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2018 Vol. 39, No. 2

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Benchmark Validation of Transport Calculation for Advanced Neutronics Lattice Code KYLIN-2
Huang Shien, Chai Xiaoming, Li Xiangyang, Chen Zhang, Yin Qiang, Lu Wei, Li Qing
2018, 39(2): 1-4. doi: 10.13832/j.jnpe.2018.02.0001
Abstract(18)
Abstract:
Several benchmark problems were validated for the transport calculation module of the advanced neutronics lattice code KYLIN-2,including homogeneous medium problem,Postma problem,BWR cell problem and C5G7 assembly problem.The validation results show that:（1） the ray spacing has limited effect on the...
Experimental Research on Effect of Nuclear Feedback on Flow Instability in Parallel Channels
Xie Feng, Xi Zhao, Yang Zumao
2018, 39(2): 5-9. doi: 10.13832/j.jnpe.2018.02.0005
Abstract(16)
Abstract:
As the existence of nuclear feedback in nuclear reactors,coupled nuclear-thermo-hydraulic instability will occur in the reactor.Coupled-nuclear-thermal-hydraulic instabilities were investigated employing experimental and theoretical methods in this paper.The experimental research on coupled nuclear-...
Research on Simulation Criterion of Steam Generator Natrual Circulation Experiment
Li Haibo, Zhao Erlei, Zan Yuanfeng, Zhuo Wenbin
2018, 39(2): 10-13. doi: 10.13832/j.jnpe.2018.00.0010
Abstract(14)
Abstract:
The research of the simulation criterion about natural circulation was introduced based on the hierarchical two-tired scaling（H2 TS） methodology.According the physical process of experiment target,the phenomena identification and ranking tables（PIRT） was established.Steam generator（SG） natural circu...
Analyses on Flow Instability of Secondary-Side Passive Residual Heat Removal System for Advanced Nuclear Heating Reactor
Liang Weihong, Xie Heng
2018, 39(2): 14-19. doi: 10.13832/j.jnpe.2018.02.0014
Abstract(28)
Abstract:
An advanced nuclear heating reactor model is provided to study the flow instability of secondary-side passive residual heat removal system.The RELAP5 program is used to simulate the transient process of PRHRS after accident.The results show that the origin for the condensate flow instability is the ...
Study on Transient Flow Characteristics of Rotor-Stator Cascade in Nuclear Main Pump
Li Yibin, Zhang Mei, Zhu Yuelong, Li Zhenggui
2018, 39(2): 20-26. doi: 10.13832/j.jnpe.2018.02.0020
Abstract(28)
Abstract:
Based on DES vortex model and block structured grid technology,the adaptability and accuracy of DES model for the CAP1400 reactor coolant pump prototype and 0.4 scale model is verified through CFD numerical analysis and experimental investigation of exterior performances.Compared with the loading di...
Effect of Fission Gas Bubble Size on Internal Characteristics of Dispersion Fuel Particles
Chen Hongsheng, Long Chongsheng, Xiao Hongxing, Wei Tianguo, Gao Wen, Zhao Yi
2018, 39(2): 27-31. doi: 10.13832/j.jnpe.2018.02.0027
Abstract(23)
Abstract:
On the basis of the fuel particles cracking model of bubble size homogeneity,the cracking model with bubble size heterogeneity was proposed according to stress equilibrium condition that the equivalent effective stresses applied by the gas bubble are identical.The effect of the fission gas bubble si...
Effect of Re-Working on Secondary Phases in Modified N18 Sheet
Chen Boquan, Qiu Shaoyu, Peng Qian, Dai Xun, Wang Pengfei, Liu Hong, Wei Tianguo
2018, 39(2): 32-36. doi: 10.13832/j.jnpe.2018.02.0032
Abstract(22)
Abstract:
A combination of various research methods including DSC,SEM and TEM was applied to systematically investigate the effect of re-working on the secondary phases.The results showed that the hot-rolling was conducted in the dual-phase region.In the process of cold-working and followed annealing,β-Zr whi...
Study on Optimization Design of Evaporator System for Radioactive Waste Disposal Center
Li Ming, Ma Xingjun, Chen Li, Gao Feng, Li Binglin, Sha Sha
2018, 39(2): 37-41. doi: 10.13832/j.jnpe.2018.02.0037
Abstract(29)
Abstract:
In this paper the evaporator system of the radioactive waste disposal center was introduced in brief firstly.Considering the problems of the traditional process such as it is unable to meet the handling capacity for the radioactive wastewater of the newly connected nuclear facilities,the design for ...
Research and Development of RPV Stud Holes Lubricating Equipment
Ren He, Weng Songfeng, Tan Hongwei, Huang Hui, Huang Xindong
2018, 39(2): 42-45. doi: 10.13832/j.jnpe.2018.02.0042
Abstract(21)
Abstract:
Aiming at the existing problem in the lubrication of reactor pressure vessel（RPV）stud holes,this paper puts forward an effective method which adopts array bristles and small caliber oil outlet to brush anti-seize lubricant on the surface of stud holes,and gives out the detailed design of RPV stud ho...
Design of a Model Predictive Controller for Load Tracking in Pressurized Water Reactor
Wang Guoxu, Wu Jie, Chen Zhijie, Zeng Bifan, Xu Zhibin, Ma Xiaoqian
2018, 39(2): 46-49. doi: 10.13832/j.jnpe.2018.02.0046
Abstract(18)
Abstract:
The state space model of the reactor core is modeled by expanding the differential equation model with Taylor series,and a model predictive controller is designed to achieve the automatic control of load tracking in pressurized water reactor（PWR） nuclear power plants.Quadratic programming（QP） is int...
Comparison of Methods for Artificial Earthquake Time History Matching Floor Response Spectra in Nuclear Power Plants
Sun Yugang, Chu Meng, Ding Zhenkun, Ge Honghui
2018, 39(2): 50-54. doi: 10.13832/j.jnpe.2018.02.0050
Abstract(25)
Abstract:
In order to satisfy the requirements of seismic analysis for the structure,system and components（SSCs） in Nuclear Power Plants（NPPs）,two typical methods（modify the Fourier amplitude spectra in frequency domain and superimposing wavelet in time domain） were used to generate the artificial time histor...
Research on Criteria of Design by Analysis of NB-3200 in ASME B＆PV Code Section Ⅲ
Gao YongJian, He YinBiao, Cao Ming, Yao Weida
2018, 39(2): 55-61. doi: 10.13832/j.jnpe.2018.02.0055
Abstract(18)
Abstract:
The ASME B＆PV Code Section Ⅲ– Division 1 – Subsection NB Design-by-Analysis（NB-3200） approach provides stress limits for the design of nuclear components.In the present paper,the failure modes in nuclear components are identified.The characteristic of Design-by-Analysis was presented.Moreover,the te...
Software Verification and Validation of Safety DCS System
Zhang Lei, Zhou Liang, Sun Yongbin, Liang Zhongqi
2018, 39(2): 62-65. doi: 10.13832/j.jnpe.2018.01.0062
Abstract(19)
Abstract:
According to the nuclear safety regulations,verification and validation（V＆V）activities should be carried out during the software development process.Based on the research of the regulations and standards and engineering experience,a suitable safety class DCS application software V＆V system is establ...
Preliminary Shielding Calculation and Analysis of OTSG for ACPR50S Small Modular Reactor
Sheng Xuanyu, Ling Chen
2018, 39(2): 66-71. doi: 10.13832/j.jnpe.2018.01.0066
Abstract(17)
Abstract:
Once-through steam generator（OTSG） is one of the main equipments of ACPR50 S small modular reactor.High flux of neutrons in OTSG will cause steam activation,increase the dose rate at main steam outlet,thus polluting entire secondary loop of the reactor.We carried out a preliminary analysis of 16N ge...
An Investigation of Trend Filter Based Collaborative Piecewise Linearization Method for Transient Curve
Bai Xiaoming, Zheng Liangang, Ai Honglei, Wang Xinju, Lu Xifeng, Zhang Rui
2018, 39(2): 72-75. doi: 10.13832/j.jnpe.2018.02.0072
Abstract(15)
Abstract:
In nuclear power plants,both the calculated design transient and monitored real transient are constructed by massive data.Due to the high computational cost,the raw transient data can not be used for fatigue analysis directly.Therefore,how to piecewise linearize the raw data is a key issue for fatig...
Numerical Simulation Research on Equivalent Thermal Conductivity Coefficient of M3 Fuel
Li Yuanming, Tang Changbing, Yu Hongxing, Chen Ping, Zhou Yi
2018, 39(2): 76-79. doi: 10.13832/j.jnpe.2018.02.0076
Abstract(19)
Abstract:
In order to research the equivalent thermal conductivity coefficient of Zirconium based dispersion micro encapsulated fuel（M3 fuel）,in this research,assuming the TRISO（three layer isotropic） particle embedded in the zirconium with body-centered cubic distribution,the simulation method for the equiva...
Analysis of Radioactive Effect in Molten Salt Reactor
Wang Dan, Wang Yahui, Ma Yu
2018, 39(2): 80-85. doi: 10.13832/j.jnpe.2018.02.0080
Abstract(20)
Abstract:
Due to the flow characteristics of molten salt in molten salt reactor（MSR）,the internal physical processes of this reactor show strongly coupling characteristic.This paper studied the coupled neutronics-thermo-hydrodynamic and radiative heat transfer processes in MSR by using finite element method a...
Study on Three-Dimensional Neutron Diffusion Heterogeneous Variational Nodal Method and Flat Source Acceleration Method
Zhang Tengfei, Wu Hongchun, Cao Liangzhi, Li Yunzhao, Xiong Jinbiao, Liu Xiaojing
2018, 39(2): 86-89. doi: 10.13832/j.jnpe.2018.02.0086
Abstract(23)
Abstract:
As a foundation of three-dimensional whole-core heterogeneous transport calculation,a 3D heterogeneous variational nodal method is proposed,which employs iso-parametric finite elements within each node and piecewise constants along nodal interfaces to explicitly describe the pin-resolved heterogeneo...
Numerical Investigation of Effect of Heat Transfer Model on Hydrogen Flow in Local Compartment
Wang Di, Tong Lili, Cao Xuewu, Zou Zhiqiang, Chen Shu
2018, 39(2): 90-95. doi: 10.13832/j.jnpe.2018.02.0090
Abstract(25)
Abstract:
During the severe accident process,the hydrogen and the steam are released into the containment and transferred to other local compartments through narrow pipes.Heat transfer between gas mixture and wall has influence on hydrogen distribution.In this paper,CFD method is adopted to build the analysis...
Inverse Problem of One-Dimensional Unsteady Heat Conduction Deducing Temperature Fluctuation of Inner Wall of Pipe
Xiong Ping, Ai Honglei, Lu Tao, Wang Xinjun
2018, 39(2): 96-100. doi: 10.13832/j.jnpe.2018.00.0096
Abstract(20)
Abstract:
A one-dimensional transient thermal conduction inverse problem mathematical model is constructed based on the finite difference method（FDM） and the optimization algorithm based on the conjugate gradient method（CGM）.The general calculation program is written in C language,and the temperature fluctuat...
Research on Rod Withdrawal Accident Analysis Method Based on RELAP5/MOD3 Variable Power Distribution
Zhang Yong, Li Songwei, Wang Wei
2018, 39(2): 101-103. doi: 10.13832/j.jnpe.2018.02.0101
Abstract(18)
Abstract:
In this paper,we use the rod-power distribution obtained by the nuclear design and the 202 tables and 205 cards of RELAP5/MOD3 to realize the change of the relative power share of each axial node in the heat of the RELAP5/MOD3.The "variable power distribution" analysis method is successfully realize...
Effect of Thermal Aging on Impact Properties in 16MND5 Forgings for Nuclear Reactor Pressure Vessel
Xing Ruisi, Chen Xu, Xie Guofu, Yang Zhihai
2018, 39(2): 104-108. doi: 10.13832/j.jnpe.2018.02.0104
Abstract(28)
Abstract:
Effects of thermal aging on Charpy V-notch impact properties in 16MND5 steel was investigated.The ductile to brittle transition temperature increased with the prolongation of thermal aging duration under impact test.According to the fracture analysis of the Charpy impact specimens with different the...
Research on In-Pile Thermal-Mechanical Behavior of UMo/Zr Monolithic Fuel Plates
Kong Xiangzhe, Ding Shurong, Tian Xu
2018, 39(2): 109-113. doi: 10.13832/j.jnpe.2018.02.0109
Abstract(28)
Abstract:
UMo/Zr monolithic fuel plates experience complex multi-field-coupling and multi-scale behavior during the in-pile radiation.In this study,the method of multi-scale simulation of the in-pile behavior in the UMo/Zr monolithic fuel plate under a homogeneous irradiation condition is built.The distributi...
Numerical Simulation of Effect of Structural Parameters of Air Relief Valve on Actuation Time for Valve
Zhang Zhenhua, Yu Deyong, Jia Li, Yang Lixin
2018, 39(2): 114-119. doi: 10.13832/j.jnpe.2018.02.0114
Abstract(33)
Abstract:
Aiming at the pre-isolation valve system of the atmospheric relief valve in a valve system of Tianwan Nuclear Power Plant,the numerical simulation technology is applied to carry out the numerical simulation of valve opening characteristics.The effect of different structural parameters and different ...
Investigation on Visualization Test Methods for Flow Field Downstream Spacer Grids
Qu Wenhai, Chen Shilong, Xiong Jinbiao, Cheng Xu, Wang Xiaoyu, Du Sijia
2018, 39(2): 120-123. doi: 10.13832/j.jnpe.2018.02.0120
Abstract(21)
Abstract:
When visualization measurement techniques,such as particle image velocimetry（PIV） and laser Doppler velocimetry（LDV）,are applied in rod bundle experiments,measurement locations are restricted due to blockage and refraction by solid rods.Employing matched index of refraction（MIR） techniques for the h...
Research of Global-Local Resonance Self-Shielding Calculation Method Based on NECP-X
He Qingming, Cao Liangzhi, Liu Zhouyu, Zu Tiejun, Wu Hongchun
2018, 39(2): 124-128. doi: 10.13832/j.jnpe.2018.02.0124
Abstract(19)
Abstract:
To meet the challenges faced in high-fidelity resonance self-shielding calculations,the global-local self-shielding calculation method is proposed.The resonance self-shielding and correlated effects are classified into global and local effects.The global effects are weak or independent of energy,whi...
RMC Solutions to Kinetic Cases of C5G7-TD Benchmark
Guo Xiaoyu, Shang Xiaotong, Qiu Yishu, Cheng Quan, Song Jing, Huang Shanfang, Wang Kan
2018, 39(2): 129-132. doi: 10.13832/j.jnpe.2018.02.0129
Abstract(22)
Abstract:
The Reactor Monte Carlo code RMC,developed by REAL group,at Tsinghua University,has a function of kinetic calculation.RMC has participated in C5G7-TD transient benchmark calculation to further optimize its kinetic algorithm,and also to validate its results.The C5G7-TD benchmark consists of 6 sets of...
Research on Preventing Xenon Oscillation in Monte Carlo Burnup Calculation Based on RMC
Li Wanlin, Yu Ganglin, Wang Kan, Li Yaodong
2018, 39(2): 133-136. doi: 10.13832/j.jnpe.2018.02.0133
Abstract(32)
Abstract:
Monte Carlo burnup calculation suffers from numerical xenon oscillation when its model is huge geometry,refers as weakly coupled neutron transport system.Numerical xenon oscillation brings in error,even aborts burnup calculation.How to effectively prevent such oscillation is a significant topic in M...
Development and Verification of Reactor Core Transient Coupling Simulation Software CTSS
Pan Junjie, An Ping, Wang Wei, Zhao Wenbo, Xing Suo, Lu Wei, Chai Xiaoming
2018, 39(2): 137-141. doi: 10.13832/j.jnpe.2018.02.0137
Abstract(34)
Abstract:
Reactor Core Transient Simulation Software（CTSS V1.0）,which is coupled by 3 D space-time kinetic program NACK V1.0,CORe Thermal-Hydraulic analysis program（CORTH V2.0） and Fuel Rod Performance Analysis Code（FUPAC V1.1）,is used to simulate the typical PWR core.NACK V1.0 calculates the core diffusion e...
Study on Semi-Implicit Scheme for Two-Fluid Seven-Equation Two-Pressure Model
Chao Fei, Shan Jianqiang, Zhang Yong, Wu Pan, Gou Junli, Li Jian
2018, 39(2): 142-148. doi: 10.13832/j.jnpe.2018.01.0142
Abstract(23)
Abstract:
Current main reactor thermal-hydraulics system analysis codes are developed based on two-fluid six-equation single pressure model.However,such six-equation model has been proved to be ill posed,which could lead to numerical oscillations.Since two-fluid seven-equation two-pressure model is unconditio...
Investigation of Flow Rate Measurement Method in Gas-Liquid Two-Phase Flow with High Gas Volume Fraction
Ma Yugao, Li Chao, Huang Shanfang, Yu Hongxing, Pan Yanzhi
2018, 39(2): 149-152. doi: 10.13832/j.jnpe.2018.02.149
Abstract(16)
Abstract:
In this paper,a slip ratio based calculation model was developed to predict the gas mass flow rate under high gas volume fraction conditions,which was verified against an experimental investigation with a Venturi meter and a gamma ray attenuation system.Experiments showed that the new model could ac...
Uncertainty Analysis for Parameters of CFAST Based on Electrical Cabinet Fire Scenario in Main Control Room
Wang Wanhong, Zhu Dahuan, Peng Changhong, Guo Yun
2018, 39(2): 153-156. doi: 10.13832/j.jnpe.2018.02.0153
Abstract(17)
Abstract:
Based on the electrical cabinet fire scenario in the main control room,The paper use software,CFAST,by coupling with Monte Carlo,to analyze the uncertainty of heat release rate and soot yield.Distributions of layer temperature and optical density is used to assess the evacuation time and its probabi...
Study on Coupling of Integral Severe Accident Analysis Code with System Code
Wu Xiaoli, Li Wei, Deng Jian, Deng Chunrui, Zhang Ming, Guo Chao, Yuan Hongsheng
2018, 39(2): 157-161. doi: 10.13832/j.jnpe.2018.02.0157
Abstract(21)
Abstract:
The integral severe accident analysis code and RELAP5 were directly coupled for improving the overall safety analysis of the containment and primary system.Two kinds of coupling were implemented,i.e.breaks on the pressure boundary between the primary system and containment,interfaces between the saf...
Experimental Investigation of Condensation Induced Water Hammer during Residual Heat Removal Process for Nuclear Reactors
Wang Lutao, Li Jian, Zhong Daotong, Ran Xu, Zhang Zhuohua, Yan Junjie
2018, 39(2): 162-165. doi: 10.13832/j.jnpe.2018.02.0162
Abstract(27)
Abstract:
Serious steam condensation induced water hammer（CIWH） phenomenon would occure when the steam directly contacts with the water during the residual heat removal process for nuclear reactors.In this paper,CIWH phenomenon was investigated experimentally in a horizontal pipe.Firstly,it was found that CIW...
System of Generic Criteria and Operational Criteria for Nuclear and Radiological Emergency of HPR1000
He Fan, Yu Hong, Mu Keliang
2018, 39(2): 166-170. doi: 10.13832/j.jnpe.2018.02.0166
Abstract(26)
Abstract:
The present system of generic intervention levels and generic action levels in China do not meet the need of the nuclear power plants such as HPR1000 for the nuclear and radiological emergency.In this paper,the system of the generic criteria and operational criteria of HPR1000 is advanced,based on t...
Experimental Study on Effect of Inlet Condition on Corrugated Plate Dryer
Mao Feng, Tian Ruifeng, Chen Bowen, He Wei
2018, 39(2): 171-175. doi: 10.13832/j.jnpe.2018.02.0171
Abstract(25)
Abstract:
The effects of inlet condition including droplet size,inlet moisture content and inlet airflow speed on corrugated plate dryer’s performance were studied in a visualized experiment system.A laser particle sizer was used to measure water particle size distribution.The results show that: the separatio...
Study on Design of Fault Diagnosis System for Nuclear Power Plants
Wang Tianshu, Yu Ren, Liu Xiaofan
2018, 39(2): 176-179. doi: 10.13832/j.jnpe.2018.02.0176
Abstract(17)
Abstract:
To improve the operation safety and reliability of nuclear power plants and reduce human error,the design of operation fault diagnosis system based on the theory of expert system and human factor engineering is researched in this paper.Firstly,the demand of the system functions is analyzed,and the d...
3D Simulation of Fuel Performance with Missing Pellet Surface Defect
Liu Zhenhai, Chen Ping, Zhou Yi, Li Wenjie, Zhang Kun, Xing Shuo, Miao Yifei
2018, 39(2): 180-184. doi: 10.13832/j.jnpe.2018.0180
Abstract(23)
Abstract:
The effect of missing pellet surface（MPS） defect on fuel performance by coupling three dimensional code and one and half dimensional code is studied.The three dimensional fuel performance code is built based on ABAQUS software and compared with COPERNIC,a one and half dimensional fuel performance co...
Investigation on Hold-down Force Calculation Model of Fuel Assembly Based on Monte Carlo Algorithm
Zhu Fawen, Pu Zengping, Chen Ping, Ma Chao, Li Yun, Zhou Xiaoyun, Zeng Xiaomin, Geng Fei
2018, 39(2): 185-188. doi: 10.13832/j.jnpe.2018.02.0185
Abstract(22)
Abstract:
The hold-down force calculation model of fuel assembly was established based on the Monte Carlo algorithm,and model validation has been demonstrated its validity.It is suitable to use 200 thousand simulation times for the hold-down system analysis when using Monte Carlo algorithm in practice.
Application and Prospect of PHM Technology in Nuclear Power Plants
Xie Guangyao, Liu Jingquan, Zeng Yuyun
2018, 39(2): 189-192. doi: 10.13832/j.jnpe.2018.02.0189
Abstract(23)
Abstract:
Prognostics and Health Management（PHM） system is able to improve the safety of nuclear power plants and reduce maintenance costs.Prognostics and Health Management system includes five modules:data collection and processing,condition monitoring,fault diagnosis,life prediction and health management.De...
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