核动力工程

Research Progress in the Application of Artificial Intelligence in Reactor Neutron Analysis

Wu Hongchun, Lei Kaihui, Shen Jingwen

2025, 46(1): 1-12. doi: 10.13832/j.jnpe.2025.01.0001

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Abstract:
Amid the global surge in scientific intelligence research and applications, artificial intelligence (AI) technology has been applied in various aspects of reactor neutron analysis to enhance its intelligence, precision and efficiency. This paper provides a comprehensive review of the research progress in the application of AI technologies within reactor neutron analysis, aiming to provide insights for advancing the digitalization of reactors and supporting future developments in this field. We first introduce AI methods' basic classifications and characteristics, and then investigate the four key aspects of neutron analysis: nuclear data evaluation, problem modeling, numerical solution of equations, and transport result application. Also critical technologies are reviewed and summarized for each aspect. Finally, the paper discusses challenges in AI-based neutron analysis, particularly in model, data, and application security. In response to these challenges, the paper proposes research recommendations.

Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN

Jiang Duoyu, Xu Peng, Jiang Xinbiao, Hu Tianliang, Wang Lipeng, Cao Lu, Li Da

2025, 46(1): 13-23. doi: 10.13832/j.jnpe.2025.01.0013

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In response to the evolving demands of advanced small reactor technologies, there is an increasing requirement for three-dimensional whole core transport calculations in reactor physics numerical simulation codes. This paper presents the development of a three-dimensional space-time kinetics neutron transport code, SAAFCGSN, based on the MOOSE platform. The code implements spatial variable discretization using the finite element method, and solves the steady-state and transient neutron transport equations as well as the delayed neutron precursor equations using a residual form approach. The Jacobian-Free Newton-Krylov (JFNK) method is employed to avoid the direct computation of the Jacobian matrix, thereby enhancing computational speed and reducing memory usage. To evaluate the transient computational capability of the code, we validated its reliability using the OECD/NEA C5G7-TD benchmark series and conducted a comparative analysis with high-fidelity deterministic neutron transport codes and Monte Carlo simulations. The study demonstrates that the SAAFCGSN code achieves high computational accuracy, effectively manages the cusping effect of control rods, and obtains detailed neutron flux distributions within the three-dimensional core. It meets the steady-state and transient neutronic calculation requirements for advanced small reactors.

Feasibility Study of 24-month Refueling for VVER

Hua Xinchao, Pan Deng, Li Yao, Peng Yipeng, Liu Jiaqing

2025, 46(1): 24-29. doi: 10.13832/j.jnpe.2025.01.0024

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In order to reduce the number of refueling overhauls within the service life of VVER, improve the unit capacity factor, and increase annual power generation, this paper studies the feasibility of realizing the 24-month refueling cycle of VVER by increasing the average ²³⁵U enrichment of fuel assemblies and the number of refueling batches using KASKAD software package, so as to design a long and short alternative 24-month long cycle core refueling scheme. The analysis of the calculation results of this scheme shows that the duration of cycle is 633.5 EFPD (effective full power days) or 667.1 EFPD, respectively, and each core characteristic parameter meets the design limit requirements and has a large margin. Compared with the current 18-month core refueling cycle scheme, this scheme has good economic benefits. When the international natural uranium purchase price does not exceed 60 dollars/pound, each unit will bring at least 12 million yuan of income per year. Therefore, the core loading scheme of the 24-month refueling equilibrium cycle studied in this paper has good safety, economy and flexibility, and has very good engineering application value.

Modeling Method and Impact Analysis of Fuel Rod Radial Thermal Expansion of PWR

Zhou Yufeng, Xie Weirong, Wan Chenghui, Cao Xin, Bai Jiahe, Guo Lin

2025, 46(1): 30-35. doi: 10.13832/j.jnpe.2025.01.0030

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Abstract:
During the power operation of PWR, fuel pellets and cladding undergo varying degrees of thermal expansion due to temperature changes, which significantly impacts the core physics calculations. In order to accurately consider the influence of fuel rod radial thermal expansion on the core physics analysis results in the "two-step" scheme, this paper proposes a precise modeling method for fuel rod radial thermal expansion based on the reactor-physics analysis software Bamboo-C. This method considers the geometric changes caused by fuel rod radial thermal expansion at both the assembly calculation and reactor core calculation levels, and determines the precise geometric expansion size through iterative convergence of the temperature field at the core calculation level. In this paper, the EPR1750 unit is taken as the research object, and the startup physical tests and power operation processes of each fuel cycle are calculated. The results show that: the average error between the calculated and measured isothermal temperature coefficients decreased from −3.065 pcm/K (1pcm=10^–5) to −1.870 pcm/K, and the average error between the calculated and measured critical boron concentrations decreased from −5.9ppm (1ppm=10^–6) and −5.7ppm to −2.5ppm and −2.7ppm. Therefore, the PWR fuel rod thermal expansion modeling method proposed in this paper can improve the calculation accuracy of key safety parameters of EPR1750 to a certain extent and has certain engineering application value.

Study on the Attenuation Method of Discrete Ordinate Ray Effect Based on Multiple Collision Source

Zheng Zheng, Li Xiang, Mei Qiliang, Li Hui, Wang Mengqi

2025, 46(1): 36-40. doi: 10.13832/j.jnpe.2025.01.0036

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This study aims to explore the use of the Monte Carlo (MC) method to calculate multi-collision source distributions, so as to mitigate the ray effect in the discrete ordinates (SN) method, particularly in models with localized sources, low-density regions, or vacuum areas. The distribution of multiple collision sources is calculated by MC method, and the method of reducing the ray effect of multiple collision sources and first collision sources is preliminarily verified and analyzed by Kobayashi benchmark problem. Through hybrid parallel technique (message passing interface MPI + open multiprocessor OpenMP), the computational efficiency is improved and the load imbalance problem is reduced. The study shows that the MC multi-collision source method effectively suppresses the ray effect in localized source problems within the SN method, significantly reducing non-physical oscillations and improving computational accuracy. When the number of collisions exceeds three, the computational deviation is reduced to below 10%. Furthermore, the use of hybrid parallel computing resulted in a parallel efficiency exceeding 80% with 120 threads. Although this method has a slight drawback in terms of computational time, it shows great potential for addressing complex geometries and physical conditions, making it a promising approach for reducing the ray effect.

Best Estimate Power Method Based on Vanadium SPND Prompt Response Currents

Shao Ruizhi, Cao Liangzhi, Li Yunzhao, Chen Lei

2025, 46(1): 41-46. doi: 10.13832/j.jnpe.2025.01.0041

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Abstract:
During operation of the third-generation pressurized water reactor (PWR), the total power can be rapidly estimated based on the response current of the Self-Powered Neutron Detector (SPND). It can be referred to as the best estimate power method. Existing techniques rely on the assumption of a linear relationship between the total SPND current of Vanadium SPND and the total power (referred to as the total current estimation method), which cannot differentiate the time response of different current components. Thus, it is not suitable for transient operating conditions. To enhance the monitoring capability during operation, a novel estimation method based on the prompt current of Vanadium SPND is proposed, utilizing the SPND response current calculation feature in the NECP-Bamboo PWR core analysis software. This method can be applicable for both steady and transient operating process. The method is quantitatively compared with existing methods. Numerical results demonstrate that: 1) the deviations of the total current estimation method and the prompt current estimation method are both less than 1% of rated full power during the steady operation; 2) the deviation of the total current estimation method exceeds 50%, while the deviation of the prompt current estimation method is less than 1% during the Rapid Power Reduction (RPR) bank insertion process.

Progress and Application of Neutron Radiography Characterization Technology for Multiphase Flow Pattern of New Working Medium Reactor

Mei Zhongkai, He Linfeng, Wen Qinglong, Qiu Zhifang, Chen Dongfeng

2025, 46(1): 47-62. doi: 10.13832/j.jnpe.2025.01.0047

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Neutron radiography shows important application potential in the visualization and measurement of multi-phase flow morphologies in new working medium reactors. This article elaborates on the basic principles of neutron radiography measurement methods, provides a comprehensive review of the research progress of neutron radiography technology in traditional light water reactors, lead-bismuth cooled fast reactors, heat pipe cooled reactors, supercritical water reactors, and sodium-cooled fast reactors. It also outlines the future development directions of neutron radiography technology applied to new working medium reactors and provides the basic methodologies for obtaining high-fidelity neutron images and measuring flow patterns.

Study on Numerical Simulation Method of Three-Dimensional Flow Field of Spiral Coil Tube Bundle in Lead-Bismuth Medium

Yin Yibo, Wen Jiming, Tian Ruifeng, Gao Puzhen, Xia Bangyang, Chen Chong, Zhang Xue, Tan Sichao

2025, 46(1): 63-72. doi: 10.13832/j.jnpe.2025.01.0063

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In the design of heat exchanger, by understanding the three-dimensional flow field characteristics such as swirling and vortical flow of the lead-bismuth medium in the tube bundle, the design of heat exchange tube can be better predicted and evaluated to ensure the safe and stable operation of heat exchanger. This study focuses on investigating the phenomenon of vortex shedding of lead-bismuth medium in a helical coil using numerical simulations. The simulations were conducted at a temperature of 200℃ and an inlet flow velocity of 0.5 m/s. The validated large eddy simulation (LES) model results were used as a benchmark to compare with different models, aiming to develop a computationally efficient solution model for this problem. The results show that detached eddy simulation (DES) SST k-ω model has higher computational efficiency than LES model. Within the study range, the spiral coil tube bundle is at least affected by vortex shedding and turbulent excitation. The interaction between shedding vortices from different tubes leads to a multi-peak feature in the lift and drag spectrum. Different from the straight tube bundle, the flow structure and velocity distribution downstream of the spiral coil tube bundle are asymmetrical. Additionally, the normal velocity of the gap between the tubes is enhanced due to the existence of the rise angle of the spiral tube bundle, and the coiling shape of the spiral tube makes the lateral velocity distribution more uneven.

Investigation of Radial Thermal-Hydraulic Characteristics of Sodium Heated Once-Through Steam Generator

Feng Zhenyu, Liu Yapeng, Wang Bo, Zhang Dalin, Li Xinyu, Tian Wenxi, Qiu Suizheng, Su Guanghui

2025, 46(1): 73-82. doi: 10.13832/j.jnpe.2025.01.0073

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The sodium heated once-through steam generator is the critical equipment of a sodium-cooled fast reactor. The traditional one-dimensional thermal-hydraulic system analysis code cannot take into account the radial thermal-hydraulic parameter distributions of the steam generator, and therefore is not applicable to fatigue and thermal stress analyses. In order to calculate the radial thermal-hydraulic parameter distribution of the steam generator, the TCOSS-2D code was developed adopting a combination of pipe network, pipeline and steam generator. Firstly, radial multilayer modeling and thermal-hydraulic calculations were carried out on the steam generator prototype of the Chinese demonstration fast reactor (CFR600) using TCOSS-2D, and the results showed that the transient response trend calculated by the code was in good agreement with the experiment, so the radial thermal hydraulic calculation capability of the code was validated. Secondly, the radial non-uniform flow distribution on the sodium side of the sodium heated once-through steam generator was considered and compared. The results show that the radial non-uniform sodium-side flow distribution has a greater effect on the radial temperature difference of the evaporator than the superheater, while the non-uniform flow distribution of the sodium-side of the transient process has a relatively small effect on the non-uniformity of the water-side. Therefore, the numerical model of sodium heated once-through steam generator developed in this study can be utilized for the analysis of its radial thermal-hydraulic characteristics.

Study on the Droplet Jet Impingement Process on Wall Based on DPM-to-VOF Method

Chen Qingshan, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Su Guanghui

2025, 46(1): 83-91. doi: 10.13832/j.jnpe.2025.01.0083

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Spray atomization is widely used in nuclear power equipment to ensure its safe and stable operation. In the spraying, some droplets will impinge on the equipment wall and expand and evolve in the form of liquid film. This study investigated droplet jet impingement based on Fluent’s Discrete Particle Model-to-Volume of Fluid (DPM-to-VOF, DTV) transition method, and the accuracy of the simulation method was validated using experimental images of two droplet falling processes. The research investigated the influence of droplet injection velocity and incident angle on the wall liquid film morphology and obtained the variation of droplet splash rate with incident conditions. Furthermore, the effects of gravity, surface tension, velocity components, and other factors on the expansion process of liquid film morphology and droplet splash rate during droplet impingement on the wall were analyzed. It is found that increasing the injection velocity could enhance the disturbance to the liquid film, significantly increasing the contact area between the liquid film and the wall and the width and height of the liquid film area. Increasing the incident angle had a significant impact on the change of liquid film morphology, which was reflected in the increase in the width and decrease in the height of the liquid film area, leading to decreased stability of the liquid film surface. The splash rate increased with the increase of injection velocity and incident angle, with the incident angle having a more significant impact on the splash rate.

Analysis and Verification of LOCUST Based on SGTR Accident of DOEL-2

Yuan Bo, Lei Xing, Wen Qinglong, Chen Kang, Xu Caihong, Li Jinggang

2025, 46(1): 92-99. doi: 10.13832/j.jnpe.2025.01.0092

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The steam generator tube rupture (SGTR) accident will cause significant fluctuations in the thermal and hydraulic parameters of the primary and secondary circuits, seriously endangering reactor safety. The thermal and hydraulic system analysis software LOCUST developed in China can be used to calculate the thermal and hydraulic parameters in the SGTR accident and predict the accident process. This study takes the SGTR accident of DOEL-2 power plant as the research object, and uses the LOCUST as the calculation tool to model and calculate the DOEL-2 power plant. The calculation results are compared with the RELAP5 calculation results and actual data to evaluate the accuracy of LOCUST software in predicting SGTR accidents. It shows that the LOCUST can effectively predict the SGTR accident process, and the calculated parameters of the primary and secondary circuits are in good agreement with the RELAP5 calculation results and actual data. This study can provide data support for the verification of LOCUST.

Development and Validation of a Four-Parameter SST k-ω-k_θ-ε_θ Model for LBE Flow and Heat Transfer

Wu Jie, Su Xingkang, Cai Jiejin, Gong Ziqi, Gu Long

2025, 46(1): 100-106. doi: 10.13832/j.jnpe.2025.01.0100

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In order to solve the problem of numerical heat transfer computation for extremely low-Prandtl-number fluids and to improve the computational accuracy of the numerical simulation of liquid lead-bismuth eutectic (LBE), a four-parameter SST k-ω-k_θ-ε_θ model for the Reynolds stress and turbulent heat flux is developed under the framework of OpenFOAM. Based on benchmark flow and heat transfer experiments of LBE in a vertical pipe and wire-wrapped 19-rod bundle, a comparison and thermal-hydraulic analysis were conducted against the turbulent Prandtl number (Pr_t) model using relevant empirical Nusselt number and friction factor correlations. The results show that the temperature predicted by the four-parameter SST k-ω-k_θ-ε_θ model agrees well with the experimental data, and the heat transfer prediction performance is better than that of the Pr_t model. The model is suitable for the numerical computation of LBE flow and heat transfer.

Numerical Simulation and Analysis of Flow and Heat Transfer Characteristics in Annular Helical Cruciform Fuel Elements

He Xing, Cheng Jie, Wu Di, Zhao Wenbin, Wang Jianjun

2025, 46(1): 107-115. doi: 10.13832/j.jnpe.2025.01.0107

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Helical cruciform fuel, as a relatively new type of fuel element, has attracted widespread attention due to its unique advantages such as strong mixing effects and self-supporting. However, helical cruciform fuel still faces issues such as high central temperature. To address these problems, this paper proposes an annular helical cruciform fuel and investigates its internal flow and heat transfer characteristics using numerical simulation methods. The results indicate that using a cosine distribution for power calculation leads to issues such as excessive axial power gradient and overestimation of fuel peak temperature compared to the actual situation. Furthermore, compared to helical cruciform fuel, annular helical cruciform fuel exhibits stronger mixing effects and lower maximum fuel temperature. Additionally, increasing inlet velocity can further enhance the mixing between coolant channels, thereby improving heat transfer capability, but the effects of inlet temperature and power density on mixing are minor. Regarding fuel temperature distribution, the location of maximum temperature at different heights of the annular helical cruciform fuel element is influenced by the magnitude of coolant transverse velocity. Higher transverse velocity results in stronger heat transfer to the fuel element, leading to lower fuel element temperatures in these regions.

Investigation on the Flow and Heat Transfer Characteristics of Liquid Lead Cross-flow Tube Bundle under Cooling Conditions

Li Liangxing, Xu Xiangyang, Xiang Zutao, Shi Shang, Lei Zhenxin

2025, 46(1): 116-127. doi: 10.13832/j.jnpe.2025.01.0116

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The flow and heat transfer characteristics of liquid metal coolant in the primary heat exchanger are of great importance to the economic and safe operation of lead-cooled fast reactor. According to the arrangement characteristics of tube bundle in a spiral coil heat exchanger, the present study established a two-dimensional analysis model for liquid lead cross-flow tube bundle. The turbulent Prandtl number model suitable for the flow of liquid lead cross-flow tube bundle is compared and analyzed, and the flow and heat transfer characteristics of liquid lead cross-flow scouring heat exchanger tube bundle are studied by numerical simulation. The results show that with the increase of inlet Reynolds number, the concave degree of the time-average velocity distribution curve of liquid lead among tubes increases, and the difference of dimensionless wall temperature enlarges in the return area, while the vortex disturbance is enhanced to the area around 180° of the tube walls. The comprehensive flow and heat transfer performance of liquid lead transversally scouring tube bundles has a positive correlation with the transverse pitch diameter ratio of the spiral coil and a negative correlation with the inlet Reynolds number. Based on the numerical simulation, the heat transfer correlation of liquid lead transversally scouring tube bundles is proposed, and its prediction error is less than 10%.

Numerical Study on Convective Heat Transfer at Low Flow and Spacer Effects in Lead-bismuth Eutectic

Yuan Bo, Sun Jie, Xiao Yao, Ding Guanqun, Gu Hanyang

2025, 46(1): 128-135. doi: 10.13832/j.jnpe.2025.01.0128

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The heat transfer characteristics of lead-bismuth eutectic (LBE), the coolant of LFR, are different from conventional fluid such as water. Therefore, the flow and heat transfer characteristics of LBE at low flow rate and its spacer effect are numerically studied in this paper. By comparing with existing experimental data and combining with previous studies, appropriate turbulence Prandtl number model and turbulence model were chosen. Based on this, the LBE low-flow convective heat transfer was calculated by computational fluid dynamics (CFD). The results show that with the enhancement of buoyancy effect, the heat transfer in upflow is firstly weakened and then strengthened, while in downflow, the heat transfer is always being strengthened, which is similar to water; However, due to the extremely low Prandtl number of LBE, its overall heat transfer strengthening and weakening degree are greatly smaller than water. In the CFD calculation of a tube with a spacer, it is found that local heat transfer enhancement occurs downstream of the spacer, and the degree of enhancement decreases and gradually disappears with the increase of the buoyancy effect. The strength and length of the stagnation zone downstream of the spacer increased first and then decreased with the increase of the buoyancy effect, and the turning point was roughly located at the boundary point between the heat transfer weakening zone rising section and the heat transfer strengthening zone. In addition, there is no heat transfer oscillation in the heat transfer enhancement zone, which is different from water.

Study on Calculation Method of Crack Leakage Rate of Valve Sealing Surface

Cao Yuming, Zhao Gang, Fang Zheng, Li Yuhuan, Zhou Du, Liu Wenshi, Wang Zheng

2025, 46(1): 136-142. doi: 10.13832/j.jnpe.2025.01.0136

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During the operation of nuclear power system, cracks may occur on the valve sealing surface due to thermal load and fatigue load, which may lead to leakage and even more serious consequences. Therefore, studying the leakage rate of cracks on the valve sealing surface will improve the safety of the reactor. Based on the Henry-Fauske critical flow calculation model and considering the resistance during the leakage process, a set of calculation methods for calculating the crack leakage rate of the valve sealing surface under different operating conditions (high temperature water and saturated steam) is developed in this paper. Finally, the experimental platform of crack leakage rate measurement is built according to the actual operating conditions, and the leakage rate measured by experiment is compared with the leakage rate calculated by the theoretical model to verify the accuracy of the leakage rate calculation method proposed in this paper. Through the comparison of experiment and theory, the calculation method of crack leakage rate proposed in this paper can predict the crack leakage rate of valve sealing surface more accurately, and has good practical application value.

Research on Disturbance Wave Characteristics of Annular Flow on Edge and Corner Rods of Rod Bundle Channel

Jin Guangyuan, Bai Jinghu, Wang Rui, Li Weilian, Du Lipeng, Zhang Wenchao

2025, 46(1): 143-151. doi: 10.13832/j.jnpe.2025.01.0143

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The research on disturbance wave characteristics of annular flow on edge and corner rods of rod bundle channel can provide theoretical supports for the steady-state operation and emergency disposal of nuclear power plants. In this study, a visualize experimental system for annular flow in rod bundle channel was set up to analyze the disturbance wave characteristics on edge and corner rod surface. The results show that the disturbance wave behaviors can be divided into the small-scale waves, the bag-shaped waves, the ligament-shaped waves and the ligament-shaped waves with liquid-loss. When the liquid superficial velocity remains constant, the average film thickness decreases with the increase of gas superficial velocity; when the liquid superficial velocity is increasing, the average film thickness is also increasing. The disturbance wave height decreases with the increasing gas superficial velocity, and the disturbance wave height on side rod is lower than that of corner rod. When the liquid superficial velocity is lower than 0.41 m/s, the wave velocity increases with the increasing gas and liquid velocity. The wave velocity is higher than the gas superficial velocity when the liquid velocity is higher than 0.41 m/s. The wave frequency becomes larger with the increasing gas superficial velocity, but the law of change with liquid superficial velocity is not obvious, which depends on the change of wave form. The average entropy value of liquid film thickness of side and corner rods increases with the wave behaviors from small-scale waves, bag-shaped waves, ligament-shaped waves to ligament-shaped waves with liquid-loss, so the wave form in rod bundle channel can be determined by the multiscale permutation entropy analysis method.

Research on the Coupling Experiment of Asymmetric Thermal Stress in Solid Core

Wang Yanpei, Tang Changbing, Li Quan, Li Tao, Li Chenxi, Qiu Bowen, Fan Hang, Li Yuanming

2025, 46(1): 152-159. doi: 10.13832/j.jnpe.2025.01.0152

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The interaction between fuel and matrix caused by asymmetric thermal stress coupling is a key problem in the analysis of solid-core reactor. In this study, a distributed cold source and heat source loading method is developed by combining numerical simulation with experiment, and the thermal stress coupling simulation and experimental study of typical solid-core reactor fuel elements at high temperature are carried out. The research results show that the high temperature strain field measured is close to the numerical simulation result, and there is no risk of failure of the core fuel matrix at 350°C. The numerical prediction method and experimental approach developed in this study for asymmetrically distributed cold and heat source thermal stress coupling can be applied to the analysis of asymmetric thermal stress distribution in solid cores, and there is no risk of failure of the fuel element matrix at low temperature difference.

Experimental Study on Uniaxial Ratcheting Fatigue Behaviour of 16MND5 Steel

Mo Xuyang, Zhu Mingliang, Zhang Shanglin, Yang Licai, Chen Yao, Xuan Fuzhen

2025, 46(1): 160-168. doi: 10.13832/j.jnpe.2025.01.0160

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The aim of this paper is to analyze the ratcheting evolution law of materials under different cyclic loads, so as to guide the life prediction and structural integrity evaluation of critical components in nuclear power plants. A series of symmetric and asymmetric stress control tests were carried out at 350°C on the domestically produced 16MND5 forged bainitic steel of reactor pressure vessel to investigate the effects of stress amplitude and mean stress on the ratcheting behavior. The results show that the alloy exhibits ratcheting effects under both symmetric and asymmetric stress cyclic loading. The increase in stress amplitude and mean stress reduces the fatigue life. The cyclic evolution exhibits initial cyclic hardening, followed by cyclic softening and finally accelerated softening. The softening is promoted by the introduction of mean stress at the same stress amplitude. Ratcheting strain does not increase monotonically with increasing tensile mean stress, and the presence of a most unfavorable mean stress leads to the most pronounced ratcheting-fatigue interaction. Fracture morphology analysis showed that, depending on the magnitude of the stress level, the specimens could be categorized into fatigue failures and ratcheting failures in which large plastic strains occurred.

Investigation of Graphite Oxidation Based on Four-step Reaction Model

Shen Teng, Wang Chengyu, Guo Shaoqiang, He Kai

2025, 46(1): 169-174. doi: 10.13832/j.jnpe.2025.01.0169

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In order to establish the analysis method of corrosion reaction between graphite and oxygen in the core of Micro Gas-Cooled Reactor, a four-step reaction model of graphite oxidation corrosion is established based on Arrhenius equation. The four-step reaction of the model represents four possible chemical reactions between graphite and oxygen, and the reactive area is increased to simulate the influence of weight loss rate on corrosion rate. Through the experiment of graphite oxidation corrosion based on gas concentration method, the oxidation corrosion rates of graphite with different gas flow rates and oxygen concentrations at 500-1100°C were obtained, and the experimental data were used for corrosion parameter fitting and model verification respectively. The results of modeling and verification analysis show that the fitted four-step reaction model can be applied to the calculation and analysis of the oxidation corrosion rate of graphite and can distinguish the reaction products, and the increase of the reactive area of the model is more conducive to accurately predicting the oxidation corrosion rate under high weight loss.

Effect of Sintering Process on Densification of UN-30%U₃Si₂ Pellets

Su Danke, Pan Xiaoqiang, Lu Yonghong, Yang Jing, Wang Ting, Duan Miaomiao

2025, 46(1): 175-182. doi: 10.13832/j.jnpe.2025.01.0175

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In this paper, UN-30%U₃Si₂ composite fuel pellets with 30wt.%U₃Si₂ content were prepared based on powder metallurgy technology. The influence of sintering atmosphere, sintering temperature, sintering time and other sintering processes on the sintering density of pellets was studied. The densification process was analyzed mainly through the chemical composition, phase composition and microstructure changes of pellets. The results show that vacuum sintering of UN-30%U₃Si₂ composite fuel pellets is more favorable to densification than argon atmosphere sintering. With the increase of sintering temperature (1600~1675°C), the density increases gradually, up to 97%T.D. When the sintering temperature is higher than the melting point of U₃Si₂, with the increase of temperature, the U₃Si₂ phase suffers losses and volatilization during the high vacuum sintering process, and the N element tends to diffuse to the U₃Si₂ phase, forming the unknown USi_xN_y phase. The increase of sintering temperature or the extension of sintering time is conducive to the formation of coating of U₃Si₂ relative to UN phase.

Progress and Considerations on Candidate Cladding Materials for Supercritical Water-Cooled Reactors

Zhang Lefu, Huang Tao, Su Haozhan, Gao Yang, Guo Xianglong, Shen Zhao, Chen Kai

2025, 46(1): 183-190. doi: 10.13832/j.jnpe.2025.01.0183

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The Supercritical Water-cooled Reactor (SCWR) is one of the six reactor types recommended by the Generation IV International Forum (GIF) due to its high thermal efficiency and simple structural design. This paper provides an overview of the design requirements and key performance challenges for SCWR cladding material, including the general corrosion, stress corrosion and irradiation properties of commercial candidate cladding materials that have been tested a lot. Ferrite/Martensite (F/M) steel, austenitic stainless steel and nickel-based alloy all have some performance deficiencies. Finally, the recent research progress of novel SCWR candidate cladding materials is reviewed, including alumina-forming austenitic (AFA) steels, oxide dispersion strengthened (ODS) steels, functionally gradient materials and microstructure improvement techniques.

Holddown Force Analysis Model Research of Fuel Assembly Leaf Springs with Varied Stiffness

Jin Yuan, Peng Yao, Li Weicai, Cai Xiaorui, Zhang Yuxiang, Hu Haixiang

2025, 46(1): 191-198. doi: 10.13832/j.jnpe.2025.01.0191

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In order to conduct the holddown force analysis of the leaf spring with varied stiffness of PWR fuel assembly, this paper firstly analyzes the loading and unloading processes of the traditional leaf spring, and then comprehensively considers the influence factors such as temperature change, irradiation growth and irradiation relaxation. The holddown force change of the leaf spring with varied stiffness during different loading and unloading processes is further analyzed, and the load-deformation curve of the leaf spring with varied stiffness in the whole process is given. The analysis results show that the irradiated growth of fuel assembly increases the deformation of the leaf spring, but does not change the loading and unloading stiffness curves. The radiation relaxation process of the leaf spring with varied stiffness can be divided into two groups of leaf springs in parallel. Their radiation relaxation amounts are considered separately, and the stiffness curve characteristics remain unchanged. When the radiation growth of fuel assembly is greater than the radiation relaxation of the leaf spring, the plastic deformation of the leaf spring will increase.

Reasearch on Corrosion Resistance of Containment Steel Liner in Simulated Concrete Pore Solution

Guo Junying, Chen Shenggang, Li Zhongcheng, Liu Jinlong, Zhou Chuanbo

2025, 46(1): 199-208. doi: 10.13832/j.jnpe.2025.01.0199

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Aiming at the special P265GH low alloy steel (CSL) for nuclear power plant containment steel liner, the passivation and de-passivation effects of CSL in saturated Ca(OH)₂ solution were investigated using electrochemical and X-ray photoelectron spectroscopy (XPS) analysis methods, and the corrosion resistance of CSL was compared with that of low carbon steel (Q235B) and 304 stainless steel (304SS). The results show that, compared with Q235B and 304SS, CSL has higher passivation efficiency in simulated concrete pore solution, but the low Fe²⁺/Fe³⁺ ratio in the passivation film leads to poor corrosion resistance, and the critical chloride ion concentration of CSL (0.16~0.2 mol/L) is much smaller than that of Q235b (0.3 mol/L). The de-passivation effect of chloride ions mainly affects the electric double layer structure on the surface of the specimen, which leads to the increase of its effective capacitance and the sharp decrease of charge transfer resistance. The passive film formed on the surface of 304SS is composed of Fe and Cr oxide and hydroxide, which has higher corrosion resistance.

Wear Analysis of Thermal Sleeve of Reactor Control Rod Drive Mechanism Based on Archard Model

Zhang Yingnan, Peng Hang, Du Hua, Yu Tianda, Yu Zhiwei, Chen Xinan, Wu Hao, Zhang Jinqiang

2025, 46(1): 209-215. doi: 10.13832/j.jnpe.2025.01.0209

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Abstract:
The thermal sleeve assembly (referred to as thermal sleeve) is one of the main components of the control rod drive mechanism of the nuclear power plant reactor. Due to the impact of jet fluid on the lower end, the flange and the pressure housing are in contact and wear, which leads to the decrease of bearing capacity and impact resistance and affects the rod dropping function of the control rod. Based on Archard model, this paper puts forward the differential equations of structural wear characteristic time and thermal sleeve wear, decouples structural wear from operating condition parameters and material physical parameters, establishes a static wear analysis model suitable for thermal sleeve, and obtains the variation law of sedimentation with structural wear characteristic time. The results show that for the CRDM structure of HPR 1000, the maximum safety operation life of the thermal sleeve is at an inclination angle with 22.6°. By giving the contrast curve between the remaining operation life and sedimentation height, the evaluation method of thermal sleeve replacement and the treatment scheme of wear defects are provided for the operators of nuclear power plants.

Maximum Deflection Prediction Method for Plastic Large Deformation of High-energy Pipelines under Impact Load

Gu Ruijie, Dong Chengwu, He Jian, Sun Xiaodan, Yao Di

2025, 46(1): 216-224. doi: 10.13832/j.jnpe.2025.01.0216

Abstract(54) HTML (37) PDF(7)

Abstract:
When high-energy pipelines under the action of impact loads, the structure will subject to large deflection deformation. In order to predict the maximum deformation deflection of high-energy pipelines, a membrane force factor applicable to the structure was derived based on the membrane force factor method (MFM). A more convenient calculation method than traditional methods for mid-span deflection was established, and its accuracy was verified by comparing it with numerical simulation results and experiment results. The research results indicate that MFM has a high accuracy in predicting the deflection of pipeline structures; The magnitude of mid-span deflection is controlled by variable factors. The mid-span deflection increases with the increase of impact load and pipeline length, and decreases with the increase of plastic ultimate bending moment. The predicted deflection of pipelines with high degree of ellipticity is larger than the actual situation beacuse of MFM not taking into account the energy dissipated by section collapse. However, due to the support of high-energy fluid inside the pipeline, the collapse effect of the high-energy pipeline is not obvious. Therefore, this method can be used to predict the large deflection deformation of high-energy pipelines. For pipelines with high degree of collapse, the MFM prediction results are modified by combining the collapse model, which greatly improves the accuracy of the prediction method.

Structure Optimization of Bottom Nozzle for Flow Resistance- Filtration-Bearing Performance

Zhang Bo, Yuan Pan, Xiao Zhong, Guang Honghao, Chai Zhenhong, Duan Xin, Xin Yong, Zhu Fawen, Sun Kun, Li Baotong

2025, 46(1): 225-231. doi: 10.13832/j.jnpe.2025.01.0225

Abstract(44) HTML (23) PDF(8)

Abstract:
To improve the comprehensive performance of the bottom nozzle, including coolant pressure drop, debris filtration, and load-bearing of structure, a multi-objective optimization approach was proposed to optimize the key dimensional parameters of the bottom nozzle with the goal of coolant pressure drop and debris filtration efficiency. Based on the approach, size optimization was conducted using the hexagonal bottom nozzle as an example. The results showed that the pressure drop and filtration efficiency were optimized by 12.5% and 6.3%, respectively. Meanwhile, the maximum stress was reduced by 14.0%, confirming a significant enhancement of the comprehensive performance. Furthermore, the multi-objective optimization approach is universal for the structural optimization of various types of bottom nozzle to improve its comprehensive performance.

Analysis and Verification of Vibration Faults for Main Pump Motor

Wang Wanjin, Yin Ziyang, Qin Fujun, Su Xiaodong, Ma Wenbo

2025, 46(1): 232-237. doi: 10.13832/j.jnpe.2025.01.0232

Abstract(54) HTML (26) PDF(16)

Abstract:
The excessive vibration seriously affects the safe operation of the main pump. To ensure the safe operation of the main pump, root cause analysis technology is adopted to sort out the possible causes of motor vibration of the main pump through fault trees. And the fault mechanism is determined through analysis and searching. Meanwhile the cause was simulated by electromagnetic vibration coupling finite element analysis. Finally, by the adjustment of the position of the stator core and the control of key links, the problem of excessive vibration of the main pump motor was completely solved. After maintenance, the main pump motor has been running continuously for two years, and it has always met the technical requirements for operation, which provides a reference for dealing with the vibration problems of the same type of equipment in the future.

Multi-Objective Optimization Design of Self-sensing Rod Position Detector End Compensation Based on MOPSO Algorithm

Zhang Yixuan, Tang Jiankai, Luo Lingyan, Wu Hao, Tang Yuan, Wang Yiming, Xu Qiwei

2025, 46(1): 238-246. doi: 10.13832/j.jnpe.2025.01.0238

Abstract(60) HTML (20) PDF(9)

Abstract:
The self-sensing rod position detector utilizes the variation characteristics of the detecting coil inductance with the rod displacement to achieve continuous rod position measurement. However, the non-uniform magnetic field distribution at the coil ends leads to a nonlinear output signal, reducing the measurement accuracy at both ends. Therefore, this paper proposes a multi-objective optimization design method for winding stepped compensating coils at both ends of the detecting coils: ① Develop a mathematical model for the inductance of the end compensation coils; ② Adopt a multi-objective particle swarm optimization (MOPSO) algorithm for the optimization of the compensation coil structure; ③ Employ the entropy weight method and fuzzy comprehensive evaluation to assign weights to multiple optimization objectives and select a compromise optimal design, thereby efficiently selecting the optimal structural parameters of the compensation coils. By comparing the results before and after compensation through finite element simulation, it was found that the end compensation improved inductance sensitivity by 28.6% and reduced the maximum linear fitting error by 45.8%. Finally, prototype experiments were conducted, which indicated that the inductance sensitivity of the end-compensated detection coil is 0.18 mH/10 mm, with a maximum linear fitting error under 0.18 mH. These results confirm a measurement accuracy of 10 mm and validate the effectiveness of the multi-objective optimization design for the coil. This study provides a theoretical foundation for the optimization of its application in modular small reactor design.

Design of Terminal Sliding Mode Controller Based on RBF Neural Network for Underwater Transportation System

Yuan Zhanhang, MA Yuxiang, LI Yunhua

2025, 46(1): 247-253. doi: 10.13832/j.jnpe.2025.01.0247

Abstract(977) HTML (30) PDF(7)

Abstract:
The underwater transportation system will be affected by the uncertain nonlinearity of water and other external disturbance when transporting loads. Aiming at the operational control of underwater transportation system, a non-singular terminal sliding mode control method based on radial basis function (RBF) neural network is designed for the underwater transport process of nuclear power plant fuel assembly. Firstly, according to Newton's second law and Morison's equation, the kinetic differential equation of the system is established and its state-space equation is derived. Secondly, a non-singular terminal sliding mode controller is designed, and the unknown nonlinear effect is estimated by RBF neural network and compensated in the controller. The adaptive updating law of network weight is derived by Lyapunov stability theory. The Lyapunov stability theory proves that the proposed control strategy can achieve asymptotic convergence for unknown nonlinear estimation and finite-time convergence for given instruction tracking. Simulations is carried out for the two conditions of upgoing with load and downgoing without load respectively, and the results verify that the controller designed has good performance.

Study on the Influence Range of Gas Cloud Explosion Around Nuclear Power Plant

Li Hui, Xiong Min, Wang Jianhua, Ding Ziang

2025, 46(1): 254-258. doi: 10.13832/j.jnpe.2025.01.0254

Abstract(42) HTML (19) PDF(8)

Abstract:
Among all kinds of accident hazard factors around the nuclear power plant, the consequence of explosion shock wave overpressure is more serious. In the nuclear safety guidelines, the TNT equivalent method is used to evaluate the overpressure of explosion shock wave, and there is no calculation model related to gas cloud explosion. By analyzing the Trinitrotoluene (TNT) equivalent method, the Netherlands Organization for Applied Scientific Research (TNO) model and the Baker-Strehlow-Tang (BST) model, this paper evaluates the applicability of the three models. The results show that the calculation results of TNT equivalent method are conservative, while the BST model can accurately evaluate the impact range of explosion shock wave overpressure of natural gas pipelines around the nuclear power plant. According to the calculation results of the BST model, the shock wave overpressure caused by the accidental explosion of three typical pipelines has little effect on the nuclear power plant.

Research on the Impact Vector Evaluation Methods of Common Cause Failure in PSA of Nuclear Power Plants

Chen Yan, Zheng Jie, Li Chaojun, Han Zhi

2025, 46(1): 259-264. doi: 10.13832/j.jnpe.2025.01.0259

Abstract(586) HTML (19) PDF(6)

Abstract:
Common cause failure analysis is an important component of probabilistic safety analysis (PSA) for nuclear power plants. The evaluation of the impact vector of common cause failure events is the foundation of common cause failure analysis. At present, the quantitative mathematical method for calculating the impact vector of common cause failure events applicable to any system sizes has not been clearly established. Therefore, the definition and classification of impact vectors are studied in this paper, the mathematical expressions for evaluating impact vectors under incomplete common cause failure events are systematically demonstrated, as well as the mapping calculation method of impact vectors of common cause failure between different system sizes. Furthermore, the evaluation process of the total common cause failure impact vectors of any system sizes is proposed, and the total common cause failure impact vectors of the system is calculated by taking the system with four redundant components as an example. Considering the impact vectors in the target system and the impact vectors mapped from the external system to the target system, the total impact vectors in the case of common cause failure of the target system are (3.03, 23.26, 11.40, 3.72, 3.07). This method can calculate the impact vector of common cause failures with any system sizes, and could provide input data for common cause parameter estimation model.

Centrifugal Pump Fault Identification Technology Based on CEEMDAN-PCA-AC-CNN Model

Li Tongxi, Liu Zhilong, Luo Qian, Zeng Zhen, Wang Qinchao, Nie Changhua

2025, 46(1): 265-272. doi: 10.13832/j.jnpe.2025.01.0265

Abstract(48) HTML (21) PDF(12)

Abstract:
In order to ensure the long-term healthy and stable operation of centrifugal pump, on-line monitoring and fault identification is particularly important. Therefore, this paper presents a fault-identification model based on adaptive noise, with the methods of Complete Ensemble Empirical Mode Decomposition with Adaptive Noise (CEEMDAN), Principal Component Analysis (PCA) and Convolution Neural Network (CNN) combined with Autocorrelation (AC). Firstly, vibration signals are collected and decomposed by CEEMDAN. Then, we discriminate the obtained Intrinsic Mode Functions (IMF) fraction and eliminate the noise fraction to reconstruct the first round of denoised signals. After that, PCA is adopted to remove noise in the denoised signals, and the signals which have been filtered twice are then processed by AC and input to CNN to train the model. Through the experimental verification of a centrifugal pump fault, the results show that compared to traditional methods like double-layer noise reduction with CNN and CEEMD-wavelet denoising-AC-CNN, the model presented in this paper is more resistant to interference and has faster convergence speed. Also, it has the advantages of higher precision and better robustness. At the same order of magnitude, its precision can reach 97.9%.

Research on the Test Equipment for Operational Characteristics of Control Rod Drive Mechanism

He Jinyu, Zhao Yang, Deng Rui, Wang Fanyu, Zhang Naixin

2025, 46(1): 273-278. doi: 10.13832/j.jnpe.2025.01.0273

Abstract(43) HTML (25) PDF(10)

Abstract:
This study addresses the problems of traditional testing methods for Control Rod Drive Mechanisms (CRDM), such as low-level automation and simple testing functions. Based on traditional testing methods, this study uses Field-Programmable Gate Array (FPGA) as the core to improve testing efficiency. Combining with PC software, this study researches and designs a set of automation testing machine for the operational characteristics of CRDM. Compared to traditional testing methods, this equipment cannot only effectively control various test parameters and collect real-time data by test cases, but also generate detailed testing reports after software processing, providing reliable data support for subsequent analysis. Testing and simulation experiments confirm that this test equipment can meet the testing requirements for CRDM operational characteristics. In conclusion, this study provides a new solution for the testing automation of CRDM. It also indicates practical significance in enhancing testing efficiency.

Numerical Investigation of Aerosol Generation by Laser Decontamination and Aerosol Removal by Spray Based on OpenFOAM

Wang Yansong, Liang Hui, Hou Yuxuan, Chen Yuqi, Sun Zhongning

2025, 46(1): 279-288. doi: 10.13832/j.jnpe.2025.01.0279

Abstract(45) HTML (17) PDF(8)

Abstract:
During the maintenance and decommissioning of nuclear facilities, laser decontamination technology can be utilized to clean the radioactively contaminated surfaces inside the primary containment vessel. However, a significant amount of sub-micron radioactive aerosol particles will be generated during the laser decontamination operations. To prevent these radioactive substances from leaking into the environment, the containment spray system can be employed to remove these submicron aerosol praticles. To investigate the physicochemical properties of the aerosols produced by laser decontamination and the efficiency of aerosol removal by spray droplets, the laser decontamination experiments were firstly conducted to measure the aerosol characteristics such as aerosol generation rates, concentrations, and particle diameter distributions. Subsequently, a CFD model was developed and implemented into OpenFOAM to simulate both the aerosol generation by laser decontamination and aerosol removal by spray droplets using Euler-Lagrange approach, and the characteristics of aerosol generation, migration and diffusion and removal by spray in an enclosed space were simulated and analyzed. The simulation results demonstrate that aerosols within the spray region can be directly removed by interacting with the spray droplets, while those in the non-spraying region need to be entrained into the spray region with the airflow movement and then removed. Furthermore, it was observed that the larger the aerosol particle size, the higher its spray removal efficiency. The numerical simulation model established in this study can be served as a foundation and technical reference for future optimizations of laser decontamination and aerosol removal by spray in enclosed space.

Investigation on Static Characteristics of Self-lubricated Aerostatic Thrust Bearing of Organic Rankine Cycle Turbine Driven by Nuclear Heat

Du Qiuwan, Liu Ming, Yan Xiao, Zhang Zhao, Zhang Cheng, Liu Wenxing, Yuan Dewen

2025, 46(1): 289-296. doi: 10.13832/j.jnpe.2025.01.0289

Abstract(1055) HTML (27) PDF(12)

Abstract:
The nuclear power system of micro reactor with organic Rankine cycle is an effective approach to achieve multiple energy supply in special scenarios. The utilization of organic working fluid lubricated air bearing in organic Rankine cycle turbine generator set can eliminate the complex lubrication oil system, which is an effective measure to improve the compactness of the system. In order to investigate the static characteristics of organic working fluid lubricated air bearing, this paper focuses on the organic working fluid aerostatic thrust bearing, and the influence of inlet pressure, inlet temperature, and rotation speed on loading capacity and mass flow rate is discussed in detail. In addition, the effects of orifice type, axial clearance, and lubricating working fluid are further analyzed. The results show that under different operating conditions, the film pressure distribution characteristics remain basically consistent. The pressure gradually decreases radially from the orifice area to the outlet. Under the conditions with high inlet pressure, high inlet temperature and low rotation speed, the loading capacity can be kept at a high level. Compared to the thrust bearing with small orifice, bearings with slots have higher loading capacity and lower mass flow rate under the same inlet area. As the axial clearance increases, the film pressure significantly decreases, the loading capacity decreases, and the mass flow rate increases. The lubricating working fluid has a certain influence on the static characteristics. In most conditions, bearing lubricated by R134a performs the best static characteristics, followed by cyclopentane bearing, and R245fa bearing presents the relatively worst performance. The results in this paper can provide vital reference for the design optimization of organic working fluid lubricating bearings.

Current Issue 2025, Volume 46, Issue 1

Current Issue
2025, Volume 46, Issue 1