核动力工程

Solving Multi-Dimensional Neutron Diffusion Equation Using Deep Machine Learning Technology Based on PINN Model

Liu Dong, Luo Qi, Tang Lei, An Ping, Yang Fan

2022, 43(2): 1-8. doi: 10.13832/j.jnpe.2022.02.0001

Abstract(3740) HTML (638) PDF(603) [Cited by] (12)

Abstract:
This paper elaborates the physics-informed neural network model (PINN), constructs a deep neural network as a trial function, substitutes it into the neutron diffusion equation to form a residual, and takes it as the weighted loss function of machine learning, and then approaches the numerical solut...

A Two-dimensional Coupled Neutron Transport Method for MOC and S_N via Boundary Fluence Rate Coupling

Zhang Sifan, Yuan Yuan, Liu Zhouyu, Zhou Xinyu, Cao Liangzhi

2022, 43(2): 9-16. doi: 10.13832/j.jnpe.2022.02.0009

Abstract(586) HTML (128) PDF(55)

Abstract:
When the method of characteristics (MOC) is used to calculate the out-of-pile detector or some special heavy water moderated light water cooled experimental reactor, the dense characteristics will lead to a large waste of computing resources because the external structural material or moderator area...

Research on Transient Multi-Group Pin-Power Reconstruction Based on Source Expansion Method

Bai Jiahe, Wan Chenghui, Li Yunzhao, Wu Hongchun

2022, 43(2): 17-21. doi: 10.13832/j.jnpe.2022.02.0017

Abstract(363) HTML (113) PDF(47) [Cited by] (1)

Abstract:
This paper studies the calculation method of transient multi-group pin-power reconstruction of PWR. Starting from the transient fixed-source equation, the transient fixed-source term is expressed in the form of biquadratic Legendre polynomial, and the pin-power reconstruction is realized through the...

Experimental Research of Bundle and Spacer Grid Arrangement on Fuel Assembly Mixing Characteristics

Cheng Cheng, Ye Tingpu, Lu Donghua, Su Qianhua, Long Biao, Chen Zhenhui

2022, 43(2): 22-27. doi: 10.13832/j.jnpe.2022.02.0022

Abstract(380) HTML (113) PDF(55) [Cited by] (1)

Abstract:
Taking the 5×5 bundle fuel assembly assembled with the spacer grid used in CPR1000 nuclear power unit as the object, experiments on the mixing characteristics of several groups of full-length bundle fuel assembly were carried out. The effects of geometric parameters such as cold-hot rod arrangement ...

Research on ¹⁰B Abundance Calculation Method of PWR Based on Bamboo-C

Liu Yu, Wan Chenghui, Huang Xing, Wu Hongchun, Zhang Shengbin, Cai Guangming

2022, 43(2): 28-31. doi: 10.13832/j.jnpe.2022.02.0028

Abstract(420) HTML (125) PDF(51) [Cited by] (2)

Abstract:
During the power operation of PWR, due to burnup effect and boronation effect, the abundance of ¹⁰B in the primary circuit boric acid solution will change continuously with time, and the nuclear power plant can not provide the real-time measured value of ¹⁰B abundance. As a result, the calculated va...

MSR Supercritical Carbon Dioxide Brayton Cycle System and Thermodynamic Analysis

Lu Heng, Zhao Heng, Dai Ye, Chen Xingwei, Jia Guobin, Zou Yang

2022, 43(2): 32-39. doi: 10.13832/j.jnpe.2022.02.0032

Abstract(971) HTML (254) PDF(106) [Cited by] (5)

Abstract:
Molten salt reactor (MSR) can realize on-line packing and post-processing, and the outlet temperature is higher, so it shall be equipped with an innovative cycle mode that matches its outlet temperature, and can achieve higher cycle efficiency. In this paper, a supercritical carbon dioxide (SCO₂) Br...

Research on the Key Influencing Factors of the Backflow Phenomenon on the Primary Side of the Inverted U-tube Steam Generator under Natural Circulation

Wang Tianshi, Wang Yuxuan, Zhao Pengcheng, Wang Xijie, Ling Yufan, Wang Yuqing, Zhu Enping

2022, 43(2): 40-46. doi: 10.13832/j.jnpe.2022.02.0040

Abstract(788) HTML (219) PDF(127) [Cited by] (1)

Abstract:
Backflow exists in the inverted U-tube steam generator (UTSG) under the condition of natural circulation, which affects the heat carrying capacity and natural circulation capacity of the primary circuit coolant system. Referring to the design parameters of UTSG in PWR thermal experimental device (PW...

Evaluation of Single-phase and Two-phase Mixing Models for Rod Bundle Channel

Ye Tingpu, Lyu Lulu, Zhang Ge, He Hui, Cheng Cheng

2022, 43(2): 47-52. doi: 10.13832/j.jnpe.2022.02.0047

Abstract(415) HTML (119) PDF(58) [Cited by] (2)

Abstract:
In this study, the subchannel program is used to evaluate the single-phase and two-phase mixing model of the rod bundle channel based on the existing experimental data. The single-phase mixing mainly considers the cross flow and turbulent mixing. The cross flow is determined by the conservation equa...

Influence of Inlet and Outlet Resistance on the Characteristics of Backflow in Inverted U-tube Bundle

Ma Songyang, Li Mingrui, Chen Wenzhen, Hao Jianli, Xiao Hongguang, Ye Lei

2022, 43(2): 53-58. doi: 10.13832/j.jnpe.2022.02.0053

Abstract(518) HTML (186) PDF(66) [Cited by] (1)

Abstract:
The inverted U-tube bundle backflow problem of the steam generator (SG), which increases the flow resistance of SG under natural circulation conditions and reduces the natural circulation capacity of the system, has a negative impact on reactor safety. In view of the above problems, taking the marin...

Stability Analysis of Double-Loop Natural Circulation under Asymmetric Conditions

Zhu Enping, Wang Ting, Liu Zijing, Zhao Pengcheng, Wang Tianshi

2022, 43(2): 59-64. doi: 10.13832/j.jnpe.2022.02.0059

Abstract(403) HTML (118) PDF(39)

Abstract:
In order to study the stability change of double loop natural circulation under asymmetric conditions, taking the single loop natural circulation heat carrying system as the starting point, the displacement term in the dimensionless governing equation set of single loop natural circulation is expand...

Study on Flow Pattern Evolution in Outlet Heat Removal Pipe of Open Natural Circulation System

Sun Yuxiang, Xu Jianjun, Zhou Huihui, Deng Zhiyong, Yuan Zhaofei, Cui Yinghuan, Huang Yanping

2022, 43(2): 65-69. doi: 10.13832/j.jnpe.2022.02.0065

Abstract(385) HTML (94) PDF(30) [Cited by] (1)

Abstract:
As the ultimate heat sink discharge loop of the new passive residual heat removal system, the safe and stable operation of the open natural circulation system is very important for removing the residual heat of the reactor core under accident conditions. In this paper, the flow pattern evolution in ...

CFD Analysis on Characteristics of High Temperature Heat Pipe

Yu Qingyuan, Zhao Pengcheng, Ma Yugao

2022, 43(2): 70-76. doi: 10.13832/j.jnpe.2022.02.0070

Abstract(1094) HTML (378) PDF(165) [Cited by] (17)

Abstract:
The analysis and prediction of the operating characteristics of high temperature heat pipe are of great significance to the design and application of heat pipe. In order to analyze the heat transfer characteristics of two-phase flow in high temperature heat pipe, first, the computational fluid dynam...

Experimental Research on Pressure Drop Characteristics of Flow Damper in Advanced Accumulator

He Yanqiu, Yuan Zhaofei, Zhang Yan, Tan Shushi, Zan Yuanfeng, Qiao Min, Hu Qiang

2022, 43(2): 77-82. doi: 10.13832/j.jnpe.2022.02.0077

Abstract(338) HTML (85) PDF(29) [Cited by] (1)

Abstract:
Two different types of pressure drop characteristics of flow damper were obtained through the test of pressure drop characteristics of flow damper in advanced accumulator. The influence law of different geometric parameters on the pressure drop coefficient is studied, and the relationship of pressur...

Study on Coupling Characteristics of Multiple Thermal Parameters during the Fuel Assembly Steady-State Irradiation in the Test Loop

Si Junping, Sun Sheng, Tong Mingyan, Lu Mengkang, Lei Jin, Jin Shuai, Wang Wanjin

2022, 43(2): 83-88. doi: 10.13832/j.jnpe.2022.02.0083

Abstract(904) HTML (118) PDF(56) [Cited by] (1)

Abstract:
The steady-state irradiation in the test loop is a key process to study the anti-irradiation performance of the fuel assembly. In view of the importance of the thermal parameters of the irradiation test loop to the test run, aiming at the steady-state irradiation in the test loop, combined with the ...

Research on Analysis for Performance and Optimization of Prismatic Dispersed Microencapsulated Fuel in Gas-Cooled Reactor

Zhao Bo, Li Quan, Li Yuanming, Huang Yongzhong, Ma Qiang, Su Min, Liu Zhenhai, Qi Feipeng, Ma Chao, Chen Hao

2022, 43(2): 89-95. doi: 10.13832/j.jnpe.2022.02.0089

Abstract(563) HTML (66) PDF(85) [Cited by] (1)

Abstract:
Prismatic dispersed microencapsulated fuel is a particle-reinforced composite fuel which is formed by TRISO fuel particles dispersed in metal or ceramic matrix. It has good structural stability, fission product inclusion capacity and irradiation stability, and it is one of the promising fuels in hig...

Research on Handling Method of Accelerometer Tripping of Fresh Fuel Assembly Transport Vessel

Jin Yuan, LI Weicai, Chen Jianxin, Zhou Yuemin

2022, 43(2): 96-101. doi: 10.13832/j.jnpe.2022.02.0096

Abstract(384) HTML (100) PDF(56)

Abstract:
Accelerometers mounted on fresh fuel assembly transport vessel are used to monitor the abnormal shock of fuel assemblies during transport. The tripping of the accelerometer indicates that there may be a load that may damage the fuel assembly during transport. In recent years, several tripping events...

Study on PUREX 1B Process Simulation Based on MPMS Calculation Model

Tang Jia, Yang Yu, Lin Mingzhang, Lu Yunyun, Xiong Wei, Guo Zifang, Wu Zhihao

2022, 43(2): 102-107. doi: 10.13832/j.jnpe.2022.02.0102

Abstract(486) HTML (232) PDF(315)

Abstract:
In order to simulate the U/Pu separation (1B) process of uranium plutonium redox extraction (PUREX), a mathematical model of 1B process with NH₃OH⁺-N₂H₄(HAN-HYD) as reducing extractant was established based on the framework of MPMS calculation model, using multi-stage mixer-settler as extraction eq...

Analysis on Wearing of Bearing for Main Pump in Nuclear Power Plant

Zhang Haiying, Huang Zhong, Jiao Shaoyang, Dai Hongling, Xu Xiaogang, Sheng Feng

2022, 43(2): 108-111. doi: 10.13832/j.jnpe.2022.02.0108

Abstract(406) HTML (185) PDF(68)

Abstract:
The heat release rate of the main pump in nuclear power plant continued to rise, it was found that the bearing surface of main pump was abnormal worn through inspection. In order to analysis the causes and the level of the wearing of the bearing surface, the physical properties of the material of th...

Three-dimensional Site Response Analysis Based on Equivalent-Linear Behavior of Foundation Soil

Qu Yunguang, Xu Zhengyu

2022, 43(2): 112-116. doi: 10.13832/j.jnpe.2022.02.0112

Abstract(347) HTML (123) PDF(43)

Abstract:
With the increasing complexity of nuclear power plant's siting conditions, soil-structure interaction (SSI) has become an important issue to be considered in the seismic analysis. At present, the classical free-field site response analysis uses the analysis of one-dimensional layered foundation soil...

Design and Sealing Performance Verification of Graphite Seal Assembly of CRDM

Luo Qingsong, Xu Huaijing, Tang Baoqiang, Han Jiaxin, He Canyun, Guo Yong, Liu Xin, Ma Zhigang

2022, 43(2): 117-121. doi: 10.13832/j.jnpe.2022.02.0117

Abstract(333) HTML (163) PDF(53)

Abstract:
The pressure housing of CRDM needs to be disassembled many times during the maintenance of control rod drive mechanism (CRDM) and core refueling. In order to solve the leakage of Ω sealing weld of the existing pressure housing and the problem that it can not be disassembled many times, the scheme of...

Study on Numerical Simulation of Drop Impact of Spent Fuel Transfer Equipment

Yuan Liang, Yang Jie

2022, 43(2): 122-125. doi: 10.13832/j.jnpe.2022.02.0122

Abstract(416) HTML (112) PDF(61) [Cited by] (3)

Abstract:
Analysis of drop of spent fuel transfer equipment in nuclear power plant is the most stringent condition in overall structural safety analysis. In order to solve the problem of dynamic impact analysis and evaluation of equipment drop, the finite element analysis and simulation software LS-DYNA is us...

Fragility Analysis of Nuclear Power Plant Containment under Near-site Vibration

Rong Hua, Jin Song, Gong Jinxin

2022, 43(2): 126-132. doi: 10.13832/j.jnpe.2022.02.0126

Abstract(476) HTML (204) PDF(50) [Cited by] (7)

Abstract:
Containment structure is one of the most important structures in nuclear power plant, and its seismic fragility is the focus of probabilistic seismic safety assessment of nuclear power plant structure. Combined with nonlinear finite element analysis technology and incremental dynamic analysis method...

Research on Test and Theoretical Analysis Methods on Stability of LBB Circumferential Through-Wall Crack in Austenitic Stainless Steel Pipe under Dynamic Load

He Feng, Yao Di, Wang Xinjun, Li Yilei, Bai Xiaoming, Xiong Furui

2022, 43(2): 133-137. doi: 10.13832/j.jnpe.2022.02.0133

Abstract(363) HTML (102) PDF(31) [Cited by] (1)

Abstract:
Judging whether the circumferential through-wall crack of pipe is stable is one of the criteria to judge whether the pipe meets the Leak-Before-Break (LBB) design criteria. In order to ensure the safety and reliability of LBB technology, the stability of circumferential through-wall crack of pipe un...

A Review of Research on Aerosol Hygroscopic Growth in Severe Nuclear Reactor Accidents

Wang Jinghong, Peng Wei, Yu Suyuan

2022, 43(2): 138-151. doi: 10.13832/j.jnpe.2022.02.0138

Abstract(783) HTML (210) PDF(133) [Cited by] (4)

Abstract:
The hygroscopic growth of soluble aerosols (hereinafter referred to as aerosols) is one of the key factors affecting the dynamic behavior of radioactive products in serious nuclear reactor accidents. In this paper, the theoretical model of hygroscopic growth, experimental measurement scheme and rese...

The Application of Modelica Simulation Technology in Micro Gas-Cooled Reactor

Liang Yangyang, Zhang Huimin, Wang Li, Li Yunlong, Yuan Yidan, Wang Jun, Du Shuhong

2022, 43(2): 152-159. doi: 10.13832/j.jnpe.2022.02.0152

Abstract(592) HTML (292) PDF(94) [Cited by] (4)

Abstract:
Compared with traditional large-scale nuclear power plants, the functions of micro reactor systems are closely coupled and restrict each other. The traditional professional decoupling design mode is difficult to deal with it, and a full range of system simulation is needed. A system simulation model...

Discussion on DEC of High-level Radioactive Waste Liquid Storage System from the Perspective of Multiple-failure

Lyu Dan, Yang Xinjing, Wang Shijun, Yang Zhiyi, Yang Hao, Xu Chunyan, Liu Xinhua

2022, 43(2): 160-166. doi: 10.13832/j.jnpe.2022.02.0160

Abstract(242) HTML (76) PDF(22)

Abstract:
The analysis of design extension condition (DEC) is an important part of the analysis of beyond-design-basis accidents of nuclear power facilities. Currently there is no such practice in the field of post-processing facilities. The high-level radioactive waste liquid storage system of the post-proce...

Analysis on Deterministic Behavior Design of Safety Digital Instrumentation and Control System

Wu Qiaofeng, Liu Hongchun, Sun Shiyan, Li Yu, Wang Lin, Zhang Junqi, Wu Kunren

2022, 43(2): 167-170. doi: 10.13832/j.jnpe.2022.02.0167

Abstract(276) HTML (170) PDF(26) [Cited by] (1)

Abstract:
The behavior logic of safety digital instrumentation and control (I&C) system is carried by software, but the software reliability evaluation is relatively difficult. Therefore, in order to ensure reproducibility and timeliness of safety digital I&C system, and ensure the reliability and saf...

Fault Diagnosis Method of Nuclear Gate Valve Based on Characteristic Analysis of Operation Process Variables

Liu Zhilong, Li Tongxi, Nie Changhua, Zhan Li, Tang Zhangchun, Liu Jie

2022, 43(2): 171-174. doi: 10.13832/j.jnpe.2022.02.0171

Abstract(333) HTML (111) PDF(41) [Cited by] (1)

Abstract:
Aiming at the sticking fault of nuclear gate valve, a fault diagnosis method of gate valve based on characteristic analysis of operation process variables is proposed. The operating process of gate valve opening and closing often contains fault characteristics and changing rules. Therefore, this met...

Experimental Study on Wind Load Performance of ACP100 Passive Containment Air Cooling System

Wang Hongliang, Yu Mingrui, Li Yunyi, Liu Changliang, Han Xu, Li Lujun

2022, 43(2): 175-180. doi: 10.13832/j.jnpe.2022.02.0175

Abstract(553) HTML (188) PDF(59) [Cited by] (5)

Abstract:
Environmental wind field has important influences on the operation of passive containment air cooling system (PAS) of ACP100, a modular pressurized water reactor nuclear power plant. A small-scale model of ACP100 is built in the wind tunnel platform to explore the influence of different environmenta...

Research on Thermal Safety of Intensive Spent Fuel Dry Storage Facility for Heavy Water Reactor

Xu Zhen, Ren Bing, Liu Zhan, Wang Zhe, Ye Qing, Guo Wei

2022, 43(2): 181-188. doi: 10.13832/j.jnpe.2022.02.0181

Abstract(322) HTML (84) PDF(42) [Cited by] (1)

Abstract:
In order to solve the problem that Qinshan No.3 Nuclear Power Co., Ltd. (hereinafter referred to as TQNPC) plans to extend the service life of the nuclear reactor, resulting in the increase of spent fuel and insufficient capacity of existing spent fuel dry storage module, an intensive spent fuel dry...

Design and Verification of Algorithm of Reactor Neutron Doubling Time Based on SCADE

Yu Heng, Wang Yinli, He Zhengxi, Huang Youjun, Jiang Tianzhi, Lin Chao, Yang Zhenlei, Zhang Mi

2022, 43(2): 189-193. doi: 10.13832/j.jnpe.2022.02.0189

Abstract(347) HTML (105) PDF(35) [Cited by] (1)

Abstract:
To realize the close monitoring of the nuclear fission rate of reactor during the period from fuel loading to power ascension, the correct and stable measurement of reactor neutron doubling time shall be implemented. Based on the statistical analysis of neutron fluence rate measurement, a doubling t...

Study on Nuclide Diffusion in Closed Environment of Marine Reactor with Large Break Loss of Coolant Accident

Zhao Fang, Zou Shuliang, Xu Shoulong, Xu Tao

2022, 43(2): 194-198. doi: 10.13832/j.jnpe.2022.02.0194

Abstract(403) HTML (91) PDF(58) [Cited by] (5)

Abstract:
Based on the research method of severe accident analysis program MELCOR coupled computational fluid dynamics software CFD-FLUENT, MELCOR is used to analyze the loss of coolant accident of marine reactor. The results are used as the initial conditions of CFD-FLUENT simulation experiment to study the ...

Researh on Vector Control Technology of Synchronous Reluctance Motor Control Rod Drive Mechanism

Peng Renyong, Wang Jinxin, Qing Xianguo, Liu Yiyi, Zhang Jianjian, Liu Yanan

2022, 43(2): 199-203. doi: 10.13832/j.jnpe.2022.02.0199

Abstract(479) HTML (123) PDF(55) [Cited by] (9)

Abstract:
Aiming at the problem that the synchronous reluctance motor control rod drive mechanism (CRDM) is highly electromagnetically coupled and it is difficult to effectively adjust the output torque linearly, this paper studies the vector control technology of the synchronous reluctance motor CRDM. The ma...

Numerical Study on Hydrogen Flow Distribution Characteristics in Small-Scale Space

Liu Hanchen, Wu Xinzhuang, Xiang Wenjuan, Liu Jie, Wu Huiping

2022, 43(2): 204-211. doi: 10.13832/j.jnpe.2022.02.0204

Abstract(395) HTML (216) PDF(39) [Cited by] (2)

Abstract:
Different from the large space of nuclear power plant containment, in small-scale space such as containment compartment and advanced small reactor, the flow of mixed gas of hydrogen and steam is limited by the wall, and the gas flow cannot fully develop, which may lead to the accumulation of hydroge...

Numerical Study on the Flow Characteristics of the Parallel Main Pumps of the Vertical Canned Motor

Zhou Xingzhu, Song Yu, Yin Junlian, Wang Dezhong, Xia Shuan, Feng Lei

2022, 43(2): 212-218. doi: 10.13832/j.jnpe.2022.02.0212

Abstract(324) HTML (87) PDF(73)

Abstract:
Taking the reactor main coolant circulation pump (referred to as the main pump) of the vertical canned motor with a scale factor of 1:4 as the research object, two kinds of geometric models with two counter-rotating parallel main pumps (model 1) and two co-rotating parallel main pumps (model 2) are ...

Fatigue Reliability Test and Evaluation of Main Pump Spindle of Nuclear Power Plant Reactor

Zhang Jianxin, Gu Jipin, Chen Shuming, Pu Enshan, Guo Xiaoxian, Li Hailiang, Fang Jinghui

2022, 43(2): 219-225. doi: 10.13832/j.jnpe.2022.02.0219

Abstract(373) HTML (89) PDF(53) [Cited by] (3)

Abstract:
In order to obtain the fatigue reliability data of the main spindle material of the nuclear power plant reactor under a given confidence level and different reliability levels, a simulation spindle with the same inner and outer diameter dimensions and processing technology as the product spindle was...

Test and Verification for Digital Nuclear Instrumentation System Prototype on Reactor

Wang Yinli, He Zhengxi, Bao Chao, Gao Zhiyu, Wu Wenchao, Luo Tingfang, Yu Heng, Luo Wei

2022, 43(2): 226-231. doi: 10.13832/j.jnpe.2022.02.0226

Abstract(348) HTML (111) PDF(38) [Cited by] (2)

Abstract:
In view of the current situation of nuclear instrumentation system equipment mainly relying on import in the current domestic nuclear power plant, a set of digital nuclear instrumentation system prototype was designed and developed. The system prototype mainly includes neutron detector assembly, sig...

Test Method on Starting Drag Torque of the Thrust Bearing of the Reactor Main Pump in Nuclear Power Plant

Zhang Jianxin, Gu Jipin, Chen Shuming, Wang Mingzheng, Liu Xiaojun

2022, 43(2): 232-236. doi: 10.13832/j.jnpe.2022.02.0232

Abstract(328) HTML (241) PDF(32) [Cited by] (2)

Abstract:
In order to obtain the ultimate starting drag torque of the thrust bearing of the reactor main pump in nuclear power plant during its lifetime, and ensure that the auxiliary motor performing the accident residual heat removal function can start the main pump under extreme conditions, a test method o...

Calibration and Verification of AFD Used for PWR Nuclear Power Plant Control System with the Mechanical Shim

Wei Guangjun

2022, 43(2): 237-241. doi: 10.13832/j.jnpe.2022.02.0237

Abstract(359) HTML (121) PDF(46)

Abstract:
The rod shadow effect of the mechanical shim of PWR nuclear power plant causes the axial flux deviation (AFD) indicated by the protection system to become the weighted value of peripheral components rather than the average value of the core. Therefore, the control system based on the average AFD of ...

Research on Pressurizer Level Measurement Based on Sub-regional Density Compensation

Liu Zhilong, Zhan Li, Nie Changhua, Liu Jie, Tang Zhangchun, Li Tongxi

2022, 43(2): 242-245. doi: 10.13832/j.jnpe.2022.02.0242

Abstract(367) HTML (85) PDF(44) [Cited by] (4)

Abstract:
In view of the large temperature difference between the upper part and the bottom of the pressurizer when the electric heating element heats the pressurizer, which leads to the large measurement error of the liquid level measurement of the traditional pressurizer differential pressure method, a pres...

Condition Prediction of Reactor Coolant Pump in Nuclear Power Plants based on the Combination of ARIMA and LSTM

Zhu Shaomin, Xia Hong, Lyu Xinzhi, Lu Chuan, Zhang Jiyu, Wang Zhichao, Yin Wenzhe

2022, 43(2): 246-253. doi: 10.13832/j.jnpe.2022.02.0246

Abstract(592) HTML (147) PDF(84) [Cited by] (27)

Abstract:
To monitor and track the operation process and improve the early warning of the reactor coolant pump (RCP) in nuclear power plants (NPPs), a hybrid RCP condition prediction approach based on autoregressive integrated moving average (ARIMA) model and long short-term memory (LSTM) neural network is pr...

Ultrasonic Inspection Process Design for CRDM Pressure Housing Weld in Nuclear Power Plant

Tang Jianbang, Yu Zhe, Wang Weiqiang, Sun Jiawei, Lyu Tianming

2022, 43(2): 254-258. doi: 10.13832/j.jnpe.2022.02.0254

Abstract(432) HTML (137) PDF(49) [Cited by] (1)

Abstract:
The pressure housing of the control rod drive mechanism (CRDM) belongs to the main circuit of the nuclear power plant, and its joint weld is the weak link of the pressure boundary of the whole radioactive circuit. Its safety and reliability directly affect the safe operation state of the reactor. In...

2022 Vol. 43, No. 2