核动力工程

Study on Monte Carlo Particle Transport Method and Application

Deng Li, Li Gang, Zhang Baoyin, Li Rui, Zhang Lingyu, Fu Yuanguang, Liu Peng, Ma Yan, Shi Dunfu, Wang Xin, Qin Guiming

2025, 46(3): 1-17. doi: 10.13832/j.jnpe.2025.02.0062

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Abstract:
Monte Carlo (MC) particle transport methodology incorporates stochastic principles derived from probability theory and mathematical statistics to establish computational frameworks. This approach facilitates the numerical resolution of complex particle transport phenomena in nuclear systems. Over the course of seven decades of development, MC particle transport theory and algorithms have reached a high level of technical maturity. This has resulted in the development of several specialized software packages, which are widely applied in fields such as nuclear radiation shielding, reactor core criticality safety analysis, nuclear detection, and radiation medicine. This study commences by establishing the theoretical framework underlying MC particle transport methodologies. Through rigorous mathematical derivation, we present the neutron flux density formulation developed via MC simulations for addressing integral-form neutron transport equations, coupled with analytical frameworks for determining associated response parameters. It also outlines the classification of deterministic approaches for solving transport equations. The study reviews the historical development and computational application of MC particle transport methods, while summarizing significant software developed domestically and internationally. Furthermore, it examines recent advancements in utilizing graphics processing unit (GPU) technology to develop MC particle transport software, highlighting current research directions and progress in this field. This paper provides a comprehensive review of recent advancements in MC particle transport methodologies and associated software, with a specific focus on key features and capabilities of the independently developed J Monte Carlo transport (JMCT) software.

Experimental Study on Start-up Characteristics of Arterial Sodium Heat Pipe with High Ratio of Length to Diameter

Xu Jun, Yu Hongxing, Deng Jian, Liu Yu, Zhang Muhao, Xia Xiaohui

2025, 46(3): 18-23. doi: 10.13832/j.jnpe.2024.10.0029

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To support the research and development of heat pipe reactors, this study designed and constructed a high-temperature compressed air cooling experimental platform to investigate the startup characteristics of high length-diameter arterial sodium heat pipes. The analysis results demonstrate the following: ① In the initial stage of the heat pipe startup process, high-temperature compressed air elevates the temperature of the condensation section, which facilitates the formation of a continuous flow of sodium vapor within the heat pipe, thereby accelerating the cold-state startup speed of the heat pipe; ② During the startup process, preheating of the condensation section enhances the temperature of the sodium vapor, effectively preventing the occurrence of sonic limit phenomenon, and consequently, increases the probability of successful heat pipe startup. The results in this paper provide data and theoretical support for the optimization of the cold start-up mode of the arterial sodium heat pipe with large length-diameter ratio.

Steady State Thermal Surrogate Model of TOPAZ-II Reactor Core

Liao Ruian, Wang Xuesong, Qi Lin, Zhang Dalin, Tian Wenxi, Su Guanghui, Qiu Suizheng

2025, 46(3): 24-33. doi: 10.13832/j.jnpe.2024.070011

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The TOPAZ-Ⅱ reactor, a space nuclear reactor power designed by the former Soviet Union, uses sodium-potassium alloy (NaK-78) as coolant and adopts thermionic conversion power generation principle to provide power for the load. To quickly and accurately calculate the steady state thermal parameters of the core, a steady state high-precision thermal surrogate model of the core was established. This study firstly used Fluent to carry out thermal calculation of core steady state, with grid node temperatures along the central longitudinal cross-section selected as sample data. Then the main features in samples were extracted by Proper Orthogonal Decomposition (POD) method, and the top 10 modes were retained based on 99.999% energy proportion for model order reduction. Finally, through Back Propagation (BP) neural network, the steady state thermal surrogate model of core was established and was compared and validated with Fluent. The results show that the maximum error of the surrogate model in calculating the temperature at grid nodes is 9.95 K, the relative error is less than 1% and the calculation time is less than 1 s. Taking the outlet temperature of the hottest coolant channel as a reference, the flow-power percentage ratio of the coolant to maintain the single-phase working state should be greater than 0.35 calculated by the thermal surrogate model. Therefore, the thermal surrogate model established in this paper can quickly and accurately calculate the steady state thermal parameters of the core, achieve simulation prediction of the core, and provide certain reference for the thermal safety analysis of the core.

Analysis and Experimental Study on Flow Field in RPV Head Plenum

Chen Yongchao, Wei Xingfang, Liu Yanwu, Fang Jian, Ran Xiaobing

2025, 46(3): 34-41. doi: 10.13832/j.jnpe.2024.070004

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This paper aims to investigate the flow characteristics in the RPV head plenum region of HPR1000 Pressurized Water Reactor (PWR), so as to provide support for for addressing the thermal sleeve wear issues in operational CPR1000 PWR nuclear power plants and optimizing the RPV head plenum structure of HPR1000. This study adopts Computational Fluid Dynamics (CFD) methods to conduct numerical simulations of the RPV head plenum region, and hydraulic simulation experiments on the RPV head plenum are carried out to obtain the flow distribution and hydraulic characteristics of key areas within the RPV head plenum. The theoretical analysis and experimental results show that: The deviation between the CFD results of the key areas in the RPV head plenum and the measured lateral velocity is within 10%. Under normal operating conditions, the overall flow velocity within the RPV head plenum is relatively low, but higher velocities are observed near the RPV head nozzles and nearby surface. The coolant water in RPV All fluid in the upper head plenum enters the upper plenum through the control rod guide tube (CRGT) flow holes without reverse flow, validating the effectiveness of the HPR1000 ‘Cold RPV head’ design. Additionally, a vortex with higher flow velocity is present near the thermal sleeve bell mouth in the central region than in the peripheral area of the RPV head plenum, resulting in more severe fluid impact and wear on the thermal sleeves.

Study on Temperature Fluctuation Characteristics of Lead- Bismuth Eutectic in Triple Jet Model Based on Large-eddy Simulation Method

Guo Chao, Xu Jiangming, Liu Songtao, Miao Yiran

2025, 46(3): 42-50. doi: 10.13832/j.jnpe.2024.060038

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To investigate the influence of flow parameters on lead-bismuth eutectic temperature fluctuations, numerical simulations were performed to study the temperature fluctuation characteristics of LBE in a triple jet model. Firstly, the temperature fluctuation phenomenon of sodium fluid was numerically simulated based on different turbulence models. The calculation results show that the large eddy simulation method can accurately analyze the temperature fluctuation phenomenon and is suitable for the numerical analysis of temperature fluctuation of liquid metal. Subsequently, the large eddy simulation method was employed to numerically calculate LBE temperature fluctuations in the triple jet model, and the time-dependent temperature variations at monitoring points were obtained. Finally, comparisons were made regarding temperature fluctuation amplitudes and frequencies at the monitoring point above the central outlet under different velocity ratios and flow velocities, and the effects of fluid velocity and velocity ratio on temperature fluctuation characteristics at various monitoring points were analyzed. The results show that both the amplitude and frequency of temperature fluctuations increase with flow velocity, primarily due to enhanced turbulent effects and increased flow disorder at higher velocities. The obtained LBE temperature fluctuation characteristics can provide references for subsequent studies on temperature fluctuations in lead-bismuth cooled fast reactors.

Experimental Study of Geyser Boiling in High-temperature Sodium Heat Pipe at Inclined Angle Condition

Yang Siyuan, Ma Yugao, Wen Qinglong, Wen Shuang, Ding Shuhua, He Linfeng, Yuan Bo

2025, 46(3): 51-60. doi: 10.13832/j.jnpe.2024.060020

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Abstract:
To study the geyser boiling phenomenon in the start-up process of high-temperature alkali metal heat pipe and provide reference operating conditions for the safe operation of the heat pipe reactor, sodium metal was used as the working medium to investigate the influence factors and mechanism of geyser boiling during the start-up process of the heat pipe. The results show that the heating power and inclination angle of the heat pipe have important effects on geyser boiling. Under the condition of 90° inclination angle, the heating power increases from 600 W to 750 W, the geyser boiling period varies from 29 s to 736 s, and the temperature amplitude ranges from 18℃ to 35℃. Geyser boiling is easy to occur under medium heating power conditions but will not happen when the inclination of the heat pipe is 0°. With the increase of inclination angle, the heating power of the start and stop of the geyser boiling decreases. When the inclination angle is 45°, 60°, 90° respectively, the geyser boiling starts at 250, 200, 150 W and stops at 600, 450, 350 W. The geyser boiling period varies under the conditions of the same heating power and different inclination angles, but the temperature amplitude changes little. The geyser boiling intensity decreases and the power range of geyser boiling advances with the shortening of the condensing section. The results of this study laid a foundation for further research on the geyser boiling mechanism of alkali metal heat pipes, and provide important data and theoretical support for the design optimization of alkali metal heat pipes and the safe operation of heat pipe reactors.

Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor

Chen Sen, Li Huaqi, Li Da, Chen Lixin, Tian Xiaoyan, Shi Leitai, Luo Xiaofei, Zhu Lei

2025, 46(3): 61-67. doi: 10.13832/j.jnpe.2024.060019

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Xi'an Pulsed Reactor (XAPR) adopts uranium-zirconium hydride fuel element, which has a good inherent safety. However, with the long-term operation of XAPR, severe accidents may occur in the core, leading to the damage of fuel element cladding, so it is necessary to carry out analysis and research on the beyond-design-basis accidents for XAPR. In this paper, based on the ISAA code, an integrated analysis code for XAPR beyond-design-basis accidents was developed by adding physical property model, fuel oxidation model, fuel rod mechanics model and point kinetics model. The accuracy and applicability of each model were validated. Besides, based on the developed code, the steady-state operation condition and large break loss of coolant accident condition were calculated and analyzed respectively, and the results were in good agreement with the literature values. The results showed that the models developed in this paper were suitable for the simulation and analysis of XAPR and could be used for the subsequent beyond-design-basis accident calculation and analysis of XAPR.

Research on Thermoelectric Coupling Characteristics of a 100 kW Silent Heat Pipe Cooled Reactor Based on Finite Element Method

Tang Simiao, Lian Qiang, Zhu Longxiang, Zhang Luteng, Ma Zaiyong

2025, 46(3): 68-77. doi: 10.13832/j.jnpe.2024.060015

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The silent heat pipe reactor adopts a energy transmission and thermoelectric conversion system that couples high-temperature heat pipes with thermoelectric power generation. It is a preferred reactor type of portable small nuclear power source in various fields such as sea, land, air, and space in the future due to its passive safety, high reliability and ultra silence. Based on the multi physical field coupling analysis platform COMSOL Multiphysics, this paper establishes a quarter model of the whole system of the heat pipe reactor according to the design scheme of a 100-kilowatt level silent heat pipe reactor, including fuel rods, core matrix, heat pipes, reflectors, control rods, sliding reflectors, thermoelectric generations and other systems. Steady-state operating conditions, single heat pipe failure conditions, and single-row thermoelectric system unloading conditions are analyzed to investigate system thermoelectric coupling characteristics. The research results indicate that due to the temperature flattening characteristics of the core matrix and the thermoelectric system matrix, the failure of a single heat pipe will not have a significant impact on the operation of the reactor and the output power of the thermoelectric system. When the local thermoelectric system unloading accident occurs in a heat pipe reactor, the core temperature will increase due to the decrease in the heat transfer capacity of the thermoelectric system. The thermoelectric system that has not been unloaded can still work normally, ensuring effective electrical energy output.

Experimental Study on Minimum Attenuation Particle Size of Submicron Aerosol Spray Removal

Tang Jiaxuan, Liu Zhuo, Zhang Luteng, Pan Liangming, Yang Yang, Li Jialong, Gao Li, Yuan Yidan

2025, 46(3): 78-85. doi: 10.13832/j.jnpe.2024.060002

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The spray removal mechanism of submicron aerosol particles exhibits significantly lower removal efficiency. Investigating the influence of spray characteristics on the minimum attenuation particle size of submicron aerosols is of paramount importance for severe accident management. This study, based on a self-constructed experimental platform for spray removal of aerosols, conducted experimental research on various dispersed submicron aerosol particles, and elucidated the relationship between spray characteristics and the minimum attenuation particle size of aerosols. The research findings indicate that the minimum attenuation particle size for submicron aerosol spray removal is concentrated in the particle size range of 0.3~0.5 μm. As the spray flow rate increases, the droplet size decreases, and consequently, the minimum attenuation particle size also decreases. For multi-component aerosol particles under the same spray characteristics, the minimum attenuation particle size varies and is related to the median particle size. Therefore, this study can be utilized to predict the minimum attenuation particle size of multi-component submicron aerosols under different spray characteristics, thus providing reliable data support for the management of severe accidents.

Nonlinear Reduced-order Analysis of Three-dimensional Thermal Stratification in the Upper Plenum of Lead-bismuth Cooled Fast Reactor Based on Graph Neural Network

Zeng Fulin, Zhao Pengcheng, Li Lingli, Liu Zijing, Li Wei

2025, 46(3): 86-94. doi: 10.13832/j.jnpe.2024.050044

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The thermal stratification phenomenon in the upper plenum of the lead-bismuth fast reactor has significant implications for the safety of internal components and the ability to remove residual heat. This paper focuses on analyzing this phenomenon. Firstly, high-precision full-order snapshots of the thermal stratification in the upper plenum of the lead-bismuth fast reactor are obtained using the computational fluid dynamics software FLUENT. Then, the graph neural network (GNN) is employed to construct a graph autoencoder (GAE) for nonlinear order reduction of the snapshots, and the reduced reconstruction results are compared with the linear reduction results obtained using Proper orthogonal decomposition (POD). Finally, a multilayer perceptron is used for online state recognition and predictive analysis of the thermal stratification snapshots. The research results demonstrate that the graph neural network, due to its high level of nonlinearity and its inherent advantage in nonlinear order reduction of large-scale CFD data, achieves comparable first-order modal reconstruction accuracy to that of POD with 30~50 basis functions. During the online process, the feature recognition and prediction of thermal stratification snapshots can be completed within a duration of 472 ms, with accuracy similar to the reconstruction accuracy. The related research results can provide a new analytical method support for the evolution mechanism analysis and consequence prediction of thermal stratification phenomenon in lead-bismuth fast reactors.

Numerical Simulation Research on Natural Circulation Flow Characteristics in Reactor Coolant System

Zhang Mingqian, Lin Run, Li Zhenguang

2025, 46(3): 95-102. doi: 10.13832/j.jnpe.2024.050038

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A high-fidelity three-dimensional numerical model of a three-loop reactor coolant system, including the reactor, steam generator, reactor coolant pump, and main pipelines, was established using computational fluid dynamics (CFD) software. A three-dimensional numerical analysis of system-level thermal-hydraulic phenomena under low-power operating conditions was conducted, and the coolant temperatures in different regions were obtained. Comparison with measured data from the operating nuclear power plant verified the rationality of the numerical model. The analysis results show that the natural circulation flow rate at this power level is 4.5% of the full power operation flow rate, and the core outlet temperature is stable, which can ensure the residual core heat being effectively removed. Local thermal convection enhances the mixing of coolant in different loops. The phenomenon of thermal stratification in the reactor pressure vessel head shows that the measured temperature value of the detecting position is not the highest temperature in this area. Rotational flow is generated at the reactor coolant pump outlet, with higher tangential velocities near the main pipe wall and localized convection in the central region. This research can further improve designers' understanding of complex system-level three-dimensional thermal-hydraulic phenomena in nuclear power plants.

Design and Optimization of Cascaded Thermoelectric Generators Based on Heat Pipe Reactor Applications

Tang Simiao, Lian Qiang, Zhu Longxiang, Zhang Luteng, Sun Wan, Ma Zaiyong, Pan Liangming

2025, 46(3): 103-110. doi: 10.13832/j.jnpe.2024.050037

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Based on the application background of silent heat pipe cooled reactor (heat pipe reactor), combined with the geometric structure and thermal boundary conditions of the energy conversion system of heat pipe reactor, the two-stage and three-stage cascaded thermoelectric generators (TEG) are designed and optimized using finite element method. The conversion efficiency and output power of cascaded TEG under different heat flux density conditions are studied. The research results indicate that the conversion efficiency can be effectively improved by optimizing the structure of different materials inside the PN legs of multi-stage cascaded TEG. For the two-stage TEG composed of skutterudite materials and half-heusler materials, the conversion efficiency can reach 15.05% under the hot-end boundary condition with a heat flux of 162.5 kW/m². For the three-stage TEG composed of bismuth telluride, skutterudite and half-heusler, the conversion efficiency can reach 15.13% under the hot end boundary condition with a heat flux of 90 kW/m².

Study on the Dynamic Melting Characteristics of Coating Exposed to Non-eutectic Corium in Nuclear Reactor

Gong Tao, Zhang Luteng, Ma Zaiyong, Sun Wan, Zhu Longxiang, Lian Qiang, Tang Simiao, Pan Liangming

2025, 46(3): 111-117. doi: 10.13832/j.jnpe.2024.050035

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To investigate the melting characteristics of the corium and coating materials in the reactor, the liquid NaNO₃-KNO₃ mixture is used as the corium and solid KNO₃ is used as the coating material to conduct experimental research on the dynamic melting characteristics of the coating. The results show that the melting process can be divided into three stages: corium quench solidification and remelting, diffusive melting of coating components and melting termination. In the process of melting, component diffusion is observed at the phase interface, and the instantaneous liquidus temperature at the phase interface is higher than that of the molten pool, resulting in the coating melting even when the phase interface temperature is lower than the melting point of the coating. On the basis of experiments, a melting characteristic model is established based on the heat and mass transfer relationship. The maximum error of component concentration and phase interface temperature is 4.5% and 11.3%, which proves the accuracy of the model.

Analysis of Heat Transfer Characteristics of Steam Generator with Axial Economizer Based on RELAP5

Huang Zhongyuan, Wang Xiaoding, Li Zhenzhong, Liu Haidong, Chen Deqi

2025, 46(3): 118-124. doi: 10.13832/j.jnpe.2024.050032

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As a crucial equipment in pressurized water reactor nuclear power plants, the vertical natural-circulation steam generator requires enhanced heat transfer performance, which is vital to the economy of the entire power plant. This study selects the steam generator of AP1000 as the research object and employs the RELAP5 system analysis code to calculate and analyze both the conventional steam generator and the steam generator with axial economizer. The heat transfer mechanism of the steam generator with axial economizer is studied, with a focus on the effects of different heights of the divider plate and the recirculated water distribution ratio on the heat transfer characteristics. The results show that the steam generator with axial economizer can significantly increase the heat transfer temperature difference between the primary and secondary sides, thereby effectively enhancing the overall heat transfer efficiency. In addition, the study finds that increasing the height of the divider plate can improve the heat transfer capacity to a certain extent, and there is an optimal height at which the heat transfer power reaches its peak. Moreover, reducing the recirculated water distribution ratio helps to increase the heat transfer temperature difference, further improving the heat transfer performance. This research provides reference for the engineering analysis and design of the steam generator with axial economizer.

Application and Experimental Study of RBF Neural Network Algorithm in Flow-Induced Vibration of Pipelines

Wang Binbin

2025, 46(3): 125-130. doi: 10.13832/j.jnpe.2024.090042

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To address the issue of time-consuming traditional fluid-structure interaction methods, which make it difficult for nuclear power plant pipeline designers to perform targeted vibration reduction calculations during the design phase, this study adopts a data-driven radial basis function (RBF) neural network algorithm for pipeline flow-induced vibration analysis. This method can quantitatively calculate the pipeline flow-induced vibration in a short period of time by training on a large amount of load data of throttling fittings in the database. Compared to traditional fluid-structure interaction methods, it greatly improves the computational efficiency of pipeline flow-induced vibration. To validate the calculation results, experimental studies were conducted on on ball valves at different opening degrees and elbow pipes. Due to the presence of external structural vibrations such as pump excitation, experiments have found that when the flow-induced vibration dominates the total vibration, the simulation results are relatively close to the experimental results. When external structural vibration dominates the total vibration, the simulation and experimental results are of the same magnitude and have consistent variation patterns. The results demonstrate that the data-driven based RBF neural network method is reliable and effective for analyzing flow-induced vibrations in pipelines.

Experimental Study on Quench Temperature Characteristics During Reflooding in an Annular Channel

Wang Jinyu, Wang Jun, Zan Yuanfeng

2025, 46(3): 131-136. doi: 10.13832/j.jnpe.2024.070064

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Quench temperature is a key indicator of quenching in the process of core reflooding, and it characterizes the initial wall temperature at the onset of quenching on the fuel element surface. Study of quench temperature is helpful to understand quench mechanism and develop quench models. Based on an experimental study of characteristics of quench temperature in a vertical annual channel, the influences of initial wall temperature, inlet coolant temperature, inlet mass flow rate, and heating power density on quench temperature were investigated. The results show that the quench temperature increases with the increase of initial wall temperature and heating power and the decrease of inlet coolant temperature. Heating power weakens the effect of inlet coolant temperature on quench temperature. At lower heating power, quench temperature increases with the increase of inlet mass flow rate; at higher heating power, quench temperature decreases with the increase of mass flow rate.

Effect of Coupled Electromagnetic Treatment on the Thermal and Mechanical Properties of Alloy 690

Zhu Yonghui, Fu Shuai, Chen Haohan, Huang Kunlan

2025, 46(3): 137-146. doi: 10.13832/j.jnpe.2024.060022

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To solve the problem that the heat exchange efficiency of alloy 690 heat transfer tube is difficult to reach the design value, this paper adopts the method of combining simulation and experiment, employs the coupled electromagnetic treatment to study the thermal conductivity and mechanical properties of alloy 690 heat transfer tube by applying electric and magnetic fields with different parameters. The results show that when the applied electromagnetic field parameters are 1.5 V-1.5 T, the thermal conductivity of alloy 690 heat transfer tube is increased by 19.6%, and the tensile strength and Vickers hardness are also increased by 6.8% and 4.3%, respectively. The thermal stress calculated by simulation is an order of magnitude larger than the modified Peierls stress, which shows that the coupled electromagnetic treatment can effectively drive the internal dislocation movement of alloy 690. EDS results showed that the coupled electromagnetic energy field could promote the precipitation of intergranular carbides (M₂₃C₆), thereby improving the thermal conductivity of alloy 690 heat transfer tubes. In this paper, the feasibility of the coupled electromagnetic treatment to improve the thermal conductivity of alloy 690 heat transfer tube is fully verified, and the heat exchange efficiency of alloy 690 heat transfer tube can be effectively improved.

Experiment Study on Rod Drop Performance of CF Series Fuel Assembly

Tian Xuelian, Zhang Ziyang, Chen Liangbin, Yu Qinglin, Jiang Yu, Guo Sibei, Nie Changhua, Zhuo Wenbin

2025, 46(3): 147-151. doi: 10.13832/j.jnpe.2024.060010

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The design and improvement characteristics of CF series fuel assemblies with completely independent intellectual property rights in China are introduced in this paper. The performance parameters of CF series fuel assemblies such as rod drop time, rod drop velocity and rod drop impact force under different conditions are obtained through 1∶1 cold and hot out-of-core simulation tests, and the differences in rod drop performance between CF2 and CF3 fuel assemblies, as well as between CF2S and CF3S fuel assemblies are compared. The experiment results show that CF3 series fuel assemblies have a longer rod drop time and a lower rod drop impact force compared with CF2 series. The improvement of guide tube structure has a greater influence on rod drop buffer time. The experimental results validate the effectiveness of the design improvements and provide support for the practical reactor application and further development of CF series fuel assemblies.

Research on Stress Corrosion Behavior of 316NG Steels in Liquid Lead-Bismuth Eutectic at 560℃

Zhang Pingping, Gong Bin, Zhao Yongfu, Gao Jun, Deng Ping, Wu Zongpei

2025, 46(3): 152-159. doi: 10.13832/j.jnpe.2024.050039

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To precisely obtain the compatibility between austenitic stainless steel and liquid metal for the service performance evaluation, this study used a high-temperature liquid lead-bismuth eutectic slow strain rate tensile test device to investigate the stress corrosion behavior of 316NG stainless steel in 560℃ Lead-Bismuth eutectic (LBE) under three dissolved oxygen concentrations: low oxygen concentration (<7×10⁻⁸%), medium oxygen concentration (2×10⁻⁶%~2×10⁻⁷%), and saturated oxygen concentration (1.0×10⁻³%~3×10⁻⁴%). The results show that, compared with argon environment, 316NG steel exhibited stress corrosion in the LBE environment. As the dissolved oxygen concentration decreased, the crack depth increased, the fracture elongation decreased, and the stress corrosion effect became more pronounced. At low and medium oxygen concentrations, the fracture mode of 316NG steel was a mixed fracture mode which involves surface intergranular cracking and matrix ductile fracture. However, at saturated oxygen concentration, the fracture mode primarily consisted of matrix ductile fracture. The main cause of stress corrosion of 316NG steel was the inability of the specimen’s surface and crack tip to form a continuous and stable oxide film. This film is essential in preventing LBE from corroding the matrix. The interaction of LBE with the steel promotes the growth of intergranular cracks. Ultimately, it leads to premature failure of the specimen.

Numerical Simulation Study on Static Buckling Behavior of Bimetallic Spacer Grids

Zhou Ming, Xiao Zhong, Ren Quanyao, Qin Mian, He Rui, Pu Zengping, Li Zhengyang

2025, 46(3): 160-165. doi: 10.13832/j.jnpe.2024.050031

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The buckling behavior of spacer grids can seriously affect the safety of fuel assemblies. To understand the static buckling evolution characteristics of spacer grids, this study established a numerical model of bimetallic spacer grids using the Finite Element Analysis (FEA) method. The model was validated by comparing computational results with experimental data, and the influence of initial clamping force on grid strength was studied. The calculation results show that the critical buckling load obtained by numerical simulation is in good agreement with the experimental results. The primary cause of spacer grid buckling is the plastic deformation occurring in straps adjacent to the central transverse row, which subsequently propagates throughout the entire structure. The decrease of spring clamping force has little effects on the critical buckling strength of the spacer grid, but has great effects on the response after buckling. The numerical analysis method in this paper can predict the static critical buckling load of bimetallic spacer grid and provide support for the structural design of new spacer grid.

Research on the Control Technology of Automatic Start-up for PWR Nuclear Power Plant

Zhang Qi, Zhang Nan, Sun Peiwei, Zhang Ruiping, Yu Wenhao, Wei Xinyu

2025, 46(3): 166-172. doi: 10.13832/j.jnpe.2024.070035

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To improve the automation level of PWR nuclear power plant units during startup, reduce the work intensity of reactor operators, shorten the start-up time, and improve the correctness and standardization of the unit start-up, this study puts forward a control technology suitable for automatic startup of nuclear power plant, which is based on the characteristics of typical PWR nuclear power plant unit system, operation management process and control requirements of automatic startup. Based on the analysis of the applicable control range, operation breakpoints, sequence control and analog control of the automatic start-up control system of PWR, an architecture for the automatic start-up control system of nuclear power plant is established, including the architecture design, the functions and design contents of each level and the interactive interface design between levels. At the same time, a typical PWR nuclear power plant automatic start-up simulation platform is established, and the automatic start-up control system is designed by taking the start-up process of nuclear power plant operation mode Ⅲ as an example, and the proposed technical scheme is simulated and verified. The simulation results show that the automatic start-up control system design can realize the automatic start-up of nuclear power plant mode Ⅲ, reducing the operation steps and workload of operators. The designed automatic start-up control system architecture in this study can provide a reference for the application of automatic start-up control system for nuclear power plant, and is of great significance for improving the automation level of start-up process of nuclear power plant units.

Study on the Impact of Different Behavioral Levels in Digital Main Control Rooms of Nuclear Power Plants on Operators’ Inhibition of Return Effect

Zheng Tengjiao, Xu Yunlong, Hou Jie, Duan Pengfei, Chen Shuai

2025, 46(3): 173-178. doi: 10.13832/j.jnpe.2024.070029

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This study aims to explore the effects of different behavior levels (skill-based and rule-based) on the Inhibition of Return (IOR) of operators in a digital main control room of nuclear power plant, as well as the relationship between workload and IOR effect. The typical accident scenarios were simulated through a simulation platform, and the IOR effects and workload of operators with different behavior levels were compared and analyzed. The study found that different behavior levels have a significant impact on the IOR effect of operators. Skill-based behavior is effective in avoiding interference from non-target information during nuclear power plant status monitoring; while rule-based behavior is easily distracted by irrelevant information, with workload showing a positive correlation with IOR effect. This research can provide theoretical and data support for training nuclear power plant operators and personnel function allocation.

Study on the Nonlinear Dynamic Model of a Single Fuel Assembly Based on the Bouc-Wen Hysteresis Model

Yang Tao, Zhang Yixiong, Cai Fengchun, Qi Huanhuan, Huang Qian, Shen Pingchuan

2025, 46(3): 179-185. doi: 10.13832/j.jnpe.2024.050041

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To simulate the nonlinear phenomena during lateral vibration of a single fuel assembly, this study introduces a Bouc-Wen hysteresis model with two-stage stiffness degradation based on the linear single-beam model. Using the finite element discretization method, we established a nonlinear dynamic model for lateral vibration of the single fuel assembly. Furthermore, combined with partial mechanical property test data, the multi-objective genetic optimization algorithm (NSGA-Ⅱ) was employed to identify unknown parameters of the model. The comparison between computational results and experimental data reveals the following: The root mean square error (RMSE) of normalized displacement during the lateral loading-unloading process is 0.027. Under five different initial conditions, the relative errors of free vibration response frequencies are less than 6%, and the relative errors of damping ratios are less than 3%. These results demonstrate that the nonlinear dynamic model effectively captures key phenomena observed in fuel assemblies, including stiffness degradation, frequency reduction, and damping ratio increase with growing lateral displacement. This study provides a methodological reference for the accurate simulation of lateral mechanical response characteristics in fuel assemblies. Furthermore, the developed nonlinear model exhibits broader applicability and may facilitate the exploration of design margins for fuel assemblies under large-deformation scenarios.

Numerical Analysis of the Performance of Lead-Bismuth Centrifugal Pump with High Temperature Closed Circuit under All Operating Conditions

Luo Changyu, Li Yibin, Ma Wensheng, Yang Youchao, Niu Teng

2025, 46(3): 186-194. doi: 10.13832/j.jnpe.2024.070033

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To investigate the thermal-hydraulic performance of lead-bismuth centrifugal pump in a closed-loop system transporting 400℃ liquid Lead-Bismuth eutectic (LBE), a joint simplified modeling approach that integrates the lead-bismuth circulation tank, inlet/outlet pipelines, and centrifugal pump was employed. Utilizing the shear stress transport (SST) k-ω turbulence model, flow characteristics within the pump under three distinct flow rate conditions were systematically analyzed. The study revealed that vortices of varying intensities in the impeller flow channels were associated with fluid force imbalances, with Coriolis forces maintaining dominant influence throughout the LBE transport process. Local entropy production rate (EPR) was primarily concentrated at the leading edge of impeller blades and rotor-stator interface regions, exhibiting a decreasing trend with increasing flow rates. Pressure signal frequencies in the impeller and guide vane channels demonstrated periodic alternations between 93.33 Hz and 116.67 Hz, while wavelet signal intensity became more pronounced near the rotor-stator interface. These findings provide important references for optimizing design and performance evaluation of centrifugal pumps in lead-bismuth reactor systems.

Feasibility Analysis on Digital Transformation of Process Instrument Protection Measurement Cabinets in Nuclear Power Plants

Guan Yue

2025, 46(3): 195-201. doi: 10.13832/j.jnpe.2024.060027

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This article provides a detailed account of the background, reasons, scope and principles, as well as the design of the new system for the safety-level nuclear island process instrument protection measurement cabinet (KRG-P) renovation of Units 1 and 2 of Qinshan Phase II Nuclear Power Plant. To address the issues of equipment supply disruption and bottlenecks in the old system, and in light of the characteristics of digital transformation of safety-level distributed control system (DCS) of in-service nuclear power plant, the reliability of the equipment and system was comprehensively enhanced through centralized arrangement of purchased components, optimization of quality bits and default values after digitalization, identification of module failure risks, and optimization of the expected transient non-trip cabinet. Through the independent research and development of safety-level hand controllers, it ensures that the safety-level DCS can stabilize the unit status in the event of a severe failure. The above optimization measures can effectively reduce the risk of the implementation of digital transformation of safety-level DCS in nuclear power plants, and also accumulate technical experience for the subsequent digital and domestic transformation of safety-level DCS in nuclear power plants.

Simulation and Experimental Study on Separation Characteristics of New Start-up Separator

Zhang Ziwei, Chen Chen, Meng Zhaoming, Cao An, Dong Chuanchang

2025, 46(3): 202-212. doi: 10.13832/j.jnpe.2024.060021

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To improve the separation efficiency and optimize the performance of the start-up separator, this paper designs a new type of startup separator with a corrugated plate separation structure as the research object and conducts simulation and experimental studies on the separation characteristics of the new separator. The software Fluent is employed for numerical simulation calculations, using a combined approach of overall simulation and local equivalence. The Eulerian two-phase flow model is utilized to simulate the separation efficiency of the gas-liquid two-phase fluid inside the start-up separator, with the corrugated plate region replaced equivalently using a porous media model in the overall simulation process. The separation characteristics of the corrugated plate in the start-up separator are investigated, and the influence of inlet gas-liquid flow velocity on separation efficiency is analyzed. The feasibility of the simulation method is verified through cold-state experiments. The results demonstrate that the combination of overall simulation with local equivalence simulation is a viable and effective approach. The introduction of a corrugated plate structure in the novel startup separator significantly enhances its separation performance. Furthermore, the study establishes the relationship between the inlet gas-phase and liquid-phase velocities and the separation efficiency of the startup separator. The findings show that increasing the inlet gas-phase velocity results in a decrease in separation efficiency, whereas increasing the inlet liquid-phase velocity improves the separation efficiency. Throughout the computational process, the separation efficiency of the novel startup separator remains consistently above 99%. The corrugated plate separation structure plays a pivotal role in improving separation efficiency, thereby optimizing the overall separation performance of the startup separator.

Study on Fluid-Thermal-Structural Coupling of Helical Tube Once-Through Steam Generator

Jiao Meng, Zhao Xinwen, Fu Shengwei, Jiang Jiahang, Ouyang Kehan

2025, 46(3): 213-219. doi: 10.13832/j.jnpe.2024.040027

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The double-sided heat exchange structure of the helical tube once-through steam generator (OTSG) effectively enhances heat transfer and has been applied in integrated reactors. The phenomenon of tube dryout in helical tube OTSG leads to a temperature jump, potentially causing damage to the tubes. This paper establishes a fluid-thermal-structural coupling analysis model of the single-channel helical tube OTSG, analyzes the void fraction and temperature distribution on the secondary side wall of the tubes, and further investigates the stress field of helical tube OTSG. The results indicate that the void fraction and temperature on the secondary side tube wall exhibit a spiral banded distribution. The temperature continuously jumps due to the banded distribution of dryout on the outer wall of the helical tube along the z direction, resulting in a corresponding jump in the equivalent stress of the tube wall. The fluctuation amplitude of the equivalent stress increases with the magnitude of the temperature jump. The stress in the dryout banded region of the tube exhibits concentration, with the maximum stress fluctuation reaching around 11 MPa. The stress distribution in the circumferential banded region also varies, with the stress fluctuation exceeding 10 MPa in the helical tube and 7 MPa in the straight tube.

Evaluation of Boundary Lubrication Status for Nuclear Coolant Pump Thrust Bearings

Ma Haoxiang, Wang Yan, Yang Jiangang, Cui Huaiming, Xu Renyi, Kuang Chengxiao

2025, 46(3): 220-223. doi: 10.13832/j.jnpe.2024.040020

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The state of lubrication significantly affects bearing wear. In this paper, a method for judging the boundary lubrication of thrust bearings of the nuclear power plant reactor coolant pump (nuclear coolant pump) based on the pressure-speed (P-ω) function curve is proposed. A thrust bearing test rig was set up, and the boundary lubrication speed under different temperatures and specific pressures was determined by monitoring torque changes, obtaining the P-ω function curve. After the specific pressure increases, the boundary lubrication speed rises. Above the curve is the fluid lubrication zone, and below is the boundary lubrication zone. A tiltable pad thrust bearing model was established based on the Reynolds equation, analyzing the relationship between speed and specific pressure under different oil inlet temperatures and film thicknesses. The study found that when the film thickness decreased to 2.5 μm, the calculated P-ω function curve was basically consistent with the experimental results. An increase in oil inlet temperature leads to a decrease in viscosity, and the boundary lubrication P-ω function curve shifts upward as a whole. This research can provide guidance for judging the lubrication state of nuclear coolant pump thrust bearings.

Design of Periodic Test Scheme for HPR1000 Nuclear Power Plant Reactor Protection System

Zhang Yu, Peng Hao, Hu Qingren, Zhou Dai

2025, 46(3): 224-228. doi: 10.13832/j.jnpe.2024.070025

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By reviewing the design requirements of periodic tests in HAF102 and GB/T5204, combing the characteristics of HPR1000 reactor protection system (RPR) in Units 1 and 2 of Zhangzhou Nuclear Power Plant, Fujian Province, and using the idea of full link coverage and test segment overlapping, a complete set of RPR periodic test design based on NASPIC for the HPR1000 is proposed. Compared with the test schemes of other nuclear power units in China, this scheme adopts automation and human-friendly design for optimization and improvement on the basis of meeting the functional requirements of RPR system, achieving automated execution of tests and reducing the risks brought by human factors. This can provide a reference for the periodic testing schemes of RPR system in subsequent projects.

General Health Assessment Method for Critical Nuclear Power Plant Equipment Based on Time-Series Characteristics of State Parameters

Ke Lishi, Du Haihu, Yang Xiaohu, Zhang Sheng, Huang Lijun

2025, 46(3): 229-235. doi: 10.13832/j.jnpe.2024.060009

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To address the issues of low accuracy and poor generalizability in existing health assessment methods for nuclear power equipment, this study establishes a generalized health assessment method for important equipment in nuclear power plant based on time-series characteristics of state parameters. By analyzing the time-series characteristics of state parameters, this method constructs an evaluation index matrix and an assessment model, and forms a general method suitable for various types of equipment in nuclear power plants. Taking circulating water pumps of different manufacturers and models as examples, the proposed method achieves over 93% accuracy in health assessment and significantly advances anomaly detection timelines. These results demonstrate that the general health condition assessment method established in this study can improve the accuracy of health status assessment of nuclear power plant equipment and is suitable for various types of equipment in nuclear power plants.

Research on Data Assimilation Technology for Nuclear Power Source Operating Conditions

Qi Lin, Wang Shuguang, Wang Xuesong, Jin Zhao

2025, 46(3): 236-243. doi: 10.13832/j.jnpe.2024.060004

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To enhance the alignment of simulation model outputs with real data for space nuclear power systems, and to achieve ground-space synchronization and digital twin implementation during the in-orbit operational phase, thereby laying the groundwork for remote diagnostics and prognostics, this study employs the Ensemble Kalman Filter assimilation technique. A data assimilation module was developed in conjunction with the Thermal-hydraulic Analysis Code of Space Thermionic Nuclear System (TASTIN). This module was tested under various transient conditions, including reactor startup, reactivity insertion, and emergency shutdown. The results demonstrate that the assimilation efficiency of operational parameters exceeds 90% across these three transient scenarios. Consequently, the data assimilation method proposed in this paper can effectively correct the simulation model.

Research and Application of AP1000 Class 1E RTD Channel Calibration Method

Gao Qifeng

2025, 46(3): 244-248. doi: 10.13832/j.jnpe.2024.060001

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To optimize the accuracy calibration method of Class 1E RTD channel in the nuclear power plant to meet the accuracy requirements of power plant instrument, this paper begins with the measurement principle of RTD and the common channel calibration methods in the nuclear power plant, elaborates the calibration and correction methods that may be used in the installation, replacement and operation test for Class 1E RTDs, puts forward a calibration correction method for RTD channels incorporating the accuracy of analog-to-digital (A/D) converters, and gives the application mode of cross-calibration data from RTDs in process system temperature platforms for RTD channel calibration. Through analysis, calculation and application verification, this RTD channel calibration accuracy correction method can effectively improve instrument channel accuracy and demonstrates strong practical value.

Ultrasonic Inspection & Data Analysis for Failed Fuel Assembly

Gan Wenjun, Cai Jiafan, Zhou Lifeng

2025, 46(3): 249-252. doi: 10.13832/j.jnpe.2024.050046

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To safely, effectively, and accurately locate failed fuel assemblies in nuclear power plants through ultrasonic inspection methods, this study utilizes the characteristic that the medium state of the inner wall of the fuel cladding before and after damage will cause differences in the attenuation of circumferential Lamb wave energy in the cladding. The propagation path of the Lamb wave in the fuel cladding during the movement of the ultrasonic probe in the gap between fuel assemblies is analyzed theoretically, and the propagation sound path is analyzed and calculated. The deviation between the calculated results and the measured values is within ±2%. Based on the principle of ultrasonic detection and the characteristics of sound beam propagation, the automatic identification algorithm and data analysis software of fuel rod cladding echo signal are developed, which can realize the rapid analysis and screening of detection data. The test results of simulated fuel assemblies and field application verify that the detection method is fast, safe and effective in detecting the damaged location of leaking rods in fuel assemblies, and the automatic signal identification algorithm and data analysis software are accurate and reliable, which can provide a basis for the targeted repair of damaged fuel assemblies in the future and improve the utilization rate of nuclear fuel.

Comparative Study on Vibration Prediction Methods of Reactor Internals Based on Neutron Noise Characteristic Frequency Time-Series Signal

Liu Yisong, Liu Caixue, Zhou Chengning, Luo Neng, Yan Jihong, Zeng Qiang

2025, 46(3): 253-259. doi: 10.13832/j.jnpe.2024.050042

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The vibration status of reactor internals is directly related to the operational safety and the maintenance node of nuclear power plant. Therefore, it is important to analyze and predict the vibration of these internals. This paper proposes a method for predicting the vibration of reactor internals based on the time-series signals of neutron noise characteristic frequency bands. The method, from two perspectives of single-cycle and double-cycle, utilizes statistical learning and machine learning methods for prediction, and an experiment was conducted using neutron noise signals collected from a nuclear power plant. The results show that, in terms of analysis methods, the processing of characteristic frequency band time-series signals can effectively utilize the temporal information in the signals. In terms of prediction methods, statistical learning models achieve higher accuracy for single-cycle prediction while machine learning models achieve higher accuracy for double-cycle prediction.Therefore, the combination of characteristic frequency band time-series signal analysis methods and appropriate prediction models can provide guidance for the prediction and determination of maintenance nodes in nuclear power plants.

Research on Calculation Method of Lower Radial Bearing Clearance of Main Pump of Nuclear Power Unit

Liu Jiaxin, Mao Lufeng, Yin Long, He Pan, Liu Yong, Yang Taibo

2025, 46(3): 260-265. doi: 10.13832/j.jnpe.2024.050036

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For the Russian million-kilowatt pressurized water reactor (VVER) nuclear power units, the lower radial bearing clearance of the main pump cannot be directly monitored with installed sensors, and these units currently use Russian-supplied encrypted software for predictions. To solve the upgrading demand caused by equipment aging, it is necessary to research the calculation method of the lower radial bearing clearance of the main pump. Firstly, the finite element model of bearing-rotor dynamics was built. The radial vibration displacement response of the pump rotor under unbalanced excitation at different positions and amplitudes was calculated, which was further fitted to obtain the rotor vibration displacement relationship expression. Then, combined with the rotor radial displacement at the measured points, the calculation method of the minimum clearance of the lower radial bearing was established, and the calculation and prediction software of the lower radial bearing clearance of the main pump was developed. A comparison between actual measurement results and software calculations showed that the maximum absolute error was 0.017 mm (maximum percentage error: 11.3%), while the minimum absolute error was 0.001 mm (minimum percentage error: 0.8%).

Application of Ultrasonic Technique in In-service Inspection of Bolts in Nuclear Power Plants

Wang Weiqiang, Ma Guanbing, Tang Jianbang, Yu Zhe, Yuan Shuxian, Ye Xin

2025, 46(3): 266-270. doi: 10.13832/j.jnpe.2024.050020

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Nuclear power plant equipment contains a large number of bolted fastening structures. Bolts may develop defects under long-term complex service conditions, and effective and reliable ultrasonic inspection of bolts is essential to ensure the safe operation of nuclear power plants. In this paper, based on two types of bolts with and without center hole, the baffle bolt and RPV stud are selected as the research object. The end-face ultrasonic inspection technology and center-hole ultrasonic inspection technology are studied, and the judgment and quantitative technology of defects are analyzed. The results on the test blocks show that the end-surface ultrasonic inspection technology and the center-hole ultrasonic inspection technology can realize the effective detection of defects and meet the requirements of in-service inspection.

Exergy Analysis of Solar-Nuclear-Storage Hybrid System under Different Operation Strategies

An Zeyi, Liu Qihong, Qiu Binbin, Ding Xu, Kang Boshi, Li Xuhui

2025, 46(3): 271-281. doi: 10.13832/j.jnpe.2024.070006

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To solve the problem that traditional PWR nuclear power units cannot meet the need to participate in grid peak regulation in the future, an Solar-Nuclear-Storage Hybrid System that couples solar energy and nuclear energy was put forward. A system model was built with thermal system simulation software EBSILON, where exergy analysis of the system under different operation strategies was carried out to study the thermodynamic performance of the system under design conditions. The exergy analysis and research under different operation strategies show that the three equipment with the highest exergy losses in the system are steam generator, solar field and steam turbine high pressure cylinder first stage, and the exergy loss of the three equipment in total is close to 50% of the total exergy loss. At the same time, the exergic efficiency of solar field is mainly affected by the change of direct normal irradiance. The exergic efficiency of electric heater is basically unchanged with the maximum change of 2% under different operation strategies.

Development of Solid State Cell Target for Isotope Irradiation Production in Research Reactor

Huang Gang, Zhang Jinsong, Si Junping, Sun Shouhua, Sun Sheng, Song Jigao, Kang Changhu, Qiu Lanlan

2025, 46(3): 282-287. doi: 10.13832/j.jnpe.2024.060041

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To utilize research reactors for the irradiation production of radioisotopes, a cell target suitable for irradiating nickel and neptunium targets is developed, along with the solid-state irradiation technology for isotope production in research reactors. A casing type, double-layer inner and outer cooled target tube structure for the target is adopted, multiple solid annular pellets are loaded in the closed cavity of the target tube, and uniform cooling water gaps are set inside and outside of the target tube, significantly improving the cooling efficiency of pellets and isotope yield. During the development process, the thermal hydraulic behavior and stress of the target were analyzed and evaluated. Out-of-pile hydraulic scouring test of the simulated target was carried out, while in-pile irradiation test on the target loaded with actual pellets was conducted. The test results demonstrate that the developed target satisfies the requirement of safe irradiation production for radioactive isotope in the research reactor.

Structure Analysis and Evaluation of the Ultrasonic Tomography System for Lead-Bismuth Two-Phase Flow

Liu Gangyang, Zhou Wenxiong, Huang Runzhi, Pan Liangming, Li Kang, Tan Xubin

2025, 46(3): 288-294. doi: 10.13832/j.jnpe.2024.050025

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Following steam generator tube rupture (SGTR) accident in a Lead-Bismuth Fast Reactor, the gas-liquid two-phase flow phenomenon emerges within the reactor core. The ultrasonic tomography method for two-phase flow detection exhibits robust anti-interference capabilities. However, conventional ultrasonic sensors become ineffective at high temperatures. Therefore, a dual-modal ultrasonic tomography system based on waveguide rods is proposed. The waveguide rod structure prevents direct contact between the sensor and the high-temperature fluid, employing reflection and transmission methods to reconstruct the two-phase distribution. In combination with numerical simulation methods, an ultrasonic tomography system utilizing a 4 MHz ultrasonic frequency, a 58 mm waveguide rod, and an array of 24 ultrasonic sensors is ultimately determined. The imaging effects of different gas-phase distributions are investigated. The results indicate that, compared to gas-liquid two-phase flows in general, the ultrasonic tomography system exhibits greater advantages in liquid Lead-Bismuth two-phase flows. This system can effectively reconstruct the two-phase distribution with mean square errors within 6% and a minimum image correlation coefficient exceeding 85%.

Research on Evaluation Method for MTBF of NPP I&C Cards during Operation

Mo Changyu, Li Gang, Li Mingli, Hu Jun, Wu Bin, Zhu Liling

2025, 46(3): 295-302. doi: 10.13832/j.jnpe.2024.050024

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The evaluation of mean time between failure (MTBF) for NPP I&C cards during operation is significant for root cause analysis, NPP maintenance and aging management, and intelligent operation and maintenance technology research. However, due to the lack of unified technical standards in the nuclear power industry at the current stage, the simple averaging method is mainly used in practical engineering applications to calculate the operational MTBF. This algorithm does not consider the definition of liability failures or the fitting of failure distribution, and it cannot provide more instructive interval estimation. To address this issue, this study proposes an MTBF evaluation method for nuclear power I&C cards based on goodness-of-fit tests for multiple failure distributions. Firstly, the definition criteria for liability failures that need to be included in the evaluation scope are provided, and fault screening and statistics are conducted based on these criteria. Secondly, likelihood equations for parameter estimation of four failure distributions are established for timed truncated data. The goodness-of-fit test and optimal distribution selection are then conducted using the Pearson chi-square test and Bayesian information criterion (BIC). Based on this, the MTBF of the card under a certain confidence level is calculated. This method was applied to the relay output cards of a nuclear power unit in commercial operation. Practical applications have shown that the MTBF evaluation results obtained using this method, including point estimation and interval estimation, comprehensively consider the contributions of failure distribution, data sample characteristics, and information loss. These results are more reasonable and accurate compared to the conventional approach. Therefore, the MTBF evaluation method established in this study for nuclear power I&C card components can be applied to engineering-based fault analysis of NPP I&C cards.

Current Issue 2025 Vol. 46, No. 3

Current Issue
2025 Vol. 46, No. 3