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Special Contribution
Practice and Learning in the Field of Structural Integrity during the HPR1000 International Regulatory and Certified Review
Mao Qing, Zhang Tao, Xu Xiao, Cen Peng, Wang Guofeng
2024, 45(2): 1-9.   doi: 10.13832/j.jnpe.2024.02.0001
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In order to promote the international market development, China General Nuclear Power Group (CGN) has carried out the Generic Design Assessment (GDA) of UK and the European Utility Requirements (EUR) assessment. The structural integrity of the structures, systems and components in nuclear power plant is the focus of international review and certification. In order to meet the requirements of British nuclear safety supervision and EUR, the project team adopted the design concept of analytical method, and completed the structural integrity analysis, evaluation and design improvement of British and European versions of HPR1000 with mechanical analysis, test and demonstration as the key means, which fully met the requirements of British nuclear safety supervision and EUR. Through the HPR1000 international regulatory and certified review, the relevant technical requirements in the field of structural integrity are studied, and the practice and learning in the review process are summarized, which can support the continuous improvement and innovation of HPR1000 technology.
Reactor Physics
Research on the Source Convergence Diagnose Method of Monte Carlo Critical Calculation for Loosely Coupled System
Zhang Yin, Cheng Yuting, Zhou Qi, Zhu Qingfu, Xia Zhaodong, Ning Tong, Zhang Zhenyang
2024, 45(2): 10-18.   doi: 10.13832/j.jnpe.2024.02.0010
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In order to improve the reliability and accuracy of Monte Carlo code in the critical safety calculation of loosely coupled systems, it is necessary to determine the source convergence of the results. In this paper, an improved Shannon entropy diagnose method considering the distribution of fission sources and the statistical deviation weight factor is proposed. The relative deviation of fission sources distribution between adjacent generations and the standard deviation of fission source distribution during source iteration are used as the weight of fission source distribution. Finally, an improved Shannon entropy convergence index is established, which makes up for the shortcomings of traditional Shannon entropy for lacking the details of local fission source. The improved Shannon entropy convergence index is applied to the benchmark problem: spent fuel rod and loosely coupled solution slabs published by OECD/NEA. The results show that compared with the traditional Shannon entropy index, the improved Shannon entropy is more sensitive to the iterative convergence process of fission source, and can more intuitively and accurately determine the convergence of fission source distribution accompanied by source iteration. For a typical case with slow convergence speed, the conclusion of pseudo-convergence is given with the traditional Shannon entropy. The improved Shannon entropy can accurately determine the number of iterations when the source iteration reaches convergence.
Thermal and Hydraulic
Study on Load-Following Characteristics of a Thermionic Space Reactor Power System
Han Xufan, Ouyang Zeyu, Wang Zhao, Shan Jianqiang
2024, 45(2): 63-71.   doi: 10.13832/j.jnpe.2024.02.0063
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In order to analyze the power load-following characteristics of thermionic space reactor power system, TOPAZ-II thermionic space reactor system code is established by using the simulation language of reusable hierarchical component model. The influence of Cs steam pressure and electrode gap on the output power is analyzed. The load tracking operation characteristics of thermionic space reactor power supply in orbit under different load changes are analyzed by using the method of core reactivity feedback and external load resistance collaborative control. The results show that the moderator temperature does not change significantly when the steady-state electric power is 5.5 kW and the electric power varies between 0.95 kW and 7.25 kW, and the reactor has self-stability. Beyond this range, the reactor loses self-stability, which is closely related to the positive temperature reactivity feedback of the moderator.
Structural Mechanics and Safety Control
Model Analysis of Fuel Assembly Grid Spring Stiffness
Jin Yuan, Zhou Sai, Chen Wei, Li Weicai, Zhang Yuxiang
2024, 45(2): 154-159.   doi: 10.13832/j.jnpe.2024.02.0154
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Grid spring is a critical component of PWR fuel assembly, which provides clamp function for the fuel rod. Stiffness is a critical characteristic of grid spring, which is related to the in-pile operational performance of the fuel rod. According to the grid spring structure and mechanical characteristics, this paper proposes a mechanical analysis model of the grid spring, and the theoretical stiffness calculation formula is obtained through deduction, and the grid spring theoretical stiffness model is also obtained. Furthermore, the finite element stiffness models of grid spring with various sizes are established in the commercial finite element code ABAQUS, and the deformation and stiffness curves of grid spring are obtained by calculation. By comparison between the FEM analysis results and the theoretical calculation ones, the rationality of the theoretical stiffness model is proved, and then the advantages and shortcomings of the theoretical stiffness model are analyzed. The theoretical stiffness model of grid spring proposed for the first time in this paper can be used to replace the finite element iterative process during the design of grid scheme, and the main design dimensions of the optimized scheme can be obtained quickly. However, the theoretical method in this paper cannot replace the experiments, and tests are still needed to determine the stiffness of the grid spring after the scheme is solidified. The grid spring theoretical model in this paper provides a new idea to quickly optimize the fuel assembly grid spring design.
Circuit Equipment and Operation Maintenance
Research and Practice on Transformation Scheme of Hand Operator in Main Control Room of Nuclear Power Plant Based on Digital System
Xu Ying, Yang Wu, Liu Shengzhi, Jiang Xiaolong, Zhao Ke, Yang Jin
2024, 45(2): 187-192.   doi: 10.13832/j.jnpe.2024.02.0187
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The instrument and control system of Daya Bay Nuclear Power Station is based on analog electronic technology, and it is planned to be digitized during the 30-year overhaul, during which the first-floor analog system will be completely digitized, but the man-machine interface equipment on the second floor will still be controlled by the hard hand operator in the main control room. According to the market research, four transformation schemes of hand operator based on digital system are preliminarily determined. Finally, the overall transformation scheme of developing new hand operator through standard I/O board interface is determined by considering many factors such as functional consistency, product reliability, technical feasibility, development cost, input/output (I/O) signal distribution efficiency and application feedback. This scheme can realize the optimal interface between the hand operator and the transfomed digital system, and effectively improve the utilization efficiency of I/O board channels. At the same time, through the rich software algorithm block function and digital system fault diagnosis function, the improvement measures such as automatic undisturbed switching of actuator and forced manual switching of process signal quality level are added, which effectively improves the reliability and stability of unit equipment control and provides an important reference scheme for similar digital transformation projects.
Column of Science and Technology on Reactor System Design Technology Laboratory
Analysis on Water Ingress Accident of a Gas Cooled Reactor
Ma Yugao, Cao Zhongbin, Wang Jinyu, Deng Jian, Bao Hui, Ding Shuhua, Cheng Kun, Hu Wenzhen
2024, 45(2): 241-247.   doi: 10.13832/j.jnpe.2024.02.0241
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The gas cooled reactor is featured with high inherent safety, small size, and a simple start-up process. However, the water ingress accident might occur due to the influence of the working environment or operating status. Based on the design scheme of the Submersion-Subcritical Safe Space(S4) reactor, this work simulated and analyzed the water ingress accident caused by the rupture of the heat transfer tube of the condenser under normal operating conditions, and studied the accident consequences such as the introduction of positive reactivity, the overpressure of the Brayton cycle. This work calculated the influence of spectral shift materials on reactivity introduction during the water ingress process with the Reactor Monte Carlo code RMC. And the temperature and Brayton cycle pressure were calculated during the water ingress process with the gas cooled reactor transient analysis code, HXRTRAN. The results show that when a water ingress accident occurs, 0.5 kg water ingress causes the pressure of the Brayton cycle to exceed 10 MPa, which may lead to larger damage to the condenser pipeline and secondary seawater injection. Meanwhile, water ingress may lead to a large amount of positive reactivity introduction. If spectral shift absorbers, Ir, are added to the fuel surface in the reactor, the core may reduce power or even subcritical shutdown spontaneously in the water ingress accident. When the amount of water vapor exceeds 5 kg, the core power quickly decreases to about 2.2% FP and gradually approaches shutdown. Therefore, the spectral shift materials, Ir, have a significant inhibition effect on the introduction of positive reactivity caused by water ingress of the core.
Development Status and Outlook for Nuclear Power in China
Zhao Chengkun
2018, 39(5): 1-3.   doi: 10.13832/j.jnpe.2018.05.0001
[Abstract](1053) [PDF 0KB](13)
Initiation and Development of Heat Pipe Cooled Reactor
Yu Hongxing, Ma Yugao, Zhang Zhuohua, Chai Xiaoming
2019, 40(4): 1-8.  
[Abstract](1398) [PDF 1128KB](425)
Present Situation and Prospect of Radioactive Waste Liquid Treatment Technology
Sun Shouhua, Ran Mingdong, Lin Li, Liu Wenlei, Li Zhenchen, Li Wenyu
2019, 40(6): 1-6.   doi: 10.13832/j.jnpe.2019.06.0001
[Abstract](1223) [PDF 178KB](716)
Reliability of Digital Pressure Control Device of Nuclear Pressurizer Based on Dynamic Fault Tree
Qian Hong, Gu Yaqi, Liu Xinjie
2019, 40(3): 103-108.   doi: 10.13832/j.jnpe.2019.03.0103
[Abstract](1001) [PDF 0KB](2)
General Technology Features of Reactor Core and Safety Systems Design of HPR1000
Yu Hongxing, Zhou Jinman, Leng Guijun, Deng Jian, Liu Yu, Wu Qing, Liu Wei
2019, 40(1): 1-7.   doi: 10.13832/j.jnpe.2019.01.0001
[Abstract](858) [PDF 0KB](15)
Design of Intelligent Wireless Vibration Sensor for Wireless Monitoring of Rotating Device Operating Condition
Yu Ren, Xie Xuyang, Qing Fatao, Peng Qiao, Wang Tianshu
2020, 41(3): 221-226.   doi: 10.13832/j.jnpe.2020.03.0221
[Abstract](252) [PDF 0KB](5)
Research on Condition Monitoring Technology for Nuclear Power Plant Equipment Based on Kernel Principal Component Analysis
Wu Tianhao, Liu Tao, Shi Haining, Zhang Tao, Tang Tang
2020, 41(5): 132-137.  
[Abstract](324) [PDF 0KB](7)
Analysis of Human Errors in Severe Accident of Nuclear Power Plant Based on Cognitive Model and Fault Tree
Zhang Li, Chen Shuai, Qing Tao, Sun Jing, Liu Zhaopeng
2020, 41(3): 137-142.   doi: 10.13832/j.jnpe.2020.03.0137
[Abstract](378) [PDF 0KB](3)
Study on Process and Properties of Pulse Laser Prepared Cr Coating for Accident Tolerant Fuel Claddings
Li Rui, Liu Tong
2019, 40(1): 74-77.   doi: 10.13832/j.jnpe.2019.01.0074
[Abstract](280) [PDF 0KB](3)
介绍了中广核研究院在事故容错燃料(ATF)包壳领域的最新成果,通过预置粉末式脉冲激光熔覆技术,在不同的功率下制备出不同厚度的锆包壳管Cr保护层;通过高温蒸汽氧化增重数据发现,采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能,证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构,采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
Research of Numerical Simulation Accuracy Based on Performance Prediction of Nuclear Main Pump
Hu Xiaodong, Wang Xiuyong, Liu Zailun, Zhang Xiaofei, Li Yibin
2019, 40(4): 127-133.  
[Abstract](327) [PDF 0KB](1)
      为提高核主泵在全工况点的数值模拟精度,研究了数值模拟过程中近壁面网格尺度、湍流模型、流动状态3种因素对计算精度的影响。结果表明,在定常状态下,重整化群(RNG) k-ε湍流模型和标准壁面函数法在近壁面网格尺度(y+)为50左右时具有较高的计算精度,并且其计算精度高于RNG k-ε增强壁面函数法、低雷诺数k-ε和剪切应力传输(SST)k-ω这3种湍流模型的计算精度,但上述不同网格尺度和湍流模型的计算结果均存在较大的计算误差;采用非定常计算时的计算精度明显高于定常计算,能够反映出扬程曲线在关死点附近的驼峰现象,效率的计算精度也有一定改善,更适合于对核主泵进行性能预测。
Evelopment Characteristics and Inspiration of Marine Nuclear Power
Lu Chuan, Wang Zhonghui, Yu Junchong
2022, 43(1): 1-6.   doi: 10.13832/j.jnpe.2022.01.0001
[Abstract](2396) [FullText HTML](369) [PDF 2825KB](369)
The marine nuclear power technology of the United States and Russia has been leading the world for a long time, and their development experience and technical context have high reference value. Through the analysis and research on the main development process and technology of marine nuclear power in the United States and Russia, this paper innovatively summarizes the common development laws of marine nuclear power in the United States and Russia, such as basic type of reactor system, general test platform and differential configuration from the aspects of technical route and trend, a series of common and differential characteristics followed by marine nuclear power technology in the United States and Russia are excavated and refined, which can provide some reference and enlightenment for the development of marine nuclear power.
Digital Reactor: Development and Challenges
Yu Hongxing, Li Wenjie, Chai Xiaoming, Li Songwei
2020, 41(4): 1-7.  
[Abstract](1167) [PDF 466KB](86)
The digital reactor is an integral numerical simulation platform for the performance of nuclear reactor systems. In the first part of this paper, the development history of the nuclear reactor simulation technology is reviewed. The three technical elements constituting the digital reactor are elaborated, including the target scenario, advanced models and multi-physics coupling technology, and the integrating environment. Although there are several challenges for the development of digital reactors, such as the difficulties in multi-physics and multi-scale computation, the complexity in design optimization, and the insufficient database, the digital reactor can help better analyze key problems that limiting the reactor performances and safety, and better understand the mechanism of  the phenomena that cannot be observed or measured experimentally.
Nuclear Power AI Applications: Status, Challenges and Opportunities
Zhang Heng, Lyu Xue, Liu Dong, Wang Guoyin, Hang Qin, Sha Rui, Guo Bin
2023, 44(1): 1-8.   doi: 10.13832/j.jnpe.2023.01.0001
[Abstract](8367) [FullText HTML](431) [PDF 2166KB](431)
In recent years, artificial intelligence (AI) technology has been widely used in the field of nuclear power to promote nuclear power plants to achieve the goal of improving production efficiency, reducing operating costs and improving operating safety through self diagnosis, self optimization and self adaptation. This paper introduces the AI technology often used in the nuclear power field, summarizes its research status in four typical application scenarios of the nuclear industry, namely, intelligent mine, intelligent design, intelligent manufacturing and intelligent operation and maintenance. Finally, it analyzes the challenges and development trends of the application of AI technology in the nuclear power field from three aspects: data samples, network security, and the explanatory nature of deep learning.
Present Situation and Prospect of Radioactive Waste Liquid Treatment Technology
Sun Shouhua, Ran Mingdong, Lin Li, Liu Wenlei, Li Zhenchen, Li Wenyu
2019, 40(6): 1-6.   doi: 10.13832/j.jnpe.2019.06.0001
[Abstract](1223) [PDF 178KB](40)
The effective disposal of radioactive waste liquid is the precondition for the rapid development of nuclear industry all over the world, and the current situation and development direction of its key technologies are the focus of attention of the nuclear industry in China. This paper introduces several traditional methods of radioactive waste liquid treatment and the emerging new technology options, summarizes the principles, advantages and disadvantages of various methods, and discusses the research direction and development trend of radioactive waste liquid treatment technology in the future.
Key Technology of ACP100: Reactor Core and Safety Design
Song Danrong, Li Qing, Qin Dong, Dang Gaojian, Zeng Chang, Li Song, Xiao Renjie, Wei Xuedong
2021, 42(4): 1-5.   doi: 10.13832/j.jnpe.2021.04.0001
[Abstract](3643) [FullText HTML](889) [PDF 4887KB](889)
Small modular reactor is a new kind of nuclear energy system. The ACP100 is a multi-purpose modular small PWR with full intellectual property in China. This paper introduces the research and development process, the main characteristics of the reactor core and safety design technology, mainly including the nuclear design, thermal-hydraulic design, safety design concept, inherent safety design, and the strategy for accidents. Through the combination of the deterministic theory and the probabilistic safety assessment, the safety of ACP100 is greatly improved and exceeds the Generation 3 nuclear power plant safety standards.
Radiation Safety System and Its Design Coordination for Nuclear-Powered Ship
Lin Xiaoling
2019, 40(5): 1-5.  
[Abstract](530) [PDF 550KB](25)
Radiation Safety is of great importance for the combat effectiveness of nuclear-powered ships, while overemphasis on radiation safety could influence the performance of carriers negatively. As a result, the optimization of radiation protection is the crucial task in the design of nuclear-powered ships. To achieve this goal, the system of radiation safety for nuclear-powered ships is constructed. The function, main problems to be considered, the control index, the relations and cooperation from inside and outside is analyzed.
Research on the Development Trend of Micro Nuclear Reactor Technology
Du Shuhong, Li Yonghua, Sun Tao, Wang Jun, Liu Xiaowen, Su Gang, Zhao Depeng
2022, 43(4): 1-4.   doi: 10.13832/j.jnpe.2022.04.0001
[Abstract](1390) [FullText HTML](172) [PDF 2053KB](172)
Micro nuclear reactors adopt Generation-IV non-light water reactors, heat pipe reactors and Generation-III light water reactors with high inherent safety, providing long-term and highly reliable power supply for innovative scenario such as remote islands, mining areas, border guard posts and bases, emergency and disaster relief, space exploration and deep-sea exploration. They have broad application prospects, being one of the important technical supports to realize the national strategy. This study summarizes the definition and main R & D reactor types of micro nuclear reactors, and describes the innovative technological characteristics of micro nuclear reactors, such as high inherent safety, easy modularization and expansion, transportability, easy deployment, independent operation and so on, analyzes the development trend of key technologies such as new fuel, integration of main loop, new thermoelectric conversion device, passive safety system, intelligent operation and maintenance and coupling of nuclear energy and other energy sources in China, providing support for the formulation of the technical route for the development of micro nuclear reactors in China.
Initiation and Development of Heat Pipe Cooled Reactor
Yu Hongxing, Ma Yugao, Zhang Zhuohua, Chai Xiaoming
2019, 40(4): 1-8.  
[Abstract](1398) [PDF 1128KB](201)
The heat pipe cooled reactor adopts the solid-state reactor design concept and passively transfer the heat out of the core through heat pipes. This paper summarizes the development history of the heat pipe cooled reactor, from the conceptual initiation, the active exploration and to the breakthrough. The technical characteristics of heat pipe cooled reactors are analyzed, including the key advantages, such as solid properties, inherent safety, simple operation, easy modularization and expansion, and transportability. In addition, this paper summarizes the technical status of heat pipe performance, material technology and energy conversion in heat pipe cooled reactors. The challenges in the further development of heat pipe cooled reactors are put forward, such as material technique, manufacturing, and operation maintainability. The future development trend of heat pipe cooled reactors is clarified, which provides a direction for the development and application of the innovative heat pipe cooled reactor technology. Overall, the heat pipe cooled reactor has broad application prospects in deep space exploration and propulsion, land-based nuclear power supply, sea exploration and other scenarios,  which may become one of the most creative technologies to change the future nuclear power patterns.
CFD Investigation on Flow and Heat Transfer Characteristics of Fuel Assembly for VVER Reactor
Wang Xiong, Du Daiquan, Zeng Xiaokang, Yang Xiaoqiang, Zan Yuanfeng
2018, 39(3): 6-9.   doi: 10.13832/j.jnpe.2018.03.0006
[Abstract](1198) [PDF 907KB](37)
The flow and heat transfer characteristics of AFA fuel assembly for VVER reactors have been investigated using computational fluid dynamics (CFD) simulation. The flow field, pressure drop and temperature distribution of the coolant in AFA under normal regime have been calculated. The results show that the pressure drop of the spacer grid of AFA is lower than that of the grid having mixing vane. The stagnation zone of coolant appears around the rim of the spacer grid and causes higher temperature in the periphery region of AFA. The power ratio of the circumferential pin around instrumental tube (Kc) with different values has a great effect on the measured temperature of the coolant at FA outlet. The results can be referred in the setting of temperature warning value (ΔTt) for the reactor core during the operation of nuclear power plants.
Solving Multi-Dimensional Neutron Diffusion Equation Using Deep Machine Learning Technology Based on PINN Model
Liu Dong, Luo Qi, Tang Lei, An Ping, Yang Fan
2022, 43(2): 1-8.   doi: 10.13832/j.jnpe.2022.02.0001
[Abstract](2831) [FullText HTML](340) [PDF 31918KB](340)
This paper elaborates the physics-informed neural network model (PINN), constructs a deep neural network as a trial function, substitutes it into the neutron diffusion equation to form a residual, and takes it as the weighted loss function of machine learning, and then approaches the numerical solution of the neutron diffusion equation by deep machine learning technique; According to the characteristics of diffusion equation, this paper puts forward innovative key technologies such as accelerated convergence method of eigenvalue equation, efficient parallel search technology of effective multiplication coefficient (keff), learning sample grid point uneven distribution strategy, and analyzes the sensitivity of key parameters such as neural network depth, neuron number, boundary condition loss function weight and so on. The verification calculation results show that the method has good accuracy, and the proposed key technology has remarkable results, and explores a new technical approach for the numerical solution of the neutron diffusion equation.