage of associated radioactive solid waste in the country, and the radiation environment supervision situation is grim. In the scope of associated radioactive mines and some problems encountered in the supervision of radiation environment, this study aims to formulate a reasonable and feasible supervision system for associated radioactive mines, standardize supervision and management activities, and ensure the safety of the radiation environment according to scientific management methods, combined with the practice of supervision of associated radioactive mines. Suggestions on the definition of the scope and development and utilization of associated radioactive mines and the management of the inventory are put forward for solving the radiation environment supervision system with consideration on the storage period of associated radioactive solid waste, the management of nearby centralized disposal and disposal facilities. As the radiation environmental supervision of associated radioactive mines is still a relatively weak area in nuclear and radiation safety supervision system in China, radiation environmental impact assessment and environmental radiation self-monitoring need to be strengthened.

BackgroundWhen steam generator tube rupture (SGTR) occurs in the lead-bismuth cooled fast reactor, high pressure water/steam flows into the primary side filled with high temperature liquid metal. According to the location and size of the rupture, the leakage behaviors of the rupture may involve leak-before-break (LBB), single-phase critical flow or two-phase critical flow. Under the action of high temperature liquid metal, different forms of heat and mass transfer behaviors occur in the two-component multiphase system of water-metal, which has an important influence on the safe operation of the lead-bismuth cooled fast reactor.PurposeThis study aims at the bubble dynamic behavior in the descending flow field of liquid lead-bismuth alloy (LBE) in the tube bundle in different stages of SGTR caused by microcrack on the surface of the heat transfer tube during the drying stage and low flow single phase steam permeates the primary side.MethodsBased on the VOF method, a numerical simulation model of steam-LBE two-phase flow and phase interface capture was established to study the bubble growth and transport behaviors from single tube or 3×3 tube bundle in the downward flow field of high temperature LBE. The SST k- ω model was employed to solve the turbulence equation. The physical law of steam bubble movement was analyzed and its influence on the heat transfer and operation stability of steam generator was evaluated.ResultsThe results show that the dynamics behaviors of steam bubbles in the descending flow field are quite different from these in static liquid or upward flow. The steam bubbles may slide along the heat transfer tube surface after departure from the crack under the actions of LBE descending flow field and buoyancy. The steam bubbles may form a steam film covering the heat transfer tube surface or accumulate by quantity in the bundle.ConclusionsThese phenomena adversely affects the flow stability of the LBE and the heat transfer of the steam generator.

BackgroundThe startup time and startup-failure are widespread in most cold redundancy equipment of nuclear power plants (NPPs). The traditional static and dynamic fault tree cannot accurately model the startup time and startup-failure.PurposeThis study aims to model the startup time and startup-failure behaviors in cold redundancy systems, and provide suggestions for the improvement of reliability assessment methods.MethodsFirst, a DFT Monte Carlo simulation method was proposed for modeling and analyzing equipment's startup time and startup-failure behaviors in a cold redundancy system. Then, the emergency diesel generator set of the nuclear power plant was taken as an example, the distribution curve of system failure probability and the sensitivity of each component were obtained. Finally, the results were compared with the static fault tree method and traditional DFT method.Results1) The proposed method can model and analyze the start-up time and start-up failure behaviors of cold redundant equipment, reflecting the real failure scenarios and actual operation status of cold redundant systems. 2) The proposed method can accurately evaluate the system failure probability, identify highly sensitive equipment parameters in different time periods, and analyze the influence of start-up time on the system failure probability.ConclusionsThe proposed method has certain theoretical significance for the optimal design of NPP's cold redundancy systems.

BackgroundThe annular fuel can increase the power density of the reactor due to its double-sided cooling, which is of great significance to the miniaturization and long-life operation of the pressurized water reactor.PurposeThis study aims at the calculation method of effective temperature of annular fuel cell by developing the thermal-hydraulic analysis code named THCAFS (Thermal-Hydraulic Code of Annular Fuel with Single channel) for annular fuel, and establishing a calculation model for the effective temperature of the annular fuel cell.MethodsBased on THCAFS, the thermal-hydraulic performance of the annular fuel cell designed by Westinghouse 4-loop PWR was analyzed, and the Code-to-Code comparison was carried out with the calculation results of VIPRE-01, TAFIX and NACAF. At the same time, the Monte Carlo code SERPENT was employed to simulate the radial power distribution and burnup process in the fuel rod, and the self-developed code THCAFS was used to simulate the thermodynamic behavior in the fuel rod, and the radial power distribution, nuclide density change and cell temperature field.Results & ConclusionsThe results show that THCAFS can be preliminarily applied to annular fuel design and thermal-hydraulic analysis. The maximum deviation of the ratio between the fitting function power and the simulated power under different fuel consumption does not exceed 2%. This effective temperature calculation method can also provide important reference value for the mechanism study of the relevant annular fuel resonance effective temperature.

BackgroundTime series usually have the characteristics of linear and nonlinear. Single model has certain limitations, which proposes mixed models for time series prediction.PurposeThis study aims to explore the application of the combination of Mann-Kendall test, differential autoregressive mobile average model (ARIMA) and long and short-term memory (LSTM) used in the prediction of the number of operational events of nuclear power plants (NPP) collected in Nuclear Safety of China and Annual Report of Nuclear Safety.MethodsFirstly, the R software was used to build ARIMA (2,1,2) model to obtain the linear part of operation events with the number of nuclear power plant operation events from 1991 to 2018, and LSTM model was developed to predict the deviation sequences, hence the nonlinear part of the number of operation events was derived from those deviation sequences. Then, combined model of ARIMA and LSTM was established to predict the number of operational events. Finally, the predicted values based on measured data were verified by actual measured data.ResultsThe verification results show that the ARIMA and LSTM combination model can be employed to improve the prediction accuracy effectively by 3%, and the predicted values of operation events in nuclear power plants from 2019 to 2020 are similar to the data collected in Annual Report of Nuclear Safety.ConclusionsThe combined model can better fit the time series of the number of operating events of NPP and correct the error of the single model.

BackgroundUN-U 3Si 2 composite fuels have a promising prospect in advanced future accident tolerant fuel elements. Its irradiation creep and thermal creep caused by in-reactor operation have an important influence on the irradiation-induced thermo-mechanical coupling behavior and safety of the fuel elements.PurposeThis study aims to develop a stochastic modeling method according to the metallographic structure of composite fuel and numerical simulation of the uniaxial tensile creep test of the UN-U 3Si 2 composite fuel (20% U 3Si 2).MethodsBased on the data of creep experiments from literatures, the dominated creep mechanisms of UN and U 3Si 2 polycrystalline fuels were analyzed, and their creep rate models considering vacancy diffusion and dislocation motion mechanisms were obtained by curve fitting. Then, the correlation model between the macroscopic creep of composite fuel and the contribution of each component was established on the basis of the homogenization theory and the removal of irradiation swelling effect. Finally, based on the metallographic structure diagram of composite fuels in the literatures, stochastic modeling method was developed and applied to the numerical simulation of the uniaxial tensile creep test of the UN-U 3Si 2 composite fuel (20% U 3Si 2).ResultsThe model predictions of UN and U 3Si 2 creep rates are in good agreement with the experimental results, validating the effectiveness of the model. The contribution of the component fuels to the macroscopic equivalent creep of the composite fuel is obtained by analysis of the underlying creep mechanism. When the fission density reaches 4.32×1027 fissions·m-3, the maximum von Mises stress at the interface between particles and matrix is about 6 times of the homogeneous tensile stress applied externally.ConclusionsThe research results indicate that the difference in the irradiation swelling of UN and U 3Si 2 will result in the strong internal mechanical interaction in the composite fuel whilst the existence of weakened stress regions leads to the negligible effect of irradiation swelling on the macroscopic equivalent creep strain of the composite fuel.

BackgroundCu-W composites are widely used in electric power, electronics, and plastic forming owing to the excellent electrical and thermal conductivities of copper combined with the outstanding strength and thermal properties of tungsten.PurposeThis study aims to investigate the extension of the solid solubility of the Cu-W immiscible system under high current pulsed electron beam (HCPEB) irradiation.MethodsFirst of all, the Cu-15W powder mixture was sintered by vacuum at 850 ℃ after ball milling for 5 h to prepare the Cu-W composites. After polishing, the Cu-W composites was exposed by a HOPE-1 type HCPEB device with 1, 5, 10 and 15 pulses, respectively. Then, Rigaku D/Max-2500/pc X-ray diffractometer (XRD) was used to analyze the phase constitutes of the milling powder, sintered samples and irradiated samples. The surface morphology was observed by JEOL JSM-7001F field emission scanning electron microscope (SEM). The energy dispersive spectrometer (EDS) in SEM was used for the examination of micro-region composition. Finally, the microstructures in the modified layer were characterized by JEOL-2100F transmission electron microscopy (TEM). The surface hardness of the sintered and irradiated samples with different pulses was measured using the HVS-1000 Vickers hardness tester.ResultsThe results reveal that Cu(W) solid solutions were formed during the process of ball milling, and the mass fraction of the solute element W in the Cu(W) solid solution reaches 0.88% after 5 h of ball milling. It was discovered that the solid solution undergoes exsolution reaction during heating using differential scanning calorimeter (DSC). The sintered sample surface was irradiated by a pulsed electron beam. The results demonstrate the formation of solid solution phase after HCPEB irradiation, whose solubility increases with the increase of the number of pulses. After 10 pulse irradiations, the mass percentage of solute element (W) in Cu(W) solid solutions reaches 1.63%. The surface hardness of irradiated samples increases significantly after HCPEB irradiation, and reaches 237.1 HV after 10-pulsed irradiation. Surface hardening is primarily caused by of solid solution strengthening and dispersion strengthening.ConclusionsThe present paper provides a meaningful instruction for preparing the immiscible alloy with high solid solubility and the excellent performance.

BackgroundRemote, large-scale, rapid and real-time detection of radioactive sources is of great significance for industrial investigation and environmental remediation. Gama camera can extract information such as the location, intensity and category of radioactive substances from a long distance, and is an effective detection tool. Due to the small detector area and the limited field of view (FOV) of the coded aperture, the detection efficiency of the current coded aperture γ cameras is low.PurposeThis study aims to propose a prototype of large area and highly sensitive coded aperture gamma imaging system based on a SPECT probe.MethodsFirst of all, the Hamamatsu BHP6601 single photon emission computed tomography (SPECT) probe with large area of 510 mm×390 mm and high intrinsic spatial resolution of 3.55 mm was selected as the detector of the system. Based on the detector by tectonic analysis, modified uniform redundant arrays (MURA) coding nesting mode was adopted with a rank of 23 and each unit aperture size of 22.1 mm×17 mm. Then, the system transmission matrix is obtained by using the gamma rays emitted from the radioactive sources at different angles of the FOV to reach the detector geometry through the coded aperture. Then, the geometric relationship between the radioactive source and the system is used to analyze and construct different radioactive source events. Next, the cumulative probability based on the system transmission matrix simulates the photon distribution on the detector. Finally, using Maximum Likelihood Expectation Maximization (MLEM) algorithm is built source image.ResultsThe results indicate that the non artifact field of view (NAFOV) is 57.32° in the X direction and 47.3° in the Y direction; and the spatial angle resolution is 2.94° in the X direction and 2.28° in the Y direction. The γ camera can accurately locate 3.7×107 Bq 137Cs single point source in 1 s at the distance of 18 m and 2 s to accurately reconstruct the position of 1.11×106 Bq 137Cs at the distance of 4 m away from the camera.ConclusionsThe γ camera of this study has a high detection capability for distant or low-activity radiation sources in large range.

BackgroundIn the process of in-situ leaching of uranium, pore blockage greatly limits the leaching efficiency of uranium, and the pore blockage is mainly due to the complex physical and chemical action of minerals and leaching agents in in-situ leaching of uranium.PurposeThis study aims at the influence mechanism of each mineral composition on pore blockage under the action of leaching agent in the process of in-situ leaching of uranium to realize the efficient exploitation of uranium resources.MethodsFirst of all, sandstones of uranium mine in Xinjiang were taken as the experimental samples, the relationship between the mineral composition and porosity in the process of acid in-situ leaching of uranium was studied through acid in-situ leaching experiment, and nuclear magnetic resonance experiment was employed every other period of time to obtain the porosity change curve in the process of uranium sandstone leaching. Then, the mineral characteristics of uranium sandstone were analyzed by X-ray diffractometer to find the changes of the mineral composition during uranium sandstone leaching. Finally, the influence of different minerals on porosity change was analyzed with the help of gray relational theory.ResultsThe results show that the pore blockage in the leaching process is mainly due to the complex physical and chemical action of many kinds of minerals, which will affect the solute transport process of other minerals in uranium sandstone and not only lead to the change of mineral composition but also affect the leaching of uranium. Through grey correlation analysis, it is found that calcium sulfate and magnesium silicate as plugging materials have a correlation degree of over 0.781 for porosity and uranium leaching rate. The physical adsorption of clay minerals can block the micropores of uranium sandstone, so its correlation with porosity and uranium leaching is 0.831 and 0.842 respectively, indicating that pore plugging has an impact on uranium leaching.ConclusionsAccording to the relationship between the change of mineral composition and pore blockage in the process of leaching, making appropriate adjustment can solve the problem of pore blockage, so as to realize the efficient exploitation of uranium resources.

BackgroundIn recent years, lead halide perovskite scintillators have received extensive attention in the field of X-ray imaging. Hard X-ray medical imaging in energy range of 20~120 keV using scintillator detectors, sensitivity and imaging spatial resolution are important performance indicators.PurposeThis study aims to explore X-ray imaging property of halide lead perovskite scintillators by simulation.MethodsFirst of all, 3D MAPbBr 3 quantum dots/polystyrene and 2D PEA 2PbBr 4 quantum dots/polystyrene scintillators were taken as research objects. Then, simulation code Geant4 was employed to establish detector model and simulate the X-ray relative detection efficiency and imaging spatial resolution of lead halide perovskite quantum dots/polymer composite scintillators. Finally, the effect of energy and the ratio of perovskite quantum dot occupation on the resolution were explained by secondary electron motion.ResultsThe results show that increasing the thickness of the composite scintillator and the proportion of perovskite quantum dots can improve the relative detection efficiency whilst reducing the thickness and increasing the proportion of perovskite quantum dots can improve the spatial resolution. When the absorption efficiency reaches 99.5%, 80% of 3D MAPbBr 3 quantum dots/polystyrene excited by 20 keV X ray obtain the same spatial resolution of 10 lp·mm-1 as CsI. When the incident energy increases to 50 keV, the spatial resolution of CsI is 8 lp·mm-1, while that of lead halide perovskite scintillators is less than 4 lp·mm-1.ConclusionsIt is shown by this study that lead halide perovskites have certain application potential in 20 keV low-energy X-ray medical imaging.

BackgroundIn recent years, electron cyclotron wave (ECW) heating and current drive (ECCD) have been widely used in tokamak discharge experiments. The inevitable presence of impurity particles in the tokamak plasma affects the ECCD through radiation energy, inhibition of turbulent transport, and change the collision rate. Changes in plasma density, temperature and other transport quantities caused by the change of impurity concentration, induce the changes of Shafranov displacement at the center of magnetic surface of the plasma.PurposeThis study aims to investigate the influence of impurity concentration changes on the ECW deposition position and current drive efficiency theoretically with consideration of all above related variations.MethodsThe One Modeling Framework for Integrated Tasks (OMFIT) platform was used to conduct integrated simulation study of the effect of impurity effect on ECW heating and current drive. The HL-2M Tokamak device parameters were combined with self-consistent coupled plasma equilibrium, external auxiliary heating and current drive, transport and other physical processes for simulation computation with the carbon ions as the unique impurity ions.ResultsThe simulation results show that when the influence of impurities on the plasma is considered, with the increase of the impurity concentration, the radial position of the ECW deposition first moves to the plasma core and then moves to the edge, and the current drive efficiency first increases and then decreases. Due to the competition between the radiation effect and the dilution effect of the impurity-plasma interaction, the radiation loss power basically increases linearly with the increase of Zeff, while the dilution effect suppresses the turbulence and improve the confinement, but the stabilization effect slows down with the increase of Zeff.ConclusionsWhen the influence of impurities on the plasma is not considered, the deposition position of ECW is basically unchanged, and the current drive efficiency decreases. Results of this study have guiding significance for electron cyclotron current drive (ECCD) to control the plasma current profile and control the instability of magnetic fluid.

BackgroundHefei Advanced Light Facility (HALF) is a fourth-generation synchrotron radiation light source based on diffraction-limited storage ring. The timing system provides trigger signals for the HALF injector, storage ring and beamline, coordinates injection and beam measurement, and achieves filling of the storage ring bucket with any designated bunch pattern.PurposeThis study aims to design a timing system for HALF to reduce the trigger signal jitter of the HALF device to less than 30 ps, and stabilize the signal drift caused by the optical fiber length change.MethodsBased on the MTCA.4 event-driven products from Micro-Research Finland Oy (MRF), the event timing technology scheme was adopted to implement the timing system. Appropriate system frequencies for electron gun, microwave system and storage ring were selected to achieve different modes of storage ring bucket filling scheme. The timing system software was developed under Experimental Physics and Industrial Control System (EPICS) to coordinate with the control system of HALF, including an EPICS driver, database records, and operation interfaces. Delay compensation was applied to signal drift caused by the variation of optical fiber length. Finally, a prototype of the timing system was developed for performance tests.Results & ConclusionsThe results show that the jitter of the trigger signal is less than 24 ps, and the signal drift is controlled at about 3 ps after delay compensation, both meet the design requirements of HALF timing system.

BackgroundThe superconducting third harmonic cavity has been developed independently in Shanghai Synchrotron Radiation Facility (SSRF) and passed beam tests. The cavity electric field needs to be precisely controlled during operation to achieve the goal of stretching beam cluster and improving beam life.PurposeThis study aims to design a digital low level radio frequency (DLLRF) controller for superconducting third harmonic cavity at SSRF.MethodsThe hardware of controller was based on a field-programmable gate array (FPGA) board and a front-end board whilst in-phase/quadrature (I/Q) demodulation techniques were implemented in the software of controller. A synergistic strategy was adopted for driving the stepper motor with slow tuning speed and piezoelectric ceramic with fast tuning speed, and an algorithm for the quench detection for passive cavity. Finally, experimental tests were performed to verify the effectiveness of this designed DLLRF.ResultsWhen the state is in top-up mode over 120 mA, the amplitude stability has improved form ±5% with open loop to less than ±1% with close loop, the voltage of piezo has varies smoothly and stably within 120 V, and the beam life has improved more than doubled.ConclusionA digital low level radio frequency controller for the superconducting third harmonic cavity has been designed and satisfies the requirements for SSRF.

Background Heat pipe cooled reactors (HPR) have good inherent safety. In the early stage of core design, heat pipe failure accident is usually one of the design basis accidents that need to be considered. Purpose This study aims to analyze the neutronic-thermalhydraulic coupling performance of a new type of megawatt heat pipe reactor. Methods Firstly the heat pipe cooled reactor system physical models, including the point kinetics model, the core and heat pipe model and radiation heat transfer model for the inner core cavity, were established according to the designed HPR prototype composed of heat pipe stack and supercritical CO2 Brayton cycle system with thermal power of 3.5 MW. Then, the finite element software FLUENT was employed to conduct neutronic-thermalhydraulic coupling calculation for the three-dimensional reactor core under the steady-state and heat pipe failure accidents. Finally, the core safety performance was evaluated by comparing the peak temperature of each component with the melting point of material. Results & Conclusions The results show the designed HPR has good safety performance under the steady state and single heat pipe failure. Radiation heat transfer in the core cavity cannot be ignored in the serious cascade three heat pipe failure accident in high power region. Meanwhile, the design cannot withstand cascading four heat pipe failure. By comparing the peak temperature of the multiple heat pipe failure with the peak temperature of the single heat pipe failure, it shows that the design has good inherent safety.

Background Radiation shielding design is an important part of reactor design, and the development of new nuclear power technology for various kinds of reactors has put forward new demands on radiation shielding optimization design methods. Purpose This study aims to overcome the shortcomings of the traditional multi-objective optimization methods for shielding structures in dealing with the optimization problem of 3D shielding structures, such as slow optimization speed, difficulty in convergence, and poor globalization. Methods Based on the non-dominated sorting genetic algorithm Ⅲ (NSGA-Ⅲ), the many-objective optimization method for 3D shielding structure design for nuclear reactor was proposed. The Monte Carlo N-Particle Transport Code (MCNP) was employed to analyze comparative performance of the NSGA-III optimization method on the basis of the 3D shielding structure model of nuclear reactors, and shield weight, volume and radiation dose rate in specific regions were taken as the optimization targets. Results & Conclusions The numerical simulation results show that the NSGA-III based optimization method for 3D shielding structure design can search for the Pareto-optimal front more efficiently and stably, providing a new idea for the optimization of radiation shielding design.

Background The corrosion of the secondary circuit has remained a challenging problem influencing the security and efficiency of a nuclear power plant. In an actual operation, an amount of alkalizer is usually added into the secondary circuit water to adjust its pH value to alleviate the corrosion of the pipelines. However, the very complex real working condition results in a significantly inhomogeneous distribution of the pH values, and the enclosed nature of the secondary circuit have frustrated various efforts to control the precise pH values at some key sites inside the circuit. So far, pH values around such key locations have been roughly estimated by either external simulating experiments or by using patented commercial software of foreign companies. However, it is difficult for such external simulations to take into account the various important working conditions. Purpose This study aims to develop a method without specific assumptions or uncontrolled approximations to calculate the distribution of pH values in the secondary circuit under its working conditions. Methods Firstly, the complex and repeated structure inside the steam generator was simplified and simulated by using one direct tube model. Except for the length, the other geometrical dimensions, support plates as well as the tube material etc. were chosen to be as similar as possible to the actual ones. Then, all temperature-dependent equilibriums involving H+ in water solution of the secondary circuit were considered with additional consideration of the equilibriums of the alkalizers in the gas and liquid phases. As an essentially field quantity defined on the space of the secondary circuit, these pH values together with other relevant parameters, such as temperature, fluid velocity etc. were depended on the position coordinates, and were calculated by using the finite element method coded in the COMSOL package. Finally, the boundary conditions of flowing rate and temperature of the water at the inlet were set as 1.0 m·s-1 and 543.2 K, respectively, and the pressure at the outlet was set as ~6.8 MPa, a stepwise linear heat flux model was used to simulate the thermal energy transfer from the primary side to the secondary one. The bubbly-flow model was used to simulate the actual steam-water fluid in the secondary side, which was assumed to be in a steady state working condition. Results The calculated pH field under the working conditions shows clearly an inhomogeneous distribution, e.g. ΔpH = ~ -0.6 from z = 0 to 3.8 m, due to the influences of the tube support plates, the temperature and the heat transfer, etc. The investigations on the ammonia/ethanolamine (ETA) binary alkalizer with different total concentrations of various NH3/ETA molar ratios show a better enhancing effect of ETA over ammonia for the pH value (ΔpH>~0.14), and reveal a saturation effect (molar ratio NH3:ETA≤ ~1:4). Conclusions The distribution of the pH values in the realistic working conditions can be calculated without resorting to empirical formulae and uncontrolled approximations. The developed method and the calculated results provide valuable information for solving the corrosion problem in the secondary circuit. The method and the model can be extended to simulate the more realistic conditions of a nuclear power plant.

Background The flow distribution in core for the liquid fuel molten salt reactor (MSR) is an important part of the thermal hydraulic design, and the hydraulic structure of reactor core plays a decisive role on flow distribution. Purpose This study aims to find out a suitable hydraulic structure design to make the core flow distribution match with the power distribution, and flatten the core temperature distribution for a 10 MW MSR. Methods First of all, a one-twelfth core model of liquid fuel MSR was established. Then, ANSYS FLUENT16.0 software was employed to conduct three-dimensional numerical simulation of the flow field. The influence of hydraulic structure of reactor core was analysed by changing the structure of upper plenum, downcomer and lower plenum, and the corresponding flow distribution characteristics of the core are obtained. Finally, a suitable structure was proposed after step-by-step improvement. Results The simulation results show that increasing of the height of upper plenum can balance the flow distribution between central and peripheral channels, increasing of the width of the downcomer can reduce the vortex flow at the downcomer outlet and flattens the flow distribution at the same time. The cylindrical lower structure with shroud in lower plenum can restrain the vortex to a certain extent and make the flow distribution more gentle. Based on the analysis above, a reasonable hydraulic structure is proposed for the molten salt reactor. Conclusions The results of this study provide important reference for the further optimization design of liquid fuel molten salt reactor.

Background The radiation accident emergency drill is an important way to maintain and improve the response ability of emergency organizations, but at present, the evaluation of the drill is mainly based on the qualitative evaluation of the surface performance of the key links, there is not yet a systematic, full-flow, accurate and targeted evaluation method of exercises. Purpose This study aims to improve the quality and real-time evaluation of radiation accident emergency drill. Methods Firstly, the alternative evaluation indicators were summarized on the basis of literature survey. The method of expert investigation was used to check the alternative indexes and establish the evaluation index system. Then, the analytic hierarchy process (AHP) method was applied to the calculation of weight distribution, and the fuzzy comprehensive evaluation (FCE) was employed to build comprehensive evaluation model for radiation accident emergency exercises. Finally, based on the AHP-PCE, comprehensive scores of the evaluation indexes at all levels of three typical examples were calculated. Results The FCE-based model synthesizes the stage of the drill preparation, the drill implementation and the drill summary, and forms the general goal of the radiation accident emergency drill effect. The evaluation system of radiation accident emergency maneuver is set up with 5 first-level indexes and 15 s-level indexes to realize the comprehensive evaluation of the radiation accident emergency drill. Conclusions The example verification shows that the AHP-FCE model has a good adaptability and rationality, and has a good application value in the evaluation of radiation accident emergency drill.

Background Compared with transistors and small-scale integrated circuits, the total ionizing dose (TID) effect and testing of multifunctional large scale integrated microprocessors are more complex. The difficulty of testing is to analyze the failure mode of microprocessor online from limited information under irradiation. Purpose This study aims to develop an extendable on-line test system for TID effect of microprocessor and carry out preliminary application. Methods The testing system was composed of control circuit, extendable signal acquisition circuit, tested sample interface, upper computer and software. Multiple parametric measurement or functional verification methods such as power consumption current, on-chip memory, communication, clock, analog-to-digital/digital-to-analog converter (ADC/DAC) and direct memory access of microprocessor were provided. Sixteen microprocessors with feature size 40 nm were irradiated with 60Co source and tested on-line. Results After the irradiation dose is accumulated to (377.44±20.34) Gy(Si), all samples are malfunctional with digital communication interruption, sudden drop in current consumption, abnormal ADC/DAC output and so on. Conclusion Based on all 12 kinds of parametric measurement or functional verification results, the TID effect of this type of microprocessor is probable to be a functional failure caused by some kernel instructions. The on-line test system of this study can provide more direct data information for the total dose failure mode analysis.

Background In order to fast and conveniently measure X, γ and neutron radiation field simultaneously, portable multi-function radiation detector is highly demanded. Purpose This study aims to design a portable multi-function radiation detection system based on LaBr3(Ce) crystal, lithium aluminum silicate oxygen (LASO) neutron detector and high range Geiger-Muller (GM) counter. Methods After the optical signal output from the LaBr3 crystal was photoelectrically converted into electronic signal by the photomultiplier tube (PMT), the integrated digital multi-channel was used for data acquisition of electronic signal, and the subsequent data processing and calculation. The front signal processing circuits, such as amplification, discrimination and shaping, were designed for both the LASO neutron detector and the GM tube counter. Finally, the digital signal processed of the LaBr3 detector was transmitted to the ARM (Advanced RISC Machine) processor in the form of TTL (Transistor-Transistor Logic) serial port, and the pulse signal formed of the neutron detector and GM counter tube was connected to the external counting port of ARM processor. Energy spectrum was processed for nuclide identification and displayed by ARM processor whilst the low dose rate measurement base on the Gamma data collected by the LaBr3 detector, the neutron detector and the high range GM tube counter were counted at fixed time and converted from the count rate into the dose rate. Results & Conclusions The designed portable multi-function radiation detector realize simultaneous measurement of wide range (48 keV~3.0 MeV) γ, low-energy (48 keV~1.25 MeV) X-ray, (0.1~100 mSv∙h-1) neutron dose rates, and nuclide identification capability of LaBr3 spectrometer, and upload the data to PC through USB interface.

Background The measurement of time-of-flight is one of the indispensable experimental contents in contemporary high-energy physics experiments and plays a vital role in exploring the essence of particle physics. Purpose This study aims to design a time to digital convertor (TDC) chip that meets the high-resolution time measurement requirements of time-of-flight detectors for high-speed flying particles in high-energy physics experiments. Methods First of all, a differential structure TDC was proposed and the main measurement part was realized by differential delay loop composed of time measurement core module, time measurement data transmission module, delay loop calibration module and clock generation module. Based on this structure, three parts of delay loop module, thermometer code generation module, and coarse count and fine count generation module were integrated into the core module of time measurement, and the 0.18 μm SMIC (Semiconductor Manufacturing International Corporation) process was adopted to achieve the TDC chip design. Results & Conclusions The designed TDC chip has a layout area of 1.35 mm×1.35 mm, a resolution of 17 ps, an accuracy of 8.5 ps (Root Mean Square, RMS), and a dynamic range of 0~210 μs. It can meet the current requirements for high-precision time measurement in high-energy physics.

Background Betatrons are widely used in non-destructive testing, cargo and vehicle security inspection systems, whilst the filament power supply is an important core component of small size betatron. Existing filament power supplies are not suitable for small size betatron. Purpose This study aims to develop a novel electron gun filament power supply for small size betatron to produce optimal radiation intensity. Methods An electron gun filament power supply composed of a filament duty cycle adjustment circuit and an injection current feedback circuit were designed on the basis of operating characteristics of a small size betatron. Two pulse-width modulation (PWM) signals were output from the digital signal processor (DSP) to drive the half-bridge circuit. Positive and negative alternating voltage pulses were output by the half-bridge circuit to the primary of the isolation transformer, and the secondary voltage of the transformer was loaded to the filament for heating of the filament. By collecting the betatron tube wall current and target current as the injection current feedback signal, the duty cycle of the filament voltage pulse was adjusted by DSP according to the feedback signal, and realizes the adjustment of the filament emission current, so that the current injected by the filament into the acceleration tube was kept at the optimal value. Results The experimental test results show that the filament power supply can keep the betatron output dose rate in the best state, and the output dose rate stability is better than 11.3%/10 min. Conclusions The filament power supply meets the requirements of small size betatron with advantages of good stability, wide adjustment range and small size. It can be applied to the small size betatron developed by the Engineering Research Center for Nuclear Technology Application of the Ministry of Education of East China University of Technology.

Background Vertical cavity surface emitting lasers (VCSEL) have very high application value in space radiation environment. Purpose This study aims to explore the degradation rule and mechanism of 850 nm VCSEL in harsh radiation environment. Methods First of all, the MULASSIS tool was employed to calculate displacement damage dose (DDD) and design experimental scheme for 850 nm multimode VCSEL samples. Then, 3 MeV and 10 MeV proton irradiation experiments were conducted to obtain the degradation rule of parameters such as light output power and threshold current with the proton fluence, and to find that the degradation degree of light output power and threshold current were equal under the same DDD. Finally, the Silvaco software was used for modeling and simulation on an experimental basis to extract microscopic parameters such as trap density, donor and acceptor ionization density, mirror loss, radiation recombination rate and photon number. Results The simulation results are in good agreement with the experimental results, these results show that each parameter changes to different degrees with the increase of proton fluence. Conclusions The parameter degradation law and radiation damage mechanism of VCSEL can be deeply explored by simulation on the basis of the experimental law, and simulation results are of great significance for understanding the degradation mechanism of VCSEL.

Background The expected signal rate of rare decay experiments, such as instance dark matter and neutrino less double beta decay experiments, is extremely low, which requires that the detector building materials have extremely low radioactivity. Low radioactive background control is one of the essential works in the rare decay experiments. 226Ra and 228Ra produced in the early decay chain of 238U and 232Th have low boiling point and high vapor pressure, removing the element Ra can break the 238U decay chain and keep a low radioactivity of 232Th-late for a long time. Purpose This study aims to investigate vacuum melting technique for low background titanium to reduce the impurity of isotopes that have negative impact on rare decay experiments and creating a low background environment for the detector running. Methods Firstly, the low background material samples were acquired by manual separation of radionuclides using physical and chemical methods. Then, radioactive impurity elements, such as K, Cs, Ra, Pb, Po and Rn with low boiling point and high vapor pressure, were volatilized in environment with high temperature and high vacuum level. Finally, radioactivity of these testing samples were measured by two sets of high-purity germanium γ spectrometer with measurement time extended to 7 days. Results Measurement results show signs of removal of radioactive isotopes by smelting-vacuum method, and the impurity in pure titanium smelted in vacuum electron beam furnace can reach the levels of (0.13±0.69) mBq∙kg-1 for 232Th-228Ac, and (0.07±0.29) mBq∙kg-1 for 238U-222Rn, respectively. Conclusions The smelting-vacuum method could provide reliable low background material for the container of the next generation PandaX detector.

Background Superconducting undulator (SCU) prototype with small magnet gap of 5 mm, long magnet length of 4 m and high magnet field of 1.58 T was being developed at Shanghai High Repetition rate XFEL and Extreme light facility (SHINE). Compared to any other superconducting undulator, there is no cryocooler being installed on the cryostat in this SCU prototype. Purpose This study aims at the cooling design for the binary current leads for SCU's normal operating. Methods Binary current leads composed of normal conductive copper leads and high temperature superconducting current leads (HTS) were adopted for SCU to connect superconducting coils inside the cryostat and outer cables. Low-temperature helium gas was used to transport independent refrigerator system to the cooling tubes inside the prototype, hence the binary current leads were cooled. Thermal conduction components installed on the middle of the thermal shield were employed to transfer heat load of normal conductive copper leads, and heat load of copper leads was optimized by simulation. Auxiliary superconducting rods were designed for connecting cold ends of HTS in the cryostat test. Results The temperature difference between hot ends of HTS and low-temperature helium gas is less than 20 K from the result of cryostat test, all binary current leads is operating normally with full current. Conclusions It is practicable to use cooling tubes with low-temperature helium gas to cool binary current leads of the SCU prototype by thermal conduction, which is different from cooling solution for current leads in any other SCU being developed presently.

As a strong candidate for the new type of non-volatile memories and artificial synaptic devices, memristor has a huge development prospect in aerospace, Mars exploration and other space science and application fields. Once large-scale application of memristor requires extremely stringent radiation resistance performance for the memristors. In order to improve the radiation resistance of memristors, it is necessary to explore the radiation effect mechanism and develop an effective radiation resistance technology. This paper summarizes the research status and trends of irradiation effects on memristors, describes the mechanism and analysis method of irradiation damage of memristor, and focuses on the irradiation effects of the memristors with transition metal oxide material system. Additionally, the possibility of scientific problems and key technologies are discussed, so as to provide some ideas for the radiation hardening and space application of memristor.

BackgroundNarrow rectangular channels are widely used in major thermal flow fields because of the compact structure and large heat transfer area.PurposeThis study aims to improve the prediction method of critical heat flux in the narrow rectangular channel and establishing the critical heat flux (CHF) mechanism model for the enhancement of reactor safety and economy.MethodsCHF experiments was carried out in the present study to identify the dominant mechanism in a narrow rectangular channel at different gap sizes. The visualization experiments were performed at pressures ranging from 1 MPa to 4 MPa, inlet subcooling from 60 K to 120 K, and mass flux from 350 kg·(m2·s)-1 to 2 000 kg·(m2·s)-1.ResultsAccording to the visual experiment results, two typical bubble behaviors are investigated in the narrow rectangular channel. Based on the bubble dynamics characteristics of narrow rectangular channels, a new CHF mechanism model was proposed, and a set of constitutive relations will be provided to close the developed model.ConclusionA comprehensive assessment of new model has been conducted and analyzed by using the experimental data for the upward flow in a vertical narrow rectangular channel and it has good accuracies of less than ±30% as relative to the experimental values.

BackgroundThe energy produced by nuclear fusion on a Tokamak device is mainly exhausted through the divertor, its service life is directly affected by the interaction between huge heat flux from core and the divertor target. The large amount of impurity produced by the heat flux hitting the target leads to the reduction of the plasma confinement performance whilst pumping is an important means to control plasma density and impurity density.PurposeThis study aims to investigate the influence of pumping on the heat load of the target plate which is of reference significance for the future experiment.MethodsBased on the experimental parameters of the HL-2A Tokamak, SOLPS-ITER code was used to study the effect of pumping on the heat load of the divertor target under different upstream electron densities. Analysis was performed through density scanning to find the sensitive threshold whilst and atom-molecular collision process was applied to the effect of pumping on the distribution of plasma and neutral particle parameters in divertor region at different upstream electron density.ResultsDensity scanning results show that pumping near the detachment threshold (TetOSP~5 eV) has a greater effect on the thermal load of the target plate. When the pumping rate is 12 m3·s-1, 36 m3·s-1 and 96 m3·s-1 respectively, the miss threshold and thermal load peak of outer target plate are 1.11, 1.24, 1.39 and 1.37, 1.96, 2.54 times of those without pumping respectively.ConclusionIt is found that the decreases of deuterium molecular density results in the energy of the collision reaction power decreases when the upstream electron density exceeds the detachment threshold, leads to the increase of the temperature and energy flow of the plasma in the target plate.

BackgroundIn accelerator driven sub-critical system (ADS), the high-energy proton beam produced by accelerator is used to strike the target nucleus, and generate spallation neutrons as external neutrons to drive and maintain its operation. The power level and the safety of ADS are susceptible to the instability of proton beam, such as beam overpower (BOP) which is treated as the typical transient accident for ADS system. In ADS core, the sudden increases of power and temperature will be caused by BOP accident, which may exceed the safety limits of materials, threatening ADS safety.PurposeThis study aims to investigate the transient safety characteristics of the eXperimental accelerator driven system (XADS) under BOP accident.MethodsThe multi-physics coupling code MPC-LBE, in which the fuel pin heat conduction (HC) model and the point reactor kinetics (PK) model were coupled with self-developed computational fluid dynamics (CFD) code, was employed to simulate BOP accident of XADS. Firstly, the MPC-LBE simulation model of XADS was constructed and then the steady state condition was established. On this basis, the double and triple BOP accident cases were simulated, and the safety boundary of BOP accident conditions was also evaluated.ResultsFor the simulation results, in the double and triple BOP cases, the reactor powers increase to only 1.88 and 2.7 times of the original ones, respectively. The maximum temperature of the cladding is about 843 K in the triple BOP case, exceeding its safety limit.ConclusionsConclusions can be drawn that the negative temperature feedback effect plays an important role in protecting the reactor from power sharp rise, and the double BOP case can be treated as the BOP safety boundary of BOP accident in XADS.

BackgroundAt present, in the key stage of the construction of China Initiative Accelerator Driven System, various research institutes have also put forward corresponding core schemes for different purposes, including the accelerators drive advanced nuclear energy system (ADANES) proposed by Institute of Modern Physics, Chinese Academy of Sciences. Detailed calculation and analysis of its scheme can provide strong technical support for the sustainable development of nuclear energy and the national energy security strategy.PurposeThis study aims to analyze the steady-state neutronics characteristics of ADANES reactor with emphasis on the preliminary typical transient analysis under the typical accident conditions of fast reactor.MethodsThe NECP-SARAX (Nuclear Engineering Computational Physics Laboratory, System for Advanced Reactor Analysis at Xi'an Jiaotong University) code system, which was based on deterministic neutron transport theory, was applied to perform the detailed analysis. The main design parameters, including core length, neutron spectrum and reactivity feedback coefficients, were calculated under various fuel types, coolant types, and reactor geometry parameters. In addition, the primary transient characteristics were simulated, including unprotected transient over power and unprotected loss of flow transient. The changes of reactor power and the maximum fuel/coolant temperature were obtained and analyzed.Results & ConclusionsThe steady-state calculation results show that ADANES could achieve 10 effective full power year for each selected case. The reactivity feedback coefficients reach -5.4×10-5 K-1 in total so that the core has inherent safety under typical accident condition conditions during the simulated transient.

BackgroundThe debris motion is an important phenomenon of a high-altitude nuclear detonation, which is also a foundation for the study of the geophysical phenomena such as the ionosphere effect and artificial radiation belt.PurposeThe study aims to clarify the debris motion characteristics and laws from a near-space nuclear detonation.MethodsFirstly, a fluid dynamics model of debris motion from a near-space nuclear detonation was established. Many influence factors were considered, such as the variation of energy dissipation, air density varies with height, gravity, air temperature rise caused by X-ray depositions and radiation cooling. Then the parameters of debris motion within the explosion equivalent of 1 kt~10 Mt and the explosion height of 30~80 km were systematically studied. The evolutions of parameters such as center height, horizontal radius, expanding velocity, ascending velocity, and shape of debris were given. Finally, the variation laws of typical characteristic parameters such as maximum ascending height and expanding radius changing with explosion height and explosion equivalent were summarized.ResultsWhen the explosion height is 30 km, the maximum rising height and the maximum horizontal radius at 5 min for a kiloton-level nuclear explosion debris are about 13~16 km and 4~5 km, the maximum rising height and the maximum horizontal radius at 5 min for a megaton-level nuclear explosion debris are about 20~40 km and 15~30 km. When the explosion height is 80 km, the maximum rising height and the maximum horizontal radius at 5 min for a kiloton-level nuclear explosion debris are about 30~50 km and 20~40 km, the maximum rising height and the maximum horizontal radius at 5 min for a megaton-level nuclear explosion debris are about 200~400 km and 110~220 km. When the explosion equivalent is small and the explosive height is low, the debris evolves into a flat ellipsoid. When the explosion equivalent is large and the explosion height is high, the debris evolves into an inverted pear shape.ConclusionsThe results show that the maximum height, horizontal radius, and speed of the debris cloud increase with the increase in the explosion height and explosion equivalent. The changes of the height, the horizontal radius, the rising time, and the shape of the debris obtained from the study are in good agreement with the literature estimation method. Those obtained motion parameters of debris can provide delayed radiation source information for the study of the geophysical phenomena such as the ionosphere effect and artificial radiation belt of nuclear explosion in near-space.

BackgroundDecommissioning of nuclear facilities generates large quantities of different types of radioactive material. According to the requirement of the waste packaging, most incompressible waste must be put into steel boxes directly, which makes it difficult to measure and obtain the activity of the radioactive material in the steel boxes.PurposeThis study aims to establish an approach for measuring radioactive waste in steel box based on the in situ objects counting system (ISOCS).MethodsIn this work, ISOCS was used to characterize the radioactive material in steel boxes. The corrections for the absorption coefficients and the geometry factors of the big bulky sources were calculated using the ISOXSW (ISOCS Calibration SoftWare) in situ efficiency calibration without a radioactive source software. The verification experiments were carried out using standard sources with similar size and geometry.ResultsThe results show that the measurement error of six symmetrical positions of steel box with standard 137Cs and 60Co radioactive source by ISOCS is less than 30%.ConclusionsThe study verifies that ISOCS software is able to accurately estimate the composition and activity of radioactive material in a steel box.

BackgroundSolid oxide cell (SOC) is the core converter for hydrogen production by high temperature electrolysis of water vapor and hydrogen fuel utilization.PurposeThis study aims to develope two kinds of aqueous casting pastes of NiO-YSZ with different components for the batch preparation of SOC without the usage of a large number of organic solvents.MethodsA 10 cm×10 cm large-scale full-scale cell was prepared by screen printing the hydrogen electrode functional layer, electrolyte layer, barrier layer and oxygen electrode layer with NiO-YSZ support film at one time casting of about 450 μm. The effect of dispersant on the microstructure of hydrogen electrode support and the stability of the pastes were analyzed by scanning electron microscope (SEM). The performance of the SOCs were tested by I-V curve and electrochemical impedance.ResultsBased on the optimized NiO-YSZ supports, the prepared planar SOCs delivers a peak power density of 0.36 W·cm-2 at 750 ℃. The electrolysis current density of SOC can reach -0.68 A·cm-2 at 1.30 V in solid oxide electrolysis cell (SOEC) model.ConclusionsThe performances of the aqueous-based SOCs can be considered highly remarkable, thus supporting the success in scaling the fabrication of SOCs using more environ-mentally friendly processes than conventional ones.

BackgroundAt present, most of the contamination detection equipments for small items in domestic nuclear power plants are manually putting in and taking out the testing items, and the operation time is long and the steps are relatively tedious. According to the on-site demand of nuclear power and feedback on the use of similar imported equipment, contamination detection equipment based on the conveyor belt can effectively solve these drawbacks.PurposeThis study aims to design a small item γ pollution measuring instrument that can be used to transmit items with a conveyor belt, hence effectively improve the detection efficiency and save the cost of manual operation.MethodsSemi-automatic two-channel control design scheme was adopted, one through the motor driver to control the conveyor belt motor running mode, the other control the digital circuit of the radiation detector and the corresponding electrical parts through the main control board. Linkage of contamination monitoring status and item convey was achieved by the status control board associated with the motor drive. The background counting rate of the equipment in the background environment, the counting rate of the radioactive source at rest in the center of the measuring chamber and the dynamic counting rate of the radioactive source moving through the measuring chamber with the conveyor belt were tested and analyzed.ResultsThe results show that the minimum net count value of the detector is 81.6% of the average count in the static state. The net value of the minimum peak value of the detector at motion state is 89.3% of the average peak count, and the minimum detectable limit is 111 Bq.ConclusionsThe test performance of prototype is better than the reference standard and meets the design requirements.

BackgroundIn recent years, with increased awareness of environmental protection and safety, the development of nuclear logging tools using non-chemical sources such as X-ray instead of chemical sources like 137Cs has become a new trend. However, X-ray source usually has a lower energy level compared to chemical source, therefore the measurement accuracy is hardly satisfying the demand of density logging tool.PurposeThis study aims to investigates the detector spacing design of a X-ray source tool based on an existing multi-detector gamma density tool.MethodsBased on a 215.9 mm diameter borehole filled with water where the logging tool was eccentrically placed in, Monte Carlo software Geant 4 was employed for the simulation of the X-ray density logging in the formation density range of 1.7~3.0 g?cm-3. According to density sensitivity, detection efficiency and depth, a series of models of this logging tool with detector-to-source distance between 135 mm and 430 mm were simulated to analyze the detector responses. Finally, based on above data, the design of source spacing for detectors was determined for the X-ray density tool.ResultsThe finalized tool includes three NaI detectors with detector-to-source distances of 160 mm, 270 mm, and 344 mm, respectively. Simulation results show that the maximum wall detection depth reaches 120 mm with the vertical resolution of 344 mm, and the density measurement accuracy is 0.014 g?cm-3.ConclusionsThe feasibility of developing a potential X-ray density logging tool is validated by this study, providing reference for future design of nonchemical source density logging tool.

BackgroundDigital measurement system based on ADCs (analog-to-digital converter) has higher requirement on the signal to noise ratio (SNR) of sampled data. Among all the factors, the jitter of sampling clock has the most prominent effect on SNR.PurposeThis study aims to design a clock circuit based on dual-loop phase-locked loop to reduce the jitter of digital measurement system input clock.MethodsFirst of all, the influence of clock jitter on digital measurement system was analyzed. Then, the LMK04610 chip with dual loop PLL architecture of Texas Instruments was employed to design and implement a dual-loop phase-locked loop jitter cleaner circuit. The cores of this design were power supply design and the loop filter design. At last, the performance of the circuit was tested by using Rodschwarz phase noise analyzer.ResultsAfter testing, the dual-loop phase-locked loop jitter cleaner circuit can reduce the jitter of the 62.475 MHz source clock from more than 7 ps to less than 2 ps with output frequency of 499.8 MHz. The SNR of the sampled data is close to the theoretical value.ConclusionsDual-loop phase-locked loop jitter cleaner circuit has a good result and can provide reference for designers of digital measurement system.

BackgroundThe solid fuel thorium element molten salt reactors (MSR) have attracted more attention recent years. A3-3 graphite is chosen as the fuel matrix for MSR, thus its irradiation behavior and mechanical property is very important before the application.PurposeThe study aims to observe the irradiation defects and hardness of A3-3 matrix graphite after ion irradiation by slow positron beam and nano-indentation, respectively.MethodsThe matrix graphite of fuel elements was irradiated with 1 MeVXe ions to fluence of 5.8×1014 ions·cm-2 and 2.9×1015 ions·cm-2 respectively at room temperature. The slow positron beam and nano-indentation were employed to investigate the effect of Xe ions irradiation on vacancy defects and hardness of matrix graphite. The changes in irradiation induced defects distribution with depth and fluence were analyzed according to the obtained positron annihilation S parameters versus positron incidence energy or depth curves, compared to SRIM (Stopping and Range of Ions in Matter) calculation.ResultsResults from slow positron beam measurement show that 1 MeV Xe ions irradiation in matrix graphite introduces a damage layer with depth of about 600 nm, and the damage peak locates at about 250~350 nm in depth, consisted with SRIM simulation. The S parameters in irradiation samples increase significantly compared to virgin sample, which suggests that a high concentration of vacancy-type defects appeared within irradiation damage layer. In addition, the S parameters increase with the irradiation fluence, which shows that the concentration or size of vacancy-type defects increases. The nano-indentation results show that the hardness of irradiated graphite matrix is enhanced.ConclusionsThe enhanced hardness of A3-3 matrix graphite after ion irradiation is ascribed to the pinning of basal plane dislocation by the high concentration of vacancy type defects introduced by irradiation, consisted with the slow positron beam analysis. Slow positron beam is a very sensitive tool to study the irradiation defects.

BackgroundHigh-performance accelerators have higher requirements for operational reliability and stability. By analyzing the historical data that is routinely saved during accelerator operation, most failures can be judged. However, when some rapid failure processes occur, due to the insufficient granularity of the historical data stored conventionally, it is impossible to effectively analyze such rapid failure processes. When a failure occurs in a particle accelerator, fast acquisition techniques are needed to collect large amounts of data from various devices with precise timestamps. The failure occurrent process can be rapidly reconstructed by using these data to locate and judge the root cause of the failure. In order to obtain data accurately when a failure occurs, hardware devices with data cache function can be used at the front-end devices, and data can be locked and obtained in the synchronous trigger mode. That is, after receiving the synchronous trigger signal, data in the cache area of the front-end hardware device can be locked, and then read and stored.PurposeThis study aims to design a failure analysis system prototype based on the event-timing technique.MethodsTwo core parts of the prototype were implemented: global high-precision timestamp implementation and data assembly and acquisition analysis. As one of the key factors, the global high-precision time stamping of failure data was applied to analyzing failure causes. Based on a high-performance rubidium atomic clock and the event-timing system, high-precision time stamps were implemented in this prototype with synchronization accuracy better than 16 ns to provide global high-precision time stamps for time data. Structured data based on the normative type of EPICS 7 was adopted for assembling and publishing the data. Essential information, including the system name, the subsystem name, the device name, the device card number, the data sampling frequency, the event timestamp, and the latched data, was obtained from the structured data.ResultsThe prototype experiment results show that the failure sequence of different equipment can be distinguished by the obtained high-precision time data, confirming the high feasibility of our proposed failure analysis system.ConclusionsThe prototype designed in this study meets the requirements for rapid failure analysis of particle accelerators. And this prototype will be applied to the CSNS accelerator in the near future. In addition, it can also be applied to EPICS-based and event-timing based accelerator control systems.

BackgroundThe Fukushima nuclear accident in 2011 exposed the shortcoming of high temperature oxidation resistance of zirconium alloy cladding. For this reason, the concept of accident tolerant fuel was proposed in the international nuclear fuel field. Cr-coated zirconium alloy cladding, as an accident tolerant fuel cladding near-term commercial technology approach, has received extensive attention.PurposeThis study aims to study the high-temperature oxidation behavior of Cr-coated Zr-4 alloys at different temperatures.MethodsCr-coated Zr-4 alloy was prepared by multi-arc ion plating, and oxidized in air atmosphere at 800?1 200 ℃ for 4 h. Scanning Electron Microscope (SEM) and Energy Dispersive Spectrometer (EDS) were used to analyze the surface and cross-sectional micro-morphologies of Cr coated Zr-4 alloy samples before and after high-temperature oxidation and the distribution of elements on the cross section. The orientation image microscopy (OIM), inverse pole figure (IPF) and pole figure (PF) of Cr coated Zr-4 alloy samples were obtained by electron backscatter diffraction (EBSD). The phase of the samples was obtained by glancing angle X-ray diffractometer (XRD).The effect of oxidation temperature on the microstructure, phase, Cr-Zr diffusion layer thickness and oxidation weight gain of Cr-coated Zr-4 alloy were investigated.ResultsThe results show that there are a large number of droplets of different sizes on the surface of the Cr-coated Zr-4 alloy samples prepared by multi-arc ion plating, and the Cr coating has a columnar crystal morphology and preferentially grows along the (110) crystal plane. After high temperature oxidation at 800~1 100 ℃ for 4 h, the surface of the sample is oxidized to different degrees, but the un-oxidized Cr coating still remains inside the coating, and no micro-cracks appear on the surface and cross-section of the sample. The thickness of the Cr-Zr diffusion layer increases linearly with the increase of the oxidation temperature, and the oxidation weight gain increases slowly. However, after high temperature oxidation at 1 200 ℃ for 4 h, the Cr coating on the surface of the sample was completely oxidized, a large number of cracks appeared on the surface and cross-section, and the thickness of the Cr-Zr diffusion layer and oxidation weight gain increased sharply.ConclusionsTherefore, the Cr-coated Zr-4 alloy prepared by multi-arc ion plating exhibited good high temperature resistance at 800~1 100 ℃, while accelerated oxidation occurred at 1 200 ℃.

BackgroundPrompt gamma-ray activation image (PGAI) is a non-destructive element imaging method for large volume samples. Most of PGAI platforms are located in research reactors, which limit their applications. From the perspective of in-field applications, attractive alternative neutron sources are isotope neutron source and neutron generator. However, the neutron fluxes of these sources are much lower than that of reactor neutron source, which leads a poor spatial resolution.PurposeThis study aims to solve this problem by implementing an approach based on multi coded-aperture collimators.MethodsFirst of all, the Monte Carlo code MCNP5 was employed to calculate spatial distribution of Cl in a known sample, and the characteristic gamma rays were produced by the thermal neutrons absorbed by the sample. Then, 36 coded-aperture collimators with random holes were used to collimate gamma rays, and 36 gamma signals were collected by high-purity germanium detectors (HPGe). Finally, the imaging of Cl was reconstructed through these data and maximum likelihood expectation maximization (MLEM) algorithm, and the relative deviation (df) and structural similarity (SSIM) were chosen to evaluate the image quality.ResultsThe spatial resolution of the imaging is 1 cm×1 cm, and the relative deviation and SSIM between the reconstructed image and the original image are 0.065 8 and 0.952 1, respectively. After neutron self-shielding correction, the relative deviation and SSIM between the reconstructed image and the original image are 0.002 3 and 0.998 4, respectively, which shows a good agreement.ConclusionsThe proposed approach is efficient to measure the distribution of Cl element, hence for element imaging of plate samples, and the reconstructed image is consistent with the set sample image.