核动力工程

Overall Study of Dock-based Floating Nuclear Power Plant

Wang Donghui, Li Qing, Song Danrong, Qin Dong, Liu Jia

2023, 44(4): 1-8. doi: 10.13832/j.jnpe.2023.04.0001

Abstract(2407) HTML (353) PDF(279) [Cited by] (4)

Abstract:
Considering the domestic and overseas development trend of floating nuclear power plant (FNPP), a dock-based ACP100S FNPP is proposed in this paper to promote the construction of FNPP in China. A preliminary evaluation is conducted in the external event, nuclear reactor design, hull design, economic...

High-fidelity Full Core Neutronics Calculation Method for Fuel Assembly Bowing and Its Application

Li Fan, Liu Zhouyu, Wang Xining, Cao Liangzhi, Wu Hongchun

2023, 44(4): 9-16. doi: 10.13832/j.jnpe.2023.04.0009

Abstract(490) HTML (80) PDF(51) [Cited by] (1)

Abstract:
In order to study the influence of bowing deformation of fuel assemblies on core power distribution, this paper proposes an equivalent method for simulating the bowing of fuel assemblies in PWR Core. That is, according to the principle of conservation of atomic number of the water gap material aroun...

Selection of Shadow Shielding Structure and Material for Space Reactor

Huang Qianming, Li Lan, Chai Xiaoming, Liu Bin, Ying Dongchuan

2023, 44(4): 17-24. doi: 10.13832/j.jnpe.2023.04.0017

Abstract(358) HTML (83) PDF(78)

Abstract:
The space reactor has strict requirements on the size and weight of radiation shielding. In order to find out a suitable shielding scheme, it is necessary to select the shielding material and structure. In this paper, firstly, the research progress of shielding targets and limits of space reactors a...

Research and Application of Single-Point Calibration Method for Ex-core Nuclear Instrument System of PWRs

Bai Jiahe, Yang Haozhe, Wan Chenghui, Pan Zefei, Li Zaipeng, Li Yunzhao, Wu Hongchun

2023, 44(4): 25-32. doi: 10.13832/j.jnpe.2023.04.0025

Abstract(307) HTML (110) PDF(53) [Cited by] (1)

Abstract:
The ex-core detector is used to indicate the reactor power and axial-power difference in commercial PWRs, and it needs to be calibrated regularly to ensure the accuracy. The conventional multi-point calibration method needs to move the control rods and measure the flux-mapping several times, which i...

Research on Direct Transport Calculation Method Based on Numerical Nuclear Reactor Physics Code SHARK

Zhao Chen, Zhao Wenbo, Zhang Hongbo, Wang Bo, Chen Zhang, Peng Xingjie, Gong Zhaohu, Zeng Wei, Li Qing

2023, 44(4): 33-40. doi: 10.13832/j.jnpe.2023.04.0033

Abstract(1408) HTML (96) PDF(108) [Cited by] (3)

Abstract:
In order to establish the next-generation reactor physics calculation method based on the numerical nuclear reactor technology and realize high-fidelity modeling, high-resolution and high-precision calculation, the research of direct transport method was conducted based on the numerical nuclear reac...

Research on Simulation of Neutron Transport with Thick Diffusion Limit in Curved Meshes

Wang Xinyu, Zhang Bin, Chen Yixue

2023, 44(4): 41-48. doi: 10.13832/j.jnpe.2023.04.0041

Abstract(227) HTML (61) PDF(26)

Abstract:
Discrete ordinate method is one of the main numerical methods for solving the problem of neutron transport with thick diffusion limit. Its commonly used spatial discrete schemes, such as finite difference scheme, are easy to cause numerical diffusion in optical thick media, and the application of di...

Research on Automatic Modeling Method of TORT Program Based on CAD Model

Xu Fangyuan, Yang Chao, Yu Tao, Chen Zhenping, Huang Guocai, Li Leiming, Li Yukun, Xian Xirui, Du Hua

2023, 44(4): 49-54. doi: 10.13832/j.jnpe.2023.04.0049

Abstract(338) HTML (98) PDF(35) [Cited by] (1)

Abstract:
Aiming at the problems of complex geometry of reactor shielding structure, limited geometric processing ability, low efficiency and error prone of traditional manual modeling, based on the multi-objective modeling and simulation platform for radiation transport (MOSRT), the volume weight homogenizat...

Numerical Study on Flow and Heat Transfer of High-pressure Sub-cooled Water Injection into High-temperature Lead-bismuth Alloy under Lead-bismuth Cooled Fast Reactor SGTR Accident

Liu Li, Yuan Junjie, Gu Hanyang, Bao Ruiqi, Liu Maolong, Wang Ke

2023, 44(4): 55-64. doi: 10.13832/j.jnpe.2023.04.0055

Abstract(558) HTML (148) PDF(105) [Cited by] (3)

Abstract:
There are high-pressure sub-cooled water and high-temperature lead-bismuth coolant on both sides of the heat transfer tube of the steam generator in the lead-bismuth fast reactor. The large pressure difference and temperature difference on both sides of the heat transfer tube and the corrosion effec...

Derivation and Evaluation of Carbon Dioxide Partial Derivative Property by Implicit Solution of Brayton Cycle System

Wen Shuang, Wen Qinglong, Hu Wenjun, Xu Shijia

2023, 44(4): 65-71. doi: 10.13832/j.jnpe.2023.04.0065

Abstract(484) HTML (99) PDF(26) [Cited by] (1)

Abstract:
To improve the accuracy of solving the Brayton Cycle equation of supercritical carbon dioxide (S-CO₂), the fully implicit or semi-implicit difference scheme is used to solve the fluid conservation equation discretely, and the partial derivative property is indispensable for the implicit solution. Th...

Development of Dimensionless Rod-bundle CHF Correlation Based on Stepwise Regression and Determination of DNBR Limit

Yin Yuan, Feng Simin, Pang Bo, Xi Yanyan, Zhang Yuxiang, Fu Xiangang

2023, 44(4): 72-78. doi: 10.13832/j.jnpe.2023.04.0072

Abstract(1170) HTML (126) PDF(46)

Abstract:
At present, the empirical correlations of critical heat flux (CHF) of advanced PWR rod-bundles at home and abroad generally have the common problems of complex mathematical form, numerous independent variable coefficients and lack of physical significance. In this study, based on 485 rod-bundle CHF ...

Study on the Effect of Inclination Angle on the Natural Convection of Molten Salt in Heat Pipe Cooled Molten Salt Reactor Core

Chen Zehan, Chen Xingwei, Dai Ye, Zou Yang

2023, 44(4): 79-87. doi: 10.13832/j.jnpe.2023.04.0079

Abstract(397) HTML (111) PDF(54)

Abstract:
The inclination angle of heat pipe-cooled molten salt reactor (MSR) core has an important influence on the core temperature distribution and local hot spots. In order to obtain the natural convection heat transfer characteristics of molten salt in the core at different inclination angles, optimize t...

Effect of the Arrangement for Grid Spacers with Mixing Vanes on the Thermal Hydraulic Characteristics of a 5×5 Rod Bundle via CFD Analysis

Su Qianhua, Fan Guanhua, Lyu Lulu, Lu Donghua, Yang Ping, Gan Fujun, Yan Binghuo, Wang Chengyue

2023, 44(4): 88-94. doi: 10.13832/j.jnpe.2023.04.0088

Abstract(364) HTML (144) PDF(64) [Cited by] (3)

Abstract:
In order to optimize the arrangement of the grid spacer for a rod bundle, the computational fluid dynamics (CFD) method is employed to analyze the flow and temperature fields for a 5×5 rod bundle equipped with three grid spacers. The effect of the axial separation between two neighboring grid spacer...

Research on Fatigue Performance of Newly Developed Supporting Structure of Spacer Grid

Guo Xiaoming, Ren Quanyao, Chen Jie, Ren Yi

2023, 44(4): 95-99. doi: 10.13832/j.jnpe.2023.04.0095

Abstract(344) HTML (148) PDF(62)

Abstract:
Aiming at the innovative supporting structure and its fatigue performance, this paper evaluates the stress state under its operating conditions by means of finite element analysis, and carries out the spring fatigue test to analyze the fatigue failure characteristics and crack initiation locations. ...

Study on Refined Calculation Method for Added Mass of Special-shaped Structures in Complex Fluid Domain

Tan Ximing, Gao Fuhai, Qi Min, Wang Yueying, Liu Zhaoyang

2023, 44(4): 100-106. doi: 10.13832/j.jnpe.2023.04.0100

Abstract(219) HTML (108) PDF(32) [Cited by] (1)

Abstract:
The engineering design of reactor internals usually uses the added mass method to simulate the dynamic effects of fluids on the structure. Taking a special-shaped pressure pipe immersed in a complex fluid domain as an example, a structural added mass iterative calculation method is proposed, which u...

Experimental Study on Mechanical Performance of Nuclear Containment Truncated Cone Region

Lan Tianyun, Wu Yuzheng, Xiao Dan, Zhou Chuanbo, Dong Zhanfa, Guo Junying, Xiong Meng

2023, 44(4): 107-115. doi: 10.13832/j.jnpe.2023.04.0107

Abstract(162) HTML (81) PDF(24) [Cited by] (2)

Abstract:
The truncated cone area of nuclear containment (the junction of shell and raft foundation) has irregular shape and complex stress. Studying the stress mechanism of this area is important to master the structural performance of the whole containment. In this study, the truncated cone area was taken a...

Multi-objective Optimization Design of the Support in Flow Induced Vibration Test of Reactor Internals

Zhang Yu, Li Pengzhou, Qiao Hongwei, Miao Yuhan, Gao Lixia, Yu Danping, Sun Lei

2023, 44(4): 116-120. doi: 10.13832/j.jnpe.2023.04.0116

Abstract(281) HTML (116) PDF(31)

Abstract:
In order to fully improve the material utilization rate and isolate the fundamental frequency of the reactor internals, a parameter optimization study based on multi-objective optimization algorithm was carried out for the support structure in the flow induced vibration (FIV) test. The finite elemen...

Analysis of Thermal Stress and Fatigue Induced by Dryout Oscillation in Once Through Steam Generator

Chen Ling, Wang Xinming, Zhang Yongfa, Zhang Liming, Jiang Lizhi, Jiao Meng, Liu Xiaoya

2023, 44(4): 121-127. doi: 10.13832/j.jnpe.2023.04.0121

Abstract(198) HTML (59) PDF(25) [Cited by] (2)

Abstract:
In order to study the damage of the heat transfer tube caused by dryout oscillation, the once through steam generator designed by Babcock&Wilcox company is taken as a prototype. Firstly, the relevant thermal and hydraulic parameters are obtained by using the method of primary and secondary side ...

Reaction Thermodynamics and Kinetics Analysis of Dry Extraction of Ba¹⁴CO₃ by Irradiated AlN

Cao Qi, Chen Yunming, Zhang Jingsong, Luo Ning, Dai Shuang, Lu Yunyun, Yang Yu

2023, 44(4): 128-132. doi: 10.13832/j.jnpe.2023.04.0128

Abstract(224) HTML (89) PDF(15)

Abstract:
The reaction thermodynamics, kinetics analysis and demonstration test of a high temperature oxidation reaction under complex irradiated AlN system was carried out by thermodynamic calculation software (HSC). The results show that it is necessary to ensure sufficient O₂ during the reaction to improve...

Probabilistic Safety Analysis Framework for Internal Events of Aqueous Homogeneous Reactor

Wang Zhe, Zhang Dan, Zou Zhiqiang, Wang Ningning, Yang Weidong, Du Yu

2023, 44(4): 133-137. doi: 10.13832/j.jnpe.2023.04.0133

Abstract(262) HTML (86) PDF(33) [Cited by] (2)

Abstract:
There are significant differences in the safety design and operation characteristics between liquid fuel reactor (aqueous homogeneous reactor) and traditional solid fuel reactor, therefore, it is impossible to carry out the safety design of aqueous homogeneous reactor only using the existing safety ...

Study on Control Strategy of Natural Circulation Lead-cooled Fast Reactor Coupled with S-CO₂ Brayton Cycle

Liu Guixiu, Yi Jingwei, Li Gen, Liang Tiebo, Fang Huawei, Chen Weixiong

2023, 44(4): 138-147. doi: 10.13832/j.jnpe.2023.04.0138

Abstract(374) HTML (119) PDF(59) [Cited by] (1)

Abstract:
The coupled power generation system of natural circulation lead cooled fast reactor with supercritical carbon dioxide (S-CO₂) Brayton cycle is the development trend of advanced nuclear energy systems in the future. Based on the software Apros, a dynamic model of the coupled power generation system w...

Uncertain Information Modeling and Processing in SPAR-H Method under Group Decision Making

Guan Xuefan, Wang Danyu, Su Xiaoyan, Xu Zhihui, Qian Hong

2023, 44(4): 148-153. doi: 10.13832/j.jnpe.2023.04.0148

Abstract(272) HTML (126) PDF(21) [Cited by] (2)

Abstract:
Considering the uncertain information in Standardized Plant Analysis of Risk-Human Reliability Analysis (SPAR-H) method and the inability to handle the assessment with multiple experts, this paper proposes a method of uncertain information modeling and processing in SPAR-H method under group decisio...

Research on Cascade Control Method of Electric Power of NUSTER-100

Pu Songmao, Huang Jiajun, Sun Peiwei, Wei Xinyu

2023, 44(4): 154-162. doi: 10.13832/j.jnpe.2023.04.0154

Abstract(187) HTML (113) PDF(36) [Cited by] (1)

Abstract:
The heat pipe cooled reactor (hereinafter referred to as heat pipe reactor) has the design concept of solid-state reactor, and the heat is passively transferred out of the core through heat pipes. It has the advantages of simple structure, high safety, low noise, compact structure and long working t...

An Approach for Dynamic Reliability Assessment of Reactor Trip System of HPR1000

Li Kunxiang, Sui Yang, Dai Tao, Yu Tao

2023, 44(4): 163-169. doi: 10.13832/j.jnpe.2023.04.0163

Abstract(217) HTML (60) PDF(47) [Cited by] (2)

Abstract:
The structure of reactor trip system (RTS) is complex, which leads to its dynamic interaction, time dependence and probability uncertainty. However, the traditional static reliability assessment methods are difficult to characterize these three characteristics. To solve this problem, a novel approac...

Optimization of Feedwater Control for Casing Steam Generator Based on Apros

Liu Haipeng, Wang Changshuo, Ye Zhu, Tian Peiyu

2023, 44(4): 170-178. doi: 10.13832/j.jnpe.2023.04.0170

Abstract(558) HTML (121) PDF(41) [Cited by] (1)

Abstract:
In allusion to the problem of water supply control caused by the strong coupling of casing steam generator, the steam-water circulation system of commercial modular small reactor with casing steam generator is taken as the research object, and the simulation model of steam-water circulation system i...

Research on Model-driven Top-level Architecture Design Method of Nuclear Engineering

Pan Xinxin, Zhuang Yaping, Song Chunjing, Lin Chao

2023, 44(4): 179-184. doi: 10.13832/j.jnpe.2023.04.0179

Abstract(928) HTML (103) PDF(67)

Abstract:
An iterative optimization modeling process of nuclear engineering top-level functional architecture and logical architecture based on MagicGrid methodology is proposed. Starting from normal conditions, functional analysis and logic construction are gradually carried out. Based on the fault mode of l...

Research on Ultrasonic Technology of Accurate Measurement of Main Feedwater Flow in Nuclear Power Plant

Bai Tian, Wang Fengning, Liu Yan, Wei Huatong, Cui Xiwei, Guo Lin, Liu Hai, Liu Li

2023, 44(4): 185-191. doi: 10.13832/j.jnpe.2023.04.0185

Abstract(254) HTML (62) PDF(48) [Cited by] (1)

Abstract:
A high-precision ultrasonic flow meter with a design measurement uncertainty of 0.3% has been investigated and developed for the measurement of the main feedwater flow in the secondary loop. This flowmeter could be used to realize the small power enhancement of PWR nuclear power units. The influence...

Prediction of RPV Irradiation Embrittlement Performance and Life Evaluation of Extended Operation in a Nuclear Power Plant

Fang Yonggang, Tong Zhenfeng, Chu Qibao, Zeng Zhen, Shen Yingcai

2023, 44(4): 192-197. doi: 10.13832/j.jnpe.2023.04.0192

Abstract(255) HTML (117) PDF(39) [Cited by] (1)

Abstract:
A domestic self-designed and built nuclear power plant has entered the stage of extended operation. A foreign prediction model was adopted for the irradiation embrittlement evaluation of the reactor pressure vessel, however, the irradiation data on which the foreign prediction model was based cannot...

Study on Issues of Transformation to Chinese Technical Specifications

Li Huwei, Zhang Yangcheng, Qian Xiaoming, Zhang Chi

2023, 44(4): 198-202. doi: 10.13832/j.jnpe.2023.04.0198

Abstract(275) HTML (102) PDF(33)

Abstract:
During the transformation of French technical specifications (TSs) to Chinese TSs for M310 NPPs such as Daya Bay Nuclear Power Plant with reference to the Standard Technical Specifications - Westinghouse Nuclear Power Plant (NUREG-1431, Rev. 4), it was found that there is no effective management of ...

Research on Automatic Calibration Algorithm of Reactor Fuel Rods

Li Xiangdong, Jiang Hesong, Wang Xueyuan, Xu Xuejin, He Xiaochun

2023, 44(4): 203-208. doi: 10.13832/j.jnpe.2023.04.0203

Abstract(166) HTML (90) PDF(23)

Abstract:
Since the nuclear reactors need frequent replacement of fuel rods, it is necessary to determine the type and installation position of the core fuel rods accurately to ensure the safe operation of the reactor. Herein, the global and local virtual two-dimensional coordinate mapping models have been es...

Design of Control System for Inspection Robot of In-Containment Refueling Water Storage Tank

Guan Chaopeng, Wu DongDong, Gui Liang

2023, 44(4): 209-213. doi: 10.13832/j.jnpe.2023.04.0209

Abstract(218) HTML (108) PDF(25) [Cited by] (1)

Abstract:
Aiming at the underwater inspection requirements of in-containment refueling water storage tank (IRWST) during the in-service overhaul of nuclear power unit and minimizing personnel radiation dose, a remote-controlled inspection robot is developed. Through the analysis of the IRWST inspection requir...

Study on Optimization of Allowable Outage Time for Essential Service Water System in HPR1000

Chen Guocai, Yang Yun, Tong Jiejuan

2023, 44(4): 214-219. doi: 10.13832/j.jnpe.2023.04.0214

Abstract(212) HTML (110) PDF(25)

Abstract:
The risk-informed integrated decision-making method combining deterministic and probability theories was adopted for the optimization analysis and demonstration of the allowed outage time (AOT) in the terms of operational technical specifications for the essential service water system (WES) of Hualo...

Experimental Study on Flow Instability in Parallel Channels with Supercritical Carbon Dioxide

Huang Jiajian, Zhou Yuan, Huang Yanping, Luo Qiao, Hu Wei

2023, 44(4): 220-225. doi: 10.13832/j.jnpe.2023.04.0220

Abstract(454) HTML (76) PDF(37)

Abstract:
Carbon dioxide has unique physical and chemical properties near the quasi-critical point. The Bretton cycle system using supercritical carbon dioxide (S-CO₂) as heat exchange medium has considerable system thermal efficiency, but the drastic change of physical properties may lead to the problem of f...

Development and Verification of 3D Code for Steam Generator Tube Rupture Accident of LBE-cooled Reactor

Gu Zhixing, Yu Hongxing, Huang Daishun, Yan Mingyu, Shen Yaou, Feng Wenpei, Gong Zhengyu

2023, 44(4): 226-233. doi: 10.13832/j.jnpe.2023.04.0226

Abstract(313) HTML (84) PDF(45) [Cited by] (2)

Abstract:
Steam Generator Tube Rupture (SGTR) accident is one of the significant safety problems that must be considered in the design of Lead-Bismuth-Eutectic (LBE) cooled reactor. With respect to the SGTR in LBE-cooled reactor, and to cope with the challenges of 3D propagation of pressure waves and 3D migra...

Study on Evaluation Method for Creep Performance of New Zirconium Alloy Cladding

Xing Shuo, Pu Zengping, Zhang Kun, Jiao Yongjun, Dai Xun, He Liang

2023, 44(4): 234-239. doi: 10.13832/j.jnpe.2023.04.0234

Abstract(390) HTML (135) PDF(59) [Cited by] (2)

Abstract:
In order to establish a creep model for new zirconium alloy, based on the data from creep test, the creep behavior of the new zirconium alloy cladding tubes at the temperature of 593-673K and the stress of 60-160 MPa was investigated. The classical zirconium alloy creep model was used to predict the...

Experimental Study on Prototype Melt under Severe Accident

Li Yang, Gong Houjun, Guo Kerong, Hu Yuwen, Yang Shengxing, Zan Yuanfeng, Yang Zumao, Huang Yanping

2023, 44(4): 240-246. doi: 10.13832/j.jnpe.2023.04.0240

Abstract(705) HTML (173) PDF(39)

Abstract:
In order to study the stratification morphology of the molten pool of the lower head of the pressure vessel under severe accident conditions, it is necessary to melt the prototype melt into liquid for experiments. In this study, CESEF experimental device is used, and the prototype melt is melted wit...

Experimental Study on Cooling Characteristics of Mixed Particle Size Debris Bed under Different Water Injection Methods

Yang Shengxing, Gong Houjun, Fang Yu, Li Yang, Hu Yuwen, Zan Yuanfeng, Yang Zumao, Zhuo Wenbin

2023, 44(4): 247-252. doi: 10.13832/j.jnpe.2023.04.0247

Abstract(525) HTML (110) PDF(29)

Abstract:
The liquid core melt interacts with the coolant and breaks into a particle bed. Effective cooling of the particle bed can realize the retention of the melt and stop the accident process. In this paper, based on the particle size distribution and porosity of the fragments after the prototype molten F...

Experimental Research on Mechanism of Aerosol Re-entrainment Behavior in Containment under Severe Accident Condition

Hu Zhen, Chen Linlin, Ji Songtao, Wei Yansong, Shi Xiaolei, Zheng Guangzong

2023, 44(4): 253-258. doi: 10.13832/j.jnpe.2023.04.0253

Abstract(511) HTML (103) PDF(39)

Abstract:
Based on the Integral test facility of aerosol re-suspension and re-entrainment, the experimental study on aerosol re-entrainment behavior in passive containment was carried out. By measuring the mass concentration, quantity concentration and particle size distribution of particulate matter, the mot...

Development and Verification of Ph Calculation Model of in-Containment Refueling Water Storage Tank under Severe Accidents

Dai Wei, Jiang Pingting, Chen Peng, He Dongyu

2023, 44(4): 259-266. doi: 10.13832/j.jnpe.2023.04.0259

Abstract(542) HTML (118) PDF(24)

Abstract:
In order to solve the problem of lacking tools for calculating the pH value of in-containment refueling water storage tank in nuclear power plants after the accident, this paper develops a direct modeling and in-time analysis model for pH value calculation. Based on Newton-Raphson method, by establi...

2023 Vol. 44, No. 4