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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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 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.
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.