核动力工程

Research and Validation of the Monte Carlo-based Multi-cycle Neutronic Calculation Methodology for High Flux Research Reactor

Xia Yi, Peng Xingjie, Kang Changhu, Ma Liyong, Qiu Liqing, Liu Runqi, Liu Chang, Song Jiyang

2025, 46(S1): 1-7. doi: 10.13832/j.jnpe.2025.S1.0001

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This paper focuses on the irradiation test in the high-flux research reactor. Based on the Monte Carlo neutron-photon transport-burnup coupling code, the physical processes of the generation, transport, leakage, and deposition of different particles in the global core environment and the local test device environment are studied, and a high-fidelity calculation model of energy deposition that can describe the physical processes of multi loops throughout the core is established based on the repetitive geometric structure. Taking into account the dynamic changes of fuel burnup, control rod position, and their coupling relationship with the particle transport process during the operation of the high-flux research reactor, the whole reactor refueling and fuel management calculations are realized, including multi-type of fuels and burnable poisons. Validation is performed using actual critical rod position, irradiation test assembly power, and point burnup measurement values under different burnup steps. The calculation results show that the error in the core neutron effective multiplication factor (k_eff) is less than 1200 pcm, and the average relative errors between calculated and measured values for irradiation test assembly power and point burnup are both less than 10%, verifying the accuracy of the multi-cycle calculation method.

POD Order Reduction Based on LBM Neutron Diffusion Equation

Chi Honghang, Wang Yahui, Ma Yu

2025, 46(S1): 8-12. doi: 10.13832/j.jnpe.2025.S1.0008

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In the unified multi-physics coupling simulation of reactor engineering, the Lattice Boltzmann Method (LBM) has good application prospects as a mature and reliable method. However, the computational resource consumption when calculating complex core structures remains an urgent issue that needs to be addressed. In order to improve computational efficiency and reduce the requirements for computing resources, this paper proposes a proper orthogonal decomposition (POD) reduction method based on the LBM neutron diffusion equation. The corresponding POD reduced-order model for LBM neutron diffusion equation is established, achieving a calculation effect of thousands of times acceleration ratio while ensuring its calculation accuracy.

High-fidelity Neutronic, Thermal-Mechanical and Heat Pipe Heat Transfer Study of Solid-state Reactors

He Ying, Qiu Meiming, Ma Yugao, Liu Guodong, Huang Shanfang, Wang Kan

2025, 46(S1): 13-20. doi: 10.13832/j.jnpe.2025.S1.0013

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Compared to traditional pressurized water reactors, solid-state reactors operate at higher temperatures, resulting in significant feedback effects due to thermal expansion. This paper proposes a high-fidelity neutronic, thermal-mechanical and heat pipe heat transfer coupling model. Based on the RMC-ANSYS coupling, the heat pipe analysis code HPTRAN is employed to calculate the axial temperature distribution of the heat pipe, providing more accurate boundary conditions for thermal calculations. By decoupling the feedback of fuel and monolith shapes, the model can accurately account for the relative positions, shapes, densities, and temperatures after thermal expansion. Applying the coupling model to multi-physics coupling of a typical solid-state reactor, the $ {k}_{\mathrm{e}\mathrm{f}\mathrm{f}} $ decreases by 570pcm (1pcm = 10⁻⁵) compared with uncoupling, the maximum fuel temperature increases by 41 K, and the maximum monolith temperature increases by 37 K. The axial temperature difference in the heat pipe can reach 200 K, and the radial temperature difference can reach 50 K. Using fixed heat pipe wall temperatures in solid-state reactor multiphysics analysis introduces substantial errors, demonstrating the necessity of introducing heat pipe coupling.

Research on the Iterative Method for Solving Neutron Diffusion Equation

Fang Chao, Li Qing, Peng Xingjie, Zhao Wenbo, Liu Kun, Chen Zhang, Wang Lianjie

2025, 46(S1): 21-25. doi: 10.13832/j.jnpe.2025.S1.0021

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To improve the computational efficiency of solving the eigenvalue of the neutron diffusion equation, this paper conducts an in-depth research on the power iteration method and the eigenvalue iteration algorithm based on the Krylov subspace idea. Firstly, in the power iteration method, an initial value setting based on fission source normalization is proposed and compared with the traditional method. Then, the number of iterations and computation time of the power iteration method and Kroylov-based iteration algorithm are compared. Finally, the preconditioning techniques of Kroylov-based iterative method are studied, including the Jacobian preconditioning, incomplete LU decomposition preconditioning, and algebraic multigrid preconditioning, and their impact on the number of iterations and computation time are analyzed. Through the calculation of the IAEA 3D benchmark problem, the results show that the Davidson type iterative algorithm combined with incomplete LU decomposition preconditioning has superior computational efficiency. For a problem with two million elements, the calculation can be completed within one minute by using a single core. Compared with traditional power iteration methods, the computational efficiency is improved by about 25 times. This achievement has significantly enhanced the computational efficiency of eigenvalue problems of the neutron diffusion equation and remarkably reduced the time cost of neutronics calculations.

Prediction of Thermal-Hydraulic Parameters in Rod Bundle Assembly Domain Based on Similarity Features

Qian Hao, Chen Guangliang, Sun Dabin, Li Jinchao, Yin Xinli, Zhang Lixuan, Zhang Yuhang, Li Rui

2025, 46(S1): 26-32. doi: 10.13832/j.jnpe.2025.S1.0026

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To accurately predict the thermal-hydraulic parameters of reactor core flow domains under high Reynolds numbers and complex geometries, and to enhance the predictive accuracy of neural networks for rapid assessment of core thermal-hydraulic conditions, this study proposes a novel auxiliary prediction method. An analysis of detailed simulation results of rod bundle channels under various operating conditions identified a similarity pattern between the macroscopic thermal-hydraulic parameters of the rod bundle assemblies and the distribution of fine-scale parameters. This pattern was then utilized to construct the input features of the neural network, enabling precise predictions of temperature, pressure, and velocity parameters. The results demonstrate that the developed surrogate model achieves a maximum mean squared error (MSE) of 7.86×10⁻⁴ and a minimum MSE of 1.39×10⁻⁴ on macroscopic parameter test data. For fine-scale parameter test data, the maximum and minimum MSE are 9.39×10⁻³ and 5.20×10⁻⁴, respectively, indicating the model’s high accuracy in predicting core thermal-hydraulic conditions. Moreover, the surrogate model obtains fine-scale thermal-hydraulic parameter fields of the core within 0.504 seconds—an efficiency improvement of 1149 times compared to traditional methods. This provides a powerful technical foundation for developing digital twins of nuclear reactor cores.

Research on Phase Distribution Characteristics of Flow Field downstream the Spacer Grid in 5$ \times $5 Rod Bundle

Cao Mingze, Yan Xiao, Zhang Junyi, Gong Suijun, Xing Dianchuan, Xu Jianjun

2025, 46(S1): 33-40. doi: 10.13832/j.jnpe.2025.S1.0033

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In order to research the phase distribution characteristics downstream the spacer grid in 5×5 rod bundle, a new type of wire mesh sensor suitable for the prototype size of 5×5 rod bundle was designed and fabricated, and the distance between measuring points was 1.05 mm. The void fraction measurement experiment of air-water two phase flow in the 5×5 rod bundle with spacer grid was carried out. The void fraction distribution characteristics within the channel were analyzed, and the phase distribution characteristics induced by gas-phase accumulation due to spacer grid mixing vanes were identified. The experimental results show that due to the overturning effect of lift, bubbles gather near the rod wall under low void fraction conditions, while bubbles appear in the center of subchannel under high void fraction conditions. The spacer grid mixing vanes cause a certain migration of the gas peak position in the channel, and gas accumulation related to the orientation of boundary mixing vanes is also observed at the side wall of the rod bundle channel. The developed wire mesh sensor can be used to measure the void fraction in the downstream flow field of more types of spacer grids, and provide reference for the structural optimization design of spacer grids.

Preliminary Study on 950℃ Coolant Outlet Temperature in HTR-PM under the OTTO Scheme

Liu Songyang, Wang Lang, Li Xuelin, Guo Ruonan, Liu wei, Luo Yong, Zhou Qin

2025, 46(S1): 41-51. doi: 10.13832/j.jnpe.2025.S1.0041

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The High-Temperature Reactor Pebble-bed Module (HTR-PM) adopts helium as the coolant, with 250/750 ℃ for core inlet/outlet temperature. Based on the public HTR-PM design parameters, this study adopts the once-through-then-out (OTTO) refueling scheme with a coolant outlet temperature of 950℃. The VSOP-THERMIX code is employed to analyze the distribution of key parameters in the HTR-PM during the equilibrium core phase. The coupled results of the neutronic–thermal hydraulics show that the maximum fuel temperature under steady-state conditions reaches 1157℃, which is below the safety limit of 1200℃, meeting the temperature criterion for retaining radioactive fission products under steady-state operating conditions. To further investigate the safety of HTR-PM with a 950℃ outlet temperature under accident conditions, a depressurized loss-of-forced-cooling (DLOFC) accident is selected to analyze the changing of maximum fuel temperature. The results indicate that 14.4 hours after the accident, the maximum fuel temperature reaches 1931.7℃, exceeding the accident temperature limit of 1620℃, but remains below the melting points of graphite and silicon carbide. Therefore, the core meltdown will not occur in the DLOFC accident. After this time point, the maximum fuel temperature gradually decreases. Moreover, the results reveal that the location of the maximum fuel temperature moved from the bottom to the upper part of the core during the DLOFC accident. To further analyze the influence of fuel enrichment in the DLOFC accident, four different fuel enrichments ranging from 8.0% to 9.5% are compared under the same refueling and operating conditions. The results show that the steady-state power peak of the equilibrium core shifts upward with increasing fuel enrichment under the OTTO scheme. Under DLOFC conditions, the maximum fuel temperatures are 1949.2, 1931.7, 1916.2℃, and 1900.8℃, respectively, showing a decreasing trend with higher enrichment levels.

Experimental and Numerical Investigation on the Hydrogen Behavior in Containment under Severe Accident

Liu Tong, Ma Le, Gong Houjun, Zan Yuanfeng

2025, 46(S1): 52-57. doi: 10.13832/j.jnpe.2025.S1.0052

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During the loss of coolant accident (LOCA) in a pressurized water reactor, a large amount of steam and hydrogen released could pose a threat to the containment integrity. Therefore, it is necessary to conduct an in-depth study on the pressure response characteristics and hydrogen behavior in the containment. This study investigates the LOCA blowdownand hydrogen release processes through experimental and numerical simulations. The high-temperature and high-pressure working fluid was spewed to the containment simulator throughthe pressurizer. A concentration measurement system was employed to measure the evolution of the steam and helium volume concentration in the containment simulator. Furthermore, the three-dimensional computational fluid dynamics code Gasflow-MPI was utilized to simulate the experimental process. Base on the experimental and numerical simulation results, the hydrogen distribution in the containment simulator was further analyzed. The experimental and numerical simulation results show that there is no obvious temperature stratification in the containment simulator, but the helium volume concentration at the top of the simulator is higher than that at the bottom with an obvious stratification.

Research on the Application of SARAX in the Design of Marine Heat Pipe Reactor

Li Fanchen, Zheng Youqi, Wang Xiayu, Wang Shidi

2025, 46(S1): 58-65. doi: 10.13832/j.jnpe.2025.S1.0058

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With the gradual maturation of heat pipe reactor technology, its potential application in unmanned underwater vehicles (UUVs) is constantly developing. In response to the stringent requirements of underwater platforms for compactness, extended operational lifetime, and high nuclear power system reliability, this paper proposes a megawatt-class heat pipe reactor specifically designed for nuclear-powered UUV, with a design lifetime of 10 years. A comprehensive neutronic analysis was conducted using SARAX, covering critical parameters such as fuel burnup trend, power distribution, reactivity coefficients, and control rod worth. The simulation results demonstrate that the proposed reactor exhibits favorable neutronic characteristics along with a well-balanced power distribution. Over its 10-year lifetime, the reactor achieves an average fuel burnup of 9.455 GW·d/t(U); the radial relative power peak of the core occurs in the outermost fuel assembly; the reactivity compensation of shim rods during the lifetime is less than 2000pcm (pcm=10⁻⁵); the worth of a single control rod does not exceed 1 β. These outcomes confirm that the reactor fulfills the design criteria of compactness, lightweight construction, and safe reactivity control. The proposed reactor is expected to provide a stable and continuous power supply for UUVs, thereby significantly enhancing their underwater endurance and mission execution capabilities.

Research on Multi-source Heterogeneous Fault Characterization Method for Reactor Coolant Pump

Xu Renyi, Wang Yan, Kuang Chengxiao, Wu Kelin, Su Shu, Tan Xin

2025, 46(S1): 66-74. doi: 10.13832/j.jnpe.2025.S1.0066

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Aiming at the problems of high-frequency sensing signal modulation, noise interference, and the low fault recognition rate and lack of evidence in single-sensor fault diagnosis for nuclear power plant reactor coolant pumps (RCPs), this paper proposes a method of multi-source heterogeneous fault characterization for RCPs based on cyclic stationary analysis and D-S evidence theory. By using time domain analysis and cyclic stationary analysis to process the high frequency sensor data, the signal demodulation and denoising are realized, and the feature parameters are calculated to construct the feature vector. And then, multi-source sensing data is fused based on D-S evidence theory, and the typical fault diagnosis of RCPs is realized at decision level according to the fusion results. The experimental verification results show that the fusion of multi-source sensing information can significantly improve the diagnosis rate of typical RCP faults, and improve the interpretability of diagnosis results. The relevant research results can provide a reference for the predictive maintenance of RCPs, and improve the operation reliability and intelligent operation and maintenance level of RCPs in nuclear power plants.

Single-channel Temperature Prediction of Heat Pipe Reactor Based on Deep Neural Network

Yu Xin, Wang Jiajun, Guo Kailun, Zhang Zeqin, Tian Wenxi, Su Guanghui, Qiu Suizheng

2025, 46(S1): 75-81. doi: 10.13832/j.jnpe.2025.S1.0075

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Heat pipe reactors have become a strong candidate for nuclear power generation due to their unique design and efficient heat conduction performance. However, accurate monitoring of the core temperature field remains a key challenge. This paper explores a novel method for rapid prediction of the core temperature field based on deep learning technology. By establishing a backpropagation neural network (BPNN) model and training a large amount of core numerical simulation data, it is possible to predict the temperature field of a single channel core section using 6 temperature measurement points. The training results of the model show that selecting the appropriate number of neurons and hidden layers can effectively improve prediction accuracy and reduce the risk of overfitting. The neural network model in this study has an average absolute error of 1.06 K on the test set, demonstrating good predictive ability and a low level of error. Errors are primarily concentrated in corner fuel rods and regions with intense heat exchange.

Configuration and Validation Plan of the Safety System for AHPR1000 Reactor

Cui Huaiming, Huang Daishun, Chen Wei, Ma Haifu, Yu Na, Lu Yili, Zhang Yu

2025, 46(S1): 82-87. doi: 10.13832/j.jnpe.2025.S1.0082

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With the batch construction of the independently developed third-generation pressurized water reactor nuclear power plant HPR1000 at home and abroad, China National Nuclear Corporation (CNNC) initiated the development of its follow-up reactor (AHPR1000) in 2019 to further enhance the safety, economy, advancement, operational reliability, environmental friendliness, and intelligence level of HPR1000. The design innovation of safety systems and facilities is a core part of this development. The safety systems and facilities of the AHPR1000 reactor are mainly used to ensure reactor safety and control/contain radioactive material releases under accident conditions. This paper proposes a design concept combining "passive and active" systems for the safety system configuration of AHPR1000. Starting from the top-level safety philosophy and design principles, it focuses on introducing the safety functions, safety configuration, accident response strategies, and experimental validation.

Physical and Thermal Analysis of the Heat Pipe Cooled Micro Nuclear Reactor Core Based on Thorium-Plutonium Mixed Fuel

Wang Feng, Sun Yuannan

2025, 46(S1): 88-94. doi: 10.13832/j.jnpe.2025.S1.0088

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Power distribution inhomogeneity is a key issue in thorium-plutonium fueled heat pipe cooled micro nuclear reactor. In order to optimize the power distribution in the core, the reflector materials such as BeO, Be, Graphite, MgO and Al₂O₃, which have better neutronics properties, are selected in this study, and the effects of these materials on the power distribution and other physical properties of the core are analyzed and compared. The results show that the use of MgO as the reflector material can effectively improve the axial and radial power distribution of the core and reduce the structural mass; at the same time, the use of MgO softens the neutron energy spectrum, improves the initial reactivity of the core, and ensures a 5-year core lifetime. Thermal analysis based on single channel model shows that the bottom temperature of the core is significantly improved with the MgO material, although the thorium-plutonium fuel has a lower thermal conductivity than the UO₂ fuel, resulting in a slightly higher overall axial temperature. However, the thorium-plutonium fuel operates consistently below its melting point in the heat pipe cooled microreactor core, thereby meeting the thermal safety requirements. This study can provide design references and theoretical support for the application of thorium-plutonium fuel in heat pipe cooled micro nuclear reactor.

Study on Modified Diffusion Layer Wall Condensation Model Considering Suction Effect

Zhu Zhizhou, Tong Lili, Cao Xuewu

2025, 46(S1): 95-102. doi: 10.13832/j.jnpe.2025.S1.0095

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Steam wall condensation is an important passive heat removal method under loss of coolant accident (LOCA). The accuracy of the wall condensation model directly affects the validity of the analysis results. Based on the computational fluid dynamics (CFD) code, a diffusion layer wall condensation model was constructed in this paper. The JERICHO experiment was selected to evaluate the model. The results show that the diffusion layer wall condensation model can accurately predict the steam wall condensation rate at low condensation rate conditions. However, the model underestimates the wall condensation rate as the steam condensation rate increases. To address this problem, the suction effect and the non-uniformity distribution of the mixed gas density along the wall-normal direction are considered. A correction relationship for the suction effect considering the influence of light gas is proposed, the condensation source term is modified, and a new diffusion layer wall condensation model is constructed. The prediction results of the modified model are verified based on the COPAIN experiment. The simulated heat flux density is in good agreement with the experimental data, and the relative error is within ±15%, which proves the accuracy of the modified diffusion layer wall condensation model.

Research on Performance Analysis Methods for Typical Lattice of Heat Pipe Reactors Based on the MOOSE Platform

Qi Min, Li Chenxi, He Yanan, Wang Yanpei, Yang Guangliang, Li Yuanming, Wang Haoyu

2025, 46(S1): 103-112. doi: 10.13832/j.jnpe.2025.S1.0103

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To address the challenge of characterizing the behavior of fuel assemblies at different radial positions in heat pipe reactors, this paper develops a typical fuel assembly code and a MultiApp code based on the MOOSE platform. The typical fuel assembly code enables detailed analysis of fuel assemblies at specific positions, while the MultiApp code is capable of performing full-core calculations. The combination of the two codes ensures consistency and accuracy in calculating the stress, strain, and temperature at different radial positions. The calculation results of the typical fuel assembly code under different boundary conditions indicate that the stress calculation results are overestimated under symmetric boundary conditions, while the stress results under free boundary conditions are more reasonable. The temperature calculations under combined boundaries and periodic boundaries are similar, but the stress results are different. By comparing the typical fuel assembly code with the full-core calculation code, it is found that the typical fuel assembly under free boundary conditions has better capability to represent the thermo-mechanical behavior of the full core under steady-state operation with a uniform power distribution.

Effect of Irradiation on Interfacial Bonding Properties of Cr-Zr Coating Cladding Materials

Xiao Wenxia, Xi Hang, Lu Chenyang, Lei Penghui, Wang Haidong, Zhang Haisheng, Wang Ziyi, Lei Yang

2025, 46(S1): 113-122. doi: 10.13832/j.jnpe.2025.S1.0113

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To study the effect of irradiation on the interfacial bonding properties of Cr-Zr coating, the room temperature Kr ion irradiation tests of Cr-Zr coating materials at 1, 5 DPa and 20 DPa were carried out in this paper. The microhardness and microstructure of Cr-Zr coating materials with different ion irradiation doses were studied by nanoindentation and TEM characterization technology. The results show that with the increase of irradiation dose, the hardness of the matrix, Cr-Zr interface and coating all increases, indicating irradiation hardening occurs to the materials; There are a large number of Kr bubbles introduced by Kr ion irradiation at the coating-matrix interface in the irradiated samples. The concentration of Kr bubbles increases with the increase of the irradiation dose, and finally the interface bonding strength decreases with the increase of the irradiation dose.

Research on Obtaining Fast Neutron Spectrum through Radiation Testing of Fuel Rods for Fast Reactor in Thermal Reactor

Wang Kaimin, Guo Yufei, Peng Xingjie, Sun Shouhua, Zhang Liang, Kang Changhu, Zheng Daji

2025, 46(S1): 123-130. doi: 10.13832/j.jnpe.2025.S1.0123

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Based on China’s High Flux Engineering Test Reactor (HFETR), research has been conducted on the neutron spectrum modification method and physical characteristics of fast reactor fuel rods. This study analyzed the effects of various neutron screen materials and multiple fuel pin irradiation devices on spectrum modification during fuel irradiation, as well as the changes in neutron screen characteristics and their influence on the reactor’s performance. Single-assembly calculations were performed using the MCNP code, and two materials (boron carbide and europium oxide) were selected from four candidate neutron screen absorber materials, while cadmium and hafnium were excluded due to their suboptimal spectrum modification results. After comparing absorbers of varying thickness, we found that a thickness of 0.3 mm or above achieved effective spectrum modification. When placed in the reactor’s central channel, device A exhibited a higher fuel rod linear power density and a relatively small negative reactivity, while device B displayed a lower fuel rod linear power density and a greater negative reactivity. Considering the impact of burnup on the absorbers under conditions accounting only for fuel and absorber material burnup, we determined that the lifetime of absorber in device A was less than 100 days, while the effective lifetime of boron carbide in device B ranged from 300 to 500 days, and europium oxide had an effective lifetime of 500 to 700 days. This paper preliminarily proposes a feasible spectrum modification scheme, demonstrating that the modification method can meet the requirements for conducting fast reactor fuel irradiation tests in a thermal-spectrum research reactor.

Design and Analysis of Integrated Heat Transfer Test Facility for Gas-Cooled Microreactor Core Components

Xue Yanfang, Wang Dingsheng, Sun Yanyu, Huang Zheng, Zhang Shuoting, Fang Jun, Han Shichao, Liu Guoming, Chen Qiaoyan

2025, 46(S1): 131-136. doi: 10.13832/j.jnpe.2025.S1.0131

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In response to the requirements for comprehensive heat transfer testing of gas-cooled microreactor core components, this study conducted detailed design and analysis of the test facility. The research encompassed the design principles of the test facility, determination of key parameters, structural design of the test section, and optimization of the startup process. Numerical simulation and pre-analysis methods were employed to systematically investigate the startup process of the test facility and the design scheme of the test section. The results indicate that the startup process control of the test facility is reasonable, and the structural design of the test section meets technical requirements, achieving design targets of 50℃ at the helium fan inlet, 489℃ at the core inlet, and 750℃ at the core outlet, while maintaining a stable helium pressure of 1.6 MPa at the test loop inlet. The comparison between experimental data and simulation results for graphite temperature demonstrates good consistency, validating the reliability of the computational model. The research results not only provide a feasible validation test platform for comprehensive heat transfer testing of gas-cooled microreactor core components, but also lay an important foundation for the subsequent development and verification of high-temperature helium equipment components.

Investigation on the Flow Characteristics and Sealing Performance of Dry Gas Seal at the Shaft End of Helium Compressor

Zhang Zhao, Du Qiuwan, Yuan Dewen, Qiu Zicheng, Zhang Cheng

2025, 46(S1): 137-144. doi: 10.13832/j.jnpe.2025.S1.0137

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The helium compressor is the core equipment of the Brayton cycle in high-temperature gas-cooled reactors, and the leakage suppression of helium working fluid at the shaft end is a key issue related to efficiency improvement. The utilization of dry gas seal with advantages such as no wear, low leakage, and non-contact is an effective approach. To investigate the flow characteristics and sealing performance of helium dry gas seal, this paper focuses on the typical spiral groove helium dry gas seal structure. Considering the influence of real gas effect, the numerical analysis is carried out to discuss the effects of structural parameters such as spiral angle, groove depth, and gas film thickness on the gas film pressure, open force and leakage. Furthermore, CO₂ and N₂ are selected to compare the flow characteristics and sealing performance with different working fluids under various inlet pressure and rotation speed conditions. The results indicate that the open force and leakage are positively correlated with the spiral angle, groove depth, and inlet pressure. As the gas film thickness increases, the open force decreases and the leakage increases. As the rotation speed increases, the open force increases and the leakage decreases for He while increases for N₂ and CO₂. Under the same operating conditions, the ranking of open force and leakage is He<N₂<CO₂. The findings of this study can provide vital reference for the design optimization of dry gas seals at the shaft end of helium compressors.

Study on the Microstructure and Nano-hardness Evolution of FeCrAl Alloy under Ion Irradiation

Pei Jingyuan, Chen Huan, Zhang Ruiqian

2025, 46(S1): 145-157. doi: 10.13832/j.jnpe.2025.S1.0145

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As an important candidate material for accident-resistant core cladding, it is necessary to deeply understand the modification mechanism of FeCrAl alloy under irradiation conditions. Au⁺ ions were used to irradiate the new Fe13Cr4.5Al alloy at room temperature and400°C. The phase structure, surface texture orientation, defects before and after irradiation, precipitated phase, amorphization and other microstructure and microhardness of the surface region of FeCrAl stainless steel before and after irradiation were systematically characterized. The correlation behavior and damage mechanism of irradiation defects, precipitated phase and hardening effect of FeCrAl stainless steel under ion irradiation were analyzed. With the increase of irradiation dose from 5 dpa to 20 dpa, the dislocation induced by irradiation has a "point-loop-line" evolution process. The density of 1/2<111> increases with the increase of irradiation dose, and the hardness of FeCrAl alloy also increases synchronously and reaches saturation after irradiation. The dislocation loop plays a leading role in hardening.

Design and Development of Reactor Coolant Pump Intelligent Monitoring and Prognosis System for Nuclear Power Plants

Xu Renyi, Wang Yan, Cui Huaiming, Kuang Chengxiao, Wu Kelin

2025, 46(S1): 158-165. doi: 10.13832/j.jnpe.2025.S1.0158

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In order to improve the intelligent operation and maintenance level of nuclear power plants, and effectively prevent and reduce equipment downtime, this paper designs and develops an intelligent monitoring and prognosis system for reactor coolant pumps (RCPs). The system integrates data acquisition and storage, online monitoring, fault diagnosis, trend prediction, fault treatment measures and prevention decision support. The verification results show that the system can track the running status of the RCP in real-time. Under fault conditions, the abnormal information of the RCP can be detected timely, and the fault mode is identified accurately. Then the O&M guidance is provided based on the current equipment status and parameter trend prediction results. Therefore, the system can track and identify the running state of reactor coolant pump in time to achieve the purpose of improving the condition monitoring capabilities and intelligent O&M level of nuclear power equipment.

Analysis of Factors Influencing the Accuracy of 3-D Flux Synthesis in Nuclear Reactor Primary Shielding

Hou Yunan, Zhang Bin

2025, 46(S1): 166-180. doi: 10.13832/j.jnpe.2025.S1.0166

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The discrete ordinates (S_N) method is one of the primary methods for calculating fast neutron flux in reactor pressure vessels (RPVs). The three-dimensional neutron flux synthesis method based on two-dimensional plus one-dimensional S_N calculations (referred to as the synthesis method) has higher computational efficiency compared with the direct three-dimensional discrete ordinates method (referred to as the three-dimensional calculation method). However, the approximations in the source and geometry processing can affect the accuracy of the synthesis method. In order to analyze the impact of source and geometric approximations on the synthesis method, a benchmark model suitable for the synthesis method was established. The relative error of fast neutron flux calculated by the synthesis method and the three-dimensional calculation of the benchmark model is used as a reference to analyze the influence of source and geometric factors on the synthesis method. In the analysis of source effects, non-uniform radial, axial, and azimuthal power distributions are introduced into the benchmark model to examine the changes in the relative error of fast neutron flux between synthetic calculations and three-dimensional calculations. In the analysis of geometric effects, the core structure of the benchmark model is altered to a square core and a stepped core, respectively, to sequentially analyze the changes in the relative error of fast neutron flux between synthetic and three-dimensional calculations. The results indicate that the maximum relative errors caused by the radial and axial power distributions are both within 1.5%, while the azimuthal power distribution leads to a relative error of 3.5% at the reactor cavity. The square and stepped core structures result in relative errors of 20% and 22%, respectively, in the cavity. In the computations for the typical pressurized water reactor HBR-2, the relative error between the synthesis method and the three-dimensional fast neutron flux calculations at the cavity was 11.65%. These findings demonstrate that the accuracy of the synthesis method for the reactor cavity region still requires improvement.

Fine Rod Power Calculation and Verification for the Entire Lifetime of a Small Lead-Bismuth Fast Reactor

Gao Jiehao, Du Xianan, Chen Wenjie, Zheng Youqi

2025, 46(S1): 181-191. doi: 10.13832/j.jnpe.2025.S1.0181

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Small lead-bismuth fast reactors, due to their compact design, impose higher requirements on reactor physics analysis codes. This paper explores the methodology of full-life cycle fine rod power calculations for the SVBR-100 small Lead-Bismuth fast reactor. The results show that for this type of reactor core, the two-step method code needs to consider the power distribution information in assembly (shape factor) obtained during the assembly homogenization process when performing fine rod power calculations. Additionally, the calculation of the shape factor must account for the influence of assembly layout and material information. Based on the above calculation method, this paper applies the SARAX from the Nuclear Engineering Computational Physics Laboratory (NECP) of Xi'an Jiaotong University to perform entire lifetime cycle fine rod power calculations for the SVBR-100 reactor core and compares the results with those from the Monte Carlo code. The calculation results demonstrate that SARAX achieves high accuracy in entire lifetime cycle fine rod power calculations for small lead-bismuth fast reactors. This work establishes a foundation for subsequent code applications in small lead-bismuth fast reactor core design and high-resolution calculations of multiphysics coupling.

Development and Validation of Assembly Calculation Module in 3D Neutron Transport Code for Molten Salt Reactors

Dai Ming, Cheng Maosong

2025, 46(S1): 192-199. doi: 10.13832/j.jnpe.2025.S1.0192

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The three-dimensional (3D) neutron transport code ThorMOC for molten salt reactors (MSRs) utilizes the non-uniform spectra modification method to provide few-group cross sections for full-core transport calculations, which relies on the multi-group effective macroscopic cross sections from assembly or super-assembly calculations. For MSR assembly or super-assembly with 3D complex shape resonance regions, an Embedded Self-Shielding Method (ESSM) based on the SHEM361 multi-group data library had been developed within ThorMOC by leveraging its existing quasi-3D method of characteristics (MOC) solver, which is enhanced by GPU parallelization and the coarse-mesh MOC-based synthetic acceleration (MSA) technique, thereby enabling MSR assembly calculations in ThorMOC. For a cylindrical channel molten salt reactor, numerical analysis is conducted on seven types of assemblies, including three 3D super-assemblies with upper and lower support plates and cavities. Compared to results from the continuous-energy Monte Carlo method, the maximum k_eff relative deviation is −110pcm (1pcm=10⁻⁵). The numerical results demonstrate that the implemented assembly calculation module is effective for the calculations of MSR assemblies with 3D complex shape resonance regions.

Software Verification for LOCUST-SIM Based on HPR1000 Verification Simulator

Tang Junming, Huang Zesong, Zhou Shuyong, Xu Caihong, Guo Hua, Zheng Wei, Gao Wei

2025, 46(S1): 200-206. doi: 10.13832/j.jnpe.2025.S1.0200

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LOCUST-SIM is an advanced thermo-hydraulic simulation code specifically developed by China General Nuclear Power Group (CGN) for PWR. In this paper, thread pool is used to optimize the calculation process of LOCUST-SIM, and the verification of the software is carried out on the HPR1000 verification simulator to evaluate its accuracy and reliability in the simulation of thermal hydraulic characteristics of nuclear power plants, so as to improve the computational efficiency of the software. The HPR1000 verification simulator was built by using the LOCUST-SIM to simulate the thermal hydraulic model and using the the simulation platform software GENUS with its integrated flow network analysis code and instrumentation & control simulation tools to simulate the process systems and control logic. The transient simulation results of three types of reactor accidents were compared and verified in depth with the results from internationally recognized thermal hydraulic simulation code RELAP5-3D. The verification results indicate that LOCUST-SIM can accurately predict the thermal-hydraulic characteristics while significantly improving computational speed, meeting the requirements for reactor thermal-hydraulic simulation applications.

Design of 40 kW Dual-drum Controlled Liquid Molten Salt Reactor in Catalogue: Neutronics and Dual-drum Worth Analysis

Zhuang Nailiang, Song Yongnian, Yin Zhengda, Zhao Hangbin

2025, 46(S1): 207-212. doi: 10.13832/j.jnpe.2025.S1.0207

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The heat source (power supply) of nuclear fission reactor is expected to be applied to deep space exploration, catalog scientific research station, interstellar navigation and other fields in the future due to its many advantages, such as no demand for solar orientation, small impact from space environment, large power range and large regulation range. Based on the fourth generation nuclear reactor technology—molten salt reactor, this paper proposes a conceptual design for a 40 kW space nuclear reactor utilizing liquid molten salt as nuclear fuel coupled with heat pipe cooling. A new reactivity control strategy combining control drum (for power regulation) and safety drum (for emergency shutdown) is proposed. The physical model of liquid molten salt space nuclear reactor is established, and the key physical characteristics of the core, such as neutron energy spectrum, neutron flux distribution, temperature effect and burnup depth, are obtained based on the Monte Carlo code MCNP and RMC analysis. The influence of control drum angle on reactivity and the reactivity control and core safety under partial failure of dual drums are further analyzed. The results show that the liquid molten salt space nuclear reactor designed in this paper can operate at full power for 10 years. The control drum arrangement can meet the core safety requirements under the failure of some control drums. This research can provide design reference for the control strategy of space liquid molten salt reactor and space reactor.

Study on Evolution Mechanism for Local Melting Accident of Dispersed Plate Fuel

Ding Wenjie, Huang Hongwen, Guo Haibing, Gao Jiao, Wang Shaohua, Ma Jimin

2025, 46(S1): 213-219. doi: 10.13832/j.jnpe.2025.S1.0213

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To understand the evolution process of plate fuel melting accident and prevent the continuous core fuel melting, the typical plate fuel research reactor JRR-3M was used as the research object, and the volume of fluid (VOF) method was coupled with enthalpy-porosity method to simulate the local melting and melt migration process of dispersed plate fuel after adjacent flow channels were blocked. The simulation results show that the evolution process of local melting accident is divided into four stages: temperature rise, melting, migration and solidification. Among them, the migration stage is extremely short and lasts less than 3 s, but it has a decisive influence on the accident expansion. In the migration stage, the melt migrates to the adjacent fuel plate in two ways: bottom gathering and middle droplet sputtering, and begins to solidify after being fully cooled by the adjacent fuel plate. After the high-temperature melt contacts the adjacent fuel plate, the temperature of the adjacent fuel plate cooling wall will rise rapidly to 500~600 K, far exceeding the boiling point of the coolant, which poses a burnout risk to the adjacent fuel plate. The established simulation method and the obtained simulation results can provide support for the safety analysis of plate fuel core melting accident.

CFD Sensitivity Study of Stirling High Temperature Components Based on Heat Pipe Heat Transfer

You Ersheng, Li Yiyi, Xing Dianchuan, Jiang Shunli, Wang Tianmi, Xu Jianjun

2025, 46(S1): 220-227. doi: 10.13832/j.jnpe.2025.S1.0220

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The heat pipe-Stirling coupling structure refers to the geometric and heat transfer interface between heat pipe bundles and the Stirling generator in the heat-pipe nuclear reactor system, responsible for transferring core heat from the heat pipes to the helium working fluid inside the Stirling engine. This study employs the computational fluid dynamics (CFD) method to numerically analyze the heat transfer process in the heat pipe-Stirling coupling structure, investigating the influence of various cold and hot boundary conditions on characteristic parameters such as effective heat transfer capacity, total temperature difference, and temperature distribution on the channel surface. The results show that the heat transfer capacity of helium side has a certain influence on the heat transfer process, and increasing the convective heat transfer coefficient or decreasing the helium temperature is beneficial to further improve the effective heat transfer. In contrast, the boundary conditions on the heat pipe side have a greater influence on the heat transfer process, which may cause a large temperature gradient at the initial position of the coupling structure and obviously increase the total heat transfer temperature difference, thus affecting the heat transfer safety of the heat pipe. Therefore, it is necessary to improve the heat transfer capability from heat pipes to Stirling high-temperature components while limiting the maximum heat flux density below 150 kW/m² to ensure thermal safety during heat pipe reactor system operation.

Numerical Study on Corrosion and Heat Transfer Coupling Characteristics of Heat Exchange Tube in LBE Environment

Wang Yifeng, Peng Tianji, Fan Xukai, Tian Wangsheng, Tang Yanze, Meng Haiyan

2025, 46(S1): 228-236. doi: 10.13832/j.jnpe.2025.S1.0228

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In order to study the phenomenon of oxidation corrosion in LBE heat exchange tube and the effect of oxide layer growth on heat transfer, this study simulates the corrosion and heat transfer process of LBE heat exchange tube in 9500 hours based on the oxidation corrosion model, mass transfer controlled corrosion model and oxide layer thermal resistance model by using the FLUENT in combination with the user-defined function (UDF). The results show that after 9500 hours of operation under baseline conditions, the average thickness of the magnetite and spinel layers reached 23.84 μm and 25.02 μm, respectively. Due to the additional thermal resistance introduced by the oxide layer growth, the average thermal resistance of the wall increased by 7.8%, and the wall temperature and outlet temperature of the heat exchange tube increased by 0.26 K and 0.2 K, respectively. The lower the inlet temperature, the smaller the thickness of the oxide layer is. However, the oxide layer thickness gradually increases with time, indicating that the oxidation layer growth process plays a dominant role compared to the removal process. The lower the inlet oxygen concentration, the smaller the thickness of the oxide layer is. When the oxygen concentration is reduced to 10⁻⁷ wt%, the magnetite layer at the inlet of the heat exchange tube appears local dissolution, and the scope of dissolution gradually expands. The spinel layer, on the other hand, continues to grow after exposure to LBE due to low removal rate and provides the main protection for the structural material.

Study on the Internal Flow Distribution Characteristics of Plate-Type Fuel Elements in THFR

Huang Yuan, Lyu Meng, Xie Heng, Shi Lei

2025, 46(S1): 237-241. doi: 10.13832/j.jnpe.2025.S1.0237

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The distribution of area averaged coolant velocities across various coolant channels within the innermost fuel element of the Tsinghua High Flux Reactor (THFR) was analyzed using the FLUENT 2022 R2 under different inlet velocity conditions. Under more conservative assumptions, for uniform velocity inlets, the maximum relative deviation of area-averaged velocities across different coolant channels from the overall area-averaged velocity was less than 0.6%; for artificially constructed non-uniform velocity inlets, the maximum relative deviation of area-averaged velocities across different coolant channels from the overall area-averaged velocity was less than 6%, providing a basis for subsequent reactor design. Non-uniform velocity inlets resulted in significant non-uniformity in the static pressure at the coolant channel inlets. In contrast, the static pressure distribution at the coolant channel outlets was more uniform, with isobaric planes present. The inconsistency in pressure drop led to non-uniform coolant velocity distributions. Therefore, in subsequent reactor structural design, it is crucial to ensure that the coolant at the core inlet is sufficiently mixed.

Development and Verification of Doppler Broadening Module in Nuclear Data Processing Code

Guo Xin, Xu Ning, Hao Chen, Yin Wen, Wang Yizhen

2025, 46(S1): 242-249. doi: 10.13832/j.jnpe.2025.S1.0242

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In order to meet the numerical simulation of various operating conditions in nuclear reactors, it is necessary to provide cross-section data at different temperature. However, the cross sections provided in the evaluated nuclear data library are the data at 0 K. Therefore, in order to meet the needs of core physics numerical simulation calculations, it is necessary to perform Doppler broadening processing based on the evaluated nuclear data in the evaluated nuclear data library to obtain continuous energy pointwise cross sections at different temperatures. Through the Kernel Broadening accurate Doppler broadening calculation method, the theoretical derivation and code development of the Doppler broadening calculation method have been completed. Based on the CENDL-3.2 evaluated nuclear data library, the broadening cross sections at different temperatures are calculated based on the doppler_broad module developed in this paper and the BROADR module in the NJOY2016, respectively, and a computational analysis is performed to verify the rationality of some convergence parameters in the Doppler broadening calculation process. The numerical results indicate that for the four temperature points of 293.6 K, 600 K, 900 K and 10⁸ K, the calculation results in this paper are in good agreement with the NJOY2016 calculation results. The value of the NMAX parameter, which determines the limit of adjacent broadening energy points, has a significant impact on the broadening cross sections. The maximum relative deviation of the broadening cross sections for different values is 1.091%.

Development and Preliminary Verification of OpenMC-PARCS Two-step Criticality and Burnup Calculation Model for Fast Reactors

Hu Henglin, Zhang Guangchun, Xiao Peng, Xia Bangyang, Wang Lianjie

2025, 46(S1): 250-259. doi: 10.13832/j.jnpe.2025.S1.0250

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Fast reactors cannot directly use PWR calculation models for neutronics analysis due to their hard spectra and complex resonance phenomena. Monte Carlo (MC) method utilizes continuous-energy neutron cross-sections, which can accurately simulate resonance interference phenomena in fast reactors, yielding highly precise homogenized few-group cross-sections. This paper, based on MC method and the Triangle-based Polynomial Expansion Nodal (TPEN) method, investigates an OpenMC-PARCS two-step method of criticality and burnup calculation for fast reactors. Based on the OpenMC one-step method calculation results, a preliminary validation of the assumed constant microscopic cross-section burnup calculation scheme is conducted using the sodium-cooled fast reactor benchmark problem MET-1000. In the initial steady-state calculation, the deviation of the core effective multiplication factor (k_eff) using the OpenMC-PARCS two-step method is −104pcm (1pcm=10⁻⁵), and the deviation in the radial power distribution is no greater than 1%. During burnup calculations, the maximum deviation of the core k_eff from the reference solution is 591.2pcm, while most major nuclide number density deviation is no greater than 1%. The preliminary validation results indicate that the OpenMC-PARCS two-step method model can be used for large metallic fast reactor core design and fuel management.

Theoretical and Numerical Analysis of Beryllium Reflector Poisoning in High-Flux Research Reactor

Li Kaiwen, Luo Hao, Liu Zhihong, She Ding, Zhao Jing

2025, 46(S1): 260-268. doi: 10.13832/j.jnpe.2025.S1.0260

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Beryllium is widely used as the reflector in high flux research reactors to improve neutron economy. To investigate the accumulation characteristics and laws of neutron poisons ⁶Li and ³He during the transmutation process in the beryllium reflector—rather than merely evaluating specific test cases—this study analyzes and solves the equations of nuclide transmutation in the beryllium reflector, obtains the relevant laws of the accumulation process of each nuclide, and evaluates the negative reactivity introduced by the beryllium reflector poisoning theoretically. It is concluded that the equilibrium concentration of ⁶Li is independent of the neutron flux level, and the upper limit of ³He accumulation rate is independent of the neutron flux level. By using RMC to calculate the beryllium reflector of the Tsinghua ‌High ‌Flux ‌Reactor (THFR) and comparing with theoretical predictions, the results are in good agreement, which validates the correctness of the theoretical analysis. The relevant conclusions can save the resource consumption of long and multi-step burnup calculations, and only a few critical calculations are needed to obtain the value of the negative reactivity, thus efficiently and accurately providing important basis for the design and replacement frequency of reflector, residual reactivity design, etc. of high flux research reactors.

Research on Multi-Group Constant with Discrete Angle and SPH Method Based on RMC

Li Yaodong, Yu Ganglin, Wang Kan

2025, 46(S1): 269-275. doi: 10.13832/j.jnpe.2025.S1.0269

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In this paper, a Monte Carlo group constant calculation method with discrete scattering angle is proposed, and the group constants are calculated and verified by using rod-type fuel assembly. First, a multi-group constant library based on fuel assembly or pin is calculated using the Reactor Monte Carlo (RMC) code, and then the group constant library is used for the neutron transport calculation of the full core. In the generation process of multi-group constant library, the neutron collision behavior is tracked based on the neutron history, which can accurately express the anisotropic scattering of neutrons. Since the group constant library is generated under real full-core conditions, no approximations are theoretically introduced during the transport calculation. For equivalent homogenization, an improved Superhomogenization (SPH) method is adopted. The above research shows that compared with the traditional Legendre scattering matrix, the method proposed in this paper avoids the generation of negative cross section; Compared with the continuous energy Monte Carlo results, the assembly calculation error is less than 70pcm (pcm=10⁻⁵); It is more accurate to calculate the group constant under real full-core conditions, with flexible geometric adaptability and excellent universality.

Processing and Verification of Shielding Library Applicable for Fast Reactor

Liu Fan, Cai Li, Yang Junwu, Lu Haoliang

2025, 46(S1): 276-281. doi: 10.13832/j.jnpe.2025.S1.0276

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In the shielding calculation of PWR, BUGLE96 or its predecessor library is generally used, which has been widely verified and applied in engineering. However, the library is made for typical PWR and BWR by considering their shielding structure and physical properties, thus is not applicable to fast reactors. Therefore, it is necessary to develop a suitable shielding library according to the fast reactor's physical properties for its shielding calculation. In this paper, the coupled 199-group neutron and 42-group photon fine-group library is made according to the nuclide types, conditions and energy spectrum characteristics of fast reactors by using the NJOY code. The fine-group library contains 65 elements and 202 nuclides commonly used in fast reactors, the cross-section data has 7 temperature points, the Legendre expansion order of the scattering cross-section is P8, and the typical fast reactor energy spectrum is used as the weight spectrum. The fine-group library is then collapsed into a 47-group neutron and 20-group photon broad-group library. Finally, the shielding library is verified by the JANUS-I fast-spectrum benchmark. The results of radial reaction rate show that for the ³²S(n,p)³²P detector, the calculated values are within ±15% error to the experimental values, and for the ¹⁰³Rh(n,n')¹⁰³Rh^m detector, the error is within ±10%. All errors can be covered by the experimental measurement deviations. The results of axial relative reaction rate show that for the ³²S(n,p)³²P detector, the calculated values are within ±10% error to the experimental values, and for the ¹⁰³Rh(n,n')¹⁰³Rh^m detector, the error is within ±15%. The error meets the requirements of engineering calculation errors. Therefore, the shielding cross-section library processed in this paper is applicable to the shielding calculation of fast reactors and can be subsequently applied to the shielding design of advanced fast reactors.

Research on Quadratic Depletion Method for Reaction Rate Based on NECP-X

Liu Runze, Liu Zhouyu

2025, 46(S1): 282-287. doi: 10.13832/j.jnpe.2025.S1.0282

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In depletion calculations, especially for gadolinium-bearing fuels, each burnup step often requires two flux calculations for predictor and corrector. However, in high-fidelity calculations, performing two physical calculations significantly reduces computational efficiency, leading to excessively high time costs in fuel cycle calculations. In this paper, quadratic depletion (QD) method is adopted in high-fidelity code NECP-X. The flux calculation in predictor is skipped, and a post-correction method is adopted to correct the number densities of seven gadolinium isotopes. In the correction step, quadratic interpolation is used for the reaction rates of gadolinium isotopes to improve the accuracy of depletion calculations of gadolinium-bearing fuels. Comparative depletion calculations are performed for both single-assembly and multi-assembly Gd-bearing fuel cases using traditional predictor-corrector method and quadratic depletion method. The results show that the quadratic depletion method cannot only improve the accuracy by at least a factor of 2, but also increase the efficiency by approximately 30%. Therefore, the proposed QD method can be well applied in Gd-bearing fuel depletion calculations.

Physics-Informed Neural Network Methods for Solving Eigenvalue Problems in Neutron Diffusion Equations

Xiao Yong, Zhou Xiafeng

2025, 46(S1): 288-295. doi: 10.13832/j.jnpe.2025.S1.0288

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To promote the practical application of Physics-Informed Neural Networks (PINNs) in core physics calculations and to achieve a deep integration of deep learning methods with nuclear physics models, thereby enhancing their potential in complex physical systems, this paper proposes a multi-group neutron diffusion eigenvalue neural network model applicable to various material arrangements. In this model, an adaptive weighting strategy is designed based on the characteristic sampling of the material region, and the flux normalization is not required. By solving the single-group multi-material case and the two-group BIBLIS benchmark problem, the calculation results show that the absolute errors of the k_eff for both cases are 529.6 pcm (1pcm=10⁻⁵) and 112.5 pcm, respectively, and the relative error of each assembly's power is less than 5%. These results preliminarily verify the accuracy and effectiveness of this model. This study, through the combination of physical constraints and neural network models, provides a new technical pathway for the numerical simulation of complex reactor cores and is expected to promote the engineering application of deep learning methods in reactor physics design, safety analysis, and multi-physics coupled calculations.

EMC Design and Verification for Class 1E Valve Position Signal Processing Cabinet

Luo Shihong, Li Junhuai, Li Yurui, He Hongyang, Zheng Wuyuan, Zhang Donglin

2025, 46(S1): 296-300. doi: 10.13832/j.jnpe.2025.S1.0296

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In order to solve the electromagnetic compatibility (EMC) issues in the current 1E valve position signal processing cabinet of nuclear power plants and ensure the operational reliability of in-service nuclear power units, this research conducts EMC design work from the aspects of power supply, input, output, internal reference ground, and structural shielding of 1E valve position signal processing cabinet. EMC qualification tests are conducted to verify whether the developed Class 1E valve position signal processing cabinet meets the technical requirements. The test results show that the EMC design of the 1E valve position signal processing cabinet achieves the expected goals. The anti-electromagnetic interference ability of key technical equipment is effectively improved, laying the technical foundation for the domestic substitution of 1E valve position signal processing cabinets in nuclear power plants.

Current Issue 2025 Vol. 46, No. S1

Current Issue
2025 Vol. 46, No. S1