核动力工程

Research on Fast Prediction Method of Neutron Flux Based on Hybrid Driven Reduced Order Model

Zhao Ziyan, Xiang Zhaocai, Zhao Pengcheng

2024, 45(4): 1-8. doi: 10.13832/j.jnpe.2024.04.0001

Abstract(307) HTML (91) PDF(84)

Abstract:
The accurate prediction of neutron flux and reactor power is very important for the safe operation of the reactor immediately after the disturbance of reactor parameters. The traditional method combining POD and Galerkin projection has the problem of low accuracy due to cumulative error. In this stu...

Research on Industrial Validation of VVER-1000 Based on PWR Core Physics Analysis Code Bamboo-C

Yang Haozhe, He Xudong, Wang Kunpeng, Wan Chenghui, Wu Hongchun

2024, 45(4): 9-16. doi: 10.13832/j.jnpe.2024.04.0009

Abstract(198) HTML (50) PDF(32)

Abstract:
This study aims to achieve precise physical analysis of VVER-1000 nuclear reactors. Based on Bamboo-C, an advanced PWR core physical analysis software developed by Xi'an Jiaotong University, a thorough methodological study is carried out. The research encompasses: in the aspect of assembly calculati...

Verification of PCM Nuclear Design Code for Whole Core Calculations

Lu Gaoqi, Ding Ming, Lan Bing, Pan Xinyi, Li Chun, Wang Chao, Ma Yunfan

2024, 45(4): 17-23. doi: 10.13832/j.jnpe.2024.04.0017

Abstract(174) HTML (44) PDF(35)

Abstract:
PCM software package, independently developed by China Nuclear Power Technology Research Institute Co. Ltd., is a nuclear design code consisting of assembly section calculation code PINE and 3D core design code COCO. To validate the whole core calculation capability and accuracy of the PCM nuclear d...

Numerical Study on DNB-Type Critical Heat Flux in Circular Tube under Rolling Condition

Fang Zheng, Du Song, Bu Shanshan, Li Zhenzhong, Chen Deqi

2024, 45(4): 24-31. doi: 10.13832/j.jnpe.2024.04.0024

Abstract(232) HTML (77) PDF(45)

Abstract:
A three-dimensional numerical calculation was carried out on the departure from nucleate boiling (DNB) type critical heat flux (CHF) in a vertical circular tube under different rolling conditions. The Euler two-phase flow model and the non-equilibrium wall boiling model were used. By comparing the s...

Research on Characteristics of Secondary Side Passive Residual Heat Removal System of Lead-bismuth Reactor under SBO

Qian Yalan, Lin Qian, Yang Zijiang, Chen Kang, Zhan Wenhui, Tang Chuntao, Yang Bo

2024, 45(4): 32-37. doi: 10.13832/j.jnpe.2024.04.0032

Abstract(163) HTML (38) PDF(35)

Abstract:
The secondary side passive residual heat removal (PRHRS) of Russian SVBR-100 lead-bismuth reactor was chosen as the research object, and the RELAP5/MOD4.0 code was used to model and evaluate the heat removal capacity and parameter sensitivity of PRHRS under station blackout (SBO) accident. The resul...

Study on Uncertainty of Two-Phase Flow Parameter Detection Based on Monte Carlo Method

Liu Li, Zhu Longxiang, Zhang Luteng, Ma Zaiyong, Sun Wan, Pan Liangming, Deng Jian

2024, 45(4): 38-44. doi: 10.13832/j.jnpe.2024.04.0038

Abstract(119) HTML (32) PDF(25)

Abstract:
Bubble velocity and bubble number are key parameters for calculating the phase characteristics such as interfacial area concentration, so it is necessary to study the uncertainty of bubble velocity and number measured by the conductivity probe. The Monte Carlo method is adopted to generate a large n...

Study on CHF Relational Expression Development Based on High-precision Subchannel Code

Wu Change, Zhang Yuxiang, Chen Changyi, Jiang Li, Shan Jianqiang, Fu Xiangang

2024, 45(4): 45-52. doi: 10.13832/j.jnpe.2024.04.0045

Abstract(89) HTML (40) PDF(24)

Abstract:
Three groups of CHF test data of non-uniform heating typical grid and guide tube grid were adopted, and the local parameters were obtained by using high precision subchannel code ATHAS. The development of CHF relation suitable for the analysis of fuel assembly deviation from nucleate boiling ratio (...

Numerical Simulation on Flow Heat Transfer Characteristics of Helium-Xenon Mixture in Tight Lattice Rod Bundle Channel

Zhang Jiaxin, Bao Hui, Cong Tenglong, Gu Hanyang

2024, 45(4): 53-60. doi: 10.13832/j.jnpe.2024.04.0053

Abstract(151) HTML (33) PDF(36)

Abstract:
In response to the design and analysis requirements for the core of the helium-xenon cooled high temperature gas cooled reactor, this study has established a comprehensive three-dimensional heat transfer model for helium-xenon mixture. This model encompasses property model, turbulent model, and turb...

Experimental Study on Boiling Two-Phase Flow Instability in a Single Helically Coiled Tube

Zheng Pengde, Tang Qifen, Li Zhenzhong, Wang Ningyuan, Chen Deqi

2024, 45(4): 61-68. doi: 10.13832/j.jnpe.2024.04.0061

Abstract(130) HTML (42) PDF(25)

Abstract:
The boiling phase change happened in the heating channel will induce flow instability. Thus, studying the two-phase flow instability in helically coiled tubes is of great significance in the design and operation of helical-coil one-through steam generators. In this paper, the boiling two-phase flow ...

Research on Prediction and Sensitivity Analysis of Minimum Film Boiling Temperature of Quenching Boiling Based on Machine Learning

Zhang Junquan, Deng Jian, Luo Yan, Lu Tao

2024, 45(4): 69-76. doi: 10.13832/j.jnpe.2024.04.0069

Abstract(131) HTML (40) PDF(20)

Abstract:
Quenching boiling is widely used in the cooling process of fuel rods after the loss of coolant accident in nuclear reactor. The determination of the minimum film boiling temperature (T_min) is very important for the safe operation of nuclear reactors. Based on the experimental data in the literature,...

Study on Jet Mixing Characteristics of Lead-Bismuth Eutectic Cooled Reactor Assembly Head Based on CFD Method

Zhang Ji, Wang Yingjie, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(4): 77-86. doi: 10.13832/j.jnpe.2024.04.0077

Abstract(692) HTML (65) PDF(40)

Abstract:
In the upper chamber of the Lead-Bismuth Eutectic (LBE) cooled reactor, fluid temperature fluctuations during the LBE mixing process from different power assemblies may lead to thermal fatigue of the solid structure, threatening the safety of LBE reactor operation. Based on the Computational Fluid D...

Numerical Study on Characteristics of Subcooled Flow Boiling with the Coupling Effect of 3×3 Petal-Shaped Fuel Rods and Coolant

Du Lipeng, Song Shangdian, Cai Weihua, Jiang Zeping, Cheng Qi, Zhang Wenchao, Jin Guangyuan

2024, 45(4): 87-95. doi: 10.13832/j.jnpe.2024.04.0087

Abstract(104) HTML (43) PDF(27)

Abstract:
In order to promote the engineering application of petal-shaped fuel rods in water-cooled reactors, it is necessary to understand the subcooled flow boiling characteristics of coolant in the sub-channels of petal-shaped fuel rod bundles. The Euler model and wall boiling model were applied to numeric...

Semi-Analytical Solution of Temperature Rise Caused by Irradiation Effect for Perforated Plates with Square Penetration Patterns in Reactor Vessel Internals

Lin Bingchi, Xu Xiao, Kong Xiaofei, Liu Pan, Jin Ting, Nie Zhaoyu, Lu Zhicheng, Yao Bowei

2024, 45(4): 96-102. doi: 10.13832/j.jnpe.2024.04.0096

Abstract(105) HTML (35) PDF(19)

Abstract:
In Reactor Vessel Internals (RVI), the temperature rise caused by the irradiation effect is significant for the thick perforated plates. Due to the complexity of the structure, its temperature rise is generally calculated by finite element method. In this paper, the semi-analytical method is used to...

Analysis on Flow Distribution Characteristics of Steam Generator under Natural Circulation Condition

Luan Xingjian, Wang Wen, Song Jiahao, Han Fei, Jiang Erhui, Cheng Kun, Yang Fan

2024, 45(4): 103-110. doi: 10.13832/j.jnpe.2024.04.0103

Abstract(570) HTML (36) PDF(28)

Abstract:
Natural circulation is a particular operating condition of nuclear power system, and reversed flow occurs in the inverted U-shaped tube bundle of the steam generator, which affects the heat transfer between the primary and secondary sides and the operation stability. In this research, In this study,...

Verification and Uncertainty Evaluation of LOCUST Reflood Model

Xu Rongshuan, Xia Hang, Xu Caihong, He Dongyu, Wang Ting, Li Jinggang

2024, 45(4): 111-117. doi: 10.13832/j.jnpe.2024.04.0111

Abstract(641) HTML (38) PDF(35)

Abstract:
The reflood stage is an important stage after a large break loss of coolant accident (LBLOCA) in pressurized water reactor. In order to evaluate the simulation ability of the system code LOCUST, the verification and uncertainty research of LOCUST reflood model are carried out. Based on the experimen...

Experimental Study of the Influences of CRUD layer on Bubble Departure Diameter and Bubble Departure Frequency on Fuel Cladding Surface

Cai Jiejin, Hu Zhiping, Deng Rining

2024, 45(4): 118-126. doi: 10.13832/j.jnpe.2024.04.0118

Abstract(129) HTML (48) PDF(18)

Abstract:
Chalk River Unidentified Deposits (CRUD) is naturally formed on the fuel cladding during the routine operation of pressurized water reactor (PWR), and its influence mechanism on the boiling heat transfer behavior of cladding is still unclear. In order to investigate the influence of CRUD on the bubb...

Study on Adaptability of Heat Transfer Model and Oxidation Relationships Based on COSINE Sub-channel Code

Cheng Yixuan, Meng Zhaocan, Zhang Hao, Zhang Yilin, Zhao Meng, Yang Yanhua

2024, 45(4): 127-133. doi: 10.13832/j.jnpe.2024.04.0127

Abstract(407) HTML (33) PDF(18)

Abstract:
In view of the urgent need of heat transfer model and oxidation relationships in pressurized water reactor nuclear subchannel software to improve the core safety and the accuracy of simulation and prediction of domestic software, we used numerical simulation technology to analyze the heat transfer m...

Preliminary Analysis of the Core Factors for Improving Efficiency of Thermionic Energy Conversion

Ni Wentao, Luo Qi, Zhong Wuye, Lyu Zheng

2024, 45(4): 134-141. doi: 10.13832/j.jnpe.2024.04.0134

Abstract(162) HTML (63) PDF(21) [Cited by] (1)

Abstract:
Efficient thermionic energy conversion technology is the key technology for improving the thermoelectric conversion efficiency of thermionic fuel elements and promoting the space thermionic reactor power supply towards higher power and longer lifetime. To explore the key factors that improve the eff...

A Reduced-Order Model of Mode Characteristics and Flow-Induced Vibration Response of Fuel Rod Based on POD Method

Min Guangyun, Jiang Naibin

2024, 45(4): 142-149. doi: 10.13832/j.jnpe.2024.04.0142

Abstract(138) HTML (43) PDF(43)

Abstract:
A Reduced-Order Model (ROM) for swiftly predicting both the modes and flow-induced vibration response of fuel rods is introduced in this study. Firstly, modes for fuel rods with varying stiffness are obtained using ANSYS-Batch, and these modes are then compiled into a Snapshots matrix. Subsequently,...

Application of Green Function Method Considering Thermal Stratification Effect in Rapid Calculation of Fatigue Monitoring System

Chen Rong, Zhang Guihe, Liang Enming

2024, 45(4): 150-154. doi: 10.13832/j.jnpe.2024.04.0150

Abstract(80) HTML (30) PDF(19)

Abstract:
This paper analyzes how to realize fast calculation when thermal stratification effect is considered in fatigue monitoring system. A new Green's function considering thermal stratification effect is proposed. At the same time, the thermal stratification stress of elbow model under assumed transient ...

Study on Load-following Operation Mode of Small Sodium-Cooled Fast Reactor Nuclear Power System

Yin Kai, Gong Lin, Duan Tianying, Hou Bin, Dai Raoqi, Liu Yong, Hu Jiayong

2024, 45(4): 155-161. doi: 10.13832/j.jnpe.2024.04.0155

Abstract(118) HTML (53) PDF(24)

Abstract:
Based on the platform of RELAP/GSE combined with MATLAB/Simulink, a simulation model is established for the scheme of coupling Stirling thermoelectric conversion module in small sodium-cooled fast reactors, so as to study the load following operation capability and operation mode of the nuclear powe...

Research on General Design Method of Emergency Communication System in Nuclear Power Plant

Yu Yun, Guo Shujie, Lu Wenkui, Wang Gaopeng, Liu Jing

2024, 45(4): 162-165. doi: 10.13832/j.jnpe.2024.04.0162

Abstract(182) HTML (59) PDF(22)

Abstract:
In order to further improve the emergency communication capability of the communication system of nuclear power plant and support the emergency response under nuclear accident conditions, this study analyzes the current relevant standards and requirements of the communication system design of nuclea...

Research on Dynamic Modeling and Control Method of Heat Pipe Reactor

Yin Shaoxuan, Yu Ren, Sheng Dongjie, Mao Wei

2024, 45(4): 166-172. doi: 10.13832/j.jnpe.2024.04.0166

Abstract(102) HTML (52) PDF(29)

Abstract:
In order to study the nuclear power control method of heat pipe reactor (HPR), the lightweight dynamic model of MegaPower HPR is constructed, and then the designed controller is verified by simulation. Based on lumped parameter method, a heat transfer model from core to heat pipe and then to heat ex...

Research on Optimization of Core Power Regulation System of Swimming Pool Reactor Based on PSO-BP Neural Network

Peng Zhiwen, Chen Xiaoliang, Zhu Jiachen, Wang Feng

2024, 45(4): 173-180. doi: 10.13832/j.jnpe.2024.04.0173

Abstract(88) HTML (30) PDF(20)

Abstract:
Based on the MATLAB/Simulink, the simulation model of the power regulation system and the primary heat transfer system of the 49-2 swimming pool reactor was constructed, and the external reactive disturbance simulation test was carried out to verify the accuracy of the model. The proportion integrat...

Study on Orthogonal Experiments of Jet Breakup and its Modeling Based on MPS Method

Peng Cheng, Meng Xianpin, Deng Jian

2024, 45(4): 181-189. doi: 10.13832/j.jnpe.2024.04.0181

Abstract(150) HTML (53) PDF(20)

Abstract:
In order to study the primary influencing factors of jet breakup length during the core melt jet breakup process and their ranking, 25 sets of experiments with 3 factors and 5 levels were designed based on the orthogonal experimental method, and the jet breakup lengths were obtained for each conditi...

Research on Efficient Verification and State Recognition Method for the Action Reliability of Manual Globe Valve

Zhou Suting, Zhang Lin, Nie Changhua, Fan Wenyutao, Huang Yanping, Liu Jie, Yuan Kai

2024, 45(4): 190-195. doi: 10.13832/j.jnpe.2024.04.0190

Abstract(619) HTML (65) PDF(56) [Cited by] (1)

Abstract:
As a typical valve in primary system, manual globe valve is of great importance to maintain system operation and protect system safety. In order to verify the action reliability of the nuclear-grade manual globe valve and determine its operation state accurately and quantitatively, this paper studie...

Design and Multi-Objective Optimization Study of Liquid Lead-Supercritical Carbon Dioxide Heat Exchanger

Li Liangxing, Shi Shang, Zhao Haoxiang, Zhao Jiayuan

2024, 45(4): 196-204. doi: 10.13832/j.jnpe.2024.04.0196

Abstract(330) HTML (54) PDF(36)

Abstract:
In order to improve the comprehensive heat transfer performance of the primary heat exchanger in lead-cooled fast reactors, the present study established a thermal-hydraulic model for a spiral-coil primary heat exchanger using liquid lead and supercritical carbon dioxide (S-CO₂) as working fluids. A...

Impact of Decay on the Transport of Radioactive Aerosols in Long Square Tubes

Liu Man, Xia Mingming, Chen Zhi

2024, 45(4): 205-212. doi: 10.13832/j.jnpe.2024.04.0205

Abstract(93) HTML (41) PDF(18)

Abstract:
Decay radiation can cause accumulation of surface charges of radioactive aerosol particles, and then affect their migration process. However, the charge effect of decay is not considered in the current radionuclide transport simulation. In this study, a particle decay charging model was established ...

Capacity Configuration and Operation Optimization of a Low-Temperature Reactor Nuclear Heating System with Heat Storage

Liu Weiqi, Wang Jinshi, Xue Kai, Sun Zhiyong, Liu Xingmin, Li Gen, Yan Junjie

2024, 45(4): 213-220. doi: 10.13832/j.jnpe.2024.04.0213

Abstract(166) HTML (82) PDF(20)

Abstract:
In order to meet the growing demand for low-carbon heating and improve the operational flexibility and economic benefits of the heating system, a nuclear heating system (DHGHS) integrating a "Yanlong" pool-type low-temperature heating reactor (DHR-400), a heat storage pool, and a gas boiler was prop...

Research on Vibration Measurement Method of Nuclear Power Plant Pipeline Based on Unmarked Vision Algorithm

He Mengfu, Zhang Yiming, Qin Manqing, Xu Zili, Liao Tongtong

2024, 45(4): 221-227. doi: 10.13832/j.jnpe.2024.04.0221

Abstract(81) HTML (24) PDF(16)

Abstract:
In order to improve the problem that the vibration response of thin-walled pipes and small branch pipes is difficult to be effectively measured by contact measurement method, this paper proposes to calculate the optical flow of adjacent frames at different times based on the camera calibration algor...

Dissolution Behavior of Steam Generator Deposit in EDTA Solution

Song Lijun, Xiao Yan, Sun Yun, Tian Zhaohui, Liu Canshuai, Zou Wei

2024, 45(4): 228-234. doi: 10.13832/j.jnpe.2024.04.0228

Abstract(83) HTML (32) PDF(11)

Abstract:
In order to explore the solubility of chemical cleaning agent EDTA on simulated deposit Fe₃O₄ and steam generator sludge, and to guide the selection of chemical cleaning processes, XRF and ICP-OES were used to analyze the dissolution effects of solution temperature, EDTA concentration, and dissoluti...

Research and Application of Influence of Black Rod and Gray Rod on Control Rod Drop Time in Nuclear Power Plant

Zhang Hengkai, Liu Hang, Liu Jikun, Liu Shuangjin, Zhao Yuntao

2024, 45(4): 235-240. doi: 10.13832/j.jnpe.2024.04.0235

Abstract(153) HTML (83) PDF(29) [Cited by] (1)

Abstract:
In order to achieve more accurate and detailed drop time of control bank in nuclear power plant, based on the force analysis of control rod assembly in Chinese Pressurized Reacter 1000 MW (CPR1000) and Advanced Chinese Pressurized Reactor 1000 MW (ACPR1000⁺) nuclear power units and the test results ...

Study on Operation Reliability of Reactor Trip Circuit Breaker in Nuclear Power Plants

Li Huwei, Zhang Yangcheng, Li Bin

2024, 45(4): 241-244. doi: 10.13832/j.jnpe.2024.04.0241

Abstract(105) HTML (30) PDF(23)

Abstract:
The reactor trip circuit breaker (RTCB) is a very important actuator in the reactor protection system (RPS) of nuclear power plant, and its reliability is directly related to whether the reactor can achieve and maintain a stable safety state. In this paper, the operation status of RTCB in domestic n...

ACP100 Radiation Streaming Shielding Design Based on Discrete Ordinates Visual Modeling Technology

Tang Songqian, Chen Xin, Liu Jiajia, Wen Xingjian, Tian Chao

2024, 45(4): 245-247. doi: 10.13832/j.jnpe.2024.04.0245

Abstract(178) HTML (110) PDF(41)

Abstract:
The design characteristics of ACP100 make the radiation steaming become one of the radiation protection problems that need to be paid attention to, and it is necessary to carry out targeted shielding design. In order to improve the accuracy of discrete ordinates method in small modular reactor model...

Research on Dynamic Parameter Model of Electrical Performance of Reactor Control Rod Drive System

Li Mengshu, Tang Shihan, Zheng Gao, He Zhengxi, Li Qing, Fu Guozhong, Peng Ziheng, Chen Shuaijun, Zhang Yun

2024, 45(4): 248-254. doi: 10.13832/j.jnpe.2024.04.0248

Abstract(741) HTML (54) PDF(37)

Abstract:
Due to the lack of accurate parameter model of electrical performance in the dynamic change process of the reaction system, the control effect of the existing control rod drive system in nuclear power plant is not good under different environmental conditions. In this paper, the static electromagnet...

Automatic Control Method of Nuclear Thermal Propulsion System Based on vPower

Ma Xinyi, Han Wenbin, Deng Jian, Huang Shanfang, Qi Zhichao

2024, 45(4): 255-261. doi: 10.13832/j.jnpe.2024.04.0255

Abstract(84) HTML (47) PDF(23)

Abstract:
Nuclear thermal propulsion has the advantages of large thrust, high specific impulse, high energy conversion efficiency and long operating time, with broad prospects in the field of deep space exploration. Automatic reactor control can reduce the misoperation accident caused by human, improve econom...

Numerical Simulation and Experiment Research on Radioisotope Thermoelectric Generator

Huang Xueliang, Li Mancang, Chen Zhang, Zhang Xinhu, Zhou Daijie, Chen Zhiyu, Xie Yunli, Guo Rui, Wang Yu

2024, 45(4): 262-266. doi: 10.13832/j.jnpe.2024.04.0262

Abstract(167) HTML (69) PDF(26)

Abstract:
Radioisotope thermoelectric generator is a device that converts the thermal energy produced by the decay of radioactive isotopes into electric energy. It involves the strong coupling of thermal-electric physical fields and is difficult to simulate accurately. In this paper, based on a ⁹⁰Sr Radioisot...

Experimental Investigation on Minimum Film Boiling Temperature during Quenching of FeCrAl

Wang Zefeng, Deng Jian, Qiu Zhifang, Chen Xi, Wang Xiaoyu, Chen Jianda, Xiong Jinbiao

2024, 45(4): 267-273. doi: 10.13832/j.jnpe.2024.04.0267

Abstract(125) HTML (66) PDF(17)

Abstract:
FeCrAl is proposed as one of the candidate materials of accident tolerant fuel (ATF) cladding, which can suppress hydrogen generation under severe accident condition, and improve reactor accident tolerance. In this paper, the boiling heat transfer behavior of FeCrAl and Zr-4 during quenching is stud...

Anti-Noise Coding Research for Next-Generation Nuclear Power DCS Communication System

Shan Weiwei, Ren Jie, Peng Weilun, Zeng Hui, Li Sixing, Xiao Anhong, Feng Jintao, Deng Yuhao

2024, 45(4): 274-279. doi: 10.13832/j.jnpe.2024.04.0274

Abstract(1258) HTML (42) PDF(31)

Abstract:
In the process of introducing wireless signals in the rupgrading of wireless circuits in nuclear power plants or the design of the next generation distributed control system (DCS), it is necessary to improve the quality of wireless communication through error-correcting codes. This paper investigate...

2024 Vol. 45, No. 4