核动力工程

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

Abstract(984) HTML (226) PDF(240)

Abstract:
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 plan...

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

Abstract(210) HTML (64) PDF(64)

Abstract:
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 s...

Research on Burnup Calculation Method Based on Splicing Fission Matrix

Gao Ruishuang, Liu Xiaojing, He Donghao, Pan Qingquan

2024, 45(2): 19-23. doi: 10.13832/j.jnpe.2024.02.0019

Abstract(180) HTML (55) PDF(37)

Abstract:
In order to realize efficient transport-burnup coupling calculation, a new calculation method based on splicing fission matrix theory is proposed in this paper. The combined fission matrix method enables to obtain the system fission matrix from a pre-calculated database, and then the database matric...

Verification and Validation of NECP-Bamboo Based on Measurement Data from Nuclear Power Plants

Liang Yilin, Li Yunzhao, Zhou Yuancheng, Li Yisong, Zhang Hengrui, Zhou Shilong, Wang Weiguo, Ou Yuxiang, Wang Songzhe, Qin Junwei, Shao Ruizhi

2024, 45(2): 24-34. doi: 10.13832/j.jnpe.2024.02.0024

Abstract(601) HTML (101) PDF(60)

Abstract:
The verification and validation (V&V) are important aspects in the life cycle of engineering software, and reflect the process of software development to practical application. They are also significant symbols of software from "bookshelf" to "goods shelf". In this paper, V&V for the softwar...

Study on New Burnable Poison Material Combinations Based on Long-life Pressurized Water Reactor

Tong Ji, Qu Lingyi, Xie Jinsen, Xu Shikun, Yu Tao

2024, 45(2): 35-41. doi: 10.13832/j.jnpe.2024.02.0035

Abstract(186) HTML (49) PDF(57)

Abstract:
Long-life pressurized water reactors using a single burnable poison assembly design is not the optimal choice. In order to meet the comprehensive requirements of burnable poisons for reactivity control in long-life pressurized water reactors, this paper conducts a combined study on new burnable pois...

Analytical Solution Study on Transient Neutron Transport in Homogeneous Bare Reactor Based on Neutron Telegraph Equation

Wang Weiguo, Li Yunzhao, Qin Junwei, Cao Liangzhi

2024, 45(2): 42-46. doi: 10.13832/j.jnpe.2024.02.0042

Abstract(533) HTML (53) PDF(27)

Abstract:
The Fick's law, as the foundation of the diffusion theory, further neglects the time derivative term of the neutron current density in the P₁ approximation equation (telegraph equation) of the neutron transport equation. Therefore, it is difficult to accurately describe the actual neutron kinetic be...

Research on Evaluation Method of Actinide Nuclide Activities in Primary Coolant System of Pressurized Water Reactor Nuclear Power Plant

Xiong Jun, Lyu Weifeng, Guo Runchun, Gao Yaoyi, Jiang Zhenyu

2024, 45(2): 47-52. doi: 10.13832/j.jnpe.2024.02.0047

Abstract(135) HTML (35) PDF(38)

Abstract:
In order to evaluate the radiation risk of workers and the damage degree of fuel cladding in PWR nuclear power plant, it is necessary to improve the evaluation approach of actinide nuclide activities in the primary coolant systems based on characteristic physical quantities. Based on the generation ...

Study on Neutron Diffusion Calculation Method Based on hp-VPINN

Zeng Fulin, Zhang Xiaolong, Zhao Pengcheng

2024, 45(2): 53-62. doi: 10.13832/j.jnpe.2024.02.0053

Abstract(249) HTML (69) PDF(45)

Abstract:
Advanced reactor simulations require the inversion of key parameters of the whole reactor based on less actual detection data. To meet this need, this paper constructs a computational model based on variational residual physics-informed neural network (hp-VPINN) with high-order polynomial domain dec...

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

Abstract(254) HTML (80) PDF(63)

Abstract:
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 o...

Experimental Study on Heat Transfer Characteristics of Pure Steam with Incomplete Condensation in Vertical Tube

Liu Jiabao, Cao Xiaxin, Yang Peixun

2024, 45(2): 72-81. doi: 10.13832/j.jnpe.2024.02.0072

Abstract(270) HTML (106) PDF(41)

Abstract:
In order to study the heat transfer characteristics of pure steam with incomplete condensation in a vertical tube, an experiment was carried out using a heat exchange tube with an inner diameter of 25 mm, with an inlet pressure of 0.1~0.3 MPa and a steam mass flux of 12~70 kg/(m²·s). The effects of ...

Numerical Simulation of Pressure and Temperature Fields in Rectangular Narrow Channel under Blistering Condition

Wang Xuejian, Zhao Yanan, Yu Tao

2024, 45(2): 82-87. doi: 10.13832/j.jnpe.2024.02.0082

Abstract(172) HTML (61) PDF(36)

Abstract:
When the rectangular narrow channel is deformed due to radiation blistering, the flow heat transfer characteristics in the channel will be changed. In this study, the fluid pressure field and temperature field in the rectangular narrow channel under blistering condition is analyzed by numerical calc...

Simulation Study on Transport Characteristics of Radioactive Aerosol in Containment

Tian Jiaming, Wang Yueshe, Li Biao, Zou Lin

2024, 45(2): 88-95. doi: 10.13832/j.jnpe.2024.02.0088

Abstract(239) HTML (193) PDF(64)

Abstract:
In order to find out the transport characteristics of radioactive aerosol in a serious accident under the actual containment size, the spatial distribution of radioactive aerosol in a serious accident was simulated by using the coupling of computational fluid dynamics and particle swarm equilibrium ...

Experimental Study on Core Flow Instability Caused by Vortex Flow in Reactor Lower Plenum

Meng Yang, Liao Hengji, Jiang Lin, Fang Ying, Zhang Jiaqi, Li Yong, Yang Zumao

2024, 45(2): 96-102. doi: 10.13832/j.jnpe.2024.02.0096

Abstract(259) HTML (87) PDF(42)

Abstract:
Strong turbulent mixing of coolant in each loop of the PWR occurs in the lower plenum, and a large number of vertex are generated in the lower plenum, which can lead to random fluctuation of the fuel assembly inlet flow and trigger transient flow instability in the core, thus affecting the reactor t...

Analysis of Steady and Transient Characteristics of Once-through Steam Generator for Lead Bismuth Fast Reactor

Huang Zhe, Liang Tiebo, Yang Wen, Lu Chuan, Li Yang, He Zhonghai, Shen Xin

2024, 45(2): 103-109. doi: 10.13832/j.jnpe.2024.02.0103

Abstract(202) HTML (54) PDF(58)

Abstract:
The safe and stable operation of lead bismuth fast reactor is closely related to the heat dissipation performance between the primary and secondary sides of the heat exchanger. In this study, the steady-state and transient coupled distribution parameter model of once-through steam generator (OTSG) b...

Study on Capillary Characteristics of High Temperature Liquid Sodium in Stainless Steel Wire Mesh Wick

Zhu Yiru, Ma Yugao, Zhang Luteng, Xi Zhiguo, Tang Simiao, Ma Zaiyong, Pan Liangming, Zhang Zhuohua, Ding Shuhua

2024, 45(2): 110-115. doi: 10.13832/j.jnpe.2024.02.0110

Abstract(836) HTML (100) PDF(53)

Abstract:
The capillary characteristics of the alkali metal high temperature heat pipe wick are of great significance to the normal operation of the heat pipe. In this paper, the capillary rise method and the weighing method were used to carry out the experimental study on the capillary characteristics of the...

Study on the Influence of Spacer Grid on Temperature Field Distribution at Rod Bundle Channel Outlet

Qiu Zicheng, Xie Shijie, Lang Xuemei, Li Pengzhou, Zhuo Wenbin

2024, 45(2): 116-122. doi: 10.13832/j.jnpe.2024.02.0116

Abstract(188) HTML (45) PDF(41)

Abstract:
To analyze the mixing ability of the spacer grid and further optimize the thermal design method of the spacer grid, this article investigates the influence of spacer grid with different structures on the temperature field distribution at the outlet of rod bundle channel through experiments. In the e...

Effect of Baffle on Flow Induced Vibration of Heat Exchange Tubes

Wang Shuaiquan, Zhang Kai, Xiong Zhenqin, Shi Linpeng

2024, 45(2): 123-129. doi: 10.13832/j.jnpe.2024.02.0123

Abstract(207) HTML (65) PDF(40)

Abstract:
The wear of tubes caused by flow induced vibration is one of the most significant factors that cause performance attenuation and malfunction of the heat exchanger. In this paper, aiming at the vibration caused by the non-uniform cross flow formed by the heat exchanger baffle, a visualized three-span...

Numerical Simulation Method of Boiling Heat Transfer and Its Application Characteristics under Multi-tube Coupled Heat Transfer

Zhang Zhenguo, Tan Sichao, Li Xiaochang, Liu Sichao, Feng Yi, Huang Yujian, Tian Ruifeng

2024, 45(2): 130-138. doi: 10.13832/j.jnpe.2024.02.0130

Abstract(498) HTML (59) PDF(51)

Abstract:
Aiming at the problems of high difficulty in simulation, low computational efficiency, and large uncertainties when applying the method of computational fluid dynamics (CFD) to the analysis of coupled heat transfer on primary and secondary sides and full-regime flow boiling heat transfer in once-thr...

Optimization and Thermal Safety Analysis of CFETR Advanced Small Sample Irradiation Capsule

Liu Chang, Liu Wenbin, Dai Yong, Sun Shouhua, Zhang Ping, Kang Changhu, Song Jiyang, Wang Kaimin, Peng Lei, Zheng Pengfei

2024, 45(2): 139-146. doi: 10.13832/j.jnpe.2024.02.0139

Abstract(166) HTML (84) PDF(14)

Abstract:
The capsule structure of the advanced material irradiation test sample in China Fusion Engineering Experimental Reactor (CFETR) is relatively complicated, and the capsule is filled with helium. The size and location of the capsule rib and the material of the filling inside the capsule have great inf...

A Porous Media Model for Thermal-hydraulic Analysis of Wire-wrapped Fuel Assembly in Sodium Cooled Fast Reactor

Wang Xinan, Zhang Dalin, Wang Ting, Qiu Suizheng, Su Guanghui

2024, 45(2): 147-153. doi: 10.13832/j.jnpe.2024.02.0147

Abstract(258) HTML (120) PDF(55)

Abstract:
In order to accurately predict the distribution of three-dimensional thermal-hydraulic parameters in the core of sodium-cooled fast reactor and reduce the demand for computing resources, a three-dimensional porous medium model for the wire-wrapped fuel assembly was established based on the concept o...

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

Abstract(329) HTML (100) PDF(67)

Abstract:
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 characterist...

Improvement of Radiolytic Gas Model for Criticality Accident Analysis of Nuclear Fuel Solution System

Sheng Huimin, He Junyi, Gou Junli, Shan Jianqiang, Liu Guoming

2024, 45(2): 160-165. doi: 10.13832/j.jnpe.2024.02.0160

Abstract(531) HTML (57) PDF(62)

Abstract:
The accurate simulation of transient criticality accident is a key factor in critical safety assessment of nuclear fuel solution system. However, the existing radiolysis gas models contain many empirical parameters, which result in significant deviations in the prediction of the power characteristic...

Analysis of Spent Fuel Cask Dropping Accident and Research on Relevant Improvement Measures

Tan Jingyao, Chen Yao, Ji Wenying, He Yingyong, Zhang Feng, Xia Yan, Ning Xinxian

2024, 45(2): 166-170. doi: 10.13832/j.jnpe.2024.02.0166

Abstract(264) HTML (84) PDF(47)

Abstract:
"Cask immersion" is generally used in domestic nuclear power plants to deliver spent fuel, thus the risk of spent fuel cask dropping accident cannot be completely eliminated. The radioactive consequences of spent fuel cask dropping accident are analyzed and calculated, and how to ensure the integrit...

Research on Inter-unit Common Cause Failure of Multi-unit PSA

Yang Chunju, Wang Ming, Lin Modi, Zhang Bing, Wang Jinkai

2024, 45(2): 171-177. doi: 10.13832/j.jnpe.2024.02.0171

Abstract(615) HTML (82) PDF(66)

Abstract:
How to reasonably estimate the contribution of inter-unit common cause failure (CCF) to the safety risk of power plant is an important technical problem that needs to be solved in multi-unit probabilistic safety analysis (PSA) modeling. In this study, analysis methods of inter-unit CCF group selecti...

Simulation Study on Steam Dump Control System of Small Pressurized Water Reactor Nuclear Power Plant

Wang Linna, Chen Chuqi, Zeng Wenjie, Chen Lekang, Li Ruokun

2024, 45(2): 178-182. doi: 10.13832/j.jnpe.2024.02.0178

Abstract(249) HTML (92) PDF(39)

Abstract:
Small pressurized water reactor nuclear power plant usually uses compact once-through steam generator, which has dramatically increased steam pressure under load rejection condition. In order to prevent damage to the system equipment caused by excessive steam pressure, based on the establishment of ...

Design and Experiment of Feed-water Control System for CEFR Steam Generator

Liu Yong, Duan Tianying, Feng Weiwei, Zhang Weiying

2024, 45(2): 183-186. doi: 10.13832/j.jnpe.2024.02.0183

Abstract(144) HTML (60) PDF(23)

Abstract:
The feed-water control system of steam generator is an important control system of China Experimental Fast Reactor (CEFR), which needs to be automatically controlled. According to the control requirements of the system, a set of feed-water control scheme is designed, and the control scheme is verifi...

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

Abstract(259) HTML (132) PDF(39)

Abstract:
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 ...

Flow Field and Blowdown Test of Secondary Side Lower Part of Steam Generator

Huang Jun, Ying Bingbin, Zheng Mingguang

2024, 45(2): 193-198. doi: 10.13832/j.jnpe.2024.02.0193

Abstract(136) HTML (48) PDF(41)

Abstract:
The lower blowdown structure located in the tubesheet of steam generator (SG) in nuclear power plant can suck out the sludge on the tubesheet secondary side surface and discharge it. In order to reasonably design this structure and improve the blowdown efficiency, based on the SG prototype structure...

Research of Ultimate Load Characteristics of Hoisting Structure Carrier for Equipment Hatch

Hou Jiru, Liu Shifeng

2024, 45(2): 199-202. doi: 10.13832/j.jnpe.2024.02.0199

Abstract(690) HTML (67) PDF(35)

Abstract:
Instability of the equipment hatch’s hoisting structure carrier occurs under the combined action of permanent and variable loads, and the accuracy of the ultimate load calculation results has an important influence on the design performance of equipment hatch. In engineering calculations, the tradit...

Primary Zinc Injection Technology Application and Zinc Injection and Dosing Equipment Design

Zou Wei, Lin Genxian, Sun Yun

2024, 45(2): 203-206. doi: 10.13832/j.jnpe.2024.02.0203

Abstract(363) HTML (131) PDF(42)

Abstract:
The application of primary zinc injection technology makes the oxide film generated on the surface of the primary equipment more stable and plays a role in reducing the collective dose of the unit. By analyzing the effect of zinc injection on the unit, and combining the target concentration of zinc ...

Research on Historical Anomaly Data Detection Technology for Nuclear Power Plant Based on Deep Auto-Encoder

Yang Jihong, Chen Ling, Wang Xiaolong, Zhang Yongfa, Gao Ming

2024, 45(2): 207-213. doi: 10.13832/j.jnpe.2024.02.0207

Abstract(149) HTML (64) PDF(29)

Abstract:
In order to solve the problem of new anomaly identification difficulties in the detection of nuclear power historical anomaly data, according to the idea of reconstruction error, an anomaly detection model based on deep auto-encoder is proposed. The model takes the normal historical data under stead...

Design and Verification of Fuel Assembly Repair System

Li Chengye, Wang Wanjin, Wang Yajun, Liu Hao, Zheng Haichuan

2024, 45(2): 214-217. doi: 10.13832/j.jnpe.2024.02.0214

Abstract(194) HTML (71) PDF(33)

Abstract:
In order to make full use of the post-irradiation fuel assembly, a fuel assembly repair process was proposed, and a repair system for post-irradiation fuel assembly was designed based on the dense and lightweight characteristics of the fuel assembly. Main parts were designed in detail and checked by...

Fault Risk Analysis Method of Instrument and Control System in Nuclear Power Plant

Luo Hui, Li Jianwei, Wang Xiangyu

2024, 45(2): 218-224. doi: 10.13832/j.jnpe.2024.02.0218

Abstract(279) HTML (92) PDF(49)

Abstract:
In order to meet the requirements of rapid accurate fault location and comprehensive reliable risk analysis of instrument and control system in nuclear power plant, a fault risk analysis method of instrument and control system based on automatic analysis model is proposed. Firstly, the traditional s...

Numerical Simulation of Displacement Process of Liquid Scintillator and Ultrapure Water in a Large Diameter Spherical Tank

Zhao Hongwei, Dong Rui, Li Junjie, Ouyang Zhengrong, Heng Yuekun

2024, 45(2): 225-230. doi: 10.13832/j.jnpe.2024.02.0225

Abstract(160) HTML (76) PDF(14)

Abstract:
On the one hand, the Neutrino experimental facility needs to shield a lot of cosmic rays and natural radioactivity from rocks, air and dust. On the other hand, it is necessary to reduce the radioactive background of the experimental facility as much as possible. In this paper, the displacement proce...

Research on Airtightness Test of Habitable Zone in Main Control Room of HPR1000

Xu Funan, Wang Guangjun, Wu Ying

2024, 45(2): 231-234. doi: 10.13832/j.jnpe.2024.02.0231

Abstract(133) HTML (53) PDF(25)

Abstract:
Based on the constant flow method, research was conducted on the airtightness test of the habitable area in the main control room from multiple dimensions such as test method selection, tracer gas injection, sampling method, equilibrium concentration, and calculation of test results. Improvements an...

Application of High Precision Laser Measurement Technology Based on Spherical Coordinate Method in Nuclear Power Plant Reactor Body Maintenance

Wang Xuefang, Lu Shaowei

2024, 45(2): 235-240. doi: 10.13832/j.jnpe.2024.02.0235

Abstract(273) HTML (95) PDF(28)

Abstract:
The maintenance of reactor body is an important work that must be done in refueling overhaul of nuclear power units. Due to the complex structure, large size, tight maintenance period and high radiation dose rate of the equipment in the reactor body, it is very difficult to measure and inspect defec...

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

Abstract(554) HTML (73) PDF(69)

Abstract:
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) reacto...

Development and Validation of DNBR On-line Monitoring System for HPR1000 Reactor

Chen Xi, Wu Qing, Deng Jian, Liu Yu, Ren Chunming, Wang Xiaoyu, Peng Huanhuan

2024, 45(2): 248-253. doi: 10.13832/j.jnpe.2024.02.0248

Abstract(338) HTML (121) PDF(61)

Abstract:
In the traditional protection system of generation II reactor, over-temperature protection signal is usually used. This protection method is conservative and indirect, however, the Departure from Nucleate Boiling Ratio (DNBR) on-line monitoring system can directly monitor the changes of safety param...

2024 Vol. 45, No. 2