核动力工程

Key Technologies of Medical Isotope Test Reactor

Li Qing, Zhang Jinsong, Zhang Yulong, Nie Huagang, Chen Yunming, Jiao Baoliang

2025, 46(5): 1-11. doi: 10.13832/j.jnpe.2025.06.0260

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Abstract:
The construction of a solution-type medical isotope test reactor for the production of isotopes such as ⁹⁹Mo and ¹³¹I is one of the important initiatives to achieve self-sufficient and controllable supply in China's medical isotope market. This article briefly introduces the application status, production principles, and production methods of medical isotopes, as well as the development overview of homogeneous solution-type reactors both domestically and internationally. It systematically elaborates on the system composition and design of the medical isotope test reactor, including the reactor and its major systems, isotope extraction process systems, and supporting systems. Furthermore, it provides a detailed explanation of the main technical issues, such as reactivity stability, radiation protection design, prevention of fuel solution precipitation, corrosion resistance of structural materials, critical safety of fuel solutions, isotope extraction processes, uranium recovery technology, fuel purification technology, and radioactive waste gas treatment technology.

Research on the Effect of Fuel Spatial Separation on Helium Production Behavior in Lithium-Cooled Fast Reactors

Wang Yue, Wei Bin, Wang Jincheng, Zhang Wenchao, Sun Jianchuang, Cai Weihua

2025, 46(5): 12-21. doi: 10.13832/j.jnpe.2024.090020

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Liquid lithium is widely utilized as a metal coolant in space fast reactors due to its smaller neutron absorption cross-section, lower density, and superior thermophysical properties, resulting in reduced core mass and enhanced heat transfer efficiency. However, under neutron irradiation, lithium reacts with neutrons to produce helium gas. The accumulation of helium increases thermal resistance in affected regions and reduces the heat transfer efficiency of system equipment. In this study, the Monte Carlo code OpenMC is employed to conduct helium production calculations for three lithium-cooled space fast reactor design models featuring different fuel space separations. The aim is to analyze the impact of these separations on the helium production behavior within the reactor core. The paper also calculates the total helium yield of the core, analyzes the helium production capacity of various nuclear reactions between liquid lithium and neutrons, and examines helium production under different ⁷Li enrichments, and the accurate calculation of helium production in the core is realized. Focusing on changes in helium production as a function of core burnup, the results indicate that larger fuel space separations result in higher helium production. Additionally, increasing ⁷Li enrichment significantly reduces helium production, with a maximum 68.76% reduction observed in helium yield when enrichment is elevated from 95% to 99%. The paper provides valuable insights for optimizing lithium-cooled space fast reactors.

Influence of Relative Position of Flow Line Before and After Radial Guide Vane on Hydraulic Performance of Reactor Coolant Pump

Li Xueqin, Wang Xiuyong, Liu Wuqing, Yang Congxin, Guo Yanlei

2025, 46(5): 22-29. doi: 10.13832/j.jnpe.2024.090012

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In order to investigate the influence of the relative positions of the radial guide vane flow lines at the front and rear cover plates on the hydraulic performance of reactor coolant pumps, non-stationary numerical calculations are carried out for five groups of models with different circumferential relative positions of the front and rear flow lines of the guide vanes, using a global structured mesh and the RNG k–ε turbulence model. The influence of the relative position changes of the front and rear flow lines of the guide vanes on the external characteristicsand the pressure pulsation characteristicsof the Reactor Coolant Pump is analyzed. The results show that when the position of the front flow line is kept unchanged while the rear flow line is deflected circumferentially in the opposite direction of the impeller rotation (i.e., the front flow line of the guide vane is positioned ahead of the rear flow line), the flow field structure in the pressurized water chamber, especially in the discharge section, is improved, and the hydraulic loss inside the guide vane and the pressurized water chamber is reduced. Compared with the original model, the head of the reactor coolant pump increases by 0.60%, the efficiency improves by 0.66%, and the average amplitude of pressure pulsations at the dominant frequency decreases by 23.08%. Thus, the hydraulic performance of the reactor coolant pump is enhanced while its vibration performance is significantly optimized.

Modeling and Applicability Analysis of Nitrogen Pressurization System Based on RELAP5

Jiang Meng, Liu Mengjuan, Zhao Meng, Huang Li, Yang Yanhua, Zhao Jiangang

2025, 46(5): 30-36. doi: 10.13832/j.jnpe.2024.090023

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In order to explore the operational characteristics of the nitrogen pressurization system, this study employed RELAP5 for computational simulation and conducted an applicability analysis of the code's nitrogen pressurization model. The results indicate that RELAP5 exhibits the following errors when simulating the nitrogen pressurization system: normalized pressurizer pressure and water level are overestimated; excessive temperature gradient in the fluid region; and excessive temperature in the gas region. Analysis shows that the reasons for the calculation errors in RELAP5 may include: the energy equation does not contain the axial heat diffusion term; the high-concentration non-condensable gas partial pressure exceeds the model's applicable range; limited applicability of wall heat transfer models; and missing material properties. Based on existing research findings, the model modification involves incorporating a thermal diffusion characterization term and modifying the nitrogen-water heat exchange model, with emphasis placed on addressing the nitrogen-water heat exchange issue which has a greater impact on the results.

Study on the Applicability of Transition Boiling Heat Transfer Models Based on LOCUST Code

Liu Huannan, Yuan Hongsheng, Ju Zhongyun, Xu Caihong, He Dongyu, Li Jinggang, Wang Ting

2025, 46(5): 37-45. doi: 10.13832/j.jnpe.2024.090046

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Transition boiling is a very important thermo-hydraulic phenomenon in the core analysis of pressurized water reactor (PWR), especially in the accident analysis. The accurate simulation of this phenomenon can improve the accuracy of core wall temperature prediction by software. Considering the small range of the transition boiling area and the large variation of the parameters, and the limitation of the test method, there is a lack of recognized and suitable transition boiling model. In order to evaluate the effect of different transition boiling models, six transition boiling models commonly used in the world are compared in this paper. Based on LOCUST, a thermo-hydraulic system analysis software independently developed by China General Nuclear Power Corporation, code development of the six transition boiling models and comparative study of software calculation results and test data were realized. The results show that the Chen relation and the Bjornard-Griffith relation are the best in simulating transition boiling phenomenon, showing good agreement with the experiment data. The results of this study lay the foundation for further exploring the differences and computational effects of various transition boiling heat transfer models, and provide reference for the selection of transition boiling heat transfer models in the development of thermo-hydraulic system analysis software.

Study on the Influence of Filling Ratio on the Startup and Heat Transfer Performance of the Separate Heat Pipe with Plate Evaporator

Chang Zhuang, Bai Bofeng, Han Xu, Wen Boyao, Luo Zhengyuan, Zhang Siliang

2025, 46(5): 46-55. doi: 10.13832/j.jnpe.2024.090044

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Separated heat pipe can remove the decay heat generated by spent fuel effectively, demonstrating significant application value in the field of passive residual heat removal in nuclear plants. A new type of Separated heat pipe, featuring a capsule-type plate for the evaporator and a serpentine finned aluminum flat tube for the condenser, was proposed by our research group, and the influence law of different filling ratios on the start-up and heat transfer performance of Separated heat pipe was experimentally analyzed. The research found that, for the separated heat pipe with plate type evaporator in this research, the differential pressure across the evaporator reflects the start-up characteristics of Separated heat pipe more accurately compared to the superheat degree at the evaporator outlet. The time required for the evaporator pressure difference to stabilize first increases and then decreases with increasing filling ratio, indicating that the filling ratio significantly affects the startup behavior of such heat pipes. The maximum heat transfer capacity of Separated heat pipe can reach 183.6 kW in stable operation, and the corresponding filling ratio is 51%. The two-phase heat exchange area in evaporator and the temperature difference between working medium and environment in condenser both reaches maximum values at this point. Furthermore, the curve of heat transfer capacity versus filling ratio shows an optimal range of 51% to 82%, within which Separated heat pipe maintains high heat transfer performance. The thermal resistance of condenser dominates the total thermal resistance of Separated heat pipe. These research results provide references for optimization design and engineering application of Separated heat pipe.

Analysis of Mixing Performance of Spacer Grid Based on 5×5 Rod Bundle Subchannels

Li Jingyi, Su Qianhua, Du Xin, Zhang Ge, Liu Zhimin, Xie Yuanlai

2025, 46(5): 56-65. doi: 10.13832/j.jnpe.2024.090065

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To analyze the influence of the spacer grid on the thermal and hydraulic performance of the fuel assembly, this study conducts numerical simulations on the 5 × 5 full length fuel assembly rod bundle subchannels. Through experimental verification and combination with the development of fluid temperature field, as well as the comprehensive consideration of key parameters such as field synergy angle, mixing factor, and temperature deviation between subchannels, the influence of temperature field development and the bending angle of the mixing vane on the thermal and hydraulic performance of the rod bundle subchannels was analyzed. The results show that the fluid temperature field downstream of the spacer grid goes through stages of development, full development, transition, and attenuation in sequence, with significant heat transfer enhancement observed during the full development and transition stages. Considering the positive impact of the bending angle on thermal-hydraulic performance and its implications for fuel safety, a bending angle in the range of 31°~36° is recommended. This study can provide a theoretical reference for the research and optimization of the fuel assembly.

Numerical Simulation of Novel Core Catcher by Using Particle Method

Lan Yicong, Zhang Yapei, Wang Xingyu, Li Yang, Gong Houjun, Tian Wenxi, Su Guanghui, Qiu Suizheng

2025, 46(5): 66-75. doi: 10.13832/j.jnpe.2024.09.0011

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In order to establish a prediction method for the entire process of a novel core catcher following lower head failure of the pressure vessel, and to provide an additional approach for the optimization analysis of the novel core catcher, Incompressible Smooth Particle Hydrodynamics (ISPH) method was used to predict the behavior of the core melt in the novel core catcher. Through numerical simulations, a series of behaviors under severe accident conditions in nuclear reactors are investigated, including: jet erosion of core melt, accumulation of melt on the stainless steel plate, melting of the stainless steel plate due to the corium's sensible heat and decay heat, flow characteristics of melt in concrete channels, molten corium-concrete interaction, and coupled heat transfer inside and outside the novel core catcher. The predicted heat fluxes and ablation rates show good agreement with experimental data and results from integrated severe accident analysis codes. Thus, this study holds significant engineering application value for optimizing the design of novel core catchers in nuclear reactors.

Topology Optimization Method for Bionic Fins Based on a Natural Convection Substitution Model

Zhang Chunbo, Hu Zongwen, Meng Zhaoming, Dong Chuanchang, He Gening, Li Donghui

2025, 46(5): 76-83. doi: 10.13832/j.jnpe.2024.09.0016

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Natural convection fin heat sinks are widely used in the nuclear field. Improving their convective heat transfer capacity is a critical measure to ensure reactor safety and enhance energy efficiency. To enhance the heat dissipation performance of heat sinks, this study conducted a topology optimization design of traditional fin heat sinks based on the engineering conditions of the fin heat sink in the Reactor Cavity Cooling System (RCCS), employing a natural convection substitution model and the optimality criterion method. The shape of the optimized structure was analyzed using the concept of fractal bionics. In this study, the optimal fin thickness was first found by numerical simulation, and then the topological optimization design of the heat sink structure with the optimal fin thickness was carried out under different volume constraints, and the heat transfer performance of the optimized structure under different volume constraints was analyzed by numerical simulation. Considering both volume and heat transfer performance, an optimized structure with a volume constraint of 0.135 was selected for further analysis. After topological optimization, the heat sink structure was vein-like, and the fins extending from the base gradually became thinner and fractal features appeared at the end of the fins. Moreover, small fins grew on the surface of the fins to further improve the convective heat transfer. The numerical simulation shows that the vein-shaped heat sink is more favorable to heat conduction and convective heat transfer than the traditional heat sinks. Finally, the fractal characteristics of the optimized fins are analyzed by using the box counting method, and the similarity between them and the veins of plants is verified, which shows that the topology optimization structure is close to the natural optimal solution.

Study on Heat Transfer Characteristics and Nuclear Thermal Coupling Effect of HFETR High Temperature and High Pressure Test Device

Liu Chang, Xia Yi, Song Jiyang, Kang Changhu, Liu Xiaosong, Qiu Liqing, He Yuhao, Liu Runqi, Guo Yufei, Zou Yutong

2025, 46(5): 84-91. doi: 10.13832/j.jnpe.2024.09.0026

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To meet the design and development requirements of new PWR fuel assemblies in the high-flux engineering test reactor (HFETR) test loop, this study aimed to address the interference of complex heat transfer phenomena caused by the presence of the downcomer and riser in the active zone coolant channels on the nuclear power measurement. A full-scale three-dimensional model of the high-temperature and high-pressure test device was constructed based on the code STAR-CCM+, and the coolant flow and heat transfer characteristics were systematically analyzed. It was found that the heat transfer power of the diversion pipe and the downcomer of the irradiation device accounted for approximately 5% of the measured thermal power of the test assembly. Meanwhile, an external interface between the two codes Monte Carlo (MC) and computational fluid dynamics (CFD) was developed using the Python language to achieve rapid data interaction and coupled calculation of nuclear and thermal data. Ultimately, the deviation between the calculated and measured thermal power values before and after the nuclear-thermal coupling was reduced by 1.5%. The research results verified the effectiveness of the CFD model and the nuclear-thermal coupling method under complex structures, providing a high-precision thermal-hydraulic analysis method for the irradiation test of new PWR fuel assemblies.

Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool

Mou Zhuoya, Zhu Longxiang, Wang Di, Ouyang Yong, Zhang Hong, Wan Lingfeng, Zhang Luteng, Tang Simiao, Pan Liangming

2025, 46(5): 92-100. doi: 10.13832/j.jnpe.2024.09.0024

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When the Steam Generator Tube Rupture (SGTR) occurs in lead-cooled fast reactors, the gas-liquid metal two-phase flow in the lead pool of primary circuit poses a severe threat to the safe operation of reactors. Understanding the flow characteristics of bubbles in the liquid metal pool and developing a drift flux model applicable to liquid metal two-phase flow are of great significance for improving the accuracy of reactor core safety predictions. An experimental study on the local distribution characteristics of phase parameters in a Woods alloy liquid pool was conducted using the double-sensor conductivity probe measurement technique. The experimental results show that the radial distribution of phase parameters such as void fraction tends to develop towards a core-peaked profile as the superficial gas velocity increases. Based on two-phase flow experimental data, a comparative analysis was conducted to evaluate the applicability of existing gas-water two-phase flow drift flux models in predicting the void fraction of liquid metal two-phase flow. It was found that only a few models showed good agreement with the experimental void fraction data. Combined with the experimental data and comparison results, the key parameters in the distribution parameter and drift velocity were modified based on the existing model. The relative calculation error of the newly developed model is within ±15%, and the absolute relative error is 5.35%.

Experimental Research on Critical Heat Flux in Rectangular Channel with Femtosecond Laser-Modified Surface

Cao Mingze, Yan Xiao, Xing Dianchuan, Zhang Yan, Xu Jianjun, Wang Yanlin, Xie Tianzhou

2025, 46(5): 101-108. doi: 10.13832/j.jnpe.2024.10.0032

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As a type of laser modification method, femtosecond laser surface modification can fabricate microstructures on surfaces of of stainless steel, zironium alloy and nickel alloy, creating specific morphological features such as honeycomb patterns, grooves, and hump structures. These features significantly influence the heat and mass transfer efficiency as well as the thermal performance limits of heating surfaces. In order to verify the enhancement effect of femtosecond laser-modified surface on critical heat flux (CHF) under pressurized flow conditions, CHF experiments were conducted at pressures ranging from 2 to 4 MPa. The experimental results indicate that both the femtosecond laser-modified surfaces and conventional surfaces exhibit an increase in CHF with higher inlet mass flux and inlet subcooling, while CHF decreases with an increase in critical steam quality. The modified surfaces demonstrate technical potential for enhancing CHF, with hump and honeycomb structures showing the most significant improvement. Although all three surface types are hydrophilic, the degree of CHF enhancement varies. This difference is attributed to the varying capabilities of different microstructures in facilitating liquid cooling of overheated surfaces.

Prediction of Cold Dropping Behavior of Control Rod Based on Fluent

Liu Jiqiu, He Xiaojun, Dai Jigao, Wu Qi, Hao Sijia

2025, 46(5): 109-114. doi: 10.13832/j.jnpe.2024.10.0039

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Focusing on a new type of fuel control rod assembly, this study investigates its drop behavior during reactor emergency shutdown. Using the software Fluent for simulation, a three-dimensional model of a single control rod was established. By adopting the dynamic mesh technique and appropriate computational parameters, the variation curves of pressure inside the guide tube and control rod velocity over time during the dropping process were obtained. Through comparative analysis with experimental results from the literature, the rationality of the research method was verified. Finally, this method was applied to calculate the drop behavior of the new fuel control rod, and the influence of different rod diameters was analyzed. The results show that the drop time of the new fuel control rod meets the acceptance criteria for drop time in current mainstream pressurized water reactors, and the drop time increases as the rod diameter grows.

Thermal-Hydraulic Performance Analysis of Tube Bundles in Economizer-type Steam Generators

Qiu Guihui, Duan Yuangang, Ran Xiaobing, Niu Wenhua, Su Xinrong

2025, 46(5): 115-123. doi: 10.13832/j.jnpe.2024.10.0043

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This paper presents a coupled thermal-hydraulic analysis of flow and heat transfer characteristics in the tube bundle region of economizer-type steam generators, specifically examining the partition plate height optimization. Based on HPR1000 standard parameters, the research examines the influence of the economizer on heat transfer performance and saturation pressure in the tube bundle region. A three-dimensional two-phase flow numerical model was developed using the porous media approach and drift-flux theory to analyze flow and heat transfer. The study reveals the quantitative impact of partition plate height on saturation pressure (with a maximum increase of 0.22 MPa) and heat transfer area (reducing approximately 1000 m²). Numerical analysis demonstrates that when the partition plate height is 5.8 m, a cold-side circulation ratio of 1.15 is obtained, and the outlet of the subcooled boiling zone in the economizer shows a high void fraction distribution, indicating that the bulk fluid temperature has reached saturation. This condition achieves optimal compatibility with the hot-side saturated boiling region. The findings provide a design basis for improving the thermal efficiency of nuclear power units.

Reliability-based Design Optimization of Thermal-Hydraulics Parameters for a 5×5 Single-Span Fuel Assembly Using CFD

Zhang Zhe, Wu Tianhao, Jiang Chao, Jin Desheng, Wang Juntao

2025, 46(5): 124-131. doi: 10.13832/j.jnpe.2024.10.0071

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To investigate the impact of thermo-hydraulic uncertainty parameters on the safety and economic performance of the fuel assembly under operational conditions, computational fluid dynamics (CFD) simulations were conducted on a 5×5 single-span fuel assembly. Additionally, a sample transfer learning method was developed for reliability analysis, enabling efficient optimization of the fuel assembly's reliability design. The results show that, with a constant total surface heat flux on the fuel rods, optimizing the inlet flow velocity and the heat flux distribution reduces the maximum surface temperature of the fuel rods by 2.6℃. Furthermore, with 99% confidence, the coolant temperature rise exceeds 2℃, and the pressure drop remains below 4400 Pa. This optimization improves thermo-hydraulic safety while maintaining economic efficiency, offering a novel approach for fuel assembly design optimization.

Study on the Sensitivity and Prevention Measures of Ductility Dip Cracking in Stainless Steel Overlay Weld

Yang Xingwang, Jiang Baiwen, Liu Gang, Shi Chunfeng, Xu Xinzhu

2025, 46(5): 132-138. doi: 10.13832/j.jnpe.2024.090032

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During the in-service inspection of the bimetallic weld in the low-pressure safety injection system pipeline of a nuclear power unit, ductility dip cracking (DDC) was identified on the overlay weld surface of ЭА-395/9 welding material. Through research on the hot cracking sensitivity of the ЭА-395/9 welding material, analysis of welding stress, and experimental study of the welding process, it was determined that when the overlay welding heat input is ≥ 16.3 kJ/cm, only 5% strain is sufficient to initiate cracking and form DDC. When the welding heat input of ЭА-395/9 welding increases, the columnar crystals become coarse and chain like precipitates form at the grain boundaries, resulting in high welding stress and providing metallurgical and mechanical conditions for the occurrence of DDC. By using a welding current of 120~130A, reducing the arc lateral swing amplitude, and controlling the interlayer temperature below 100℃, it is possible to effectively prevent the formation of DDC in the overlayweld of ЭА-395/9 welding material.

Optimization Design of Rod Drawing Wear Process Parameters Based on Multi-objective Optimization Algorithm

He Fuchun, Fu Chunming, Zong Benyang, Chen Kai, Tang Dewen, Zhang Zhe

2025, 46(5): 139-147. doi: 10.13832/j.jnpe.2024.090035

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In the process of loading fuel rods into the fuel assembly, wear on the surface of the fuel rods is an unavoidable consequence. Over prolonged reactor operation, such wear can potentially lead to damage to the fuel cladding, resulting in the leakage of nuclear fuel into the primary circuit and thus posing a significant threat to reactor safety. To address the issue of fuel rod wear, this study performs a comprehensive analysis and optimization from two perspectives: the fuel rod loading speed and the geometric cross-sectional width of the rigidly convex contact surface. The research establishes a finite element analysis model based on Archard wear theory. To enhance computational efficiency, a Kriging model is developed from the analysis results. Additionally, an improved Multi-Objective Dwarf Mongoose Optimization Algorithm (MDMO) based on dominance relationships is applied to solve the Pareto front of the rod wear problem, using the Kriging model as the objective function. The findings indicate that the three sets of optimized process parameters, derived from the Pareto front characteristics, significantly reduce the depth of wear on the fuel rod surfaces during rod drawing. This study provides a valuable reference for mitigating surface wear of fuel rods during rod loading operations.

Research on the Determination Method of Preloading Force for Main Equipment Supporting Anchor Bolts in Nuclear Power Plants

Chen Sun, Zhou Yupeng, Xie Honghu, Zhao Xiaohong

2025, 46(5): 148-151. doi: 10.13832/j.jnpe.2024.090001

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This study addresses the calculation of preloading force of anchor bolts for main equipment supports in nuclear power plants. By analyzing the loads on these bolts and evaluating the factors influencing the preloading force under actual operating conditions, a calculation method tailored to determine the preloading force of such bolts is proposed. Validation against design data from similar overseas projects shows a deviation of less than 5%.

Experimental Study on the Strength Probability Distribution of a Domestic Fine-Grained Nuclear Graphite

Qian Hao, Lan Tianbao, Yan Peng, Liu Hetong

2025, 46(5): 152-160. doi: 10.13832/j.jnpe.2024.10.0053

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Five strength tests are conducted for a candidate domestic fine-grained nuclear graphite used in gas-cooled micro-reactors, including the uniaxial tensile test as per the standard of Deutsche Industrie Norm (DIN), as well as the three-point flexural test, the Brazilian disc splitting tensile test, the uniaxial compressive test, and the uniaxial tensile test as per the standards of American Society for Testing and Materials (ASTM). Based on the test results, the strength probability distributions are systematically analyzed. It was found that the fitting results of the two-parameter Weibull distribution to the five types of strength data all pass the Anderson-Darling test (A-D test). Compared to the normal distribution and the three-parameter Weibull distribution, the fitting results of the two-parameter Weibull distribution are more conservative at low probabilities. Therefore, it is reasonable to use the two-parameter Weibull distribution to describe the strength probability distribution of the domestic fine-grained graphite studied. The characteristic strength of the nuclear graphite is closely related to the stress state and gradient. The uniaxial compressive strength is much higher than the uniaxial tensile strength, while the latter is clearly higher than the Brazilian disc splitting tensile strength. The three-point flexural strength is obviously higher than the uniaxial tensile strength. In addition, the Weibull modulus, which reflects the strength dispersion, is closely related to the stress state. The strength dispersion under the uniaxial tensile stress state is much higher than that under the uniaxial compressive stress state. The strength dispersion of Brazilian disc splitting tensile strength is between the above two, because the stress state in the high-stress critical region of the Brazilian disc, where the stress reaches more than 90% of the maximum value of the center, is between the uniaxial tensile and uniaxial compressive stress states. This observation suggests that the Weibull modulus should be considered as a function of stress states.

Evaluation of Irradiation Displacement Damage and Radioactivity of High Entropy Alloys under Typical Neutron Energy Spectrum

Yin Xianpeng, Chen Da

2025, 46(5): 161-170. doi: 10.13832/j.jnpe.2024.09.0010

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The structural materials in nuclear reactors are subjected to neutron irradiation, which leads to lattice displacement damage and neutron activation. These two neutron irradiation effects are correlated with the material's elemental composition, neutron fluence, and irradiation energy spectrum. This study focuses on high-entropy alloys, employing Monte Carlo neutronics calculation methods to evaluate radioactivity levels, equivalent dose rates, displacement damage, and helium production in three typical reactor spectra (pressurized water reactor, fast reactor, and fusion reactor). The results demonstrate that compared with 15-15Ti austenitic stainless steel and Eurofer97 reduced-activation steel, both FeCoNiCr and NbZrTiV high-entropy alloys exhibit higher neutron-induced radioactivity. These alloys show greater displacement damage efficiency in fast reactor spectra but relatively lower helium production. Future research should focus on targeted optimization of elemental compositions in high-entropy alloys according to specific service environment, with particular emphasis on minimizing easily activated elements.

Research on Control Strategy of the Coolant System for Low-Temperature Heating Reactor

Jiang Qingfeng, Hong Hao, Lyu Hong, Wang Pengfei

2025, 46(5): 171-179. doi: 10.13832/j.jnpe.2024.090030

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To ensure the operation safety of low-temperature heating reactors, it is crucial to study reactor power control strategies that meet their operational requirements. To this end, this study takes the low-temperature heating reactor coolant system as the research object and proposes two control strategies: a dual-feedback control strategy and a cascade control strategy. Their respective effects on the coupled control of reactor power and core coolant outlet temperature are investigated. The simulation results show that both control strategies can effectively control the low-temperature heating reactor coolant system under step and linear load change conditions. Under the cascade control strategy, the coolant outlet temperature is the main controlled variable, whose change magnitude is small in the load change conditions, but the settling time of reactor power is prolonged. Therefore, the cascade control strategy is more suitable for the linear load change conditions. Under the double-feedback control strategy, the reactor power control channel and the temperature control channel are in a parallel configuration, allowing both control performances to be taken into account. Therefore, the double-feedback control strategy is more suitable for the step load change conditions. This study can provide a reference for the development and optimization of low-temperature heating reactor coolant system control strategy.

Research and Analysis on the Performance of Differential Transformer Type Rod Position Detector

Jin Shuwu, Li Zewen, Zhou Guofeng, Hu Lunbao, Li Yuanyuan, Lu Zhaohui

2025, 46(5): 180-186. doi: 10.13832/j.jnpe.2024.09.0002

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This study analyzes the causes of the rod position jump caused by the gourd-shaped waveform in induction voltage of the secondary A-code coil of the rod position detector, and proposes a solution to improve the performance of the rod position detector. Using a 1/2 calculation model, the secondary A-code coil was split and reconfigured. The study found that due to the placement of the secondary A-code coil #1 at the end of the primary coil, it is significantly affected by the large leakage flux at the primary coil end. This results in a relatively low induced voltage in the coil #1, which cannot effectively offset the induced voltage in the differentially connected secondary A-code coil #3. Consequently, a substantial residual voltage remains in the circuit at the zero-position voltage. To solve this problem, this paper proposes to increase the number of turns of the the secondary A-code coil #1 to compensate for the leakage flux at the primary coil end. This solution was verified through numerical calculation and prototype test. The results show that when the number of turns of secondary A-code coil #1 is more than 1.24 times, the small gourd-shaped waveform can be eliminated, the rod position jump caused by the small gourd-shaped is avoided, and the performance of the rod position detector is improved.

Design of Common Cause Failure Prevention for Priority Actuation and Control System in VVER-1200 Nuclear Power Plant

Xia Limin

2025, 46(5): 187-194. doi: 10.13832/j.jnpe.2024.09.0013

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In the Vodo-Vodyanoi Energetichesky Reactor (VVER-1200) nuclear power plants, an independent priority actuation and control system is used to realize the control function of safety actuators with different levels of defense in depth and safety classes from different systems/equipment. During the design of PACS, it is necessary to adopt corresponding measures to mitigate the impact of common cause failure on the performance of safety functions. Based on the domestically developed platform of the main instrumentation and control system for VVER-1200 nuclear power plants, the design principles and methods of PACS for preventing common-cause failures are introduced, and the feasible schemes to reduce the probability of common-cause failure occurrence and the probability of simultaneous triggering or propagation of common-cause failures are proposed. Various means such as diversity, testability and independence are combined to improve the level of common-cause failure prevention of the system, and these measures have been successfully applied in engineering projects. This design approach holds significant reference value for the development, research, and improvement of common-cause failure prevention strategies in PACS for various reactor types.

Modeling and Control System Design of a New Gas-Cooled Microreactor System

Li Ying, Jiang Mingyue, Wu Changhao, Liao Shengyong, Zhao Chenkai, Fan Sui

2025, 46(5): 195-204. doi: 10.13832/j.jnpe.2024.10.0033

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Compared to traditional pressurized water reactors, the new gas-cooled microreactor exhibits significant differences in system characteristics and control methods in the unique core structure and thermal conduction mechanism. To understand the dynamic characteristics of the 4 MW new gas-cooled microreactor system and design a load-following control system, this paper establishes a mechanistic model of the new gas-cooled microreactor through theoretical derivation. The model was validated under both steady-state and transient conditions. The results show that the steady-state error is less than 1% and the transient results were consistent with the mechanistic model. It is demonstratd that the established mechanistic model can accurately reflect the dynamic characteristics of the new gas-cooled microreactor. Based on this, a dynamic control model was established to analyze the dynamic characteristics. At the 100% power level, 10% step decrease in load level is introduced to simulate and validate the established control scheme. The results show that the designed control system can track changes in the load setpoint, ensuring core outlet temperature and secondary side outlet air temperature of the intermediate heat exchanger constantly. The new gas-cooled microreactor model and control system design studied in this paper can provide technical support for the development of new gas-cooled microreactor technology.

Development Opportunities and Challenges for China's Nuclear Power Equipment under the "Dual Carbon" Goals

Tang Chuanbao, Chai Xiaoming, Zhu Yonghui, He Xiaoqiang, Fu Guozhong, Yu Zeyuan, Li Rui

2025, 46(5): 205-210. doi: 10.13832/j.jnpe.2025.04.0184

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Abstract:
Nuclear power is a safe, economical, and efficient clean energy source that serves as a strong support for China's national strategy of "Carbon Peaking and Carbon Neutrality". Nuclear power equipment is the core enabler of nuclear power plants' critical functions. This paper analyzes the development status of nuclear power in China and elaborates on the current state of nuclear power equipment. It examines development opportunities in two major directions: power generation and multi-purpose applications. The challenges faced by nuclear power equipment are identified, including performance improvements of existing equipment, R&D of equipment for advanced reactor types, and the establishment and transformation of the industrial chain. Development suggestions are provided, which may serve as a reference for the high-quality development of China's nuclear power equipment.

Research on Differential Pressure Level Measurement Methods for Pressurizers

Lu Qing, Wang Yu, Jiang Xiaowei, Chen Zhihui, He Liang, Jiang Guanfu, liang Yu, Pan Junjie

2025, 46(5): 211-216. doi: 10.13832/j.jnpe.2024.09.0005

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Accurate measurement of pressurizer level is crucial for the safe operation of nuclear reactors. Based on the fundamental principles of differential pressure level measurement and considering the high-temperature, high-pressure operating environment of nuclear reactor pressurizers, this study investigates five level measurement schemes to eliminate the effects of system temperature, pressure variations, and ambient temperature variations on measurement accuracy. Three typical schemes, i.e., single internal reference tube, double internal reference tubes, and external reference tube, were selected for modeling and simulation. The simulation results indicate that the the double internal reference tube scheme effectively mitigates the adverse impact of ambient temperature on measurements and better reflects the dynamic level changes within the pressurizer. Therefore, the double internal reference tube level measurement scheme is recommended.

Identification Technology of Weak Collision Mechanical Noise of Reactor Lower Grating Plate Base on WPD and Kurtosis

Liu Jiaxin, Bao Yufeng, Zhe Na, Wang Jin, Duan Zhiyong, Liu Caixue, Yang Taibo

2025, 46(5): 217-223. doi: 10.13832/j.jnpe.2024.10.0049

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The loose parts carried by the coolant in the reactor primary circuit can move to the lower grating plate, causing collisions with the lower grating plate and further blocking the diversion hole. After being transmitted to the top cover of the reactor pressure vessel through internal structures, the collision mechanical noise of the lower grating plate experiences signal attenuation and is masked by the reactor background noise, making direct identification impossible. Therefore, this study first conducted simulation tests to obtain background noise data and weak collision mechanical noise data of the lower grating plate. Then, wavelet packet decomposition (WPD) combined with a kurtosis threshold method was used to denoise the weak collision mechanical noise submerged in background noise. Finally, based on the denoised collision signal, loose parts collision identification was performed, and a code for identifying collision events of the lower grating plate in the reactor was developed. Test results show that the proposed denoising method is effective, and the developed code can effectively identify the signals of lower grating plate collision events submerged in background noise.

Research Status of Deep Learning in Equipment Condition Assessment and Its Application Prospects in Nuclear Field

Luo Neng, Liu Caixue, Zhu Jun, Zhou Zhengping, Luo Ting, Zhou Chengning

2025, 46(5): 224-233. doi: 10.13832/j.jnpe.2024.10.0069

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Abstract:
With the wide application of nuclear energy in various industries around the world, condition assessment of reactor electromechanical equipment is facing higher requirements in terms of coverage, real-time, economy and accuracy. Traditional methods based on physical simulation, expert knowledge or data-driven approaches are difficult to meet these new challenges. Starting from the research idea of deep learning in equipment condition assessment, this paper centers on the four core tasks of equipment condition identification, fault diagnosis, fault prediction, and deployment and application of condition assessment models, systematically sorts out the status and shortcomings of current research, and reviews and looks forward to the application of deep learning in equipment condition assessment in the nuclear field. On this basis, an equipment condition assessment technology system based on knowledge and data fusion iteration is proposed, which clarifies the direction of future technology research and provides new ideas and methods for intelligent operation and maintenance of nuclear energy equipment.

Research on Intelligent Online Monitoring and Robust Self-Correction for Nuclear Reactor Sensors

Xu Fenqin, Yan Xiaoyu, Pang Bo, Zhao Dou, Tu Yan

2025, 46(5): 234-242. doi: 10.13832/j.jnpe.2024.090019

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A novel training dataset processing method with high robustness was proposed to address the poor robustness in data-driven analytic redundant sensor models, and an sensor online monitoring and self-correction method was developed based on a auto-associative multivariate Long Short-Term Memory (LSTM) artificial neural network model. The method was validated using actual sensor measurement data retrieved from a pressurized water reactor engineering test facility. The results indicate that this research method can achieve high-precision and robust reconstruction of sensor signals, hence meets the requirements of online monitoring and robust self-correction of nuclear reactor sensors.

Experimental Study of Bubble Distribution Characteristics in a 10×10 Rod Bundle Subchannel Based on Mask R-CNN

Sun Niujia, Zhang Heng, Hang Qin, Wang Wencong, Liu Caixue, Zhang Peilai, Li Jiayi, Wang Xiangfan

2025, 46(5): 243-248. doi: 10.13832/j.jnpe.2024.090054

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Abstract:
The rod bundle structure is widely used in critical equipment such as reactor cores and steam generators, and the geometric and physical parameters of the bubbles in their subchannels play a crucial role in the numerical prediction of mass transfer and heat transfer processes. This paper conducts an experimental study on the air-water two-phase flow behavior in the subchannels of a 10×10 rod bundle, using high-speed imaging combined with Mask R-CNN to investigate the effects of gas flow rate and nozzle diameter at different heights on bubble size, shape, and void fraction. The results show that under the constraint of narrow gaps, approximately 45% of the bubbles maintain a stable ellipsoidal shape. When the nozzle diameter reaches 0.3 mm, about 30% of the bubbles undergo significant deformation. Due to the limited gap space, the bubble diameter peaks around 2 mm, but as the gas flow rate and nozzle diameter increase, the bubble generation frequency rises, and the void fraction increases accordingly. During the bubble ascent, the diameter slightly increases, but the bubble number decreases along the axis, ultimately leading to a reduction in the void fraction.

Point Cloud Modeling Method for Deep Learning Numerical Calculation of Nuclear Reactor Core

Li Manyuan, Liu Yanli, Liu Dong, Lyu Hengye, Mu Qiao, An Ping, Xing Guanyu, Yang Hongyu, Tu Xiaolan, Pang Zhixin

2025, 46(5): 249-257. doi: 10.13832/j.jnpe.2024.090048

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With the development of deep learning technology, the application of deep learning computational methods to solve multi-disciplinary equations in reactor physics, thermal hydraulics, and other fields has become a hot research topic and an important future direction in reactor numerical computation. To address the challenge that traditional mesh structures cannot be directly used for nuclear reactor simulation, this paper investigates a point cloud-based modeling method for nuclear reactor cores at three levels: fuel cell, assembly, and core. The study achieves hierarchical modeling from fuel cells to assemblies and finally to the core, improving the reusability of model data. Based on this modeling approach, a point cloud-based reactor modeling software is developed, featuring a series of functions such as nuclear reactor structure design, point cloud sampling, boundary separation, and attribute visualization. This work represents the first publicly documented reactor modeling method tailored for deep learning numerical computation, providing effective and reusable nuclear reactor data for deep learning simulations. Using real reactor core data, this paper validates the constructed core model through deep learning numerical simulations, confirming the correctness and effectiveness of the models generated by the software.

Research on Data Fusion Methods for Digital Twins of Steam Generators

Song Houde, Song Meiqi, Liu Xiaojing

2025, 46(5): 258-266. doi: 10.13832/j.jnpe.2024.090059

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Digital twin technology provides innovative means for the research and safe operation of critical equipment in nuclear power plants. Among them, the steam generator, as the core equipment for energy exchange between the primary and secondary circuits, has complex thermal-hydraulic phenomena inside that are crucial for nuclear power safety. However, in the process of constructing a digital twin for the steam generator, there is a challenge of scarce high-fidelity data. By integrating data of varying fidelity and scales, the value of low-fidelity data can be enhanced, alleviating the problem of data scarcity. To address this, this study proposes a data fusion model based on neural networks to establish connections between data of different fidelity. First, a large set of low-fidelity data samples is used to train an approximate model. Then, a small set of high-fidelity data is employed to learn the relationship between low- and high-fidelity data, resulting in a final high-fidelity surrogate model. The proposed method was validated using 2291 low-fidelity and 582 high-fidelity samples from the steam generator simulation with RELAP5 of PKL B3.1 case. The results show that the test set normalized error (MSE) of the data fusion model can fall below 2.0×10⁻⁴, which can achieve similar or even higher accuracy compared to the direct modeling relying on only high-fidelity data. This method is conducive to solving the insufficient accuracy caused by data scarcity in the process of digital twin construction.

Research on Adaptive Control Method of Turbine in Nuclear Power Plant based on Fuzzy Neural Network

Zhang Jiacheng, Jiang Guanfu, Wei Xinyu, Sun Peiwei

2025, 46(5): 267-273. doi: 10.13832/j.jnpe.2024.090061

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Aiming at the Digital Electro-Hydraulic (DEH) Control System faults that affect the stable operation of nuclear power plant, this study conducts fault simulation analysis of electro-hydraulic servo valves, oil motors, and displacement sensors in the DEH system. An adaptive control scheme based on a fuzzy neural network (FNN) with fault information adjustment is proposed. The FNN is used to optimize the parameters of the nuclear power plant turbine controller, enabling adaptive adjustments according to changes in actuator characteristics. Simulation results after optimization show that the designed control system has a good control effect on the turbine system both in fault-free and fault-present conditions. Consequently, the FNN adaptive control method proposed in this study can be applied to mitigate DEH system faults and provides valuable reference for the design of automatic turbine control systems in nuclear power plants.

Research on Rapid Prediction of 3D Steady-State Temperature Field of Steam Generators Based on Deep Learning

Ye Yibo, He Shaopeng, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Su Guanghui

2025, 46(5): 274-284. doi: 10.13832/j.jnpe.2024.090064

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To establish a fast prediction method for the three-dimensional steady-state temperature field of a steam generator and enable fast computation of the steady-state response of key nuclear reactor equipment under different operating conditions, this study employs Convolutional Neural Network (CNN) and Transformer algorithms to predict the 3D steady-state temperature field of vertical natural circulation U-tube steam generators. Steady-state temperature distributions across multiple planes of the steam generator under different boundary conditions are obtained through Computational Fluid Dynamics (CFD) simulations. These results are then processed with interpolation methods to create training and testing datasets suitable for deep learning algorithms. The aforementioned deep learning algorithms are trained using the training dataset, and their performance is evaluated using the testing dataset. The results show that when predicting the 3D steady-state temperature field of steam generators, deep learning algorithms can reduce prediction time to 0.3 seconds, with temperature prediction errors not exceeding 1 K, enabling real-time prediction of the 3D steady-state temperature field.

Multi-objective Optimization and Objective Decision Analysis of Supercritical Carbon Dioxide Nuclear Reactor System Based on CRITIC-TOPSIS Method

Shen Yu, Zhou Xiafeng, Zhang Fan, Chen Chong, Ren Huan

2025, 46(5): 285-293. doi: 10.13832/j.jnpe.2024.090022

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Supercritical carbon dioxide (S-CO₂) nuclear reactor system, due to its high efficiency and compact design, holds significant potential for applications in miniaturized reactor domains such as maritime vessels and space exploration. This study conducts a multi-dimensional comprehensive performance optimization of the system, covering overall volume, efficiency, and specific work, and proposes a data-driven objective weighting method to reduce the impact of subjective judgment on the optimization results. In this study, we first construct models for the S-CO₂ nuclear reactor recompression direct cycle system and the calculation of various optimization objectives. Subsequently, using the genetic optimization algorithm (NSGA-II), we develop a multi-objective optimization and objective decision analysis framework for the S-CO₂ reactor system based on the Criteria Importance Through Intercriteria Correlation (CRITIC) method and the Technique for Order Preference by Similarity to Ideal Solution (TOPSIS). With maximum thermal efficiency, maximum specific work, and minimum volume as optimization objectives, the objective weights of each target are fully derived from sample data. Finally, we conduct a systematic multi-objective performance optimization and decision analysis of the entire system. The results indicate that within the given range of optimization variables, system volume carries a relatively large objective weight. A multi-criteria decision analysis is performed on the optimized Pareto frontier using the derived objective weights, leading to the identification of the optimal multi-objective parameter values under the TOPSIS decision scheme. This study provides a theoretical reference for a comprehensive and in-depth analysis of the S-CO₂ reactor system's performance.

Development of Overall Design Method and Modeling Platform for Nuclear Reactor Based on MBSE

Chen Zhao, Yan Bo, Xu Weifeng, Yuan Weiquan, Duan Chengjie, Lin Jiming, Lan Nianwu, Zhao Pengcheng

2025, 46(5): 294-302. doi: 10.13832/j.jnpe.2024.10.0066

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To enhance the overall design capability of nuclear reactors and support the digital transformation of nuclear reactor development, this study conducts research on the overall design method of nuclear reactors and the development of a modeling platform. The Model-Based Systems Engineering (MBSE) methodology is integrated with the engineering design process of nuclear reactors to establish a MBSE model system and modeling process for nuclear reactor overall design. An MBSE collaborative modeling platform, M-Reactor, is developed for nuclear reactors. Based on the M-Reactor platform, MBSE modeling is conducted for the innovative concept of heavy oil thermal extraction liquid heavy metal reactor as an example, and the requirement model, functional model, architecture model, and verification model are constructed. The closed-loop verification and validation are conducted at the system level, and the nuclear reactor design process from requirement analysis to overall solution design is practiced. The study shows that the MBSE method can accurately describe and convey the requirements, functions, and architectures of nuclear reactors and enable data exchange based on digital models. It can quickly verify and validate the overall scheme of a new nuclear reactor in the early design stage, significantly improving the efficiency and quality of design demonstration for the overall scheme of new reactors. The results of this study provide theoretical guidance and tool support for further application of MBSE in the field of nuclear reactor development.

Current Issue 2025, Volume 46, Issue 5

Current Issue
2025, Volume 46, Issue 5