BackgroundThe exploration of mineral resources and geological structures is crucial for the sustainable development of the economy, society, and environment. Cosmic ray muons, a type of natural background radiation, can be utilized in muon transmission imaging which is based on density differences of transmitted target.PurposeThis study aims to enable non-contact, long-range, and non-destructive imaging of target objects, making it a powerful complement to traditional exploration methods for mineral resources.MethodsThe Geant4 software was utilized to simulate the physical processes of cosmic ray muons with materials of varying densities, and the shallow and deep geological structures were explored using a "telescope" configuration for muon transmission imaging. Firstly, the discriminability of muon imaging technology for substances with varying percentage density differences in rocks. For the simulation of, a volcanic model was constructed and four muon detectors was employed for simulating the imaging of the shallow geological structures from different angles, ensuring coverage of the entire mountain. Then, muon detectors located 600 m underground were utilized to extend exploration above unexplored areas with varying scales of undiscovered gold ore to obtain deep geological structures. Muons with energy lower than the minimum penetrating energy along their paths were absorbed by objects whilst muons that reached the detectors carried information about the materials along their paths. Meanwhile, the collected ray information was utilized to establish a density inversion model to obtain the minimum penetrating energy for each path, enabling the deduction of opacity distribution. Finally, the density distribution of volcanic model was determined by combining the geometric structure of the detected object, and individual detection points enabled two-dimensional monitoring whilst multiple detection points allowed for three-dimensional monitoring.ResultsThe imaging results of simulation show that the muon transmission imaging method can differentiate between different geological structures when the density difference exceeds 5%. In deep geological exploration, due to the low muon flux, imaging requires more time to accumulate sufficient muon events. Muon transmission imaging technology can effectively identify mineral deposits within deep rock formations when the difference in opacity between the path of muon penetration through the gold ore and the surrounding rock is greater than 4%.ConclusionsResults of his study demonstrate that the cosmic ray muon transmission imaging technology can be applied to geological exploration to achieve non-destructive exploration and obtain higher imaging accuracy when there is a reasonable density difference in the mineral resources and geological structures of the exploration area.

BackgroundWith the development of materials science and technology, more and more novel neutron shielding materials have been studied and prepared. The test of thermal neutron shielding performance of materials is an important measure of evaluating the shielding performance of novel neutron shielding materials, and provides important guidance for the improvement and application of materials. Special attention has been paid to the thermal neutron shielding properties of some thin materials, such as radiation protection fibers, shielding coatings, radiation protection rubber, composites with thermal neutron absorbers. At present, it is not easy to obtain pure thermal neutron field based on isotope neutron source, and the commonly used thermal neutron detector has a wide energy response range, and there is no unified test standard, hence the accuracy of the test results of thermal neutron shielding performance of materials needs to be further improved.PurposeThis study aims to propose an accurate method for testing the thermal neutron shielding properties of materials and to design a corresponding test platform.MethodsA thermal neutron shielding performance testing method based on the "cadmium filter method" was proposed and a corresponding testing platform was established. Firstly, the platform was based on a 252Cf isotope neutron source and a 3He proportional counter, with the design principle of minimizing errors. Then, Monte Carlo simulation method was used to optimize the materials and dimensions of the radiation shield, neutron moderator, collimator, and detector shield of the platform. The system error caused by the testing principle was reduced by further optimizing the cadmium filter method, and a collimating neutron beam launcher was designed to improve the collimation effect of the neutron beam, increase the thermal neutron fluence rate and thermal neutron share, and reduce the random error and the system error caused by the system design. Finally, the thermal neutron shielding performance of the materials commonly used in the nuclear field were tested and corresponding simulation calculations were carried out to verify the feasibility and accuracy of the test method. Five factors affecting shielding performance testing, i.e., detector type, neutron source intensity, distance to detector, detection system centerline offset, and neutron source type were analyzed simultaneously.ResultsThe actual test results are in good agreement with the simulation results, with excellent repeatability and reproducibility under different influencing factors.ConclusionsThis study provides an effective means to accurately evaluate the thermal neutron shielding properties of materials.

BackgroundPhotonuclear reactions and compact neutron sources have emerged as promising tools for the production of medical isotopes, providing alternatives to conventional reactor-based high-enriched uranium methods. East China University of Technology (ECUT) is currently constructing an electron accelerator-driven photoneutron source (ECANS) for medical isotope production research.PurposeThis study aims to investigate the photonuclear reaction with 100Mo isotope and utilize the generated neutrons for isotopic production.MethodsFirstly, the photonuclear reactions of 100Mo was analyzed in details. The neutron spectrum and activation yield of 99Mo within a high purity 100Mo target were investigated. Then, a new model to produce medical isotopes was established on the basis of the ECANS photonuclear source, comprising neutron energy modulation layer and neutron reflection layer. Finally, the production yields of 99Mo, 177Lu, and 90Y in various natural oxides were calculated and the feasibility of using photonuclear sources for medical isotope production was assessed. The content of radioactive impurities in natural oxides under irradiation conditions was also analyzed.ResultsSimulation results demonstrate that photo-nuclear reactions can effectively produce medical isotopes such as 99Mo, 177Lu, and 90Y, with respective activities of 0.64 TBq·d-1, 0.67 TBq·d-1, and 2.11 TBq·d-1. And in the high purity 100Mo target, the daily output of 99Mo reaches 2.00 TBq·d-1.ConclusionsThis study demonstrates the feasibility of using the photodisintegration reaction of 100Mo as a neutron source for secondary production of medical isotopes, offering the potential to enhance the economic viability of isotope production. The approach of this study provides preliminary insights for subsequent separation and purification processes, hence has certain reference value for the development of tools for radioactive isotope production.

BackgroundEfficiency calibration is an important prerequisite for γ-spectroscopy measurement, and its accuracy directly affects the reliability of measurement results. The commonly used efficiency calibration methods are active efficiency calibration based on experimental measurement and passive efficiency calibration based on Monte Carlo (MC) simulations.PurposeThis study aims to propose a sourceless efficiency calibration method for γ spectrum based on numerical analysis.MethodsBased on theoretical analysis and numerical calculation, a passive efficiency calibration method based on numerical analysis was established, and applied to a constructed geometric model for typical cylindrical LaBr3(Ce) and NaI(Tl) detector with a diameter and length of 3.8 cm. The γ-ray path trajectory inside the detector, relative position of the source and detector, and peak-to-total ratio were comprehensively analyzed. Efficiency formulas for radioactive point, surface, and volume radioactive sources were formulated. Then, a numerical integration program based on Simpson's algorithm was implemented in MATLAB to solve the numerical analytical formula, and applied to the calculation of detection efficiencies of the 3.8 cm cylindrical LaBr3(Ce) and NaI(Tl) detectors for isotropic 137Cs point, surface, and volume sources. The calculation was performed again by changing the position of the point source and the size of the surface and volume sources. Simultaneously, the Monte Carlo (MC) simulation software MCNP5 was employed to establish the physical model of the detector, and the F8 pulse amplitude card was used to calculate the detection efficiency of the detector for specific energy γ-rays. Finally, detection efficiency obtained by numerical calculation was compared with that of MC simulation to verify the efficiency of proposed sourceless efficiency calibration method for γ spectrum.ResultsCompared with the MC simulation results, the maximum error of numerical calculation does not exceed 3.5%, and the maximum relative errors of the point source efficiency, surface source efficiency and volume source efficiency are 3.26%, 3.33% and 3.36%, repectively, which proves the reliability and accuracy of the numerical analysis method in calculating the detection efficiency.ConclusionsThe passive efficiency calibration method proposed in this study is accurate and fast, avoids complicated MC simulation software, which tends to be difficult to understand and operate. Additionally, this work can be extended to the efficiency calibration of nuclear radiation detectors with different shapes, providing a new pathway for the efficiency calibration of detectors.

BackgroundBoron neutron capture therapy (BNCT) is a highly promising and precise cancer treatment technique with substantial application prospects worldwide. The neutron transport process directly influences beam characteristics and accuracy of treatment planning.PurposeThis study aims to use the advanced Monte Carlo software, NECP-MCX, to investigate the neutron transport in an accelerator-driven BNCT (AB-BNCT) device.MethodsThe advanced Monte Carlo software, NECP-MCX, was used to investigate the neutron transport in an AB-BNCT device and calculate the beam parameters at the exit of the beam-shaping assembly (BSA), and results were compared with that calculated using the mainstream Monte Carlo software MCNP software. The relative biological dose deposition in Snyder head phantoms using various databases was calculated using both NECP-MCX and MCNP for comparative analysis. [Results andConclusions] The results show that in the design of the designated beam-shaping assembly, the neutron beam parameters obtained from NECP-MCX simulations minimally deviated from those calculated using the mainstream Monte Carlo software. The values satisfy the specifications of the International Atomic Energy Agency, confirming the applicability of NECP-MCX for neutron transport simulation in AB-BNCT. Regarding the Snyder head phantom, the relative biological dose parameters obtained by matching NECP-MCX with different databases satisfy clinical treatment standards, providing a foundation for selecting databases in the biological-based treatment planning system.

BackgroundThe Shanghai soft X-ray Free Electron Laser (SXFEL) and the Shanghai High repetition rate XFEL and Extreme light facility (SHINE) are equipped with various high-power microwave components, including traveling wave accelerating tubes, deflection cavities, pulse compressors, etc.PurposeThis study aims to develop two high-power stainless steel absorbing loads, so as to meet the testing and operational requirements of these components at high power levels.MethodsFirstly, an initial load model was designed through simulation methods, and the microwave parameters were optimized. Then, the convection heat transfer coefficient in the waterway was calculated using theoretical calculations. Based on this information, thermal analysis was conducted on the mechanical model of the load to determine its temperature distribution under working conditions. Finally, the loads were manufactured and their RF parameters were measured using a vector network analyzer both in the clamping state and after welding.ResultsThe two X-band loads feature a waveguide structure with periodic grooves, and are operated at 11.424 GHz and 11.988 GHz respectively. The simulation results show that the bandwidth of two loads below -20 dB near the center frequency reaches over 100 MHz. The experimental test results are in good agreement with the simulation calculation results, meeting the design requirements.ConclusionsThe developed high power X-band dry loads described in this study satisfy operating requirements for SXFEL and SHINE.

BackgroundHigh-purity Ge (HPGe) detectors are widely used in nuclear science, technology, and national defense because of their excellent energy resolution capability, which enables them to fingerprint gamma rays and accurately determine nuclide types and intensities. However, HPGe detectors typically need to be cooled to low temperatures for normal operation to prevent excessive thermal noise at room temperature.PurposeThis study aims to investigate the heat transfer process and temperature distribution law inside a HPGe detector to ensure the stable operation of the detector in a low-temperature environment.MethodsThrough the analysis of the heat transfer mechanism of the high-purity germanium (HPGe) detector, the internal heat transfer structure and a three-dimensional heat transfer model were constructed. COMSOL software was utilized to simulate the internal temperature distribution of the HPGe detector under liquid nitrogen refrigeration, as well as to investigate the effects of varying refrigeration times and packaging structures on this distribution. Based on these simulations, the packaging structure of the HPGe detector was optimized. Additionally, a temperature testing platform for the detector was constructed, and the simulation results were compared with experimental data to validate the model's accuracy.ResultsThe simulation results demonstrate that the detector model achieves dynamic equilibrium after 6 h of cooling, with the minimum internal temperature at the tip of the cold finger approximately -175 ℃. The utilization of oxygen-free copper as the cold chain material, combined with a low heat leakage and high-strength support material for the cold finger, enhances refrigeration efficiency. Additionally, the Dewar is designed with a sidewall thickness of 1.5 mm, and the spacing between the sidewall and the crystal cylinder is set at 3 mm. These design features collectively facilitate the detector's attainment of a lower limiting refrigeration temperature.ConclusionsThe modeling and temperature field simulation methods for HPGe detectors are validated through a consistency comparison between simulated and experimental data. Theoretical support is obtained for the further optimization and improvement of the design parameters of liquid nitrogen and electric cooling HPGe detectors.

BackgroundMeasurements of neutron energy spectra and doses during discharges of fusion devices are crucial for radiation monitoring and protection, and Bonner sphere spectrometers are generally used to measure neutron energy spectra and doses.PurposeThis study aims to investigate the neutron energy spectrum and dose equivalent of an Experimental Advanced Superconducting Tokamak (EAST) under experimental neutral beam injection (NBI) heating conditions.MethodsFirst, a Bonner neutron sphere spectrometer consisting of a 6Li-coated 4H-SiC semiconductor detector and eight moderated spheres was developed, and measurements in NBI heating and discharge experiments were conducted. Subsequently, the experimental results were normalized with the neutron flux monitoring results of the fission ionization chamber (ZZ3) on the EAST device and combined with the neutron response functions obtained from the Monte Carlo simulation for each moderated sphere. Then, the maximum entropy program was used to calculate the neutron energy spectrum distribution of the experimental positions inside the hall. Finally, the neutron dose equivalent at this location was obtained using the fluence dose conversion coefficient and compared with the neutron energy spectrum and dose equivalent at the corresponding location in the main hall, simulated using an existing EAST Monte Carlo model.ResultsComparison results show that the overall agreement between the simulated and experimental energy spectrum distributions of the measurement position is good, and the neutron flux at the measurement position under the normalized fusion neutron source is 1.38×10-7 cm-2, with a ratio of 0.98 to the simulated value. The surrounding dose equivalent is about 2.27×10-11 μSv, with a ratio of 1.05 to the simulated value, which shows consistency.ConclusionsThe results of this study demonstrate that the multisphere spectrometer is reliable for measuring neutron spectra and dose equivalents in the EAST hall and providing data for radiation protection in fusion devices. It can also provide a reference for neutron spectrum measurements in high-power fusion devices.

BackgroundThe vanadium self-powered neutron detector (SPND) generates significant gamma noise current in the mixed radiation field of a nuclear reactor, which adversely affects the accuracy of neutron flux measurements and online monitoring of reactor core.PurposeThis study aims to explore the gamma effect of vanadium SPND through a combination of theoretical calculation method and experimental verification.MethodsFirst of all, a current component separation model was established according to the vanadium detector response mechanism, and the current calculation method of SPND in the pulsed research reactor was constructed. Then, multi-step Monte Carlo method was used to realize the coupling calculation of the core scale and the detector signal. Based on the temporal characteristics of different current components, the prompt γ current component was quantified by using the shutdown attenuation measurement experiment at the 200 kW radiation chamber of a research reactor. Finally, the measured signals were used to verify the sensitivity component separation model, and the dynamic response tracking and prompt γ current simulation capabilities of the detector numerical model were further verified by the vanadium SPND simulation of the commercial pressurized water reactor load rejection test process.ResultsThe established vanadium SPND calculation model can be used to effectively distinguish the detector response current component with a relative deviation of 2.27% for steady-state current calculation, and the calculated results agree well with the experimental data in terms of calculating the steady-state current and the prompt component of the steady-state current for the vanadium detector.ConclusionsThis study has shown that the gamma effect of vanadium SPND needs to be considered in both theoretical calculations and measurement analysis, and there is a mutual compensation effect among different current components in the gamma response of the actual reactor core.

BackgroundWhen performing gamma-ray spectroscopy analysis of samples with low levels of radioactive nuclide content, the weak peaks are difficult to be identified.PurposeThis study aims to propose a new method for identifying peaks in γ spectra by utilizing singular value decomposition (SVD) to improve the detection efficiency of weak peaks.MethodsFirstly, the matrix construction of spectrum data was improved by transforming the γ spectrum into a two-way cyclic matrix, and singular value decomposition of matrix was performed to get singular values and singular vectors. Then, the second singular value was selected to reconstruct the matrix and perform peak finding. Finally, the γ spectrum of the radioactive source 152Eu was used as the experimental object, the peak finding performance of proposed method was compared with that of first-order derivative peak finding, symmetric zero-area peak finding, and singular value decomposition peak finding.ResultsComparison result show that the bidirectional circular matrix SVD peaking method has higher recall rate, precision rate, and F1 value, achieving 100%, 87% and 0.94, respectively.ConclusionsThe approach of this study can optimize weak peak detection and offer additional options for peak finding methods.

BackgroundDuring the long-term operation of a nuclear reactor, the contact between zirconium alloy cladding and cooling water results in oxidation reactions and hydrogen uptake-induced embrittlement behavior, which deteriorates the thermal and mechanical properties of the cladding, posing a threat to the safety characteristics of fuel elements. Therefore, conducting research on the oxidation and hydrogen uptake behavior of rod-shaped fuels is of significant importance. MOOSE is an object-oriented finite element multi-physics coupling platform developed using the C++ programming language. BEEs, developed based on MOOSE, is programmed in C++ and operates under the Linux system.PurposeThis study aims to integrate a corrosion model into MOOSE-BEEs fuel performance code and verify its adaptability, consisting of an oxidation corrosion model and a hydrogen absorption corrosion model.MethodsFirstly, a corrosion calculation model for pressurized water reactor rod-shaped fuel in the MOOSE-BEEs program was developed and integrated into the MOOSE platform to enhance the functionality of the BEEs program. The corrosion model primarily included an oxidation corrosion model and a hydrogen absorption corrosion model. The oxidation model served as the boundary of the hydrogen absorption model to provide hydrogen uptake. The hydrogen at the boundary diffused under the action of concentration gradient and temperature gradient. Then, according to the relationship between the concentration in the region and the terminal solid solubility, predictions was made regarding the occurrence of precipitation phenomena at this location. The terminal solid solubility and precipitation rate are related to temperature. Subsequently, simple geometric structures were established to perform coupled calculations of fuel thermal conductivity, oxidation, hydrogen absorption corrosion, hydrogen diffusion and precipitation. Finally, the calculated results were compared with the BISON program and experimental values, and the hydrogen precipitation was verified in terms of terminal solid solubility and precipitation rate.ResultsBased on experimental data and computational results from the BISON program, separate models and coupled models for oxidation corrosion, hydrogen diffusion and hydrogen precipitation have been validated. The oxidation corrosion model is in good agreement with REP Na10 experiment results and Katheren calculation results. Hydrogen diffusion verification includes concentration gradient verification and temperature gradient verification. The diffusion model and hydrogen precipitation model are in good agreement with the results of BISON simulation and Kammenzind experiment. The coupling model of oxidation and hydrogen absorption corrosion is in good agreement with the results of BISON simulation and Gravelines reactor experiment. The difference between the calculated results of most corrosion models and the experimental values and BISON program is less than 10%.ConclusionsThe validation results demonstrate that the BEEs predictions are in good agreement with the experimental data and BISON program, indicating that BEEs is capable of accurately simulating the oxidation and hydrogen absorption behavior of fuel rods.

BackgroundCompared with traditional components, novel dual cooled assemblies for liquid-fuel molten salt reactors demonstrate a lower graphite hot spot temperature owing to their increased heat transfer area and reduced graphite thermal conduction distance. However, owing to the distinct geometric characteristics and spontaneous heating of liquid fuel, unique heat division and flow distribution patterns occur between the internal and external channels in dual cooled assemblies.PurposeThis study aims to develop new computational analysis tools to estimate the thermal-hydraulic performance of these assemblies.MethodsThis study developed a one-dimensional steady-state thermal-hydraulic analysis code, namely Thermal-Hydraulic Analysis Code for Dual Cooled Assembly-Molten Salt Reactor (THDA-MSR) using the MATLAB platform. Considering the spontaneous heating of the fuel salt, a one-dimensional temperature distribution model for liquid fuel and a he at transfer model between the fuel salt and graphite were established. A flow distribution model was developed based on the principle of equal pressure drops in parallel channels. Additionally, numerical simulations were performed using computational fluid dynamics (CFD) software to validate the code results.ResultsThe study demonstrates a strong agreement between the THDA-MSR and CFD results, with a pressure drop deviation and maximum graphite temperature deviation below 4.84% and 0.15%, respectively. The analysis shows that the ratio of the external channel is the key parameter affecting the maximum outlet temperature of the fuel salt and the maximum temperature of graphite.ConclusionsThe model is applicable for thermal-hydraulic calculations and analysis of dual cooled assemblies in liquid-fuel molten salt reactors.

BackgroundIn a pressurized water reactor, the corrosion chemical reaction between zirconium alloy cladding and water will adversely affect the mechanical properties of the cladding, thus limiting the service life of the fuel elements. In order to slow down the oxidation rate of the cladding and prevent the potential risk of hydrogen explosion, a conceptual design of accident tolerant fuel was proposed. Chromium metal has excellent corrosion and oxidation resistance, and has been widely used as cladding coatings in the field of nuclear power. At present, the micro-mechanism of corrosion and oxidation resistance of chromium coating at high temperature is not clear, so it is urgent to carry out relevant research.PurposeThis study aims to investigate the diffusion behavior of oxygen and hydrogen in coating on fuel cladding.MethodsThe diffusion mechanism of oxygen and hydrogen in chromium crystals was investigated on the electronic scale by using the first principles method. The Arrhenius diffusion equation was employed to obtain the diffusion coefficients of O and H at different temperatures. In addition, the reaction-diffusion paths and migration energy barriers of oxygen and hydrogen were calculated by elastic band method.ResultsSimulation results show that oxygen occupies the most stable position in the octahedral interstitial site (OIS), and hydrogen tends to occupy the tetrahedral interstitial site (TIS). The oxygen atoms diffuses from the reaction path TIS to TIS with the diffusion energy barrier 0.79 eV whilst the oxygen atoms diffusion along the TIS to OIS reaction path has the diffusion energy barrier 0.65 eV. It suggests that there is a preferential diffusion pathway from TIS to OIS for oxygen atom due to its lower diffusion energy barrier. Notably, hydrogen demonstrates comparable diffusion energy barriers (0.17 eV) when moving along the reaction path from TIS to TIS and from TIS to OIS, respectively. The diffusion coefficients of oxygen and hydrogen increase linearly with the increase of tempterature, respectively. And for, the diffusion coefficients of both oxygen atom and hydrogen atoms along the TIS to OIS reaction path are higher than that of TIS to TIS reaction path at the different temperatures.ConclusionsThe solubility of hydrogen is much lower than that of oxygen. The negative dissolution energy of oxygen in the interstitial site indicates that there is a strong mutual attraction between oxygen and the first nearest neighbor chromium. A fitted relationship established between temperature and diffusion coefficient in this study provides theoretical support for investigation coating corrosion properties at elevated temperatures.

BackgroundFor high-fidelity simulations of fluid dynamics in molten salt reactor (MSR), even though a supercomputer is able to suppress the period of each simulation, the consequent expense is still prohibitively costly. A possible way to overcome this limitation is the use of Reduced Order Modelling (ROM) techniques.PurposeThis study aims to evaluate the accuracy of the ROM methods for reconstructing the velocity and pressure fields.MethodsTwo ROM methods based on the Proper Orthogonal Decomposition (POD) with both Galerkin projection, namely FV-ROM (ROM based on Finite Volume approximation) and SUP-ROM (ROM with supremizer stabilization), were established for fluid dynamics of MSR. Then, both methods were tested on the unsteady cases of liquid-fueled molten salt reactor (LFMSR) for comparison and applicability analysis.ResultsThe FV-ROM demonstrates notable advantages in both velocity prediction and computational efficiency. For laminar and turbulent transient simulations, the average velocity L2 relative errors are less than 0.5% and 0.6%, respectively, with acceleration ratios of approximately 1 500 and 1 000 times for single time steps. Conversely, the SUP-ROM scheme demonstrates significant prowess in pressure prediction, achieving remarkably low pressure average L2 relative errors of 0.20% and 0.38% for laminar and turbulent transient scenario, respectively.ConclusionsResults of this study indicate that combination of SUP-ROM and FV-ROM for fluid dynamics computations of MSR can significantly enhance computational efficiency and ensure reliability and accuracy of transient simulation.

BackgroundHelical coiled tube steam generator is the core equipment for energy transfer in a liquid metal fast reactor (LMFR), which transfers the heat released from the core on the primary side to the working mass on the secondary side, generates steam and pushes the turbine to do work. The stability and safety of its operation have a crucial impact on the operational safety, economy and reliability of nuclear power plants.PurposeThis study aims to propose a numerical simulation method using computational fluid dynamics (CFD) software for the coupled heat transfer calculation of two-phase fluids in the steam generator of LMFR.MethodsFirst of all, a three-dimensional numerical model of coupled primary and secondary heat transfer in the steam generator of LMFR was constructed, and the correlation equations of liquid metal and water-vapor variability were established based on the OECD (The Organisation for Economic Co-operation and Development) physical property handbook and the NIST (National Institute of Standards and Technology) database. Then, the Lee phase transition model was used to calculate the mass transfer between the two phases during the evaporation of water-vapor on the secondary side. Finally, the lead-bismuth fast reactor was taken as an object, the coupled heat transfer characteristics between the primary and secondary sides of the steam generator under different primary-side inlet parameters were investigated and compared with the conventional water reactors.Results & ConclusionsThe results show that, under the same conditions, compared with the traditional water reactor, the wall heat flux between the primary and secondary sides is significantly increased when lead-bismuth liquid metal is used in the primary side, and the peak heat flux can reach 1 439.97 kW?m-2, which is 5~6 times higher than that of the corresponding value of the water reactor, which leads to a significant intensification of the vapor evaporation process in the tube of the secondary side, and the volumetric vapor volume fraction rate rises sharply. Simutaneously, the along-track heat flux distribution between the primary and secondary sides is more heterogeneous, which leads to an increase of the vapor volume fraction rate. Meanwhile, the relative deviation of the heat flux distribution along the heat flux is 3~4 times larger than the corresponding value of water reactor. With the increase of the inlet lead-bismuth temperature on the primary side from 350 ℃ to 450 ℃, the wall heat flux between the primary and secondary sides increases, and the corresponding peak heat flux increases from 950.7 kW?m-2 to 1 439.97 kW?m-2. The distribution of the along-range heat flux between the primary and secondary sides is more inhomogeneous, and the inhomogeneity is increased by 20%.

BackgroundIn case of transient changes or minor accidents during reactor operation, the fuel temperature may temporarily exceed the critical threshold, thus forming bubbles on the fuel plate. Bubbling can significantly affect the temperature distribution and mass flow balance in rectangular channel of the fuel plate, which may lead to the rupture of the fuel plate and even the damage of the whole reactor core. The phenomenon of bubbling in plate-type fuel assemblies within nuclear reactors includes fission gas bubbles and solid bubbles.PurposeThis study aims to investigate the effects of air gap on the flow and heat transfer behavior in rectangular channel of fuel plate during bubbling conditions.MethodsFirstly, a fuel plate and two adjacent flow channels were selected as the calculation domain, and Fluent software with dynamic mesh technology was utilized to simulate gas bubbling and solid bubbling phenomena within nuclear reactor fuel plates. Then, the dynamic mesh was employed to accurately adapt to the geometric changes during bubble formation and development, and the Realizable k-ε turbulence model was used to handle complex fluid dynamics, with boundary conditions set as inlet velocity and outlet pressure to reflect real operational environments. Finally, the differences between fission gas bubbling and solid bubbling were compared, and all solid surfaces were designated as no-slip and adiabatic, enhancing the predictions of interactions between heat transfer and fluid flow.ResultsThe findings reveal that gas bubbles cause a local increase in temperature, with the heat flux around the bubbles tripling, though the overall heat flux of the fuel plate remains largely unchanged. The formation of bubbles locally enhances heat transfer capability by approximately 10%, with a 4% increase in heat flux on the bubble side. Under conditions of high flow rates, the presence of bubbles leads to a significant pressure difference across the fuel plate, causing deformation of the fuel plate and potentially leading to the blockage of the flow channel.ConclusionsResults of this study provide significant references for the design and safety assessment of nuclear fuel plates, highlighting the importance of considering the effects of gas bubbling on thermal-hydraulic characteristics in the design and operation of nuclear reactors.

BackgroundFlexible neutron protection composite materials are of great significance for the protection of special-shaped components and personnel. Their performances are closely related to the properties of functional fillers, but the effect of size of 2D functional fillers on the performance of composites is not yet clear.PurposeThis study aims to explore the influence of two-dimensional (2D) functional filler size on the mechanical properties and neutron shielding performance of composite materials.MethodsFirstly, ethylene propylene diene monomer (EPDM) rubber with good performance such as high temperature resistance, weather resistance, radiation stability, mechanical properties, and high hydrogen content was used as flexible substrate material, and layered boron bitride (BN) with a high thermal neutron absorption cross-section was taken as 2D functional filler, and the surface of BN was modified by mercapto group through two-step chemical grafting. BN/EPDM flexible neutron protection composite materials were prepared by the process of plasticizing, mixing and hot pressing and vulcanization, and the content of BN was controlled at 20~100 phr (parts per hundred parts of rubber), and azodiisobutyronitrile (AIBN) acted as an initiator to promote the interfacial bonding between mercaptylated boron nitride (BN-SH) and EPDM substrate. Then, the microstructures of the materials such as functional group and microscopic morphology were characterized and analyzed using Fourier transform infrared spectrum (FTIR) and scanning electron microscope (SEM). Subsequently, the tensile properties including tensile strength and elongation at break were characterized by universal testing machine, according to GB/T 528―2009 standard. The surface hardness was measured by Shore A hardness tester, according to GB/T 531.1―2008 standard. Finally, the neutron shielding performance was tested based on cadmium sheet difference method, where americium-beryllium source with moderation and collimation was used as narrow beam measurement geometry, and 3He detector was used to count neutrons.ResultsThe experimental results indicate that the reduction of BN size is conducive to improving the tensile strength and thermal neutron shielding performance of composite materials. When the thickness of the material is 2 mm, its tensile strength can reach up to 8.13 MPa, and the thermal neutron shielding rate can be increased by up to about 5%, which is related to the amount of nano BN.ConclusionsThis study confirms the size effect of two-dimensional functional fillers in neutron shielding composite materials, and can provide the basis for the design and preparation of polymer-based flexible neutron protection composites.

BackgroundThe Scanning Transmission X-ray Microscopy (STXM) endstation, located at the Shanghai Synchrotron Radiation Facility (SSRF), stands as China's sole STXM device. It boasts the capability for high spatial resolution imaging down to a 15 nm scale. Utilizing the scanning coherent diffraction imaging method (i.e. ptychography), this station achieves an optimal resolution of 7.32 nm, necessitating utmost system stability. However, the STXM system operates within a complex environment where the impact of external vibrations on imaging quality is increasingly problematic.PurposeThis study aims to develop a closed loop control method for the vibration of STXM system to improve imaging quality.MethodsFirst of all, system vibration between the Fresnel zone plate (FZP) and the sample was analyzed in detail, so did the vibration suppression methods. Then, a software-driven closed-loop control system was developed on Field Programmable Gate Array (FPGA) platform by leveraging these insights, and a Fast Fourier Transform (FFT) based approach was implemented to process precise (picometer-scale) and rapid (1 kHz) positional data captured by a laser interferometer. Finally, the internal sensor of motor controller was replaced by laser interferometer to realize the closed-loop control to suppress vertical vibration, and vibration data were applied to a piezo motor for effectively replacing its internal sensors to suppress system vibrations.ResultsThe optimal performance achieved by the closed-loop control scheme shows that the root mean square (RMS) value of the vibration is reduced to 4.833 nm, which is around one third of the value before.ConclusionsThe successful implementation of this closed-loop control scheme has furnished robust technical foundations for further enhancements in the imaging resolution of the STXM system.