核动力工程

Development and Prospect of Advanced Nuclear Energy Technology

Wang Conglin, Chai Xiaoming, Yang Bo, Li Zhongchun

2023, 44(5): 1-5. doi: 10.13832/j.jnpe.2023.05.0001

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Abstract:
The proposal of the "carbon peaking and carbon neutrality" target has a profound impact on the future energy system development in China. As a stable and clean energy source, nuclear energy can play an important role in the energy system, with great application potential and demand in electricity generation, heating, hydrogen energy, and other fields. After more than 60 years of development, nuclear energy has established a complete industrial chain. Research and development have formed third-generation large-scale commercial pressurized water reactor nuclear power technology brands with completely independent intellectual property rights, such as “Hualong-1” and multi-purpose modular small reactors “Linglong-1” with international advanced level. Moreover, China actively explores fourth-generation advanced nuclear energy technologies, such as sodium-cooled fast reactors, ultra-high temperature gas-cooled reactors, and molten salt reactors, and continues to carry out research on nuclear fusion energy. However, China's nuclear energy development also faces some challenges, and advanced nuclear energy technologies are in urgent need of breakthrough. To address these challenges, a development strategy and path for advanced nuclear energy technologies have been put forward, including intelligent operation management of in-service nuclear power plants, batch deployment of third-generation nuclear power plants, research and development of inherent safety fast reactor technology, active research and development of ultra-high temperature gas-cooled reactors for high-efficiency hydrogen production, active exploration of small modular reactors for industrial heating and platform power supply, and domestic and international cooperation in developing critical advanced nuclear energy technologies. These six aspects provide specific research goals and directions for the development of China’s advanced nuclear energy technology.

The Deep Learning Method to Search Effective Multiplication Factor of Nuclear Reactor Directly

Liu Dong, Tang Lei, An Ping, Zhang Bin, Jiang Yong

2023, 44(5): 6-14. doi: 10.13832/j.jnpe.2023.05.0006

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Abstract:
It's a basic problem in the reactor criticality calculation to determine the effective multiplication factor k_eff. At present, the source iteration method is widely used in the industry. Based on the theory of artificial intelligence deep learning method to solve differential equations, this paper puts forward a new method, which takes k_eff and the weights of neural network neurons as machine learning optimization parameters, calculates the weighted loss function formed by substituting neural network functions into neutron differential equations, and simultaneously approaches neutron fluence rate and searches for k_eff directly. The mathematical form of eigenvalue of neutron differential equation, the setting method of initial neural network, the weighting factor of loss function, convergence criterion and other important factors affecting deep learning performance and the corresponding performance improvement schemes are discussed. The correctness of the method and the effectiveness of the learning performance improvement schemes are verified by numerical calculation of various examples. The results have explored a new technical way to calculate the k_eff of nuclear reactors, which is an important scientific problem in neutron physics.

Pipelined Parallel JFNK Method and its Application in Neutron k Eigenvalue Problem

Liu Lixun, Zhang Han, Wu Yingjie, Guo Jiong, Li Fu

2023, 44(5): 15-22. doi: 10.13832/j.jnpe.2023.05.0015

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Abstract:
JFNK (Jacobian-free Newton-Krylov) is an efficient acceleration method for solving nonlinear problems such as neutron k eigenvalue and coupling of multiple physical fields in the reactor, and the generalized minimum residual (GMRES) algorithm is commonly used in Krylov iteration. The parallel JFNK method is a necessary means for solving large-scale problems, and its main problem lies in the low parallel efficiency of Gram-Schmidt (GS) orthogonalization procedure in GMRES, which causes massive collective communications. In this paper, the parallel JFNK method based on the parallel programming model of message passing interface and spatial domain decomposition technology is developed for the three-dimensional neutron k eigenvalue problem. Aiming at the poor parallel scalability of GS orthogonalization procedure, the pipelined method is studied to improve the parallel efficiency of parallel JFNK. Then, the computation time and parallel efficiency of parallel JFNK using classical GS orthogonalization, modified GS orthogonalization and pipelined method are compared. The IAEA-3D three-dimensional diffusion benchmark problem was used for numerical test. The results show that the parallel efficiency of pipelined parallel JFNK is significantly superior to that using classical or modified GS orthogonalization, and the convergence of pipelined parallel JFNK is not affected.

Evaluation of Thermal Neutron Scattering Data for H in ZrH_x Based on Deterministic Method

Wang Lipeng, Zhang Xinyi, Jiang Duoyu, Hu Tianliang, Cao Liangzhi, Wu Hongchun, Cao Lu

2023, 44(5): 23-29. doi: 10.13832/j.jnpe.2023.05.0023

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Abstract:
ZrH_x has been widely used in nuclear technology because of its high H content and good moderating property. However, the thermal neutron scattering data generation of H in ZrH_x mostly uses direct numerical simulation method, which has not been evaluated with experimental data. In this paper, a fast determinstic evaluation method of thermal neutron scattering data of H in ZrH_x is presented, which is related to semi-empirical phonon model. It is completed by adjusting the thermal scattering data simulated by computer with the experimental data by using generalized least square method. The experimental data are based on the total cross-section measurements and the effective multiplication coefficient (k_eff) of the benchmark reactor. The results show that the optical range of the H phonon density in the modified ZrH_x is shifted to high energy region, and the total cross-section values are in good agreement with the experimental data. The accuracy of k_eff calculation for the two TRIGA reactors after adjustment has been improved.

Influence of Depletion on Reactivity Insertion Transients for Xi’an Pulsed Reactor

Zhang Xinyi, Wang Lipeng, Wang Yongping, Jiang Xinbiao, Yin Huabei, He Bin, Li Da, Zhang Liang, Jiang Duoyu, Hu Tianliang

2023, 44(5): 30-38. doi: 10.13832/j.jnpe.2023.05.0030

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Abstract:
In this paper, a new method is presented to accurately simulate the transient behavior of reactor burnup. The core transient calculation code based on three-dimensional physical and thermal coupling is coupled with the three-dimensional precise micro burnup calculation code. According to the control rod movement process of Xi'an Pulsed Reactor (XAPR), the transient characteristics of different parameters such as core power and fuel temperature due to the different form and magnitude of reactivity insertion at different burnup depth are analyzed. The sensitivity of core pulse power response to reactivity insertion is analyzed, and the change mechanism of core dynamic parameters is studied. The numerical results show that with the burnup of nuclear fuel, the peak pulse power tends to decrease when the core is arranged with pulse state, while the peak pulse power increases obviously with the steady state core arrangement, and the time to reach the peak pulse power is advanced.

Study on JFNK Global Solution Method of of Full-core Thermal Sub-channel Model

Zhang Yunshan, Zhou Xiafeng

2023, 44(5): 39-46. doi: 10.13832/j.jnpe.2023.05.0039

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Abstract:
The thermal sub-channel model of reactor takes into account many coupling factors such as axial flow, lateral mixing and turbulent mixing in detail, and it is the key model of thermal hydraulic analysis of the core. However, these factors bring difficulties and challenges to the numerical simulation of sub-channel. In order to improve the computational efficiency and convergence of the thermal sub-channel model, this paper develops a global solution framework for the thermal sub-channel model of the whole core based on the Jacobian-Free Newton-Krylov (JFNK) global solution method (hereinafter referred to as JFNK method), and establishes a residual system based on physical pretreatment according to the existing code model and framework to enhance the convergence rate of JFNK method. Numerical solutions show that the speedup of JFNK methods is about five compared with the fixed-point iterative method used in the original sub-channel models and there are higher efficiency for JFNK methods as convergence precision improves, which indicates the potential and efficiency advantages of JFNK methods for complicated thermal sub-channel model.

Development and Application of Cooling Characteristic Analysis Code for Molten Debris Bed

Fang Yu, Yang Shengxing, Gong Houjun, Zan Yuanfeng, Yang Zumao, Zhuo Wenbin

2023, 44(5): 47-53. doi: 10.13832/j.jnpe.2023.05.0047

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Abstract:
To analyze the cooling characteristics of debris bed formed in the late stage of severe accident in pressurized water reactor, an analysis code for molten debris bed cooling was developed. Based on the one-dimensional six-equation two-phase flow model, the physical process of two-phase flow and heat transfer in the debris bed was described by using the porous medium flow boiling heat transfer model, and the equations were discretized and solved by using the controlled volume integral method, semi-implicit method and first-order upwind scheme. The results of TUTU, COOLOCE and STYX experiments were used to validate the model from aspects: two-phase flow and dry-out heat flux (DHF). It was found that Hu&Theofanous model and Reed model had better prediction results for the two-phase flow of debris bed with relatively large particle size, while Lipinski model had higher prediction accuracy for the low-pressure DHF of small particle debris bed. The cooling capacity of molten debris bed in severe accident of PWR was predicted by the code. At a particle heat release rate of 1 MW/m³, the cooling height of debris bed was 0.56 m under top-flooding condition; and the height increased to 0.85 m by using bottom-flooding driven by natural circulation.

Analysis of Correlation Between Reaction and Flow Loss of Helium Compressor

Sun Zeqin, Zhang Jingxuan

2023, 44(5): 54-63. doi: 10.13832/j.jnpe.2023.05.0054

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Abstract:
Multi-stage axial helium compressor is one of the key components of helium turbine. High reaction design is usually used to enhance the enthalpy gain of a single stage to reduce the number of stages, but experience shows that high reaction often leads to a reduction in efficiency, so it is necessary to clarify the effect of reaction on flow loss when using high reaction designs. In this paper, the correlation between the profile loss, corner loss and tip leakage loss inside the compressor vane channel based on helium flow and the reaction degree is analyzed in detail using CFD method for helium compressors with different reactions. The results show that the profile loss of 100% reaction rotor and 50% reaction stator increases due to the high relative velocity. As the reaction increases, the high load in the rotor leads to the increase of the loss at the tip leakage of the rotor. And, the increase of the stator bending angel and spanwise differential pressure make the separation of the stator angle more serious. At 100% reaction, the tip leakage loss accounts for 53% of the rotor loss and the corner loss accounts for 60% of the stator loss. Based on the air compressor profile loss model, the profile loss model suitable for helium compressor is optimized through physical property analysis.

Study on Prediction Method for Accident Parameters of Lead-bismuth Reactor Based on Coupling Multivariable LSTM and Optimization Algorithm

Ji Nan, Yang Junkang, Zhao Pengcheng, Wang Kai

2023, 44(5): 64-70. doi: 10.13832/j.jnpe.2023.05.0064

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Abstract:
Accurate prediction of key parameters of lead-bismuth reactor under accident conditions is an important content of reactor safety analysis, which is of great significance to improve the safety of the reactor under accident conditions. In this work, an optimization algorithm is used to improve the prediction performance of the Long Short Term Memory (LSTM) neural network by hyperparameter optimization, and a parameter prediction method based on the coupled optimization algorithm of multivariate LSTM neural network is proposed. For the parameter prediction problem of lead-bismuth fast reactor MARS-3 under unprotected loss of flow accident conditions, a comprehensive evaluation of the proposed method is performed using Technique for Order Preference by Similarity to Ideal Solution (TOPSIS) method after data samples generated by the sub-channel code SUBCHANFLOW. The results show that the prediction performance of the multivariate LSTM neural network coupled with the Particle Swarm optimization method is optimal, and its computational efficiency can be improved to 438 times that of SUBCHANFLOW. The relevant research results can help improve the efficiency of predicting key thermal parameters of lead-bismuth reactors and improve the emergency response capability of lead-bismuth reactors.

Numerical Study on Bending Angle Optimization of Rod Bundle Channel Spacer Grid Mixing Vane Based on Subcooled Boiling CFD

Zheng Xu, Sun Lanxin, Huang Junfeng, Zhu Renren, Duan Zhongping, Cheng Xiaoyu, Li Jinrong

2023, 44(5): 71-79. doi: 10.13832/j.jnpe.2023.05.0071

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Abstract:
The spacer grid is a key component of the fuel assembly. It not only fixes and supports the fuel assembly, but also affects the thermal and hydraulic performance of the fuel assembly. In this paper, the Rensselaer Polytechnic Institute(RPI) wall boiling model based on the Euler-Euler six equations of two fluids is used to carry out a numerical simulation study on the working conditions when subcooled boiling occurs in the 5×5 fuel rod bundle channel; the flow field in the rod bundle channel, the distribution of circumferential wall temperature at different axial heights of the central rod bundle and the distribution of void are obtained. The numerical study results show that the pressure loss of the rod bundle channel increases with the increase of the bending angle of the mixing vane; when the bending angle of the mixing vane is 30°, the proportion of void is relatively small, and the peak temperature of the central rod bundle is relatively low in the downstream of the spacer grid.

Development of Fatigue Monitoring and Transient Counting System in HPR1000

Cui Huaiming, Tang Chuanbao, Bai Xiaoming, Ai Honglei, Liu Jia

2023, 44(5): 80-84. doi: 10.13832/j.jnpe.2023.05.0080

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Abstract:
HPR1000 is one of the 3rd generation nuclear power plants in China. The fatigue monitoring and transient counting system is an important monitoring system in HPR1000, which plays a positive role in the safe and economical operation of HPR1000. Nuclear Power Institute of China has developed a fatigue monitoring and transient counting system with independent intellectual property rights. The system consists of 25 temperature measuring components, a signal processing cabinet and a computing workstation. The models for thermal inverse problem, green function stress calculation, coolant environmental fatigue calculation, and transient counting have been developed, which allow the functions of monitoring the fatigue state of the primary system and automatically identifying and counting the running transient. Through the development of principle prototype and engineering prototype, the related key technologies have been solidified through verification, and the technology is mature and has engineering application conditions.

Fragility Analysis of Main Aftershock by Nuclear Power Plant SSC Coupling System Model Considering SSI Effect

Zhao Jinyi, Song Lei, Zhou Zhiguang

2023, 44(5): 85-94. doi: 10.13832/j.jnpe.2023.05.0085

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Abstract:
Seismic fragility analysis of NPP systems can reflect the failure probability of coupled structures, systems and components (SSCs) under different earthquake intensities, in which the soil-structure interaction (SSI) and the main aftershock effect are two very important factors. In this paper, the AP1000 NPP SSC coupled system model is established; the typical soft rock foundation is selected as the site condition; the main aftershock records are selected according to the AP1000 design spectrum; and the coupled model is analyzed for seismic fragility considering the SSI effect using the IDA calculation method. It is calculated and analyzed that the damage to the structure and equipment from the main aftershock effect may be greater than the effect of a single mainshock. Considering the SSI effect generally increases the conditional failure probability of SSCs under main aftershocks. From the typical SSC seismic performance results, the failure mode of the coupled system is that the concrete of the shield building cracks first, followed by the yielding of the steam generator piping, and finally the main steam piping enters yielding. Considering SSI effect, the values of high confidence low failure probability (HCLPF) of the three in the limit state between basically intact failure state and general failure state are 0.48g, 0.68g and 0.92g respectively. The research results indicate that the effects of the SSI and the main aftershock should not be neglected in the fragility assessment of nuclear power plants.

Study on Shock Response of ACP100S Floating Nuclear Power Plant Subjected to Ship Collision

Wang Donghui, Li Qing, Zhang Yanming, Zeng Qingna, Dong Leilei

2023, 44(5): 95-103. doi: 10.13832/j.jnpe.2023.05.0095

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Abstract:
Ship collision is a significant external event for the design of nuclear floating facilities, and it has a great impact on safety. In this paper, based on the crashworthy design review of nuclear merchant ships, the numerical simulation techniques suitable for the ship-collision analysis are established and validated by the available experimental result in the literature. Using the validated numerical method, the collision between a supply ship and the ACP100S floating nuclear power plant (FNPP) is simulated for bow- and broadside-collision scenarios, and the shock response of the key shipboard equipment is obtained. The calculation and analysis results show that the shock acceleration of the key equipment in the broadside-collision case exceeds 1g, which is the design basis load for nuclear merchant ships. The research results of this study have certain guiding significance for the shock resistance design of floating nuclear power plant hull and reactor key equipment.

Local Stiffness Calculation and Characteristics Analysis of Cylinder and Lug Connection Structure

Jiang Xiaozhou, Liao Guojiang, Ye Xianhui, Liu Shuai, Huang Xuan, Peng Xiangfeng

2023, 44(5): 104-109. doi: 10.13832/j.jnpe.2023.05.0104

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Abstract:
It is of great engineering significance to accurately calculate the local stiffness of typical cylinder and lug connection structures in nuclear power plant reactor system. In this paper, a three-dimensional finite element model of cylinder and lug connection structure is established. The total displacement can be acquired by finite element calculation, and the calculation method is respectively introduced to solve the beam displacement due to concentrated force and torque, which will be used to calculate the final local stiffness of cylinder and lug connection structure. The separate model of cylinder and lug connection structure is established, which is also used to calculate the final local stiffness, and the local stiffness difference is discussed by comparing with the calculation results of whole model. The whole model calculation method is considered to calculate the local stiffness by selecting different boundary planes, and the variation law of local stiffness is obtained. At the same time, the angle of lug is adjusted to calculate and study the influence on local stiffness characteristics. This paper can provide reference for engineering design of cylinder and lug connection structure.

Fast Determination of Containment Leakage Rate Based on Pressure Rise Monitoring

Wu Kuibo, Liang Zhaorui, Zhang Tengfei, Cao Qing, Song Yongjun, Xue Yu, Liu Junjie

2023, 44(5): 110-115. doi: 10.13832/j.jnpe.2023.05.0110

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Abstract:
In this paper, a qualitative and quantitative method for fast determination of containment leakage at the start-up stage of the unit is proposed. Based on the rising trend of containment pressure within 24 h during the unit start-up, the set leakage rate limit is taken as the assumed leakage rate, and the calculated pressure rise is obtained according to the ideal gas state equation and the gas mass balance relationship in the containment, and the leakage situation is qualitatively determined by comparing the calculated pressure rise with the monitored pressure rise. According to the gas mass balance relationship in the containment, the quantitative method can reversely solve the containment leakage rate matching with the monitoring pressure rise. Using the above method, the containment leakage rate of five units in a nuclear power plant is analyzed. The qualitative determination results show that the leakage rate of only one unit exceeds the limit of 5 Nm³/h (Nm³ is the gas volume at 0℃ and 1 standard atmospheric pressure). The quantitative calculation results show that the leakage rate of the unit is 5.11 Nm³/h, which is close to the first daily leakage rate of 4.98 Nm³/h, indicating that the method proposed in this paper has good accuracy.

Evaluation, Diagnosis and Optimization Analysis of Control Performance of Nuclear Power Unit Transient Process

Liu Daoguang, Luan Zhenhua, Liang Jun, Liu Peng, Zhao Yuntao, Zhou Chuangbin

2023, 44(5): 116-123. doi: 10.13832/j.jnpe.2023.05.0116

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Abstract:
The complex control system is an important component of the instrument and control system in nuclear power plants, and its function is to ensure the smooth operation of the system. Therefore, the performance verification and evaluation of the control system is an important part of unit debugging. By constructing a performance evaluation standard system for unit control system debugging and testing, a control system performance evaluation and defect diagnosis method is proposed to diagnose and optimize the specific transient operation process of the unit. This research achievement has been applied in multiple nuclear power bases, and the logic faults and performance defects of complex control systems have been fully verified in the start-up stage of the unit, which has improved the control performance and safe operation level of the unit.

Research of Signal Risk under DCS Net-Node Failure in Nuclear Power Plant

Yang Liang, Zhou Weichang, Bian Xiushi, Deng Jijie

2023, 44(5): 124-129. doi: 10.13832/j.jnpe.2023.05.0124

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Abstract:
After a single net-node failure in the nuclear power plant, the conservative decision must also consider the net node failure at other locations and make an enveloping risk analysis. Based on the modeling and analysis of Distributed Control System（DCS） network structure and control logic data of a nuclear power plant, the calculation method of the physical path of network signals and the judgment rules of signal redundancy are given, the failure probability of the remaining network nodes is calculated by constructing the state transition matrix under single node failure, the network signal risk analysis and evaluation under single or superimposed failure are carried out, and finally the risk control and improvement suggestions are given for the identified high-risk network signals. This model and analysis method have been realized by computer software, which can quickly identify design defects such as false redundancy and unreasonable layout, and can quickly analyze the affected signals and risks under single or superimposed faults and output them explicitly. This study can provide guidance for related maintenance operations of nuclear power plants, or provide reference input for DCS configuration design or function optimization.

Investigation on Application of Intelligent Control in SGTR Procedure

Yuan Rui, Hao Zulong, Yuan Jinxiao, Li Mingqian, Deng Shiguang

2023, 44(5): 130-135. doi: 10.13832/j.jnpe.2023.05.0130

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Abstract:
After the occurrence of steam generator heat transfer tube rupture (SGTR) accident in nuclear power plant, operators shall follow the corresponding procedures. However, in accident sequence processing, the cooling and depressurizing effects of primary circuit coolant mostly depend on the operating experience of the operator. To solve this problem, this paper proposes a SGTR operating procedure optimization method based on LSTM and adaptive proportional-integral-differential (PID), and gives the implementation process. The SGTR accident simulation experiment was carried out by a 1000 MW simulator. Based on the simulation experiment data, the proposed method was tested. Simulation results show that, compared with manual operation, the proposed intelligent operation method has better cooling and depressurization effects. The proposed method can provide a new idea for intelligent operation of accident procedures in nuclear power plants.

Development of Transient Computational Model for Stirling Engine Used in Liquid Metal Reactor

Zhou Linjie, Zhang Donghui, Yang Jun, Wang Xiaokun, Wang Jin, Guo Zhongxiao

2023, 44(5): 136-143. doi: 10.13832/j.jnpe.2023.05.0136

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Abstract:
In order to explore the matching characteristics of the Stirling engine and the liquid metal nuclear reactor, a transient code for Stirling engine operation simulation is developed based on the Stirling cycle third-order analysis method. In this model, the working cavity and the heat exchanger of Stirling engine are modeled by the one-dimensional hydrodynamics method, and the simulation of operating Stirling engine in the liquid metal nuclear reactor is realized. In this paper, the steady state operation of the fixed wall temperature boundary condition is simulated by using the test data of GPU-3 Stirling engine, and the results show good agreement with the test data. After connecting to the heat exchanger solver module, the code has successfully realized the transient simulation of the Stirling engine under constant and transient boundary conditions. This indicates that the model can be used for the working condition analysis of the coupling system between the liquid metal nuclear reactor and the Stirling engine.

Design and Process Validation of Non-planar Annular Linear Array Cylindrical Guided Wave Transducer

Zhao Yang, Wang Fei, Xu An, Wang Libo, Sun Haixuan, Wang Zicheng, Yang Zhoubin

2023, 44(5): 144-150. doi: 10.13832/j.jnpe.2023.05.0144

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Abstract:
In order to realize online detection of nuclear safety level fasteners on T-boss end face in radiation control area, the cylinder guided wave in M56 mm×416 mm fastener is simulated using COMSOL software. The non-planar annular linear array cylindrical guided wave transducer is designed and developed, and the process validation is carried out on the simulation test block for designing and processing artificial reflectors. The results show that the dispersion curve of high frequency cylindrical guided wave tends to be consistent, and the T-shaped high-low step delayed excitation mode and the design of the chip size of 32 mm inner diameter and 50 mm outer diameter have optimal sound field response, which can effectively detect artificial grooves with a depth of 1 mm, and the depth error is within ±2 mm. The non-planar annular linear array cylindrical guided wave transducer designed and developed in this paper can realize efficient and portable on-line detection of fasteners in radiation control area.

Kinematics Analysis and Positioning Solution of Three Degrees of Freedom Manipulator for Eddy Current Inspection of SG Heat Transfer Tube

Xiao Xiang, Gao Sanjie, Jiang Weiyu, Liu Li, Li Shuliang

2023, 44(5): 151-155. doi: 10.13832/j.jnpe.2023.05.0151

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In-service automatic eddy current inspection device for heat transfer tube of steam generator requires high positioning accuracy. If positioning deviation exists, the probe will wear worse or even get stuck in the heat transfer tube. In order to solve the problem, based on the three degrees of freedom detection manipulator, the kinematics model of the manipulator joint is established, and the forward and inverse kinematics are solved in this paper. Besides, the method of error elimination is proposed for solving the position deviation of the base when installing the manipulator. The test results show that the positioning calculation method is accurate, stable and reliable.

Development and Application of Automatic Eddy Current Inspection System for Heat Transfer Tube of Steam Generator in Nuclear Power Plant

Sun Liming, Pei Xibao, Chen Xiaoliang, Li Yang, Ma Qiang, He Peng

2023, 44(5): 156-162. doi: 10.13832/j.jnpe.2023.05.0156

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Abstract:
Steam generator （SG） is one of the key equipment in nuclear power plant, and its heat transfer tube is the weakest link in the primary pressure boundary. In order to solve the problem that SG small curvature heat transfer tube cannot realize one-time full tube eddy current detection due to the influence of bending tube curvature, which leads to the extension of overhaul period and the increase of radiation risk of personnel, a pneumatic probe with spring skeleton structure and air damping device is designed in this paper. The probe drive retraction device and the pneumatic drive device for internal and external air supply are adopted innovatively, and a new type of automatic eddy current inspection system for SG heat transfer tube is developed, which can realize one-time full-tube eddy current detection of SG small curvature heat transfer tube.

Study on Passivation of Primary Circuit of Nuclear Power Plant

Yu Miao, Gu Yu, Zeng Xiaomin, Xu Gang, Gui Luting, Wu Tong

2023, 44(5): 163-168. doi: 10.13832/j.jnpe.2023.05.0163

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It is an important work to forming a dense passivation film on the surface of the primary equipment in the hot functional test (HFT). The quality of the passivation film directly affects the radiation source term during the nuclear operation of the power plant. Therefore, it is of great significance to do a good job in the passivation of primary circuit during the HFT. Chemical control determines the quality of passivation film. This paper first briefly analyzes the characteristics of water chemistry of various types of nuclear power reactors in China, and then analyzes the influencing factors of pH value (pH_T) and boron-alkali curve at the operating temperature of primary coolant. The passivation application examples of four types of reactors in China are emphatically analyzed, and the chemical control mechanism of passivation film formation is discussed. Finally, the improvement measures for chemical control during primary circuit passivation are given.

Optimization and Application of Polarization Voltage Selection Method for ²³⁵U Micro Fission Ionization Chamber

Zhang Hengkai, Wang Yu, Zheng Junwei, Wei Qiao, Cheng Xiongwei, Deng Sen, Liu Jikun, Ren Yi

2023, 44(5): 169-174. doi: 10.13832/j.jnpe.2023.05.0169

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Abstract:
In order to solve the problem that the plateau slope is bigger with a wrong polarization voltage when ²³⁵U micro fission ionization chamber works continuously in different neutron flux measurement channels, this study combines the characteristics that the saturation current of ²³⁵U fission ionization chamber and the initial voltage entering the saturation region increase with the increase of neutron fluence, and performs the plateau curve test in the lowest and highest measurement paths respectively. Based on this, the plateau slope characteristic curves of the ²³⁵U fission ionization chamber in the two measurement channels are established, and the range where the plateau slope changes slightly is defined as the optimal polarization voltage range. The actual polarization voltage is selected by intersecting the optimal polarization voltage range and the actual allowable adjustment range of polarization voltage, so that the plateau slope of all measurement channels between the lowest and highest neutron fluence can be kept at a relatively low level. The practical application proves that the actual polarization voltage selected by the intersection method according to the optimal polarization voltage range is balanced and optimal for all neutron flux level measurement channels.

Experimental Investigation on Measuring Leakage Rate of Nuclear Power Plant Containment by Constant Pressure Method

Li Jianfa, Hua Yongzhen, Liu Feng, Sun Zhongning, Meng Zhaoming

2023, 44(5): 175-180. doi: 10.13832/j.jnpe.2023.05.0175

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Abstract:
In order to further optimize the containment leakage rate measurement scheme and improve the stability and reliability of the measured value, a new leakage rate measurement method, namely the constant pressure method, was developed based on the LabVIEW code in this study. The applicability of the constant pressure method to the leakage rate measurement of the inner containment and the outer containment was confirmed for the first time by carrying out the leakage rate measurement experiment of the large-scale containment simulant. The research shows that the leakage rate measured by the constant pressure method is in good agreement with those measured by traditional pressure drop method, and the relative deviation is less than 5%. For the measurement of the leakage rate of the outer containment, compared with the pressure drop method, the constant pressure method can greatly reduce the measurement time from 10 h to less than 0.8 h on the premise of ensuring the measurement accuracy. The research conclusions can provide theoretical basis for the application of constant pressure method.

Sensitivity Analysis of Heat Transfer Tube Plug Structure of Steam Generator

Li Jinghuai, Han Yile, Zhou Quan, Ying Bingbin, Chao Mengke, Yang Xing

2023, 44(5): 181-187. doi: 10.13832/j.jnpe.2023.05.0181

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Abstract:
The heat transfer tube plug of steam generator (SG) in nuclear power plant is a common component to maintain the heat transfer tube. Two-dimensional axisymmetric model elastic-plastic analysis was used to simulate the plugging process. The influence and sensitivity analysis of seven parameters (stretching distance, friction coefficient between expander and plug, inclination of plug inner wall and expansion block, tooth width, tooth height, groove width, and expander length) on the tension rod force, expansion block contact force and maximum contact force during plugging process were studied, and the prediction formula for the sensitivity change of each parameter to the plugging process was obtained. The sensitivity effect engineering estimation formula of all sensitivity parameters on the plug plugging process was established. The research results can provide reference for the structural design of SG heat transfer tube plug.

Research on Anomaly Detection and Correction of Nuclear Power Plant Operation Data Based on GRU-MLP

Wang Tianshu, Yu Ren, Mao Wei, Song Xiaosen, Ma Jie

2023, 44(5): 188-194. doi: 10.13832/j.jnpe.2023.05.0188

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Abstract:
In order to solve the data quality problems such as missing, drifting and jumping in the operation data collected or stored by the instrument & control system of nuclear power plant and provide more reliable input for operational data analysis and automatic controllers, a hybrid model based on Gated Recurrent Unit and Multilayer Perception (GRU-MLP) is proposed to detect and correct the abnormal operation monitoring parameters data of nuclear power plant. Firstly, the short-term prediction algorithm of operation data based on GRU model is studied to provide reference for anomaly detection and correction of operation data. Secondly, in order to improve the prediction accuracy of GRU model for the operation data containing anomalies, the real-time correction mechanism is used for optimization. Then, using the nonlinear fitting ability of MLP model, the fixed threshold used in the "prediction-anomaly detection" mechanism is optimized to the dynamic threshold, which improves the anomaly detection accuracy of the proposed method. Finally, the accuracy and feasibility of the proposed algorithm are verified through experiments based on the operation data of a certain nuclear power plant.

Precision Validation on the Flaws in Pipe Blocks of Performance Demonstration Inspection for Nuclear Power Plant

Yu Zhaohui, Zhou Lusheng, Chen Shu, Yang Tao, Du Yida

2023, 44(5): 195-200. doi: 10.13832/j.jnpe.2023.05.0195

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Abstract:
The radiographic testing (RT), ultrasonic testing (UT) and anatomical measurement were employed to measure the embodied real flaws in self-developed pipe blocks of performance demonstration inspection (PDI), and the differences between the geometric dimensions and position coordinate parameters of the embodied flaws and the design values were analyzed. The test involved four different specifications of pipe blocks, with a total of 16 defects, all of which are cracks. The results show that the track of embodying weld is eliminated completely, the cracks are featured with natural and high closure, and the tips are distinguishable with no secondary propagation. The embodied crack forms a complete metallurgical bond with the parent body of the block, and there is no abnormal signal reflection interface or secondary flaw in the embodied weld. After RT and UT, it is confirmed that the average absolute value deviation of the geometric size and position coordinates of the defects is less than 2.0 mm, and after anatomical measurement, the average absolute value deviation is less than 1.0 mm. The flaw precision meets the requirement for PDI of nuclear power plant.

Research on Interpretable Diagnosis Method of Reactor Accidents Based on Representation Extraction

Li Chengyuan, Li Meifu, Qiu Zhifang

2023, 44(5): 201-209. doi: 10.13832/j.jnpe.2023.05.0201

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Abstract:
This study proposes a diagnostic framework that is interpretable and based on representation extraction to achieve accurate, robust and reliable reactor accident diagnosis. Firstly, the Denoising Padded AutoEncoder (DPAE) deep learning model is introduced. Through self-supervised learning on a simulated dataset with varying break sizes and positions, the DPAE encoder can automatically extract low-dimensional representation vectors of monitoring parameters from partially missing data and noise data, which can then be used for downstream diagnostic tasks that involve classification and regression algorithms. Then, a parameter importance calculation method based on posteriori interpretability algorithm is introduced to analyze the contribution of monitoring parameters to diagnostic results. The proposed diagnostic method is validated using HPR1000 as the research object under LOCA conditions. The experimental results show that the trained DPAE model can still obtain effective data representation under Gaussian noise with signal-to-noise ratio of 30 dB and random masking ratio of 0.3. In addition, under the interference of signal-to-noise ratio of 20 dB and masking ratio of 0.2, compared with the "end-to-end" diagnosis model, the "upstream and downstream" diagnosis model proposed in this study performs better in the diagnosis of break position and size, and can identify monitoring parameters with greater contributions to the diagnostic results. The reactor accident diagnosis method proposed in this study is helpful to build an accurate, stable and reliable intelligent reactor operation and maintenance system.

Prediction Model for Growth of Chalk River Unidentified Deposit on the Surface of PWR Fuel Cladding

Chen Jiajie, Liu Xiaojing, Du Sijia, Wang Jiageng, He Hui

2023, 44(5): 210-215. doi: 10.13832/j.jnpe.2023.05.0210

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Abstract:
In order to establish a prediction method for the thickness of chalk river unidentified deposit (CRUD) on the surface of PWR fuel cladding, this paper takes the primary circuit of typical PWR as the research object. Aiming at the influence of water chemistry and physical conditions in PWR on CRUD, a deposition model is established. Compared with the actual operation data of Sizewell B nuclear power plant, the predicted results of the model have the same growth order and trend, which indicates that the model can be used to quantitatively predict the thickness of CRUD on the surface of PWR fuel cladding. The effects of heat flux, H₂ and Li concentration on the accumulation of corrosion products are studied, the predicted results show that the CRUD thickness of the non-boiling section is less affected by the heat flux, while the CRUD thickness of the boiling section increases with the increase of heat flux. The CRUD thickness increases with the increase of H₂ concentration, and the increase of Li concentration in the system is helpful to restrain the deposition of oxidation corrosion products.

Numerical Study on Boiling Heat Transfer and Thermal-Mechanical Characteristics of Helical Cruciform Fuel Assembly

Cong Tenglong, Gao Yong, Cheng Yi, Cai Mengke, Zhang Qi, Xiao Yao, Liu Maolong, Gu Hanyang

2023, 44(5): 216-222. doi: 10.13832/j.jnpe.2023.05.0216

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Abstract:
The helical cruciform fuel (HCF) assembly is an innovative design for nuclear reactor fuel. Compared to the traditional cylinder or plate type fuel, the HCF assembly has the features of intensive mixing, self-support and no requirement of spacer grid. The two-phase conjugate heat transfer model for HCF assembly and the thermal-mechanical analysis model for HCF rod were built to evaluate the performances of HCF from the aspects of thermohydraulics and mechanics, respectively. The distributions of fuel rod and coolant temperature and coolant void fraction under two-phase conditions, and the fuel rod stress distribution under different burnup were obtained. The results show that the HCF assembly can reduce the fuel center temperature and the average surface heat flux compared with traditional circular rod fuel assembly. When boiling crisis was triggered, the linear heat flux of HCF assembly was higher than that of traditional fuel assembly. The preliminary thermal-mechanical analysis shows that the fuel rod is subject to great stress under irradiation for HCF elements with metal core without air gap.

Research and Development on Thermal Hydraulic and Safety of Supercritical Water-cooled Reactor

Zhao Xuebin, Huang Yanping, Zang Jinguang

2023, 44(5): 223-231. doi: 10.13832/j.jnpe.2023.05.0223

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Abstract:
Supercritical water cooled reactor (SCWR) is the only water-cooled reactor among the six advanced reactor types identified by the Generation IV International Forum. Supercritical water has a unique performance in thermal hydraulics due to its utilization as a coolant and unique properties in physical phase. This paper introduces the general requirements of SCWR thermal hydraulics, and describes the basic characteristics of typical thermal hydraulic processes and the main research and development progress at present. Focusing on the SCWR engineering application, this paper puts forward the follow-up research and development tasks and suggestions for future development of SCWR.

Experimental Study on Critical Flow of Supercritical Carbon Dioxide at Transient State

Zhang Dongxu, Li Weiqing, Zhao Minfu, Liang Peng, Xu Yongwang, Li Qingyuan, Duan Minghui

2023, 44(5): 232-236. doi: 10.13832/j.jnpe.2023.05.0232

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Abstract:
When a break accident occurs in a supercritical system, the high-temperature and high-pressure fluid in the system will have a critical flow at the break, and the transient critical flow characteristics have a great influence on the process of the accident. Compared with the steady state of critical flow, the pressure relief speed in the process of transient blowout has an important influence on the critical mass flow rate. Therefore, a transient test of critical flow with supercritical carbon dioxide (SCO₂) as working medium was carried out. A nozzle with a diameter of 2 mm, a length diameter ratio (L/D) of 3 and a rounded inlet was used as the test section. The initial pressure was 7.70~7.98 MPa and the initial temperature was 35.5-40.6℃. Transient tests with four depressurization velocities were carried out. The experimental results show that the greater the velocity of depressurization, the smaller the critical mass flow rate at the initial stage of blowout. However, at the later stage of blowout, the greater the velocity of depressurization, the greater the critical mass flow rate.

Experimental Study on Low-Cycle Fatigue Properties of Supercritical Water-cooled Reactor Candidate Cladding Material

Zhao Yuxiang, Xiong Ru, Liang Bo, Tang Rui

2023, 44(5): 237-243. doi: 10.13832/j.jnpe.2023.05.0237

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Abstract:
In order to obtain the low cycle fatigue performance data of 20Cr-25Ni, a candidate cladding material for supercritical water cooled reactor (SCWR), and to provide technical reference for the design, development and engineering of SCWR, the low cycle fatigue tests of 20Cr-25Ni at room temperature, 500℃, 650℃ and 800℃ were carried out on MTS809 testing machine, and the multi-level cyclic hysteresis curve, cyclic stress-strain curve, the relationship curve between stress amplitude and cyclic fraction, cyclic stress-strain model and low cycle fatigue Manson-Coffin model parameters were obtained. The results of low cycle fatigue test of 20Cr-25Ni show that the fatigue resistance of 20Cr-25Ni stainless steel pipe decreases with the increase of temperature, and shows obvious cyclic hardening at 500℃ and 650℃. Therefore, the cladding material has good fatigue resistance below 650℃. When the maximum core temperature of SCWR is higher than 650℃, it is necessary to be careful to use this material as cladding material.

Corrosion Behavior of a Novel Alumina-forming-austenitic Stainless Steel Exposed to Supercritical Water

Sun Dayun, Gao Yang, Zhang Lefu, Han Zhongli, Guo Xianglong

2023, 44(5): 244-250. doi: 10.13832/j.jnpe.2023.05.0244

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Abstract:
In order to enrich the research data on the corrosion behavior of alumina-forming-austenitic stainless steel (AFAs) in the supercritical water environment and provide support for SCWR cladding material evaluation, this study conducted the supercritical water corrosion test at 600℃/25 MPa for the independently designed AFAs, and analyzed its corrosion behavior by combining the microscopic analysis and characterization methods. The results show that the corrosion weight gain of AFAs exposed for 1000 h is 34 mg/dm², about half of that of 310S steel under the same conditions. There is a double-layer oxide forming on surface of AFAs. The outer layer mainly consists of Fe₂O₃ and Ni-rich spinel, and the inner layer is mainly composed of Cr₂O₃. Alumina exists in the layer as discrete particles, hindering the diffusion process and promoting the formation of Cr₂O₃. Therefore, AFAs in this study showed excellent corrosion resistance.

Effect of Microstructure on Corrosion Behavior of Alloy 800H in Supercritical Water

Huang Tao, Su Haozhan, Zhang Lefu, Chen kai

2023, 44(5): 251-258. doi: 10.13832/j.jnpe.2023.05.0251

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Abstract:
Alloy 800H is listed as one of the main candidate nuclear fuel cladding materials in the design of supercritical water-cooled reactor (SCWR), but its corrosion performance under application conditions is significantly affected by processing conditions. In this paper, the corrosion behavior of Alloy 800H in different states in supercritical water is studied by autoclave immersion test, microscopic characterization and mechanism analysis, and the effects of surface grinding and polishing state, cold deformation and grain size on its general corrosion behavior are obtained. The results show that surface rough grinding, cold deformation and grain refinement can significantly reduce the corrosion rate and cause the law of corrosion weight gain to change from parabolic to linear. Grain refinement improves the grain boundary density of the material, and the high diffusion rate of Cr near the grain boundary is conducive to the formation of the Cr₂O₃ protective layer, thus improving the corrosion resistance of the material. The shallow surface deformation layer left after surface rough grinding can be recrystallized into high-density nanocrystals at high temperature, which is conducive to the rapid formation of the surface Cr₂O₃ protective layer and has a significant inhibitory effect on the initial corrosion behavior. The cold deformation caused by rolling improves the grain boundary and dislocation density of material, which obviously enhances the long-term corrosion resistance of the cladding tube.

Microstructure and Tensile Properties of ODS-310 Austenitic Steel

Yin Chenxin, Jia Haodong, Zhou Zhangjian, Zheng Wenyue

2023, 44(5): 259-266. doi: 10.13832/j.jnpe.2023.05.0259

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Abstract:
In order to provide reliable nuclear fuel cladding materials for supercritical water-cooled reactors, ODS-310 austenitic steel with ultrafine grains and a large number of dispersed nano-oxide particles was prepared by mechanical alloying (MA) and hot isostatic pressing (HIP). The microstructure of the material after different heat treatment was analyzed by scanning electron microscopy (SEM), energy dispersive spectrometer (EDS) and transmission electron microscopy (TEM), and its tensile properties were tested. The results show that the dispersion strengthening particles in the material are spherical and mainly distributed in the grain and at the grain boundary. The average size is below 10 nm. It can be determined as Y₂Al₅O₁₂ by composition analysis and high resolution calibration. The grain structure of the sample can be obviously regulated by hot rolling plastic deformation combined with heat treatment. After heat treatment at 1100℃/120 h, the size and composition of the dispersed particles remain stable, and the particles have obvious pinning effect on dislocations. The prepared ODS-310 austenitic steel has high tensile strength and good thermal stability. The tensile strength of the samples before and after heat treatment at different temperatures is about 850 MPa, and the plasticity of the samples after heat treatment at 1100°C/120 h is significantly improved. This study shows that the tensile properties of ODS-310 austenitic steel are good, and the grain structure can be regulated by heat treatment. It can provide valuable data support for the study of ODS austenitic steel properties.

Effect of Water Chemistry on the Performance of Alloy 800H in Supercritical Water-cooled Reactor

Su Haozhan, Huang Tao, Zhang Lefu, Chen Kai

2023, 44(5): 267-274. doi: 10.13832/j.jnpe.2023.05.0267

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Abstract:
In order to study the influence of water chemistry factors such as temperature and dissolved oxygen on the service performance of cladding materials in supercritical water environment, and to find out the water chemistry control policy of supercritical water-cooled reactor, Alloy 800H was used as the experimental material in this paper, and the corrosion weight gain and slow strain rate tensile curve of the material in supercritical water under different temperature and water chemistry conditions were measured. Increasing the temperature would accelerate the corrosion rate of Alloy 800H, and the corrosion activation energy was about 159 kJ/mol. Increasing the temperature from 550℃ to 650℃, the yield strength of the material did not change significantly, which was about 175 MPa. However, the ratio of yield strength over tensile strength decreased significantly, showing an obvious softness trend of the material. At 650℃, raising the dissolved oxygen concentration from 0 μg·kg^–1 to 500 μg·kg^–1 resulted in an increase of corrosion weight gain of about 30%, but the water chemistry control method using hydrazine deoxygenation could reduce the corrosion rate of Alloy 800H. The dissolved oxygen concentration did not have a significant effect on the results of slow strain rate tensile testing, mainly because the stress corrosion failure of the material in supercritical water environment is dominated by creep process. The results indicate that appropriate controlling of temperature and dissolved oxygen in supercritical water environment can be helpful to reduce the corrosion rate of Alloy 800H and maintain its mechanical properties.

Effect of N and Al Addition on Microstructure and Mechanical Properties of Modified 25Ni-20Cr Austenitic Stainless Steel Aged at 700℃

Wang Qi, Chen Guoshuai, Zhou Zhangjian, Xiong Ru, Zheng Jiyun, Tang Rui, Zhang Lefu

2023, 44(5): 275-283. doi: 10.13832/j.jnpe.2023.05.0275

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Abstract:
The high temperature strength of the 25Ni-20Cr (S35140) austenitic stainless steel needs to be improved to meet the application requirements of supercritical water reactor (SCWR) for cladding materials. In this study, the properties of S35140 Steel were improved by microalloying, adding N and Al and aging at 700℃. The results showed that nano-NbN phase precipitated in N-added steel, and dislocation was pinched. With aging, the tensile strength at room temperature increased slightly, the elongation at room temperature almost remained unchanged, and the tensile strength at high temperature decreased slightly; however, the elongation at high temperature increased to 65%, and the impact energy still reached 111.39 J after aging for 120 h. A large number of NiAl and Laves phases precipitated in Al-added steel. With aging, the tensile strength at room temperature and high temperature increased significantly, and the tensile strength at room temperature even reached 1000 MPa, while the ductility and impact toughness decreased significantly. Therefore, in S35140 steel, adding N increased the ductility and toughness, while adding Al increased the strength, both of which significantly improved the mechanical properties of S35140 steel.

Reaxff-MD Simulation of the Effect of Incoherent Grain Boundaries and Its Segregation on Oxidation of 3C-SiC in Supercritical Carbon Dioxide

Zhou Qiyin, Liu Zhu, Zhang Lefu, Long Jiachen, Guo Xianglong

2023, 44(5): 284-289. doi: 10.13832/j.jnpe.2023.05.0284

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Abstract:
To understand the corrosion failure mechanism of silicon carbide (SiC) material in supercritical carbon dioxide (sCO₂) reactors, this study investigated the oxidation behavior of 3C-SiC in supercritical CO₂ environment through molecular dynamics simulation, and explored in depth the effect of element segregation at incoherent grain boundaries (GBs) on oxidation. The results show that the oxidation rate near the incoherent GBs is faster than that of single crystals, and the segregation of either silicon or carbon elements intensifies the oxidation at the incoherent GBs. The accelerated oxidation near incoherent GBs is attributed to the incompletely coordinated silicon atoms at GBs, and these silicon atoms are more likely to bond with oxygen atoms. Elemental segregation at incoherent GBs further enhances the oxidation rate of silicon carbide at incoherent GBs, where the segregation of silicon elements makes it more difficult to fully coordinate silicon atoms, which results in more silicon atoms at GBs with a lower positive charge, while the partial segregation of carbon leads to a larger free volume at the GBs, so the oxygen atoms can bond with the deeper silicon atoms. This study revealed the corrosion mechanism of 3C-SiC in sCO₂ and the reasons for the accelerated corrosion at incoherent GBs, providing theoretical support for the degradation mechanism of SiC materials in sCO₂ reactors.

Study on General Corrosion Behavior of Two Alumina-forming Austenitic Stainless Steels in Supercritical Carbon Dioxide

Liu Zhu, Zhou Qiyin, Zhang Lefu, Long Jiachen, Gao Yang, Guo Xianglong

2023, 44(5): 290-297. doi: 10.13832/j.jnpe.2023.05.0290

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Abstract:
To evaluate the application prospect of alumina-forming austenitic stainless steels in supercritical carbon dioxide nuclear reactors, the general corrosion behavior of two alumina-forming austenitic stainless steels (904L-2.5Al and 904L-3.5Al stainless steels) and its base steel (904L stainless steel) in supercritical carbon dioxide (sCO₂) at 600℃ and 20 MPa was investigated through experiments. The corrosion kinetics of materials are evaluated by weight gain method, and the morphology, structure and chemical composition of materials before and after corrosion are analyzed by scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. The results show that the weight gain of all materials approximates the parabolic growth law. The weight gain is lowered by increasing the content of Al, and 904L-3.5Al stainless steel has the lowest weight gain. After exposure, Fe-rich oxide is formed on the surface of 904L stainless steel, and carburization occurs. Continuous Cr/Al-rich oxide scales are formed on the surface of 904L-2.5Al and 904L-3.5Al stainless steels, while no carburization is occurred. Higher Al content facilitates the formation of protective Cr/Al-rich oxides on the material surface, which improves the oxidation and carburization resistance of materials after exposure to sCO₂.

2023 Vol. 44, No. 5