核动力工程

Study on Dynamic Characteristics of Xi’an Pulsed Reactor under Non-Pulse Transient Condition

Zhang Liang, Yuan Jianxin, Zhao Wei, Wang Baosheng, Zhang Qiang, Zhu Guangning, Jiang Xinbiao, Chen Lixin

2020, 41(5): 1-7.

Abstract(554) PDF(401) [Cited by] (4)

Abstract:
The instrument and control system of Xi’an Pulsed Reactor is under the digitized reconstruction, and the code for analyzing the dynamic characteristics of non-pulse transient conditions is essential to provide the real-time variety parameters such as power and fuel temperature. Based on the previous...

Correction of CPPF Factor in QINSHAN CANDU Reactor

Wang Jun

2020, 41(5): 8-11.

Abstract(328)

Abstract:
With the aging of the unit, the core fuel cooling performance of QINSHAN CANDU reactor decreases. To ensure the safe operation of the reactor, it needs to do regular analysis of ROP TSP. The analysis result is applied by correcting the CPPF which is used to calibrate ROP detectors to continuously tr...

A Brief Method for Derivation of Streaming Term Expression for Neutron Transport Equation in Spherical Geometry Coordinate System

Yang Benlin, Chen Shi, Wan Like, Wang Ziguan, Zhao Xinyue, Hu Longxiang, Yang Song

2020, 41(5): 12-14.

Abstract(194)

Abstract:
An ingenious method has been discussed to provide a brief way for the derivation of the streaming term expression in the spherical geometry coordinate system. Instead of finding the differential relationship between the dihedral angle and the neutron transport distance, finding the geometry relation...

Development and Validation of Non LOCA Thermal Hydraulics and Three Dimensional Neutronics Coupling Code GINKGO/COCO

He Qingyun, Luo Jingyi, Chen Jun, Ren Zhihao, Peng Sitao, Zhou Zhou, Shan Jianqiang

2020, 41(5): 15-19.

Abstract(307) [Cited by] (5)

Abstract:
For the needs of more detailed and accurate core modeling and thermal hydraulic analysis, based on the Non LOCA thermal hydraulic analysis code GINKGO developed on self-reliance and three-dimensional core physical code COCO, GINKGO/COCO coupling code has been developed by using Dynamic Link Library ...

Three-Dimensional Flow Field Calculation of Pressure Vessel in Small Pressurized Water Reactor

Wang Kun, Dong Xiucheng, Liu Haipeng, Zhang Xin, Yuan Jiangtao

2020, 41(5): 20-23.

Abstract(377) [Cited by] (6)

Abstract:
In the reactor safety analysis process, it is important to obtain an accurate flow field inside the pressure vessel. Taking the small pressurized water reactor as the research object, the computational fluid dynamics (CFD) method was used to calculate and analyze the internal flow field of the react...

Analysis of Coupled Flow and Heat Transfer in Primary and Secondary Sides of Helical Coil Once-Through Tube Steam Generator

Liu Fayu, Zhang Xiaoying, Chen Jiayue, Chen Hu, ong

2020, 41(5): 24-29.

Abstract(504) [Cited by] (7)

Abstract:
To study the flow and heat transfer characteristics of the primary and secondary sides of the helical coil once-through tube steam generator (HCOTSG) under steady state conditions, taking HCOTSG of International Reactor Innovative and Secure (IRIS) as the research object, a primary and secondary sid...

Experimental Study on Enhancement of Pool Boiling Critical Heat Flux on SA508 Carbon Steel with Cold Spray Coating

Qin Fei, Liu Hanzhou, Hu Lian, Chen Deqi, Zhong Dawen

2020, 41(5): 30-34.

Abstract(274) [Cited by] (3)

Abstract:
In this study, a coating technique known as “cold spray” was developed to form a micro-porous coating on the SA508 Gr3 carbon steel heater, which was the prototypical material for the reactor vessel. On the rotatable experimental bench, downward facing pool boiling CHF experiments using a plate test...

Research on Method of Generating Monte Carlo Multigroup Library for Fast Reactor

Zhu Shuaitao, Ma Xubo, Xu Qian, Cao Bo, Chen Yixue

2020, 41(5): 35-39.

Abstract(236) [Cited by] (2)

Abstract:
Based on the discrete angle method, a Monte Carlo multi-group cross section generation program MGXSMC was developed. This program can read the cross section data from an input file or read the cross section from a library in a specified format to generate the multi-group cross section for MCNP or RM...

Study on Thermal Conductivity Model of Dispersion Fuel

Ren Qisen, Liao Yehong, Chen Mengteng, Zhang Yongdong, Xie Yiran, Liu Tong, Liu Weiqiang

2020, 41(5): 40-43.

Abstract(391) [Cited by] (1)

Abstract:
The effective thermal conductivity (ETC) of dispersion fuels plays an important role in nuclear reactor safety analysis and fuel performance evaluation. In this study, based on the theory of porous body, considering the relativity of dispersion particle distributions, an ETC model of dispersion fuel...

Design and Verification of Relaxation Test Device for Hold-down Leaf Spring

Luo Wenguang, Wang Yajun, Wang Wanjin, Wu Rui, Zhang Xianmeng

2020, 41(5): 44-48.

Abstract(254)

Abstract:
A relaxation test device for the hold-down leaf spring was designed considering the characteristics of the hold-down leaf spring for AFA3G fuel assembly. The reliability of the device is verified by a series of simulation tests. Finally, four groups of different recycled AFA3G fuel assemblies were s...

Prediction Method for Long-Term Smt of 316 Stainless Steel

Li Changxiang, Mo Jintao, Duan Chunhui

2020, 41(5): 49-52.

Abstract(356) [Cited by] (1)

Abstract:
General primary membrane stress (S_mt ) of material is the important parameter used for the mechanical analysis of high temperature reactor design, and the data of Smt at 300000 hours in the ASME code and RCC-MR code cannot meet the needs of long-life nuclear reactor design. Based on the data of allo...

Creep Crack Growth Behavior of N18 Zircaloy Thin-Walled Tubes with Double Edged Axial-Notched Cracks

Li Zhihao, Bao Chen, Wang Bo, Liu Xiaokun

2020, 41(5): 53-59.

Abstract(241) PDF(135) [Cited by] (1)

Abstract:
Through the design of double edged axial-notched tube (DEAT) specimen and fixture, the expression of C* integral of DEAT specimen was obtained based on the principles of energy equivalent and load separation. A test method for the creep crack growth rate of thin-walled tubes with axial cracks has be...

Analysis of Fatigue Time Limit Aging of Reactor Pressure Vessels

Shao Xuejiao, Xie Hai, Zhang Liping, YangYu, Du Juan, Tian Jun, Kuang Linyuan, Gao Shiqing

2020, 41(5): 60-64.

Abstract(384) [Cited by] (2)

Abstract:
Based on two methods of evaluating the influence of the coolant environment on the fatigue life of equipment proposed by U.S.NRC in the management guideline RG1.207, the effects on different environmental fatigue correction factor（F_en） expressions and boundary conditions were compared between the NU...

Study on Nonlinear Beam Model of PWR Fuel Assembly

Gu Chenglong, Yang Yuying, Guo Yan

2020, 41(5): 65-69.

Abstract(301) [Cited by] (3)

Abstract:
In order to describe the nonlinear characteristics of the fuel assembly, a lateral beam model of assembly is built by the finite element method. The beam model is embedded with the hysteretic model to simulate the nonlinear effect occurred at lateral deformation. The calculation shows that the bendi...

Effect of Main Bolt Break on Seal Performance, Stress and Fatigue of Reactor Pressure Vessel

Zheng Liangang, Bai Xiaoming, Shi Kaikai, Du Juan

2020, 41(5): 70-73.

Abstract(218) [Cited by] (5)

Abstract:
Base on the mechanics theory and numerical simulation technique, a method to analyze the effect of the main bolt break on the stress, fatigue and seal is studied in this paper, and is adopted to evaluate and analyze the fracture influence of main bolt. The results show that this method is applicable...

Study of Application of Level 2 Probabilistic Safety Analysis in Severe Accident Management

Zhang Jiajia, Ni Man, Xiao Jun, Gong Yu, Qian Hongtao

2020, 41(5): 74-78.

Abstract(216) [Cited by] (1)

Abstract:
Level 2 Probabilistic Safety Analysis (PSA) can be used to quantitatively assess the risk of severe accident and is a good tool to evaluate the severe accident management. By studying the general method and procedure for the application of level 2 PSA in severe accident management, taking an improve...

A Brief Introduction of Design Optimization for AP1000 Condensation Return

Ma Baisong, Zhuang Yaping, Qie Weiqing

2020, 41(5): 79-83.

Abstract(271)

Abstract:
During any non-LOCA event, the average temperature of the reactor cooling system (RCS) should be decreased to 215.6℃within 36 hours in AP1000 nuclear power plant. However, this goal cannot be achieved because the rate of condensate return was far lower than that expected. Through the analysis and ve...

Simulated Analysis of Pressure Control and Overpressure Problem of Reactor Cold Start-up with Nuclear Heating

Qing Xianguo, Xiao Kai, Huang Ke, Chen Guanyu, Li Yiliang, Chen Zhi

2020, 41(5): 84-88.

Abstract(200) [Cited by] (1)

Abstract:
Based on the control requirements of the reactor cold start-up process with nuclear heating, the automatic pressure control of reactor cold start-up with nuclear heating is studied in this paper, and the method for the system pressure automatic control in the process of cold start-up with nuclear he...

Analysis of Reactor Power Supply System on Floating Nuclear Power Plants

Chen Qiang, Guo Xiang, Zhu Chenghua

2020, 41(5): 89-93.

Abstract(205) [Cited by] (2)

Abstract:
The safety of the floating nuclear power plant is closely related to the merits of the reactor power supply system. In order to improve the safety factor of floating nuclear power plant, it is necessary to analyze the reactor power supply system. In this paper, the configuration of the reactor power...

Research on Monitoring of Reactor Cavity Cooling Status for NPPs

He Peng, Chen Jing, Li Xiaofen, He Zhengxi, Zhu Jialiang, Xu Tao, Li Hongxia

2020, 41(5): 94-98.

Abstract(277) [Cited by] (3)

Abstract:
In order to justify the accident progresses of the cavity and the implementation effect of the cavity injection strategy that is activated under the severe accident condition, the evolution sequence of several cavity physical property parameters of different injection water velocity under severe acc...

Data Reliability Analysis during Containment Leakage Test Based on Statistical Software R

Shen Dongming, Cai Jiantao, He Rui, Huang Xiaoming

2020, 41(5): 99-103.

Abstract(262) [Cited by] (10)

Abstract:
The most important part in the calculation of the containment leakage is to perform the linear regression on time for a series of data measured at different times. The significance test of the regression and residual analysis are the substantial means to evaluate the test results. This paper analyze...

Application of Requirement Modeling in Nuclear Requirements Analysis

Zhu Junzhi, Yang Jue, Wan Lei, Cui Jun, Liu Yongkang, Liu Qingsong

2020, 41(5): 104-109.

Abstract(361) [Cited by] (12)

Abstract:
At present, it is increasingly difficult for the nuclear power design products to meet user expectations for there is no effective method to collect and manage the requirement information, and it is unable to perform early the requirement verification and to change the requirements in the research a...

Simulation Analysis of Boron Concentration Difference between Primary Circuit and Pressurizer of ACPR1000

Jiang Xialan, Li Hui, Qin Zhiguo

2020, 41(5): 110-115.

Abstract(200) [Cited by] (3)

Abstract:
In order to solve the problem of excessively large boron concentration difference between the primary circuit and the pressurizer in the normal dilution and boration process of Improved Chinese Pressurized Water Reactor (ACPR1000), and to prevent the accidental violation of the operating technical s...

Numerical Simulation of Airborne Radioactive Exclusion in Spent Fuel Storage Tank Ventilation Mode

Gui Ting, Xian Chunmei, Fang Zhen, Dong Changqing, An Jing, He Meikui

2020, 41(5): 116-121.

Abstract(207) [Cited by] (1)

Abstract:
In order to protect the workers in the spent fuel storage compartment from internal radiation, the airborne radioactive concentration in the spent fuel storage compartment needs to be controlled. Exhaustion of the airborne radiation is mainly realized through the ventilation system. According to the...

Research and Risk Analysis of Partial Cooldown Test of EPR Nuclear Power Plant

Zeng Huan, Zhao Xin, Duan Shengzhi

2020, 41(5): 122-126.

Abstract(140)

Abstract:
As the first reactor test of the European Pressurized Water Reactor (EPR) Units, partial cooldown test would cause enormous thermal shocks in the primary and secondary coolant circuits, which are not allowed to happen more than 15 times during the service life of a nuclear power station. In order to...

Design and Test Study on a High Temperature Molten Salt Loop

Kong Xiangbo, Wang Naxiu, Lin Liangcheng, Lu Huiju, Fu Yuan, Wang Xiao

2020, 41(5): 127-131.

Abstract(340) [Cited by] (3)

Abstract:
The Thorium Molten Salt Reactor (TMSR) project plans to construct a 2 MWt liquid fuel molten salt reactor. After successful R&D of the proto-type equipment like pump, heat exchanger and freeze valve, construction of a high temperature fluoride loop is designed and constructed to test them. The r...

Research on Condition Monitoring Technology for Nuclear Power Plant Equipment Based on Kernel Principal Component Analysis

Wu Tianhao, Liu Tao, Shi Haining, Zhang Tao, Tang Tang

2020, 41(5): 132-137.

Abstract(491) [Cited by] (28)

Abstract:
In order to solve the limitations of the traditional monitoring methods for nuclear power plants, this paper proposes to introduce Kernel Principal Component Analysis (KPCA) into the online monitoring field of nuclear power plant equipment, and design the monitoring method and online monitoring stra...

Research on Lower C Seal Cutting Technology for AP1000 Large Canned Pumps

Yang Zhiye, Li Tao, Wang Binyuan, Li Song, Zhao Mingshen

2020, 41(5): 138-141.

Abstract(193)

Abstract:
The canned motor reactor coolant pumps are used in the Unit 1and Unit 2 in Sanmen Nuclear Power Plant. The lower C seal has to be cut apart by the special way and tool during the pump disassembly. According to the structure of the pump, the functional requirements of the cutting scheme are defined a...

Model Building Approach for Nuclear Power Operation Procedure Based on Ontology

Xu Yu, Huang Yuanyuan, Xiong Lihong, Leng Shan, Zhu Xiaoliang

2020, 41(5): 142-145.

Abstract(607) [Cited by] (4)

Abstract:
Nuclear power operating procedures are the operational procedures that must be followed for the safe operation of nuclear power plants. Having a lot of texts, the operation procedures are inconvenient for users to browse. However, the current electronic procedures have problems such as lack of simpl...

Study on Measures to Prevent Pressurizer Overfill During Loss of Normal Feedwater for AP1000 NPP

Ma Baisong, Guo Hongen

2020, 41(5): 146-149.

Abstract(226)

Abstract:
In an accident of loss of feedwater in an AP1000 plant, the pressurizer was filled with water for a series of improper operations, and the safety valves may not be qualified to re-close following multiple cycles of opening, which is not acceptable in Condition Ⅱ events. The paper analyzes the causes...

Research on Impact Factors of Depth Quantification in Eddy Current Test for Thimble Tubes Wear of Nuclear Power Plants

Ma Qiang, Chen Cheng, Li Pingren, Kong Yuying, Ding Boyuan, Zhao Hongqiang, Yang Hongbo

2020, 41(5): 150-154.

Abstract(161) [Cited by] (2)

Abstract:
The result of the eddy current inspection for the thimble tube of a nuclear reactor neutron flux measurement system is the reference for the nuclear power plant to take maintenance actions. Based on the currently-used eddy current inspection method, the effect of the parameter variation of the tube ...

Analysis of Effect of Cone Angle on Performance of Vapor-Anode AMTEC Conical Evaporator

Zhu Lei, Jiang Xinbiao, Li Huaqi, Chen Sen, Tian Xiaoyan, Qiu Suizheng

2020, 41(5): 155-161.

Abstract(195) [Cited by] (2)

Abstract:
To analyze the effect of the cone angle on the performance of the evaporator in vapor-anode Alkali Metal Thermal to Electric Converter (AMTEC), a steady-state two-dimensional thermal hydraulic model for liquid-return wick and conical evaporator was developed. The effects of different cone angles on ...

Effects of Internal Leakage Pathwayon Main Control Room Habitable Dose in Accident Condition of Nuclear Power Plants

Wang Qi, Wang Kai, Wang Jianhua

2020, 41(5): 162-167.

Abstract(285) [Cited by] (3)

Abstract:
In order to fully consider the effects of internal leakage pathway on the radiation safety of the operators in the main control room habitable area during potential accident condition, a radioactive material migration model is established and compartment model is modified considering internal leakag...

Research and Application of Calibration Test Method for PTR Liquid Level Gauge in EPR Units

Yuan Meichun, Sun Dongmei, Nan Xiayu

2020, 41(5): 168-172.

Abstract(215)

Abstract:
Aiming at the calibration test of the fuel pool cooling and purification system (PTR) liquid level gauge in EPR units, a test method based on the principle of the connector was proposed, and a special test device was designed. The results show that the new test method can shorten the construction pe...

Requirements for Monte Carlo Method during Radiation Shielding Optimization Design of Advanced PWR Nuclear Power Plants

Lyu Weifeng, Xiong Jun, Liu Jie, Tang Shaohua

2020, 41(5): 173-177.

Abstract(233) [Cited by] (2)

Abstract:
Based on the analysis of the problems resulting from the tool limitations during the radiation shielding design of M310 nuclear power plant and the requirements presented by the radiation shielding design of HPR1000 nuclear power plant, the requirements for Monte Carlo(MC) method during the optimiza...

Research of Multi-Objective Optimization Method of Nuclear Reactor Radiation Shielding

Zhang Zehuan, Song Yingming, Lu Chuan, Tang Songqian, Xiao Feng, Lyu Huanwen, Yang Junyun, Mao Jie

2020, 41(5): 178-184.

Abstract(409) [Cited by] (12)

Abstract:
To overcome the disadvantages in the efficiency and applicability of the traditional shielding optimization method based on the Monte Carlo method, in this paper, we studied the reactor radiation shielding optimization method by the non-dominated sorting genetic algorithm (NSGA-Ⅱ) based on the eliti...

Design Improvement of Reactor Vessel Shielding Component

Zhuang Yaping, Ma Baisong

2020, 41(5): 185-188.

Abstract(189) [Cited by] (4)

Abstract:
During the hot functional test of one NPP, the neutron shielding material was heated and released from the reactor vessel shielding blocks. The structure and layout of the block were redesigned, and B4C was adopted as the neutron shielding material. This paper analyzes the improved design scheme in ...

Overall Design and Verification of ACP100S Floating Nuclear Power Plant

Li Qing, Song Danrong, Zeng Wei, Chen Zhang, Liu Jia, Wang Donghui, Xiao Renjie

2020, 41(5): 189-192.

Abstract(1003) [Cited by] (16)

Abstract:
Floating nuclear power plant (FNPP) is a movable nuclear power plant built on the floating platform. FNPP belongs to the category of small reactor based on the classification of electric power. It can be used for power generation, desalination, heating, and can satisfy the special needs of regional ...

Research on Debris In-Core Cooling and Retention Characteristics

Song Jian, Xiang Qingan, Deng Jian, Yu Hongxing, Du Juan, Bi Jinsheng

2020, 41(5): 193-196.

Abstract(260) [Cited by] (4)

Abstract:
The severe accident analysis model of the small modular reactor ACP100 is built using MELCOR code, and the core heat removed process through the barrel and wall of reactor pressure vessel (RPV) is analyzed by the cavity injection system (CIS). The collapse behavior of the fuel assemblies is estimate...

In-pile Performance Simulation and Structure Design of Fully Ceramics Microencapsulated Fuel

Zhou Yi, Liu Shichao, Chen Ping, Li Yuanming, Xin Yong, Liu Zhenhai, Zhang Lin, Gu Mingfei, Zhao Yanli, Le Yunlin

2020, 41(5): 197-200.

Abstract(229) [Cited by] (3)

Abstract:
The thermal mechanical performance of the fully ceramics microencapsulated fuel (FCM) with different non-fuel part size was simulated using two-dimensional characteristic unit. When the fissile loading meet the requirements of the reactor core, the stress condition of SiC matrix and SiC layers were ...

2020 Vol. 41, No. 5