核动力工程

Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor

Cao Liangzhi, Zou Xiaoyang, Liu Zhouyu, Wan Chenghui, Wu Hongchun

2021, 42(2): 1-15. doi: 10.13832/j.jnpe.2021.02.0001

Abstract(2033) PDF(376) [Cited by] (13)

Abstract:
Numerical reactor based on high fidelity model and method is with the characteristics of high precision and high resolution, but the inherent uncertainty of nuclear data and other parameters will seriously affect the uncertainty of numerical reactor analysis results. Based on the review of the resea...

Experimental Study on Effect of Diffusion Angle on Turbulence in Venturi Tube

Shentu Yunqi, Song Yuchen, Yin Junlian, Yuan Hong, Wang Dezhong

2021, 42(2): 16-22. doi: 10.13832/j.jnpe.2021.02.0016

Abstract(489) [Cited by] (12)

Abstract:
In order to study the effect of the diffusion angle on the turbulence in the venturi tube, in this work, the stereo-particle image velocimetry is used to measure the transient velocity of the diffusion part of venturi tube, which diffusion angle is 10°, 12.5°, 15°and 20°, respectively, and the time-...

Conceptual Design of a Reduced Moderation Long-Life Heavy Water Cooled Thorium Small Modular Reactor

Sun Qizheng, Wang Lianjie, Zhang Tengfei, Li Xiangyang, Liu Xiaojing

2021, 42(2): 23-28. doi: 10.13832/j.jnpe.2021.02.0023

Abstract(258)

Abstract:
This paper proposes a conceptual design for an innovative reactor, reduced moderation small modular reactor (RMSMR), based on the technology of conventional PWRs. In this paper we firstly establish a two-dimensional model to analyze the influences of fuel types, coolant types and other parameters of...

Study on Separation Characteristics and Mechanism of a New Swirl Vane Separator

Qian Yalan, Yang Lingfang, Zhang Tingting, Yin Junlian, Wang Dezhong

2021, 42(2): 29-34. doi: 10.13832/j.jnpe.2021.02.0029

Abstract(317) [Cited by] (5)

Abstract:
In order to obtain the bubble separation behavior in the swirl vane separator of the molten salt reactor quantitatively, the liquid flow field distribution inside the separator was obtained by computational fluid dynamics method. Based on the bubble separation model that coupling the vortex model an...

Investigations on Supercritical CO2 Critical Flow through Mini Tubes

Wang Junfeng, Wang Yangle, Zhou Yuan, Huang Yanping

2021, 42(2): 35-38. doi: 10.13832/j.jnpe.2021.02.0035

Abstract(239) [Cited by] (8)

Abstract:
In this paper, an experimental supercritical CO2 critical flow investigation was carried out. The effect of upstream stagnation parameters and length to diameter ratio on the critical flow rate was obtained. A prediction correlation of mass flow rate equation was proposed based on the experimental d...

Development of 3-D Parallel First Collision Source Module

ang Chao, Yu Tao, Deng Li, Cheng Tangpei

2021, 42(2): 39-42. doi: 10.13832/j.jnpe.2021.02.0039

Abstract(143) [Cited by] (1)

Abstract:
The computational efficiency of the first collision method is low for large-scale mesh problems, and meanwhile there is a large calculation error in the near source region. In order to solve these problems, 3-D parallel first collision source module is developed in this paper. The spatial domain par...

Research on Ring Reactivity-Equivalent Physical Transformation Method for Homogenization of Double-Heterogeneous Systems

Lou Lei, Peng Xingjie, Chai Xiaoming, Yao Dong, Li Mancang, Yu Yingrui, Wang Lianjie

2021, 42(2): 43-49. doi: 10.13832/j.jnpe.2021.02.0043

Abstract(230) PDF(63) [Cited by] (1)

Abstract:
Because of the double heterogeneity of the dispersed particle-type fuel and the burnable poison, the traditional volumetric homogenization method (VHM) on double-heterogeneous (DH) systems will bring unacceptable calculation deviation. The reactivity-equivalent physical transformation (RPT) method a...

Fine CFD Simulation Study of Whole Core for HPR1000 Reactor Vessel

Bi Shumao, Liu Yu, Liu Luguo, Xu Youyou, Deng Jian, Miao Yifei, Wu Lingyan

2021, 42(2): 50-54. doi: 10.13832/j.jnpe.2021.02.0050

Abstract(520) [Cited by] (11)

Abstract:
Fine and large scale CFD simulation of the whole reactor core is an important methodology in the research and design of HPR1000 and digital reactor. Based on a series of reasonable simplification of the core, the geometry model of the whole core were built for HPR1000. Meanwhile, the fuel assembly i...

Analysis of Enhanced Convection Heat Transfer of Coaxial Cross Triple Twisted Tape for Single Phase Fluid

Liu Xiaoya, Zhang Yongfa, Yang Zimu, Ding Ming

2021, 42(2): 55-59. doi: 10.13832/j.jnpe.2021.02.0055

Abstract(228) [Cited by] (7)

Abstract:
A large number of heat exchangers are used in the passive safety system of nuclear power plants. Under the condition of natural circulation, the flow of the fluid in the heat exchanger is weak, that may be in laminar flow and leads to the low heat transfer coefficient. In order to enhance the heat t...

Experimental Investigation of Counter-Current Flow in Downcomer of CAP1400

Fei Likai, Zhang Peng, He D, an, Yuan Haowei, Hu Fuquan, Cui Lei, Shen Feng, Liu Lifang

2021, 42(2): 60-64. doi: 10.13832/j.jnpe.2021.02.0060

Abstract(446)

Abstract:
In order to study the phenomenon of counter-current flow in the downcomer of the CAP1400 reactor under real working conditions, a test bench with the height and diameter ratio of 1:1 of the downcomer of the pressure vessel and a 60° slice was built using CAP1400 as a prototype. The test working medi...

Flow Instability in Parallel Narrow Rectangular Channel under Different Radial Heat Distribution

Wang Yanlin, Zhou Lei, Zan Yuanfeng, Yang Zumao, Yan Xiao

2021, 42(2): 65-68. doi: 10.13832/j.jnpe.2021.02.0065

Abstract(236) [Cited by] (3)

Abstract:
The flow instability in the parallel channels under different radial heat distribution was investigated in the following condition: the system pressure is 0.89~1.32 MPa, the mass flux is 500-750 kg/(m2?s) and the inlet temperature is 58.5℃~132.3℃ with the upflow deionized water. The test section con...

Study on Physical Characteristics of Mid-Beryllium Control Rod of HFETR

Kang Changhu, Liu Shuiqing, Chen Qibing, Xiang Yuxin, Zou Peng, Song Jiyang, Song Yuge, Liu Wenbin

2021, 42(2): 69-71. doi: 10.13832/j.jnpe.2021.02.0069

Abstract(139) [Cited by] (1)

Abstract:
In this paper, the physical characteristic of the mid-beryllium control rod of HFETR is studied. Firstly, the few group cross section of every assembly is calculated by CELL code. Secondly, the core physical calculations of the setting up control rod cell and mid-beryllium control rod cell are carri...

Numerical Analysis of Density Wave Instability Phenomena of Supercritical Water in Two Parallel Channels

Zang Jinguang, Yan Xiao, Huang Yanping

2021, 42(2): 72-76. doi: 10.13832/j.jnpe.2021.02.0072

Abstract(307) [Cited by] (3)

Abstract:
The flow instability of supercritical water is an important aspect of thermal-hydraulic design of Supercritical Water-cooled Reactor (SCWR). The validated numerical simulation method helps to obtain the internal mechanism of flow instability, a wider instability boundary, and further deepen the unde...

Formation Mechanism for Turbulence Induced Disturbance Wave on Liquid Film of Annular Flow

He Hui, Ren Quanyao, Ye Tingpu, Wu Yao, Pan Liangming

2021, 42(2): 77-81. doi: 10.13832/j.jnpe.2021.02.0077

Abstract(283) [Cited by] (10)

Abstract:
To explore the formation mechanism of the disturbance wave from the perspective of accurately predicting the properties of the liquid droplet/film mass transport and dryout scenario, the high-speed camera and conductivity-based techniques are employed to visualize and measure the evolution behaviors...

Theoretical Study of Bubble Departure Diameter in Subcooled Flow Boiling

in Di, Xiong Jinbiao, Cheng Xu

2021, 42(2): 82-87. doi: 10.13832/j.jnpe.2021.02.0082

Abstract(309) [Cited by] (3)

Abstract:
Bubble departure diameter model is an important sub-model for wall boiling calculation. In order to predict the wall heat transfer accurately, this paper models the bubble departure diameter in subcooled flow boiling by using the force balance method with a new improved bubble growth model. Bubble g...

Experimental Study of Single-Phase Flow Characteristics in a Narrow Rectangular Channel with Vertical upward Flow

Song Mingliang, Ma Jian, Huang Yanping, Zhang Yu, Lin Zhenxia

2021, 42(2): 88-92. doi: 10.13832/j.jnpe.2021.02.0088

Abstract(243) [Cited by] (2)

Abstract:
In this paper, a vertical upward narrow gap rectangular channel is taken as the research object, and the experimental study of single-phase flow characteristics under constant heat flux density of heating conditions is carried out. According to the measured temperature, flow rate, pressure drop and ...

Research on Design Method of HPR1000 Emergency Residual Heat Removal System Based on System Engineering Method

Chen Guocai, Li Feng, Tang Huapeng, Qiu Zhifang, DengJian

2021, 42(2): 93-98. doi: 10.13832/j.jnpe.2021.02.0093

Abstract(210) [Cited by] (2)

Abstract:
Based on the system engineering, the system design is conducted for the emergency core residual removal system of HPR1000 nuclear power plant. Considering the requirements of safety, the economic efficiency and technology maturity, the systematic evaluation index system is constructed with the objec...

Study on Design of Semi-Physical Simulation System for Xi’an Pulsed Reactor

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

2021, 42(2): 99-104. doi: 10.13832/j.jnpe.2021.02.0099

Abstract(247) [Cited by] (1)

Abstract:
A semi-physical simulation system of Xi’an Pulsed Reactor is developed for the purpose of providing the analog signal source, verifying the power regulation method and personnel training for the digital instrument and control system of Xi’an pulse reactor under construction. The design philosophy of...

Improvement and Test Verification of Reactor-Core Neutron and Temperature Detector

Ma Xiaoyu, Deng Tao

2021, 42(2): 105-109. doi: 10.13832/j.jnpe.2021.02.0105

Abstract(355) [Cited by] (3)

Abstract:
The reactor-core neutron and temperature detector is an integrated detector integrating rhodium self powered neutron detector and thermocouple thermometer. The detector can simultaneously measure the neutron fluence rate of the core and the outlet temperature of the fuel assembly. This paper focuses...

Development and Application of Fuzzy Switching Control System for Nuclear Reactor Core

Jiang Qingfeng, Zeng Wenjie

2021, 42(2): 110-114. doi: 10.13832/j.jnpe.2021.02.0110

Abstract(278) [Cited by] (4)

Abstract:
In order to make use of the advantages of different types of controllers, based on the fuzzy multi-model of reactor core, a fuzzy switching controller is designed by using PID controller and fuzzy controller with the T-S fuzzy rules. Taking the pressurized water reactor core of the Three Mile Island...

Design and Verification on Nuclear Safety Class Digital Instrument Control System Based on FPGA

Ma Xiaoyu, Huang Xiaojin, Wang Dong

2021, 42(2): 115-120. doi: 10.13832/j.jnpe.2021.02.0115

Abstract(310) [Cited by] (4)

Abstract:
The common cause failure(CCF) due to the software design of the digital safety class instrument and control (I&C) system based on micro-processor technology for NPP is resolved by the diversity measures implemented based on FPGA, which could prevent the failure of the reactor protection system (...

Study on Temperature Field of Control Rod Drive Mechanism Based on Natural Convection

Li Wei, Peng Hang, Yu Ti, a, Fu Guozhong, Wu Hao, Tang Yuan, Tang Jiankai

2021, 42(2): 121-124. doi: 10.13832/j.jnpe.2021.02.0121

Abstract(289) [Cited by] (3)

Abstract:
After the improvement of the temperature resistance level of the control rod drive mechanism (CRDM) of the reactor, in order to analyze whether it is feasible to eliminate the forced ventilation system of CRDM, a calculation model of the airflow field intended for single CRDM is established to deter...

Research on Accuracy of Hydraulic Performance Calculation under Full Working Conditions of a Nuclear Main Pump Based on Detached Eddy Simulation

Wang Xiuyong, Liu Zhiyuan, Liu Yuning, Li Yibin

2021, 42(2): 125-130. doi: 10.13832/j.jnpe.2021.02.0125

Abstract(305) [Cited by] (2)

Abstract:
In order to study the effect of the detached eddy simulation model on the accuracy of hydraulic performance prediction for a nuclear main pump, with the hexahedral structured grid, the detached-eddy simulation (DES), delayed detached-eddy simulation (DDES) and improved delayed detached-eddy simulati...

Thermal-Fluid-Solid Coupling Characteristics of a Circular Tube in Two Degrees of Freedom Flow-Induced Vibration

Ding Lin, He Haoyu, Yang Zuomei, Zhang Li, Yang Lin

2021, 42(2): 131-136. doi: 10.13832/j.jnpe.2021.02.0131

Abstract(142) [Cited by] (6)

Abstract:
This paper studies the fluid-induced vibration-heat transfer coupling characteristics of a single circular tube with constant temperature, and the variation rules of the cylinder vibration response, temperature field, average Nusselt number and maximum local Nusselt number are analyzed. The results ...

Design and Simulation Verification of Boron Concentration Algorithm of Boron Meter

Zheng Junwei

2021, 42(2): 137-143. doi: 10.13832/j.jnpe.2021.02.0137

Abstract(291) [Cited by] (2)

Abstract:
In order to solve the problem of the deviation of the total boron concentration of the coolant of nuclear island primary circuit measured by the off-line boron meter (OFBM) from the chemical titration boron concentration in the primary coolant in pressurized water reactor nuclear power plant exceeds...

Development of Ultrasonic Inspection System for Control Rod Guide Split Pins in Nuclear Power Plants

Ma Guanbing, Ding Hui, Wang Weiqiang, Yan Jingli, Ma Chao, Wang Bin

2021, 42(2): 144-147. doi: 10.13832/j.jnpe.2021.02.0144

Abstract(337) [Cited by] (2)

Abstract:
The control rod guide split pin is a key connecting part of the nuclear power plant reactor internals. Stress corrosion cracking is an important potential danger that affects the quality of the split pin after long-term operation. In order to achieve an efficient and accurate inspection of the contr...

Research on PA Ultrasonic Inspection Technology in RPV

Hu Chenxu, Sun Jiawei, Li Bingqian

2021, 42(2): 148-152. doi: 10.13832/j.jnpe.2021.02.0148

Abstract(266) [Cited by] (6)

Abstract:
By building the RPV model, the phase array (PA) ultrasonic detection technology was developed to replace the conventional ultrasonic detection technology, and then the detection and quantitative capability of the two detection technologies were compared and evaluated. The verification tests show tha...

Study on Key Parameters of the Selection of Steam Piping Traps in Nuclear Power Plant

Liu Chongyu, Duan Zhengqiang, Wu Wei

2021, 42(2): 153-156. doi: 10.13832/j.jnpe.2021.02.0153

Abstract(182)

Abstract:
In order to gain quickly the quantity of the drain of the nuclear power plant steam pipe, by calculating the quantity of the drain of the steam pipe in the start period and normal operation period, the heat dissipation picture of steam pipes with various piping specification are analyzed and drawn. ...

Application of Bayesian Estimation Method in Parameter Estimation of Alpha Factor Model

An Jin, Yan Lin

2021, 42(2): 157-160. doi: 10.13832/j.jnpe.2021.02.0157

Abstract(300) [Cited by] (1)

Abstract:
The probabilistic safety analysis (PSA) results of the nuclear power plant shows that common cause failures (CCF) plays an important role in the system reliability. In China, generic data are often used for common cause failure data in PSA. It is difficult to reflect the operation characteristics of...

Simulation Research on Dynamic Characteristics and Small Break Fault of Once-through Steam Generator

Xu Yu, Huangfu Zeyu, Xu Jianqun, Tian Peiyu, Li Chunmei, Cheng Xiang, Yan Siwei, Liao Xianwei

2021, 42(2): 161-167. doi: 10.13832/j.jnpe.2021.02.0161

Abstract(579) [Cited by] (7)

Abstract:
Taking a once-through steam generator designed by Babcock & Wilcox as the study object, a graphical model was established, based on the basic module of supporting platform APROS. On this basis, the steady-state and dynamic simulation were carried out under varying operating conditions. According...

Research of UO2-(Zr0.8Ca0.2)O1.8 FCC Solid Solution Spheres Preparation

Long Dijun, Zeng Cheng, Lu Zhangxian, Li Jia

2021, 42(2): 168-172. doi: 10.13832/j.jnpe.2021.02.0168

Abstract(232)

Abstract:
In order to develop the single - phase ceramic fuel with good radiation resistance, the UO2-(Zr0.8Ca0.2)O1.8 microspheres with different uranium contents (30mol%, 50mol%, 70mol%) were prepared by Inner-Sol-Gel technique. The technique is with five processes including sol preparation, casting and gel...

Qinshan-I Reactor Shielding Simulation and Sensitivity Analysis Based on JMCT Monte Carlo Code

Deng Li, Li Rui, Ding Qianxue, Qiu Youheng, Li Gang, Fu Yuanguang, Shi Dunfu, Liu Peng

2021, 42(2): 173-179. doi: 10.13832/j.jnpe.2021.02.0173

Abstract(447) [Cited by] (8)

Abstract:
Qinshan-I reactor has been operated for over thirty years. A great amount of data is obtained from the design, building, operation, life-extension and decommission and can provide the valued reference for the design, building, operation and maintenance of the nuclear power plants with new generation...

Research on Contact Shape Function and Contact Model of Planetary Roller Screw Drive Pair

Liu Jia, Peng Hang, Luo Ying, Zhang Yixiong, Zhu Zihao, Yan Dapeng, Deng Qiang

2021, 42(2): 180-182. doi: 10.13832/j.jnpe.2021.02.0180

Abstract(125) [Cited by] (4)

Abstract:
Based on the static contact state of planetary roller screw drive pair, a contact characteristic model and contact force analysis method are proposed by the shape function analysis of the drive pair and Hertz contact theory. The results of finite element analysis show that the method is reasonable, ...

Study on Online Monitoring System Technology for Safety Instruments in Nuclear Power Plants

Zhao Yao, Huo Yujia, Wang Jun, Yu Junhui

2021, 42(2): 183-187. doi: 10.13832/j.jnpe.2021.02.0183

Abstract(211) [Cited by] (6)

Abstract:
At present, in the nuclear power plants in China, the periodic calibration is often adopted to manage the drift of safety instruments, however it is too conservative and uneconomical. Thus, this paper studies the technology of online monitoring system for safety instruments. First of all, the actual...

Preliminary Study of Influencing Factors of Power-to-Mass Ratio of Heat Pipe Reactor Power System Based on Simple Open Brayton Cycl

Wang Jinyu, Yu Hongxing, Zhang zhuohua, Ma Yugao, Chai Xiaoming, Chen Wei, Yi Jingwei, Zeng Chang, Su Dongchuan, Xiao Cong

2021, 42(2): 188-192. doi: 10.13832/j.jnpe.2021.02.0188

Abstract(287) [Cited by] (2)

Abstract:
The heat pipe reactor system based on the simple open Brayton cycle is with the characteristics of simple structure, inherent safety, and low risk of radioactive leakage, and is a potential leading technology option to meet the demand for more practical and innovative nuclear reactor designs capable...

Research and Design of Flow Measurement Tests of Reactor Coolant System

Huang Zongren, Wang Mingli, Li Feng

2021, 42(2): 193-196. doi: 10.13832/j.jnpe.2021.02.0193

Abstract(320) [Cited by] (3)

Abstract:
The reactor coolant pump inlet and outlet differential pressure gauge was set in HPR1000 to measure the loop flow of the reactor coolant system, and the elbow flowmeter set in the second generation progressive nuclear power unit was eliminated. The change of the flow measurement means directly affec...

Preliminary Neutronics Analysis for Fluorine-Salt-Cooled Pebble-Bed Reactor

Li Zhifeng, Zhao Changyou, Zhang Guangchun, Han Song, Huang Jie

2021, 42(2): 197-201. doi: 10.13832/j.jnpe.2021.02.0197

Abstract(238)

Abstract:
In order to obtain the key neutronics parameters of the pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR), the random packing method is used to obtain the coordinates of the three layers of isotropic coated particles (TRISO) particles, and the discrete element method is adopted to th...

Research on Hybrid Monte-Carlo-Deterministic and Weight-Window Mesh-Coarsening Method Based on NECP-MCX

Zheng Qi, Shen Wei, He Qingming, Li Jie, Cao Liangzhi

2021, 42(2): 202-207. doi: 10.13832/j.jnpe.2021.02.0202

Abstract(281) [Cited by] (11)

Abstract:
In the Monte Carlo simulation of a deep-penetration problem, only a small number of particles can penetrate the shielding layer and reach the target region, resulting in a very low computational efficiency. In order to solve the deep-penetration problem, the hybrid Monte-Carlo-Deterministic method i...

Preliminary Research on a Multi-Physics Coupling Platform for Heat Pipe Reactors

Li Xiangyue, Xiao Wei, Zhang Tengfei, Li Peijie, Liu Xiaojing

2021, 42(2): 208-212. doi: 10.13832/j.jnpe.2021.02.0208

Abstract(452) [Cited by] (7)

Abstract:
In order to achieve high-precision and high-fidelity numerical simulations of nuclear energy system and to explore the true physical process therein, a coupled three-dimensional high-fidelity calculation platform named MPCH for neutronics/thermal-conduction/stress-analysis calculation has been devel...

Development of Advanced Neutronics Code SCAP-N for Reactor Core High Fidelity Simulation

Peng Lianghui, Tang Chuntao, Yang Weiyan

2021, 42(2): 213-218. doi: 10.13832/j.jnpe.2021.02.0213

Abstract(382) [Cited by] (8)

Abstract:
Reactor core neutronics calculation is important for the reactor design and analysis. In order to improve the resolution and accuracy of the reactor core neutron transport simulation, the advanced neutronics code SCAP-N was developed. In this code, firstly, the core was divided into layers according...

2021 Vol. 42, No. 2