核动力工程

Evalution and Application of Probabilistic Safety Assessment in Nuclear Energy Safety Analysis

Yu Hongxing, Wu Lingjun, Deng Chunrui, Deng Jian, Lu Yili, Zhang Hang, Peng Huanhuan, Wang Xiaoji

2020, 41(6): 1-7.

Abstract(753) PDF(391)

Abstract:
Probabilistic safety assessment (PSA) is one of the two major safety analysis methods in the field of nuclear energy safety analysis. Starting from the concept of PSA, this paper firstly compares the difference between deterministic safety assessment and probabilistic safety assessment from the theo...

Numerical Simulation of Steady-State and Transient Characteristics of Once Through Steam Generators

Zhao Xiao, Bai Yufei, Zhang Zhen, Yang Xingtuan

2020, 41(6): 8-13.

Abstract(383)

Abstract:
In once through steam generators, the secondary fluid is heated by the primary fluid. Therefore, the axial heat flux distribution is related to the operational parameters of the fluids in both sides. The RELAP5/MOD4.0 code is adopted to perform the static and transient simulations of a once through ...

Development and Verification of Coolability Analysis Program for Debris Bed of Sodium-Cooled Fast Reactor in Long-Term Cooling Phase

He Lianren, Zhang Bin, Teng Chunming, Shan Jianqiang

2020, 41(6): 14-18.

Abstract(260)

Abstract:
In order to accurately analyze the cooling performance and temperature distribution of the sodium-cooled fast reactor debris bed in the boiling, drying and channel drying phases, and to improve the calculation efficiency, COMMEN-LT was developed based on the COMMEN program and the DEBRIS-HT program,...

Study on Local Three Dimension Hydrogen Risk Using GOTHIC

Huang Gaofeng, Gong Yu, Fu Tingzao, Wang Jiayun, Zhang Kun, Fang Likai

2020, 41(6): 19-23.

Abstract(246)

Abstract:
Integrated severe accident code is used to analyze the hydrogen risk in current safety assessment. After Fukushima accident, higher requirements are placed on hydrogen risk analysis. In order to supplement the lumped parameter analysis, three dimension hydrogen risk analysis method using GOTHIC is s...

Effect of PWR Cladding Materials in Different Nuclear Evaluated Data Libraries on Reactivity and Its Analysis

Xiao Xiang, Chen Yixue, Yang Tongrui, Wu Jun

2020, 41(6): 24-30.

Abstract(190)

Abstract:
Pressurized water reactor cladding material, Nuclear evaluated data library, NJOY2016

Heterogeneous Discontinuity Factors in Heterogeneous Variational Nodal Method to Eliminate Control Rod Cusping Effect

Liang Boning, Wu Hongchun, Li Yunzhao

2020, 41(6): 31-35.

Abstract(265)

Abstract:
To handle the control rod cusping effect in pressurized water reactor (PWR) fuel management calculation, the variational nodal method (VNM) in the fuel management calculation code system NECP-Bamboo has been extended to tread the heterogeneous cross section distribution by expanding the volumetric c...

Research on Multi-Physics Coupling Method of Pool Type Fast Reactor Based on CFD

Zhao Pengcheng Liu Zijing, Yu Tao, Liu Peiqi, Xie Jinsen, Chen Zhenping

2020, 41(6): 36-44.

Abstract(302) PDF(163)

Abstract:
Based on the critical/subcritical point kinetics model, the fuel pin heat transfer model, and the auxiliary thermal hydraulic models such as the heat exchanger model and the porous media model, a multi-physical coupling code CFD/PF was developed by means of the explicit iteration method, dynamic lin...

Poisoning in a Reactor Stimulated by Single Group Point Reactor Model and its Experimental Verification

He Ran, Zhang Kuan, You Haixiang, Zheng Xiaomin

2020, 41(6): 45-51.

Abstract(208)

Abstract:
To predict the poisoning quickly and precisely, a method to determine the parameters of the single group point reactor model with appropriate boundary conditions is developed, and the poisoning is predicted with the single group reactor model. And then, the evolution of Xenon poisoning and Samarium ...

Experimental Study on Critical Heat Flux in Deformed Flow Channel between Pressure Vessel and Insulation

Liu Yusheng, Xue Yanfang, Wang Kunpeng, Wen Shuang, Zhang Zhengxin, Li Congxin

2020, 41(6): 52-57.

Abstract(164)

Abstract:
Aiming at the deformation issue of flow gap between reactor pressure vessel (RPV) and insulation in external reactor vessel cooling (ERVC), the effects of insulation deformation on the critical heat flux (CHF) of the bottom head were investigated with FIMR test facility within the same flow rate ran...

Analysis on Effect of Control Rod Worth in Extrapolation in Criticality Experiment

Song Jingkai, Wang Wencong, Yuan Wei, Huang Liyuan, Wang Yongchao

2020, 41(6): 58-61.

Abstract(282)

Abstract:
The application effect of the control rod worth in the criticality extrapolation experiment is studied. The criticality extrapolation experiment is conducted on a zero power reactor, and the analysis is carried out by 2 extrapolation methods, i.e., with and without the consideration of the control r...

Monte Carlo Neutron Transport Simulation for Dispersion Fuel Based on Chord Length Sampling Method

Chen Zhenping, Guo Qian, Yu Tao, Zhang Zhenyu, Ma Huiqiang, Xie Jinsen

2020, 41(6): 62-68.

Abstract(200)

Abstract:
The dispersion fuel is with the advantages of high burnup, strong ability of containing fission products and good thermal conductivity. It is widely used as an advanced fuel element in new types of nuclear reactors. However, the dispersion fuel element in which the fuel particles statistically distr...

Theoretical Research on Effective Thermal Conductivity of TRISO Particle

Qian Libo, Yu Hongxing, Sun Yufa, Deng Jian, Chen Wei, Liu Yu, Du Sijia, Shen Danhong

2020, 41(6): 69-74.

Abstract(407)

Abstract:
TRISO (tri-structural isotropic) fuel particle consists of a fuel kernel in the center coated with four layers, with good fission product retention capability. The effective thermal conductivity of TRISO fuel particle is an important basis for calculating the effective thermal conductivity of disper...

Study on Heat Treatment Process for Improving the Toughness of 1Cr14Co14Mo5 Stainless Steel

Shu Ming, Wang Hao, Wu Songling, Liu Xiao, Wang Li, Kong Fanya, Xu Dianxin

2020, 41(6): 75-79.

Abstract(197)

Abstract:
In view of the brittle fracture failure of the end of ball screw caused by insufficient material toughness, this paper carries out the improvement research of martensitic precipitation hardening stainless steel 1Cr14Co14Mo5. Based on the local composition adjustment of 1Cr14Co14Mo5, the study adjust...

Research Progress of FeCrAl Cladding Materials Strengthened by Particles for ATF

Wan Haiyi, Wang Hui, Zha Wusheng, An Xuguang, Kong Qingquan, Chen Xiuli

2020, 41(6): 80-84.

Abstract(264)

Abstract:
Since the Fukushima Daiichi accident in Japan, the safety of zirconium alloy as a nuclear fuel cladding material has been questioned. Therefore, the development of accident tolerant fuel (ATF) has been proposed by many countries in the world. FeCrAl alloy has become one of the important materials in...

Effects of Thermal Creep on Blister Behavior in UMo/Zr Monolithic Fuel Plates

Yan Feng, Jian Xiaobin, Ding Shurong, Xin Yong, Tang Changbing, Li Yuanming

2020, 41(6): 85-91.

Abstract(374)

Abstract:
For a UMo/Zr monolithic fuel plate with a gas space, a method is developed to simulate the macroscale blister behavior considering the thermal creep effects of the cladding, in which the calculation of cladding deformation is coupled with the gas space pressure. Based on the developed simulation met...

Research on Prediction Model of Irradiation Embrittlement of RPV Materials Based on Artificial Neural Network

Kang Jing, Sun Kai, Mi Xiaoxi, Wu Lu, Mao Jianjun, Zhang Shuo, Lei Yang, Pan Rongjian, Tang Aitao

2020, 41(6): 92-95.

Abstract(356)

Abstract:
Based on the analysis of a certain amount of on-site test samples, this paper constructs a high-precision artificial neural network model for the ductile-brittle transition temperature prediction of RPV materials. Then we use the model to explore the influence of neutron fluence and neutron fluence ...

Optimization of Transmission Gear of Personnel Airlock for In-Service Nuclear Power Plant

He Yingyong, Li Qiang, Xie Honghu, Zhang Feng, Liu Xiaohua

2020, 41(6): 96-100.

Abstract(162)

Abstract:
ANSYS finite element analysis program was used to analyze the stresses of the transmission gear for an in-service nuclear power plant in China, and the causes for tooth rupture were found. The transmission torque was so large that the calculated stress exceeded the allowable stress limit of the mate...

Analysis of Flow-Induced Vibration of HP-Cooler Coil on Reactor Coolant Pump

Feng Xiaodong, Su Wentao, Ma Yu, Wang Lei, Li Xiaobin

2020, 41(6): 101-105.

Abstract(231)

Abstract:
To verify that the high-pressure cooler structure of the reactor coolant pump (reffered to as the main pump) can avoid the flow-induced vibration under normal operating conditions, this work analyzes the influence of the shell-side fluid on the vibration of the intermediate coil from three aspects, ...

Research of Ultimate Bearing Characteristics for Equipment Hatch of Nuclear Power Plant under External Pressure

Du Kun, Zuo Yongde, Yuan Liang

2020, 41(6): 106-110.

Abstract(259)

Abstract:
The ultimate bearing characteristics of the equipment hatch under the external pressure was studied by the finite element analysis method. The parametric calculation model was established in ANSYS, and the external pressure ultimate load of the equipment gate head was obtained by nonlinear buckling ...

Numerical Simulation Research on Coupling Vibration of Heat Transfer Tube Bundles under Lateral Action of Single-Phase Fluid

Bao Shiyi Zhu Hai, Tang Di, Yuan Wei, Huang Xipeng

2020, 41(6): 111-115.

Abstract(170)

Abstract:
The heat exchange tube bundle is an important part of the steam generator, and its reliability directly affects the safe operation of the nuclear power plant reactor. Based on the research in related fields, a computational fluid dynamics(CFD)/computational structural dynamics(CSD) coupling calculat...

Study on Flow-Induced Vibration Damping Simulation of Heat Exchanger Tube in Non-Uniform Two-Phase Flow

Shen Pingchuan, Liu Qing, Qi Huanhuan, Huang Xuan, Liu Jian, Chen Guo

2020, 41(6): 116-119.

Abstract(252)

Abstract:
In the flow-induced vibration analysis of the heat exchanger tube of the stream generator, the damping of each position on the tube is different, since the secondary side of the tube is the two-phase flow (stream-water) and the void-fraction is gradually increased from bottom to top. It is necessary...

Development of Symptom Based Emergency Operation Procedure for HPR1000

Ran Xu, Yu Na, Li Feng, Qian Libo, Chen Wei, Zhang Ming, Wu Qing, Liu Changwen, Leng Guijun

2020, 41(6): 121-125.

Abstract(444)

Abstract:
In order to compensate for the defects of event-oriented emergency procedure (EOP) and state-oriented emergency procedure (SOP), HPR1000 nuclear power technology takes the advantages of the two operation procedures. Considering probabilistic safety analysis (PSA), a new symptom based emergency opera...

Application of T-S Fuzzy Switching Controller in Core Power Control

Jiang Qingfeng, Zeng Wenjie

2020, 41(6): 126-130.

Abstract(211)

Abstract:
The traditional PID controller is used to control the core power, which has the problems of large overshoot and long regulating time in the control process. In order to solve this problem, based on the core transfer function model, the PD controller, the PID controller and the fuzzy controller are w...

A New Uncertainty Analysis Method for Tolerance Limit Evaluation

Guo Jiafeng, Lu Chuan, Mao Huihui, Sun Zhongning, Wang Jianjun, Wang Xiaolie

2020, 41(6): 131-137.

Abstract(310)

Abstract:
Uncertainty analysis methods based on WILKS formula is most generally applied for its advantages in reducing the calculation amount. However, the high resolution calculation is required in the nuclear reactor design and safty evaluation , which demand the uncertainty analysis methods with higher eff...

Dynamic Reliability Analysis of AP1000 Equipment Cooling Water System Based on BDMP

Zhou Shiliang, Chen Xiyu, E Wanjiang, Zhang Lei

2020, 41(6): 138-142.

Abstract(275)

Abstract:
CCS is a kind of repairable systems with double redundancy. The factors such as the alternate operation of redundant equipment and the repair of faulty equipment have a great influence on the reliability analysis results. Due to the lack of the description of time factors in traditional fault tree a...

Adaptive Predictive Control for Core Power Based on CPSO Rolling Optimization

Pan Yuekai, Qian Hong, Jiang cheng, Liu Xiaojing

2020, 41(6): 143-149.

Abstract(229)

Abstract:
In view of the nonlinear and reactive constraint of nuclear reactors in the process of variable power, this paper proposes an improved generalized predictive control (JGPC) for core power control. The JGPC calculates the predicted output value by predicting the model parameters and recursive relatio...

Development of Operation Procedure for Secondary Pipe Break Accident Integrating State Oriented and Event Oriented

Yu Na, Ran Xu, Xian Lin, Li Feng, Zhang Zhuohua, Wu Qing, Liu Changwen, Leng Guijun, Chen Wei, Fang Hongyu, Cheng Hongxia

2020, 41(6): 150-154.

Abstract(218)

Abstract:
HPR1000 adopts the symptom based emergency operating procedures (SEOP) to deal with accidents. In this paper, the related procedures for secondary pipe break accidents in SEOP are studied, including the development of procedures and supporting verification. In the development process of the procedur...

Flow Rate Research of TXRIS007 Supervision Requirement Criterion

Qiu Yanfei, Wu Shungui, Yang Zijun, Lu Yang

2020, 41(6): 155-161.

Abstract(115)

Abstract:
In order to solve the problem that the check valve function verification tests (TXRIS007) of multi-unit accumulators in a nuclear power plant do not meet the acceptance criteria in most cases, the one-dimensional fluid simulation software (Flowmaster) was used to establish the test model to carry ou...

Development of Intelligent On-Line Monitoring Device for Fuel Cladding Defect in Nuclear Power Plants

Xiao Ming, Cheng Xiaoqiang, Wang Yangyi, Song Yun

2020, 41(6): 162-166.

Abstract(166)

Abstract:
An intelligent on-line monitoring device for fuel cladding defect has been developed. HPGe with anti-Compton scattering detection system is used to measure the activity of characteristic radionuclides in primary cooling water, and the multi-nuclide group coupled analysis method is used to diagnose t...

Pre-Tightening Force Calibration Test Research for Stud without Measuring Rod of HTR-PM

Jin Dongjie, Wang Yin, Jin Gang, Geng Baojie, Guo Yunlong

2020, 41(6): 167-171.

Abstract(218)

Abstract:
In order to control the final pre-tightening force of the stud without measuring rod in the high temperature gas cooled reactor Pebble Module (HTR-PM) and ensure the sealing of the reactor primary circuit pressure boundary, it is necessary to calibrate the pre-tightening force of the stud. Taking th...

Rearserch of Localization Substitution of Reactor Pressure Vessel Seal Ring

Hu Wensheng, Hong Jun

2020, 41(6): 172-176.

Abstract(321)

Abstract:
C-ring is the core part for seal of the reactor pressure vessel top cover and cylinder, and the sealing performance is directly related to the safety and stably operation of the nuclear power plant. For a long time, the manufacturing technology of C-ring was monopolized by foreign company, with high...

Study on Fluid Flow and Heat Transfer in Seal Chamber of Hydrodynamic Mechanical Seals in Reactor Coolant Pump

Xiang Xianbao, Yang Quanchao, Wen Xue, Zheng Jiarong

2020, 41(6): 177-181.

Abstract(297)

Abstract:
To study the distribution of the flow and temperature field in a type of hydrodynamic mechanical seal of a nuclear reactor coolant pump, a three-dimensional model of the mechanical seal and the seal chamber is established based on the software Pro/E. The N-S equations and energy equation coupling wi...

Analysis and Improvement of the Increasing of Sodium Ion in Secondary Loop Caused by Condensate Polishing Plant in Nuclear Power Plants

Cheng Zhenhua, Wang Guoliang

2020, 41(6): 182-186.

Abstract(313)

Abstract:
The content of sodium ion in one steam generator blowdown system cannot satisfy the WANO chemical indices. This paper analyzes the possible causes of sodium ions in the secondary loop of the nuclear power plant and the difference of water quality indicators before and after the operation of the cond...

Research on Evaluation Method of Nuclear Power Pipeline Fracture Size under Random Missing Data

Zhao Xin, Cai Qi, Zhang Liming, Zhao Xinwen, Wang Xiaolong, Li Haicui

2020, 41(6): 187-193.

Abstract(197)

Abstract:
The monitoring parameters of the nuclear power system are randomly lost due to noise interference,which affects the judgement of the operators onthe severity of the accident. A diagnosis model of fracture size with tolerance parameter loss is proposed. The multiple time series which fracture size is...

Numerical Simulation of Gas Injected Bubble Dynamics from Single Submerged Orifice

Shen Lanting, Chai Xiang, Cheng Xu

2020, 41(6): 194-197.

Abstract(458)

Abstract:
In severe accidents of a nuclear power plant, the released radioactive aerosols can be removed by pool scrubbing effect. Two-phase numerical simulation of the pool scrubbing process is necessary. The boundary conditions at the bubble injection point need to be determined before using the two-phase C...

Research and Application of Spent Fuel Storage Racks Positioning Test Method Based on Machine Vision

Cheng Wei, Wang Lei, Hu Jianhua, Wang Zhiming, Zhang Peng, Ji Dapeng

2020, 41(6): 198-201.

Abstract(262)

Abstract:
The positioning test of spent fuel racks is a key test in the commissioning of nuclear fuel handling and storage system. It directly affects the safety and efficiency of the receiving and storage of nuclear fuel assemblies. In this paper, the traditional manual positioning test method is studied, an...

Effect of PWHT on Mechanical Properties and Microstructure of SA517 Gr.F Welded Joint

Li Juan, Wang Lulu, Yu Jie, Liu Hongpeng

2020, 41(6): 202-206.

Abstract(374)

Abstract:
In order to better understand the effect of PWHT (post weld heat treatment) on SA517 Gr.F quenched and tempered steel welded joint, the mechanical properties and microstructure distribution characteristics of SA517 Gr.F welded joint by SMAW (shielding metal arc welding) before and after PWHT were co...

Research on Scaling Analysis Method for Natural Circulation Test Facility of Lead-Based Fast Reactor

Zhao Pengcheng, Zhu Enping, Yu Hongxing, Zhai Pengdi, Deng Shengwen, Xia Bangyang, Chen Baowen

2020, 41(6): 207-213.

Abstract(410)

Abstract:
Lead-based fast reactors have good natural circulation capabilities, and its natural circulation characteristics is of great value to improve the inherent safety of the reactor, and the scaling analysis method is the theoretical basis for establishing a reasonable and feasible lead-based fast reacto...

Study on Measurement Method of Subcritical Reactivity

Tang Xiao, Xiao Peng, Liu Tongxian, Liao Hongkuan, Huang Can, Zhao Dehua, Liu Mingquan, Lu Di, Li Mancang

2020, 41(6): 214-217.

Abstract(187)

Abstract:

Dynamics Numerical Calculation for Control Rod Drop

Zhang Jibin, Gao Xilong, He Hangxing, Gong Ruzhi, Ma Chao, Yue Ning

2020, 41(6): 218-223.

Abstract(406)

Abstract:
The dropping time and profile of control rods are important parameters during the safety evaluation of nuclear power plants. CFD method and dynamic mesh were used to study the dropping profile of the control rod and flow evolution. The changes of displacement, velocity and acceleration versus time d...

2020 Vol. 41, No. 6