核动力工程

Study on Characteristics of Ex-core Detector Response on Primary Neutron Source and Criticality Extrapolation Method during Start-up Experiment

Han Song, Shi Xiuan, Fu Xuefeng

2014, 35(4): 1-4. doi: 10.13832/j.jnpe.2014.04.0001

Abstract(13) PDF(0)

Abstract:
MCNP5 was used to simulate the ex-core source range detector（SRD） response to the primary source（PNS） at the stage of reactor start-up experiment. The relationship among k_eff, k_s and the detector counts C has been analyzed. Through the prediction of SRD response factor a, traditional criticality extrapolation method 1/C is optimized by a’/C in order to obtain a more linear extrapolated curve in the boron dilution process. The result shows that a’/C gives a better match with 1-k_eff as soon as the core approaches to critical.

Wielandt Iteration Combined with Multi-Group GMRES for Accelerating Matrix MOC Convergence

Wu Wenbin, Li Qing, Wang Kan

2014, 35(4): 5-9. doi: 10.13832/j.jnpe.2014.04.0005

Abstract(27) PDF(0)

Abstract:
In the Matrix MOC, a linear algebraic equation system can be constructed by sweeping only once, and then solving the linear system takes the place of repeatedly characteristics sweeping. Traditionally, keff is computed by power iteration（PI）, whose convergence rate depends on the dominance ratio deeply. Large problems of practical interest often have dominance ratios close to 1, leading to slow convergence of PI. Combined with multi-group GMRES coupling all groups directly, Wielandt iteration is studied for accelerating Matrix MOC. Numerical results of several benchmarks demonstrate that Wielandt iteration combined with multi-group GMRES can obtain good accuracy and higher efficiency compared with PI.

Study on Measurement Method of ²³⁵U Isotope Abundance of Fuel Assembly for Research Reactor Based on a Compact Neutron Generator

Zhang Min, Cao Fangfang, Que Ji

2014, 35(4): 10-12. doi: 10.13832/j.jnpe.2014.04.0010

Abstract(18) PDF(0)

Abstract:
Neutron activation analysis and mass spectrometry are used as 235 U isotope abundance of fuel assembly. Neutron activation analysis and mass spectrometry can’t be used as an effective measurement means of the 235 U Isotope Abundance of Fuel Assembly. Meanwhile they aren’t also used as a verification mean of 235 U isotopic abundance the fuel assembly management and transportation because of their shortcomings of large measurement equipment and sampling. So based on a compact neutron generator, a rapid and portable measurement method for 235 U Isotope Abundance of Fuel Assembly is significant. According to the linear correlation relating the ratio of average yield for the specific fissile nuclides to 235 U Isotope Abundance, the samples whose 235 U Isotope Abundance is 10%~90% are successfully measured. It provides the experimental basis for the verification of 235 U Isotope Abundance of the fuel assembly for the research reactor.

Dynamic Analysis of CRDM Loads during Stepping Motion

Wei Qiaoyuan, Li Yuezhong, Ran Xiaobing

2014, 35(4): 13-16. doi: 10.13832/j.jnpe.2014.04.0013

Abstract(29) PDF(0)

Abstract:
CRDM stepping loads act on the parts of CRDM during stepping motions. In this study the CPR1000 CRDM stepping loads are calculated through kinematics analysis and Adams dynamic simulation. The stepping loads of latch, pin, and coupling are got. Conclusions show that the method is an effective method to calculate CRDM stepping loads and the results are consistent with engineering experience. At the same time comparison and analysis of the impact of hydraulic camping force on insertion motion are studied.

Flow-Induced Vibration Analysis of Steam Generator Heat Transfer Tube Based on ANSYS

Zhu Yong, Qin Jiaming, ReN Hongbing, Zuo Chaoping, Han Tongxing

2014, 35(4): 17-20. doi: 10.13832/j.jnpe.2014.04.0017

Abstract(28) PDF(0)

Abstract:
Based on general finite element software ANSYS using APDL language, steam generator heat transfer tube linear flow induced vibration analysis program is developed. The paper established tube finite element model with 3-D beam element, complete tube mode analysis, and the fluid-elastic instability ratio and turbulence response of the tube is calculated. Comparison of mode analysis result and fluid-elastic ratio calculated by ANSYS-APDL and program-specific shows that the two results are consistent, and the turbulence response calculated by ANSYS-APDL is more conservative. The computation program is based on the general finite element program ANSYS, it can greatly improve the efficiency of the practice engineering analysis with modeling convenience, strong readability and wide application range.

Application Survey and Prospect of Base-Seismic Isolation Technology for Nuclear Power Plants

Li Zhongcheng, Zhang Tao, Xu Botao

2014, 35(4): 21-26. doi: 10.13832/j.jnpe.2014.04.0021

Abstract(20) PDF(0)

Abstract:
Seismic safety has always been an important aspect of nuclear safety, while base seismic-isolation technology is an effective means to improve seismic safety of nuclear facilities. This paper, taking China’s mega kilowatt pressurized reactor as an example, provides the overall seismic isolation solution of nuclear island as well as the analysis of the isolation effects from the angle of design standardization of nuclear power station. Meanwhile, combined with the engineering application, major problems and solutions facing the nuclear power engineering in scale-applying seismic isolation technology are discussed in the paper. Therefore, the paper is of certain reference value and guiding significance to the application of seismic isolation technology in the field of nuclear power engineering.

Experimental Study on Performance Characteristics of Servo-Piston Hydraulic Control Rod Driving Mechanism

Yu Mingrui, Han Weishi, Wang Ge, Zhou Jie, Liu Chunyu, Yang Zhida

2014, 35(4): 27-30. doi: 10.13832/j.jnpe.2014.04.0027

Abstract(24) PDF(0)

Abstract:
An experimental study on the performance characteristics of the servo-piston hydraulic control rod driving mechanism is carried out, the dynamic processes of the driving mechanism are obtained through the experiments in different working conditions. Combined with the structure characteristics of the driving mechanism, the change rule between the characteristics parameters and the working condition is analyzed. The results indicate that the traction of the servo-tube decreases quickly at first, then slowly and finally trends to be a constant with the working pressure increasing, the tractions are the largest in the startup and deboost phases. The under pressure of the drive cylinder rises slowly and the upper pressure decreases rapidly at the beginning of the rise, the variation trend is opposite in the falling stage. There exists quick and clear flow change processes in the startup and deboost phases, the flow mutation value reduces and the mutation time changes a little with the working pressure increasing. The driving mechanism runs stable and has high sensitivity precision, the load does not vibrate at all when working conditions has small disturbance, a steady transform can be realized among every condition.

Three-Dimentional Finite Element Analysis for Bolted Flange Joint Structure of Primary Manway of Steam Generator

CheN Tao, CheN Rong, Zhang Yunbo, Lu Yan

2014, 35(4): 31-34. doi: 10.13832/j.jnpe.2014.04.0031

Abstract(34) PDF(0)

Abstract:
Three-dimentional finite element model for bolted flange joint structure of primary manway of steam generator was built considering temperature lag of the bolt, turning angle of the flange and contact between gasket and the flange, nut and the cover. Transient analysis was carried out under normal and upset conditions. Problem of air heat trsfer in the bolt hole and between the cover and the flange were solved. Bolt pretension load was applied, and the change of the bolt load was obtained along with the time. Mechanical and heat haracterics of the joint structure was simulated. Leaktightness was validated and fatigue stress and displacement of the joint structure were estimated according to regulations.

Development of Piping Support Structure Design Software Based on PDMS

Tang Yongtao, Guan Hui, Su Rongfu, Huang Wei, Mao Huihui

2014, 35(4): 35-38. doi: 10.13832/j.jnpe.2014.04.0035

Abstract(30) PDF(0)

Abstract:
In order to enhance the efficiency of nuclear power process system piping support design, the veracity of interface with support, piping and anchor, and decrease the clash between supports and other disciplines, developed piping support structure design software NPHS based on PDMS independently. That achieved the seamless integration of PDMS and NPHS by method of embedded development, reduce the size of program code, improve the running efficiency; That predigested the 3D modeling and information storage for support parts, that increased the support database opening and maintenance using the special mechanism and configuration of database. The support modeling efficiency due to setting of the connection key point of support parts is improved. Practices in several real nuclear power projects proved that NPHS software is provided with such outstanding performances: quick running, strong stability, accurate data, easy to operate and maintain, and output results satisfied the engineering requirements.

Numerical Analysis of Effects of Grid Spacer on Resistance Characteristics within Sub-Channel of SCWR

Zhu Xiaojing, Teng Dawei, Shen Shengqiang

2014, 35(4): 39-42. doi: 10.13832/j.jnpe.2014.04.0039

Abstract(23) PDF(0)

Abstract:
Numerical analysis is carried out for the effects of four kinds of grid spacer on the resistance characteristics within sub-channels of supercritical reactor core with commercial CFD code using unstructured polygonal mesh. The results show that the pressure loss induced by the grid spacer includes the frictional drag and form drag. The frictional drag is mainly determined by the length of grid strap while the form drag is determined by the blockage ratio of grid spacer and the structure of flow enhancing feature. Due to the more streamlined structure, the mixing-van type grid spacer induces less pressure loss when it is used to enhance forced convection within sub-channel. Although the mixing-van type grid spacer has more complex structure in comparison with standard egg-crate grid spacer, it has obvious advantages in the area of flow enhancing and can be primarily considered in the optimal design of grid spacer used in supercritical water cooled reactor core.

Numerical Simulation of Ex-Vessel Steam Explosion Based on 1000 MW Level PWR Nuclear Power Plant

ZhoNG Mingjun, LiN Meng, Zhang Zhengming, Yang Yanhua

2014, 35(4): 43-47. doi: 10.13832/j.jnpe.2014.04.0043

Abstract(18) PDF(0)

Abstract:
In order to investigate the pressure distribution outside the vessel during the process of steam explosion including both the premixing stage and explosion stage, a three dimensional multiphase fluid analysis code MC3 D has been used based on the structural and environmental characteristics of the sever accident state of 1000 MW level PWR Nuclear Power Plant with the consideration of both pressure variation and impulse variation. Analysis of effect of break’s size and its location is also included. The results indicate that the explosion stage creates a high level and sharp pressure peak which threatens the integrity of the reactor cavity structure. In addition, when the intersection angle between axis of the vessel and location of the break is 45° and size of the break is 0.7 m, the risk of steam explosion is the greatest.

Analysis on Method of ReactorMetal Reflector By-pass Flow

Shi Lin, Fang Jian, Ran Xiaobing, Dai Zhangnian

2014, 35(4): 48-51. doi: 10.13832/j.jnpe.2014.04.0048

Abstract(22) PDF(0)

Abstract:
The metal is much more than the water for the heavy reflector. By-pass flow should be calculated, which can optimize structure and provide input for the thermal analysis. CFD can be used for analysis the portion of metal reflector by-pass flow, and then results can be analyzed and evaluated. Finally, use the results to comparewith the theoretical analysis and experimental results. The result show that the complex channel and the characteristics of the flow field can be simulated by CFD, results are credible.

Analysis of Static Flow Excursion in Natural Circulation Using CATHARE Code

Zhang Yan, Yan Xiao, PeNG Chuanxin, Zhang Zhen

2014, 35(4): 52-55. doi: 10.13832/j.jnpe.2014.04.0052

Abstract(17) PDF(0)

Abstract:
This paper based on the characteristic of the static flow excursion under middle and low pressure condition, proposed the research results. CATHARE code was adopted for simulation of the natural circulation system, while different parameters such pressure and resistance positions were taken into account, looking for the influence factors. The results show that lower pressure easily lead to static flow excursion, as well as the higher resistance in the outlet of heater domain. Resistance distribution in the outlet of the heater section made a great impact for the whole natural circulation system. Besides, characteristics of static bifurcation and hysteresis effect are given in this paper.

Investigation of Flow and Heat Transfer of Subcooled Boiling of Water in Vertical Narrow Channel at Atmosphere Pressure

Guo Ang, Yang Lixin, CheN Weihong

2014, 35(4): 56-59. doi: 10.13832/j.jnpe.2014.04.0056

Abstract(25) PDF(0)

Abstract:
In the present paper, the flow and heat transfer characteristics of subcooled boiling of deionized water was investigated in a vertical rectangular narrow channel at atmospheric pressure. Flow visualization experiments were conducted to study the bubble nucleation on the heating wall. Based on the experimental observations, empirical models were proposed to correlate the active nucleation site density, bubble departure diameter and bubble departure frequency with given flow conditions for 2mm vertical narrow channel. Then, computational fluid dynamics（CFD） models were developed to simulate the subcooled flow boiling in the narrow channel by combining the two-fluid model, the Rensselaer Polytechnic Institute（RPI） wall heat flux partitioning model and the proposed bubble nucleation correlations. The numerical results show excellent agreement with the experimental data under the typical test conditions simulated.

Study on Operating Characteristics of Flashing Driving Natural Circulation System

Liu Xinkai, Wang Jianjun

2014, 35(4): 60-64. doi: 10.13832/j.jnpe.2014.04.0060

Abstract(20) PDF(0)

Abstract:
We deal with a proposed natural circulation system under atmospheric pressure. The codes are developed by us for the purpose of analyzing the performance of the system. The startup, operating process are simulated through the codes. Furthermore, the influence of flashing on the system behaviors is studied. The results show that the system can build up natural circulation quite quickly from beginning during startup phase and can provide reliable natural circulation capacity for long time. It can also be concluded that the flashing of the water is dominant for the system in long run. The sensitive analysis also proves the height difference between heat sink and heat source plays a very little role on the performance of the system.

Discussion on Control Room Habitability Assessment

Li Bing, CheN Yingying, Xiao Jun, Yang Duanjie, Cui Hao

2014, 35(4): 65-69. doi: 10.13832/j.jnpe.2014.04.0065

Abstract(22) PDF(0)

Abstract:
The discussion on control room envelope integrity, source term analysis in habitability assessments and other impact factors for dose consequence is provided combined with regulatory requirements and the current status of domestic NPPs. Considering that the infiltration is an important factor for control room habitability assessment, CRE integrity test should be performed to demonstrate the CRE’s infiltration characteristics. The consequence assessment should be performed based on different DBAs and different pathways, such as pathways internal to the plant.

Design of Multifunctional Temperature Controller for NPP Ventilation System

Xiao Xipeng, He Li, YiN Xiaolong, GoNG Jianjun

2014, 35(4): 70-73. doi: 10.13832/j.jnpe.2014.04.0070

Abstract(23) PDF(0)

Abstract:
Based on full study on the original analog temperature controller of ventilation system in nuclear power plant, for the interface definition fixation, parameters preset as well as multiple output combination of the temperature controller circuit, a digital temperature circuit based on the front-end input conditioning module + multi channel ADC+STM32 architecture is proposed. The hardware structure and software process are described in details. The sampling data processing, phase-cut voltage output as well as control principle are focused, and some part of software program is presented. The experimental results show that the temperature controller can stably and reliably control the temperature of ventilation system and display digital parameters, and it can completely replace and upgrade the original analog controller.

Selecting of Key Safety Parameters in Reactor Nuclear Safety Supervision

He Fan, Yu Hong

2014, 35(4): 74-77. doi: 10.13832/j.jnpe.2014.04.0074

Abstract(18) PDF(0)

Abstract:
The safety parameters indicate the operational states and safety of research reactor are the basis of nuclear safety supervision institution to carry out effective supervision to nuclear facilities. In this paper, the selecting of key safety parameters presented by the research reactor operating unit to National Nuclear Safety Administration that can express the research reactor operational states and safety when operational occurrence or nuclear accident happens, and the interrelationship between them are discussed. Analysis shows that, the key parameters to nuclear safety supervision of research reactor including design limits, operational limits and conditions, safety system settings, safety limits, acceptable limits and emergency action level etc.

Development and Application of Nuclear Power Plant Operational Safety Performance Indicator System

Chen Lu, Zhang Binbin, Zhang Qinfang

2014, 35(4): 78-81. doi: 10.13832/j.jnpe.2014.04.0078

Abstract(20) PDF(0)

Abstract:
The operational safety performance indicator（SPI） is a popular tool to assess the operation safety of nuclear power plant. It has been more and more regarded recently for their clear and effective characteristic. This thesis introduced the development and application of SPI system, included establishment of SPI framework, define the assessment criteria of performance indicator and establishment of software for SPI system, and give some suggestions for improving SPI system.

Measurement Device for Multi-Channel Neutron Flux Relative Distribution

Zhao Xiuliang, Huang Shun, Liu Liyan, Liu Caixue, ZoNG Xuncheng

2014, 35(4): 82-85. doi: 10.13832/j.jnpe.2014.04.0082

Abstract(25) PDF(0)

Abstract:
To realize the relative distribution measurement of neutron flux in reactor core, the measurement device for multi-channel neutron flux relative distribution has been developed, based on the radioactive counting of activated detection slice measured by NaI detector. The measurement device is made up of PTMC12 data-acquisition board, industrial PC and MNFDAS software, which can automatically achieve the timing and counting function in both cyclical and non-cyclical models, meanwhile the measuring results can be saved as both graphic and data files. The test shows that the stability of the measurement device is excellent with relative deviation within 1%, which can guarantee the requirement of real-time and long-term stable measurement of relative distribution measurement of neutron flux in reactor.

Eddy Current Data Analysis and Simulation for Wear Indications within Structural Supports of Nuclear Power Plant Steam Generator Tubes

Bei Yayao, SoNG Tao, Wang Xiaogang, CheN Yusu, LiN Ge

2014, 35(4): 86-89. doi: 10.13832/j.jnpe.2014.04.0086

Abstract(20) PDF(0)

Abstract:
This paper studied the eddy current inspection for wear indications within structural supports of nuclear power plant steam generator tubes and discussed the data analysis method of two different kinds of wear mechanisms. Meanwhile, We measured or calculated the depth, length, location and volume loss of wear indication through eddy current data analysis method and model computation. Three different depth of wear indication were studied using CIVA simulation platform. It was found that eddy current signals wound give a good prediction for the morphology of wear, and the volume loss of the wear indication could be calculate through the combination of model computation and data analysis method. CIVA simulation results were agreed with the experimental data, the simulated data amplitude-depth curve was a linear equation when the flaws were regular and standard artificial defects.

Analysis of Spent Fuel Pool Cooling Problems of CPR1000 NPP

Wang Chuang, XioNG Dongqing, Xu Guangzhen, Li Juan

2014, 35(4): 90-93. doi: 10.13832/j.jnpe.2014.04.0090

Abstract(23) PDF(0)

Abstract:
The spent fuel pool of the CPR1000 type nuclear power plant（pressure water reactor）may loses it’s cooling function in some conditions. The author uses Ling Dong NPP as the example to analysis the design basis, system functions and defects of PTR, and gives some suggestions to improve the PTR with considering the requirements of technical specification. The results show that the spent fuel pool cooling problem is because that the designer does not consider the redundancy of the PTR completely. The author suggests to improve PTR cooling capacity and reduces the risk of losing cooling of spent fuel pool.

Nuclear Power Commissioning Problem Treatment and Design Analysis

Zhang Wenfeng, Duan Yongqiang, ZeNG Chang

2014, 35(4): 94-97. doi: 10.13832/j.jnpe.2014.04.0094

Abstract(25) PDF(0)

Abstract:
During nuclear power plant construction, commissioning is a very important stage. Through commissioning it finds the problems of the whole system, verifies the correction of design, equipments, and the installation, so to guarantee plant work as the performance requirements. This article states the commissioning purples, commissioning program management, commissioning results analysis and treatment. Using the NSSS ASG commissioning as example, it states the main content of commissioning and utilizes the flowmaster software handling the problem of deaerator flowrate excessing base requirement. That deals with the problem quickly and fast.

Study on Pressure Leak Test for Post-irradiation Fuel Rods

Kuang Liuwei, Jiang Linzhi, Ren Liang, Guo Chengming, Yu Feiyang, YiN Chunyan

2014, 35(4): 98-101. doi: 10.13832/j.jnpe.2014.04.0098

Abstract(21) PDF(0)

Abstract:
Carry out the research on the technology of pressure leak test to define the track mediator and experiment pressure; Considering the sealing, retain and fix, Leakage monitoring system, tracer medium and pressure of the post-irradiation fuel rods and other factors in hot cell, designing and processing the pressure leak device, then proving the pressure tightness indoor and outdoor hot cell; through the pressure leak test to the artificial simulative fuel rod, which indicated that the pressure leak device was effective and the method was feasibility.

Study on Low Temperature Thermal Aging Behavior of 316LN Austenitic Stainless Steel Weld

Luo Qiang, CheN Yong, Qiu Shaoyu, Zhou Jun, He Kun

2014, 35(4): 102-105. doi: 10.13832/j.jnpe.2014.04.0102

Abstract(25) PDF(0)

Abstract:
Austenitic stainless steel welds in PWR main pipe are susceptible to thermal aging embrittlement at reactor operating temperature of 280~325 ℃after a long service life.In order to obtain thermal aging embrittlement of 316 LN austenitic stainless steel weld at reactor operating temperature, In this study, thermal aging embrittlement of types 316 LN austenitic stainless steel weld with 10.7% ferrite was investigated on basis of embrittlement properties after aging up to 6,000 h at 325,365 and 400℃. The result show that,no change in the main alloy elements contents in ferrite and austenite.Large increase in the microhardness of weld was observed with no change in austenite hardness. The impact toughness was significantly degraded after aging.

Analysis for T2 Test Scheme of Reactor Protection System Based on MELTAC Platform

Wang Qiang, Zhu Wen, Zhang Lan, Wang Gang

2014, 35(4): 106-109. doi: 10.13832/j.jnpe.2014.04.0106

Abstract(17) PDF(0)

Abstract:
Hongyanhe nuclear power plant is the first CPR1000 reactor, whose reactor Protection System（RPS） is realized by digital I＆C technology called MELTAC platform. T2 processing channel test is one of important part of periodic test for RPS, whether it is necessary or not and how to perform have become the controversial topic all the time. The article describes the final agreed T2 test scheme of RPS based on MELTAC, furthermore, the correctness of RPS T2 test is demonstrated on the basis of IEC60671-2007 in details.

Evaluation of Turbulence Model Performance in Predicting 3×3 Bundle Flows

Wang Hongyan, XioNG Jinbiao, Liu Yu, CheNG Xu

2014, 35(4): 110-113. doi: 10.13832/j.jnpe.2014.04.0110

Abstract(19) PDF(0)

Abstract:
In this paper, the flow characteristics throughout a PWR 3x3 rod bundles was investigated, and four turbulence models including SST,SSG-RSM,BSL-RSM and ω-RSM were used to simulate in steady state and then evaluated with the laser Doppler measurement data of the flow. The comparisons demonstrate that BSL-RSM and SSG perform better relatively though still having some errors in predicting the turbulent fluctuation.

Optimization Analysis of High-Energy Piping System

Wang Xinjun, Lu Xifeng, ai Honglei, He Feng, Yuan Yanli, LÜ Yongbo

2014, 35(4): 114-118. doi: 10.13832/j.jnpe.2014.04.0114

Abstract(21) PDF(0)

Abstract:
The mechanical analysis results usually cannot match the requirement of criterion or specification for new layout high-energy piping system. Then the design of the piping system or the location or function of supports on the piping should be improved until requirements of criterion and specification are matched. Then the space should be checked to insure that the piping and its supports can be laid down. Repeat improvement and harmonization are needed in order to find a satisfying piping layout. An optimization analysis method which can reduce the number of improvement and harmonization is introduced in this paper. This method is used in a new high-energy piping system mechanical analysis. The process of high-energy piping mechanical analysis, the method for special stress indices calculation and the evaluation of analysis results are also introduced in this paper. The results show that the optimization analysis method is an effective and practical method used for new high energy piping system analysis.

Code Development for Debris Bed Cool Ability Characteristics

Wu Xiaoli, Liu Bo, Zhang Yapei, Tian Wenxi, Su Guanghui, Qiu Suizheng

2014, 35(4): 119-122. doi: 10.13832/j.jnpe.2014.04.0119

Abstract(18) PDF(0)

Abstract:
In this paper, we developed a two-dimensional steady-state core debris bed analysis program based on the finite element method from the basic mass and momentum equations for porous media debris bed.The effects of debris bed’s parameters such as the system pressure, particle diameter, porosity and other relevant parameters on the cool ability of the debris bed were also analyzed. Then provide a theoretical basis of the integrity of the pressure vessel.

Automatic Control Method for Reactivity Insertion from Nuclear Reactor Cold Start-up to Critical

Yang Huailei, Yu Ren, CheN Zhi, Liao Longtao

2014, 35(4): 123-128. doi: 10.13832/j.jnpe.2014.04.0123

Abstract(22) PDF(0)

Abstract:
Based on the manual operational experience in the reactor cold start-up, the paper presents two automatic control methods for reactivity insertion from nuclear reactor cold start-up to critical. The effect of the control algorithm is verified by a simulation experiment, with which the point reactor model with six delayed neutrons is utilized. The results show that the designed method can control the reactor to be critical automatically by inserting the reactivity timely and properly.

Research of Optimization Design of Jet Pump’s Hydraulic Characteristics

Zhao Xuecen, Wang Jintao, Liu Songya, Liu Lizhi

2014, 35(4): 129-132. doi: 10.13832/j.jnpe.2014.04.0129

Abstract(20) PDF(0)

Abstract:
The jet pump liquid in limited space with flow in irregular area belongs to high Reynolds number strong shear turbulent jet, therefore at present there is no complete and accurate formula that can express the flow of two kinds of different performance fluid mixing and diffusion in the jet pump. Based on the study of design and optimization method of jet pump by scholars at home and abroad, this paper selected the area ratio, throat length to diameter ratio and distance between throat and nozzle as the key parameters,with the goal to improve the efficiency and carry out the numerical calculations,and finally achieve a jet pump design,which efficiency increased by 6% than the initial design results. Consequently, the pump power is reduced.

Numerical Simulation Research of Velocity and Temperature Fields in Lead-bismuth Eutectic

Zhou Tao, Zou Wenzhong, Huo Qijun, Li Yunbo

2014, 35(4): 133-136. doi: 10.13832/j.jnpe.2014.04.0133

Abstract(25) PDF(0)

Abstract:
On the basis of two-phase flow and the heat transfer theory, the velocity and temperature of the lead-bismuth was calculated with gas-injection or without gas-injection by FLUENT code, in the calculation segment. The radial velocity profiles and temperature profiles were acquired at different cross section. Analysis results show that: no matter the air is existence or not, the velocity distribution and the temperature distribution have a good symmetry. Temperature rise of the lead-bismuth liquid with gas injected is larger than without air. It is shown that the gas-injection is good for lead-bismuth liquid heat transfer at the calculation segment.

Study on Two-color Planar Laser Induced Fluorescence Thermometry

Li Shaodan, LiN Yuansheng, Tan Sichao, Gao Puzhen

2014, 35(4): 137-141. doi: 10.13832/j.jnpe.2014.04.0137

Abstract(26) PDF(0)

Abstract:
Many of the convection heat transfer process are involved in the research of nuclear reactor thermal hydraulics. To experimentally determine the variation of the temperature filed in those processes is important for the design and safety operation of the nuclear reactor. The application of the two-color planar laser induced fluorescence（PLIF） in the measurements of fluid temperature distribution is discussed in the paper. The laser dyes used here is rhodamine B（RhB） with negative temperature coefficient and fluorescein 27（Fl27） with positive temperature coefficient. The beam of the laser light is adjusted to laser sheet by using the lens group. The fluid with dyes is excited by this laser sheet in a specific plane and temperature dependent fluorescence is released. The temperature field of the plane can be determined through the intensity information. Some technical aspects encountered in the application of the two-laser PLIF are discussed in the paper, such as the spectra characteristic of the dyes and the separation of the spectra. The calibration temperature is higher than the water saturation temperature（at atmosphere pressure）.

Research on Tally Data Decomposition Algorithms Based on Reactor Monte Carlo Code RMC

Liang Jingang, Wang Kan, Yu Ganglin, She Ding, Chai Xiaoming, Qiang Shenglong, Yao Dong

2014, 35(4): 142-146. doi: 10.13832/j.jnpe.2014.04.0142

Abstract(33) PDF(0)

Abstract:
The applications of Monte Carlo method in reactor physics analysis are restricted due to excessive memory demand in solving large-scale problems, while data decomposition is supposed to be a remedy. In this paper quantitative memory requirements in MC simulation are analyzed. Two types of tally data decomposition algorithms, which utilize synchronous and asynchronous communications, are designed and implemented in Reactor Monte Carlo code（RMC）. Numerical tests are carried out to evaluate performance of new algorithms. It is shown that tally data decomposition algorithm can reduce memory size effectively while parallel efficiency of the code is not affected.

A Comprehensive Centralized Control System for Radiation Waste Treatment Facility

KoNG Jingsong

2014, 35(4): 147-150. doi: 10.13832/j.jnpe.2014.04.0147

Abstract(18) PDF(0)

Abstract:
A comprehensive centralized control system is designed for the radiation waste treatment facility that lacking of coordinated operational mechanism for the radiation waste treatment. The centralized control and alarm linkage of various systems is implemented to ensure effectively the safety of nuclear facility and materials, improve the integral control ability through advanced informatization ways.

Application of Explosive Cutting and Control of Radioactive Dust in Decommissioning of Nuclear Facilities

KoNG Jingsong

2014, 35(4): 151-154. doi: 10.13832/j.jnpe.2014.04.0151

Abstract(26) PDF(0)

Abstract:
The characteristics of nuclear facilities decommissioning are analyzed, the application of controlled blasting techniques is described in demolition activities of nuclear facilities decommissioning, then the mechanism, characteristics and application instance of explosive cutting are elaborated. The problem of radioactive dust during explosive cutting is discussed, including source and quantity of radioactive dust. The controlling and protective measures are raised at the same time.

Treatment and Disposal of Radioactive Contamination Boric Acid Wastewater

KoNG Jingsong

2014, 35(4): 155-158. doi: 10.13832/j.jnpe.2014.04.0155

Abstract(27) PDF(0)

Abstract:
The characteristics and environmental effects of the radioactive contamination boric acid Wastwater are described in this paper. Three schemes of treatment and disposal are raised on radioactive contamination boric acid wastewater,and are analyzed and compared in aspects of technology and economics.

Design of Integrated Natural Cycle Evaporator for Low and Medial Level Radioactive Liquid Waste

MA Xingjun, Li Binglin, Chen Li, CheN Xianlin, Zhang Yongkang

2014, 35(4): 159-162. doi: 10.13832/j.jnpe.2014.04.0159

Abstract(25) PDF(0)

Abstract:
The paper mainly introduces a new natural cycle evaporator with central downcomer, which is used to process the low and medial level radioactive wastewater in the reactor operation. The designed evaporator is integrated with the process of evaporated concentration, cleaning demister, and entrainment eliminator. All the assemblies are set in a single pressure vessel. Through the integrated collocation, the cleaning water form demister is reasonably treated, the equipment required less space, the whole evaporated system is simplified and can be handled easily. Through the multi-effect purification, the quantity of entrained mist and volatile product in the produced steam significantly reduced. The proposed evaporator can improve the purification and the decontamination factor of radioactive wastewater, which is accorded with the radioactive waste minimization principle.

Measurement of Crack Growth Rate of Alloy X-750 Based on DCPD Method

Du Donghai, Zhang Lefu, Chen Kai, Yu Lun, Shi Xiuqiang, Xu Xuelian

2014, 35(4): 163-167. doi: 10.13832/j.jnpe.2014.04.0163

Abstract(18) PDF(0)

Abstract:
This paper studies the fatigue cracking in air and stress corrosion cracking（SCC） in high temperature and high pressure water of alloy X-750 respectively based on direct current potential drop（DCPD） method. The testing method using compact tensile specimens is briefly described, and the mechanism and process that control the crack tip chemistry are discussed according to the slip-film rupture-oxidation model. The effects of dissolved oxygen and SO₄^2- impurities on crack tip chemistry are also discussed.

Analysis and Evaluation of Airborne Radioactivity Concentration in PWR Nuclear Plants

Bai Ying, Wang Yiwei, Zhou Wei

2014, 35(4): 168-171. doi: 10.13832/j.jnpe.2014.04.0168

Abstract(27) PDF(0)

Abstract:
This paper introduces the basic calculation method of PWR airborne radioactivity concentration, and analyzes the airborne concentration of noble gases, fission products, corrosion products and tritium during each phase such as normal operation, residual heat removal and RPV opened in terms of relative 1000 MWe PWR’s design experiences. Especially, the detailed calculations are provided in that concerning the phenomenon of increased airborne radioactivity concentration during residual heat removal due to the iodine spike by fuel cladding failure and the corrosion products burst by oxidation operation respectively. Finally, this paper describes some improvement suggestions for internal exposure protection and ventilation design based on the tendency of PWR airborne radioactivity concentration during each phase.

Estimation of Graphite Dust Resuspension in Steam Generator of HTR-10

Zhang Tianqi, PeNG Wei, Yu Suyuan

2014, 35(4): 172-176. doi: 10.13832/j.jnpe.2014.04.0172

Abstract(26) PDF(0)

Abstract:
According to the design instruction of HTR-10, helium flow velocity along with physical parameters are calculated in cold heating period, nucleate boiling, film boiling and overheating heating region of steam generator- the most remarkable amount of graphite dust deposition equipment. Improved Vainshtein model is applied to estimate resuspension fraction. Results show that the friction velocity along with resuspension fraction increases gradually in the height direction; time-related resuspension fraction has a short-term effect, and larger graphite particles turns out to be more easily removed in the same velocity.

2014 Vol. 35, No. 4