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2021 Vol. 42, No. 3

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Research of Optimization Technology for Equilibrium Cycle with Gadolinium
Ma Zirong, Su Jian, Zhou Sheng
2021, 42(3): 1-5. doi: 10.13832/j.jnpe.2021.03.0001
Abstract(316) PDF(98)
Abstract:
The design optimization of gadolinium rods is determined by analyzing the fuel management technology used for equilibrium cycle, and the optimization methods are to raise the bearing enrichment of the UO2-Gd2O3 pellets and loading fully enriched UO2 pellets at both ends of the gadolinium rods. The l...
Development and Verification of Fast Reactor Multi-Group Cross Section Database Processing Code MGGC1.0
Huang Zifeng, Ma Xubo, Zhu Runze, Li Yaozhou, Zhang Bin
2021, 42(3): 6-13. doi: 10.13832/j.jnpe.2021.03.0006
Abstract(412) PDF(74)
Abstract:
为产生高精度的快堆截面数据，基于一致性N阶的勒让德函数（PN）近似方法与临界曲率搜索方法，开发了快堆多群截面处理程序MGGC1.0，并进行了多方面基准验证。通过对均匀混合介质的宏观截面验证表明，中子产生截面的相对偏差均小于0.1%，裂变能谱的相对偏差均小于0.25%，总截面由于修正方式不同导致偏差稍大，但绝大多数能群的相对偏差都在0.5%以内。在临界基准实验中与蒙特卡罗程序RMC采用连续点截面的计算结果相比，78%的基准题的偏差都在100 pcm（1 pcm=10-5）以内，表明MGGC1.0处理截面的精度较好。在此基础上，采用钠冷快堆基准题BN-600进行计算，与基准题参考计算结果相比，输运...
Development and Validation of Integrate CHF Correlation Development System
Liu Wei, Li Zhigang, Lu Qi, Du Sijia, Liu Yu, Deng Jian, Hu Ying
2021, 42(3): 14-17. doi: 10.13832/j.jnpe.2021.03.0014
Abstract(389) PDF(60)
Abstract:
In the manual development of rod bundle CHF correlation, there are shortcomings and deficiencies such as numerous links, tedious process, large amount of data, high error rate and long computing time, so the integrate CHF correlation development system (ICODES) is researched and implemented by Nucle...
Prediction of Natural Convection Characteristics of Spent Fuel Vessel with Multi-Fluid Domains and High Temperature Difference Based on Non-Boussinesq Model
Long Teng, Zhang Guihe, Gao Chen, Liu Pan, Deng Xiaoyun, Jin Ting, Xiong Guangming
2021, 42(3): 18-25. doi: 10.13832/j.jnpe.2021.03.0018
Abstract(320) PDF(57)
Abstract:
Non-Boussinesq model is established based on Boussinesq model and full buoyancy model, to simulate the natural convection flow in multi-fluid domains with high temperature difference. The comparison between predicted profiles and experimental data in a single-fluid domain and in multi-fluid domains ...
Experimental and Numerical Investigation of Cavitation-Induced Choked Flow in Venturi Tube      
Zhou Junjie, Song Yuchen, Wang Dezhong, Yin Junlian
2021, 42(3): 26-33. doi: 10.13832/j.jnpe.2021.03.0025
Abstract(588) PDF(119)
Abstract:
When an over-flow condition occurs in a PWR nuclear power plant, the venturi flowmeter downstream of the make-up pump is required to form a cavitation-induced choked flow to protect the flowrate from exceeding the limit. Using the method of FLUENT numerical simulation and high-speed camera experimen...
Numerical Simulation of Leakage Characteristics of Pressurized Water Reactor Nuclear Power Plant
Yin Songtao, Wang Ningning, Wang Haijun, Zhu Mengxin
2021, 42(3): 32-37. doi: 10.13832/j.jnpe.2021.03.0032
Abstract(361) PDF(73)
Abstract:
This paper aims to propose a two-phase critical flow model to accurately evaluate the leakage rate of pipeline breaks. The critical flow model considering the nucleation of metastable liquids and non-equilibrium mass transfer is based on a two-fluid model coupled with an isothermal bubble growth mod...
Investigation on Reversed Flow Characteristics of U-Tube Steam Generator under Ocean Conditions
He Gening, Li Xiaojia, Cong Tenglong, Chen Yiran, Li Donghui, Wu Ge
2021, 42(3): 37-42. doi: 10.13832/j.jnpe.2021.03.0037
Abstract(149) PDF(45)
Abstract:
Based on RELAP5 program, a dynamic simulation model of U-tube steam generator (SG) under typical ocean conditions is established and verified. In the reactor start-up period, the effect of different rolling conditions on the flow in the U-tube of SG under low flow forced cycle condition is studied. ...
Investigation on Chaotic Evolution of Natural Circulation Pressure Drop Oscillation under Rolling Condition Based on Maximal Lyapunov Exponent
Lin Yuqi, Gao Puzhen, Li Zongyang, Zhang Yinxing, Wang Zhongyi
2021, 42(3): 42-48. doi: 10.13832/j.jnpe.2021.03.0042
Abstract(316) PDF(32)
Abstract:
In this paper, a series of experiments on the chaotic evolution of pressure drop oscillations (PDO) under rolling motion were designed and performed. The chaotic characteristic is analyzed based on maximal Lyapunov exponent (λ). It is concluded that the chaotic evolution route can be divided into 3 ...
Study on Effect of Anisotropic Scattering Cross Section on Sensitivity Coefficient Calculation for Fast Reactors
Wang Dongyong, Ma Xubo, Zhu Runze, Zhang Bin, Peng Xingjie, Wang Lianjie
2021, 42(3): 48-55. doi: 10.13832/j.jnpe.2021.03.0048
Abstract(283) PDF(47)
Abstract:
Due to the high energy of neutrons in fast-spectrum reactors, the anisotropic scattering of neutrons is with great effect on the calculation results. In this study, the effects of high-order scattering cross-section perturbation on the calculation of elastic and inelastic cross-section sensitivity c...
Study on Irradiation Effect for Value of Ag-In-Cd Control Rods
Zhang Lidong, Zhao Jun
2021, 42(3): 55-59. doi: 10.13832/j.jnpe.2021.03.0055
Abstract(307) PDF(76)
Abstract:
In order to study the main neutron absorber nuclides of the Ag-In-Cd control rods in the reactor and its effect on the control rod value, Monte Carlo method is used to simulate the burnup of the main nuclides in Ag-In-Cd control rods during the reactor operation, and the neutron flux in the control ...
Research of Water Packing Numerical Problem in Thermal Hydraulic System Code
Guo Yingran, Li Jiangkuan, Lin Meng, Yang Yanhua, Huang Tao
2021, 42(3): 59-63. doi: 10.13832/j.jnpe.2021.03.0059
Abstract(539) PDF(109)
Abstract:
Because of the discontinuous change in the compressibility between a two-phase mixture with a small void fraction and the pure liquid phase and the discrete method of the discrete momentum equation, the thermal hydraulic system program based on the two-fluid six-equation method may calculate a ficti...
Strategies in Station Blackout Accident for Small Modular Reactors
Qiu Zhifang, Li Feng, Deng Jian, Cheng Kun, Du Zhengyu, Wu Lingyan
2021, 42(3): 64-69. doi: 10.13832/j.jnpe.2021.03.0064
Abstract(277) PDF(46)
Abstract:
The ability of nuclear power plants to deal with Station Blackout (SBO) accident has attracted much attention after Fukushima accident. Whether there is a sufficient ability to mitigate SBO accident has become crucial to measure the safety performance of a nuclear power plant. As a new type of react...
Analysis of High-Temperature Transients in Severe Accident Depressurization Valve of 100 MW Pressurized Water Reactors
Wang Xiaoji, Wu Lingjun, Wu Qing, Liu Lili, Peng Huanhuan, Zou Zhiqiang
2021, 42(3): 69-74.
Abstract(284) PDF(46)
Abstract:
Due to the severe conditions in severe accidents of the nuclear power plants, the depressurization valves may experience a high temperature transient that the valves cannot withstand and may fail during the depressurization process. In this paper, the typical severe accident sequences that have a ce...
Static Elastoplastic Model of Metal Matrix Dispersion Fuel Element under Unstable Swelling Condition
Chen Hongsheng, Long Chongsheng, Xiao Hongxing
2021, 42(3): 74-80. doi: 10.13832/j.jnpe.2021.03.0074
Abstract(378) PDF(67)
Abstract:
For the unstable swelling of metal matrix dispersion fuel element induced by the cracking of metal matrix, the static elastoplastic model of crack surface was established without considering viscoplastic deformation, and this model was verified by the finite element simulation. The primary deformati...
Evaluation of Effect of Fuel Pellet Manufacturing Parameters on Fuel Rod Performance by Numerical Fitting Method
Wang Kun, Zhang Kun, Xing Shuo, He Liang, Yin Mingyang
2021, 42(3): 80-85. doi: 10.13832/j.jnpe.2021.03.0080
Abstract(292) PDF(86)
Abstract:
Based on the theoretical calculation model, the manufacturing parameters related to the performance of fuel rods are obtained. With the help of FUPAC fuel rod performance analysis software, the sensitivity analysis of the parameters is carried out one by one, and the key parameters affecting the per...
Development of Ultrasonic Transducer for Rod Cluster Control Assembly
Jin Xiaoming, Sun Jiawei, Li Bingqian, Hu Chenxu
2021, 42(3): 85-89. doi: 10.13832/j.jnpe.2021.03.0085
Abstract(364) PDF(64)
Abstract:
It is necessary to detect the defects of Rod Cluster Control Assembly used in nuclear power plant reactors to ensure the successful implementation in service inspection and reduce the cost of nondestructive testing. The ultrasonic probe for Rod Cluster Control Assembly (RCCA) inspection is developed...
A Hybrid CFD and Quasi-Static Theory Method for Fluidelastic Instability Prediction of a Tube Bundle
Song Lekun, Zhao Xielin, Zhou Jinxiong, Ye Xianhui, Feng Zhipeng, Xiong Furui
2021, 42(3): 89-94. doi: 10.13832/j.jnpe.2021.03.0090
Abstract(350) PDF(59)
Abstract:
In order to develop a method for predicting the fluidelastic instability of the tube bundle without relying on experiments, a hybrid fluidelastic instability prediction method for tube bundle is proposed by using CFD to capture drag and lift coefficients and their spatial derivatives, and substituti...
Comparative Study on Fatigue Crack Growth Performance of Nuclear Stainless Steel Weld Joints and Base Metal
Chang Haijun
2021, 42(3): 96-103. doi: 10.13832/j.jnpe.2021.03.0096
Abstract(288) PDF(40)
Abstract:
Welded joints are widely used on the pipe sockets in nuclear power plants, and fatigue cracks are one of the important causes resulting in the failure of welded joints. Therefore, it is of great significance to study the fatigue crack propagation and life prediction methods for welding zone material...
Research and Application of 3D Visual Reinforcement System for Nuclear Power Containment Shell
Zhang Jie, Zhou Jianqiu, Xu Xinwei, Liu Quanchang
2021, 42(3): 103-108. doi: 10.13832/j.jnpe.2021.03.0103
Abstract(166) PDF(45)
Abstract:
Based on the nuclear power project under construction, a three-dimensional visualized reinforcement system for nuclear power containment is designed and developed on the basis of PDMS which is an international general plant design software, and a digitized reinforcement algorithm is studied. The vis...
Study on Countermeasures for Effect of Flow-Induced Vibration Test of Reactor Internals in First Demonstration Project of HPR1000 on Total Construction Period
Wang Qilong, Ma Ying, Xing Hui, Sun Chuanyi
2021, 42(3): 108-116. doi: 10.13832/j.jnpe.2021.03.0108
Abstract(241) PDF(52)
Abstract:
As the FOAK, HPR1000 is required to carry out the in-reactor measurement of flow induced vibration of reactor internals based on the regulation Comprehensive Vibration Assessment Program for Reactor Internals during Pre-operational and Start-up Testing (RG1.20) . The evaluation shows that this will ...
Study on Small Power Improving of HPR1000
Xiang Meiqiong, Zhu Jialiang, Liu Yanyang, Qing Xianguo, He Zhengxi, Wu Qian, Zhu Biwei, Lyu Xin
2021, 42(3): 115-122. doi: 10.13832/j.jnpe.2021.03.0115
Abstract(431) PDF(85)
Abstract:
The thermal power accuracy of HPR1000 was analyzed, and the contribution of steam generator outlet pressure measurement accuracy, feed water temperature measurement accuracy and feed water flow measurement accuracy to HPR1000 thermal power accuracy was calculated. The quantitative data proves that t...
Research on Double Containment Annulus Leakage Rate Test
He Rui, Shen Dongming, Li Shaochun, Chen Wei, Huang Xiaoming
2021, 42(3): 121-126. doi: 10.13832/j.jnpe.2021.03.0121
Abstract(334) PDF(42)
Abstract:
The tightness of the double containment annulus is critical to the safety of nuclear power plants, and the test procedure and data analysis should be reliable and reasonable. The formulation of the pressure in double containment annulus versus time is deduced in this paper with the foundation of the...
Application of SPAR-H Method in Human Reliability Analysis of  Digital Nuclear Power Plants
Qing Tao, , Liu Zhaopeng, Zhang Li, Tang Yaqin, Hu Hong, Zang Jing
2021, 42(3): 126-132. doi: 10.13832/j.jnpe.2021.03.0126
Abstract(584) PDF(73)
Abstract:
The applicability of SPAR-H method in the digital nuclear power plant has not been fully studied. This paper studied the operator behavior characteristics of the digital nuclear power plant and the application of SPAR-H method in Lingdong Nuclear Power Plant. The study results show that SPAR-H metho...
Research on Covert Attack Method in Large Pressurized Heavy Water Reactors
Zhang Yan, Fan Dengning, Huang Yu, Wang Dongfeng, Xu Peihao
2021, 42(3): 132-140. doi: 10.13832/j.jnpe.2021.03.0132
Abstract(472) PDF(41)
Abstract:
In order to facilitate the research and development of the security defense system for large pressurized heavy water reactors (PHWR), this paper studies the potential attack mode in PHWR networked control system and proposes a covert attack approach based on Gaussian process regression model optimiz...
Research on Feedforward Compensation for Steam Generator Level Control System Manual/Automatic Switch
Xu Ying, Chen Jiancai, Yu Hang, Wang Zhixian
2021, 42(3): 140-145. doi: 10.13832/j.jnpe.2021.03.0140
Abstract(299) PDF(56)
Abstract:
Based on the current design of the replication loop for the steam generator level control system, after the main feed water flow regulating valve is switched from manual mode to automatic mode, the calculation basis of the level controller is the steam water mismatch signal during switching, which c...
Research of Quantification Method of Risk in Seismic Probabilistic Safety Analysis in Nuclear Power Plants
Jing Xu, Xiao Jun
2021, 42(3): 145-150. doi: 10.13832/j.jnpe.2021.03.0145
Abstract(338) PDF(46)
Abstract:
The current status of quantification method and tools for seismic probabilistic safety assessment (PSA) in nuclear power plants was discussed, and the challenges faced by quantitative tools and the issues need to be resolved was suggested. A quantitative method based on the nature of probability the...
Study on Level 2 PSA Release Categories and Selection of Representative Accident Sequence in Nuclear Power Plants
Zhang Jiajia, He Dongyu, Gong Yu, Luo Yong, Chen Peng, Chen Yingying
2021, 42(3): 149-154. doi: 10.13832/j.jnpe.2021.03.0150
Abstract(250) PDF(43)
Abstract:
Domestic nuclear power projects such as AP1000, EPR and HPR1000 have used the level 2 PSA source terms for the emergency input, but there is no clear operational methods for the division of level 2 PSA release categories and the selection of representative accident sequences for each release categor...
Model Study on Hydrogen Stratification Behavior within a Containment
Peng Cheng, Deng Jian
2021, 42(3): 155-160. doi: 10.13832/j.jnpe.2021.03.0155
Abstract(638) PDF(54)
Abstract:
In this paper, with theoretical modeling and experimental correlation, a semi-empirical model which can predict the hydrogen distribution characteristics has been proposed, based on the dominant factors of interaction among the inertial force, viscous force and buoyancy under the injection of both s...
Study on Online Monitoring of Equipment Condition Based on Local Outlier Factor and Artificial Neural Networks Model
Shen Jiangfei, Li Huaizhou, Huang Lijun, Mao Xiaoming, Zhang Sheng
2021, 42(3): 160-166. doi: 10.13832/j.jnpe.2021.03.0160
Abstract(441) PDF(72)
Abstract:
The centralized online monitoring technology plays the most important role in nuclear power plants for the safety of major equipments and economic operation. In order to solve the false alarm and alarm failure problems in the traditional online monitoring, a new artificial intelligence monitoring me...
Disposal of Abnormal Tension Event of Main Pump Snubber Tie Rod
Chen Sun, Zhang Yihan, Zhao Xiaohong, Li Shilei
2021, 42(3): 166-171. doi: 10.13832/j.jnpe.2021.03.0166
Abstract(243) PDF(37)
Abstract:
Abnormal tension occurs on the main pump snubber tie rod of a nuclear station when it is loaded by bolt stretcher. In order to solve this problem, this paper uses the methods of cause elimination and analytical calculation to diagnoze the fault. It is identified that the main reason for abnormal ten...
Preliminary Study of General Design of Floating Nuclear Power Plants
Chen Yanxia, Zhu Chenghua, Guo Jian, You Xiaojian, Zhang Jincai, Tan Mei, Li Pengfan
2021, 42(3): 171-177. doi: 10.13832/j.jnpe.2021.03.0171
Abstract(429) PDF(88)
Abstract:
Based on the analysis of the advantages and disadvantages of the existing reactor types in the world and the application on ships, this paper suggests that the mature PWR can be used in the floating nuclear power plants on the sea, and the principle suggestions on the reactor power and refueling cyc...
Study on Design and Safety Analysis of Spent Fuel Storage Grid for Marine Nuclear Power Platform
Mei Zhen, Sun Fujiang, Zhu Gang, Yu Ying, Chen Juan, Lu You
2021, 42(3): 177-183. doi: 10.13832/j.jnpe.2021.03.0177
Abstract(165) PDF(37)
Abstract:
For the problem in the safety assurance for long-term maritime storage of spent fuel for marine nuclear power platform, this paper improved the fixed form between the fuel assembly and the storage cell, optimized the connection form between the storage cell and the spent fuel storage grid frame body...
Fault Diagnosis of Vibration Induced by Fluid of 100D Main Pump for CPR1000 Unit
Shu Xiangting, Yang Zhang, Xu Yizhe, Jiang Yanlong
2021, 42(3): 183-188. doi: 10.13832/j.jnpe.2021.03.0183
Abstract(254) PDF(62)
Abstract:
The vibration phenomenon of 100D main pump under various operation modes for CPR1000 unit show that, when the main pump motor is in normal shutdown condition of steam generator cooling or residual heat removal cooling, the amplitude value of the main pump motor tile often has a large range of impact...
Comprehensive Leakage Diagnosis Technology of Primary Loop Pressure Boundary of Nuclear Power Plants Based on Leakage Monitoring Data Synthesis
Ling Jun, Yang Yutao, Li Hongxia, Zang Yiming, Tan Ke, Yuan Jingqi
2021, 42(3): 188-193. doi: 10.13832/j.jnpe.2021.03.0188
Abstract(429) PDF(78)
Abstract:
The leakage monitoring system is used to monitor the integrity of the reactor coolant system pressure boundary (RCPB), and is also a prerequisite for the application of leak before break (LBB) technology. Leakage comprehensive diagnosis is the core function of leakage monitoring system. In this pape...
Research on Efficient Compound Decontamination Technology for  Decommissioning of Highly Radioactive Hot Cell
Jia Haopeng, Teng Lei, , Wang Shuai
2021, 42(3): 193-197. doi: 10.13832/j.jnpe.2021.03.0193
Abstract(286) PDF(53)
Abstract:
The highly radioactive hot cell is used as an auxiliary facility for the irradiation inspection of reactor materials, and the hot cells are with high radiation level, complicated structure and great difficulty in decontamination. In view of the particularity and complexity of the decontamination of ...
Research on Accident Diagnosis Method for Reactor Primary Circuit System Based on SDG and PCA
Ma Jie, Zhang Longfei, Yu Ren, Peng Qiao, Hu Pengfei
2021, 42(3): 197-203. doi: 10.13832/j.jnpe.2021.03.0197
Abstract(327) PDF(60)
Abstract:
Research on Hot Leg Temperature Mixing and Measurement Characteristics of Generation III Advanced Reactor
Ren Chunming, Du Sijia, Deng Jian, Wu Qing, Xin Sufang, Hu Ying, Liu Xiaobo
2021, 42(3): 203-207. doi: 10.13832/j.jnpe.2021.03.0203
Abstract(251) PDF(43)
Abstract:
In order to predict the rationality of thermometers setting on hot leg of Generation Ⅲ Advanced Reactor,the hot leg temperature mixing and measurement characteristics was studied for conditions with different core outlet temperature and flowrate distributions, using computational fluid dynamic (CFD)...
First Collision Compensation Technique for Leakage Source
Tang Xiao, Li Qing, Chen Zhang, Chai Xiaoming, Tu Xiaolan, Wang Liangzi, Li Mancang
2021, 42(3): 207-211. doi: 10.13832/j.jnpe.2021.03.0207
Abstract(176) PDF(40)
Abstract:
In order to solve the problem of convergence instability of two-dimensional / one-dimensional neutron transport calculation, the first collision compensation for leakage source technique is proposed. The source term is equivalent to the scattering source of each region by the first collision method,...
Review on Development of Critical Heat Flux Mechanistic Model
Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Deng Jian, Hu Ying, Liu Xiaobo
2021, 42(3): 211-218. doi: 10.13832/j.jnpe.2021.03.0211
Abstract(1832) PDF(227)
Abstract:
In order to clarify the development of critical heat flux (CHF) mechanistic model and promote the CHF experimental and theoretical research, the existing achievements and progress of CHF mechanistic model are systematically reviewed and the basic assumptions and detailed modeling process of each mod...
Development and Application of COSINE Reactor Monte Carlo Code cosRMC
Yu Hui, Quan Guoping, Qin Yao, Yan Yiman, Chen Yixue
2021, 42(3): 218-224. doi: 10.13832/j.jnpe.2021.03.0218
Abstract(397) PDF(62)
Abstract:
The self-developed COSINE Monte Carlo code cosRMC is designed for the analysis of the reactor core and the radiation shielding. The typical functions as transport simulation, burnup calculation, group constants generation, sensitivity & uncertainty analysis and visual modeling have been sufficie...
Research on Optimization of Core Fuel Management for  Units 1-4 of Tianwan NPP
Guo Zhipeng, Wu Jinying, Xu Min, Zhang Haoran, Yi Xuan, Huang Peng, Ye Liusuo
2021, 42(3): 224-229. doi: 10.13832/j.jnpe.2021.03.0224
Abstract(387) PDF(269)
Abstract:
This paper has done research on the optimization of core fuel management for units 1-4 of Tianwan NPP by KASKAD program package to unify the fuel management schemes. In the optimizing schemes, less varieties and quantities of fresh fuel assembles are used, however the average enrichment of new FAs i...
Generalized Perturbation-Theory-Based Sensitivity Analysis  with CMFD Acceleration
Wu Qu, Peng Xingjie, Yu Yingrui, Li Qing
2021, 42(3): 229-233.
Abstract(321) PDF(39)
Abstract:
To perform the generalized sensitivity analysis for nuclear data in a reactor physics design code, KYLIN-Ⅱ, the generalized perturbation theory is adopted and several generalized fix-source equations with the orthogonal definite condition need to be solved when the sensitivity coefficients are figur...
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