核动力工程

Production of Pseudo Fission-Product for Transmutation of Fast Reactor

HU Wen-chao, LIU Bin, OUYANG Xiao-ping, HUANG Liming, WANG Kai, FU Juan, MENG Hai-yan

2013, 34(6): 1-4.

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Abstract:
Firstly, the NJOY nuclear Date Processing System is used to process the fissional nuclides from ENDF/B-VII for sections of fast reactors cross. We use the Matlab software to process the cross section date from NJOY and get multigroup pseudo fission-product for fast reactor transmutation. The fissile parent nuclides are U-235.The cross section data include total cross section, elastic scattering cross section, radioactive capture cross section, the cross section of producing an alpha particle（n, a）, the cross section of producing a proton（n, p）, the cross section of producing one neutron（n, n） and the cross section of producing two neutrons（n, 2n）. The paper uses the neutron energy spectrum to deal with the multigroup cross sections from a fast reactor and validates the multigroup cross section from MCNP and the date form NJOY. It is concluded that the processed cross section date can be used for the accurate calculations of fast reactor transmutation.

Coupled Neutronics/Thermal-Hydraulics Analysis of PT-SCWR Fuel Assembly

LIU Wei, BAI Ning, DAN Jian-qiang, ZHU Yuan-bing, ZHANG Bo, LI Jing-gang

2013, 34(6): 5-9.

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The neutronics code Monte Carlo N-Particle code（MCNP） and the sub-channel analysis code Advanced Thermal-Hydraulics Analysis Sub-channel（ATHAS） are used in a coupled way to better understand the design characteristics of PT-SCWR fuel assembly. The developed coupled code is reasonable and effective. The results show that: The distribution of PT-SCWR fuel enrichment has strong effect on radial power distribution. It is available to flatten the radial power distribution by changing the fuel enrichment of the inner and outer fuel rods. Moderator thickness has strong effect on the axial power distribution: when it is 25 cm, the axial power distribution approaches a cosine configuration.

Deep Subcritical Levels Measurements Dependents upon Kinetic Distortion Factors

PAN Shi-biao, LI Xiang, FU Guo-en, HUANG Li-yuan, MU Ke-liang

2013, 34(6): 10-12.

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The measurement of deep subcritical levels, with the increase of subcriticality, showed that the results impact on the kinetic distortion effect, along with neutron flux strongly deteriorated. Using the diffusion theory, calculations have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that epithermal neutron distributions are strongly affected by kinetic distortion. Subcriticality measurements in four different rod-state combination at the zero power device was carried out. The test data analysis shows that, with increasing subcriticality, kinetic distortion effect correction factor gradually increases from 1.052 to 1.065, corresponding reactive correction amount of 0.78β_eff~3.01β_eff. Thus, it is necessary to consider the kinetic distortion effect in the deep subcritical reactivity measurements.

Analysis on CEA93 Library

GUO Fengchen, WANG Jia-chong, LU Wei, YAO Dong, LIU Xu-dong

2013, 34(6): 13-17.

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In this paper, CEA93 library was studied, which provides the basic nuclear data for APOLLO2-F program of science system. The properties of structure for CEA93 library and the methods of maintenance were introduced. By comparing with the WIMS library, the results show that CEA93 library can be read more quickly than WIMS library because of index table and be maintained easily by job command stream.

Study on Development of ASCFR1.0/MC and Initial Calculation of Moderator Temperature Effect for ASCFR

LI Zhi-feng, YU Tao, XIE Jin-sen

2013, 34(6): 18-23.

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In order to develop the temperature-dependent point-wise cross section library for the advanced supercritical water cooled fast reactor, the JEZEBEL fast neutron benchmark was used to analyze the important parameters of the NJOY code and compare the different effects of the input parameters. Then the most reasonable parameters were selected to develop the ASCFR1.0/MC which was based on the ENDF/B-VII.1. Finally, the Doppler coefficient benchmark was applied to test and verify the ASCFR.10/MC. In conclusion, the precision of the ASCFR1.0/MC were perfect. The resulted library can be used in the analysis and verification of the temperature effect in the Advanced Supercritical Fast Reactor（ASCFR）. Finally, the moderator effect of ASCFR was calculated with MCNP using the ASCFR1.0/MC library, and the moderator effect of the ASCFR is positive.

Study on 3D Nonlinear Seismic Analysis of Reactor Structure

LIU Wen-jin, WU Wan-jun, LAN Bin, ZHANG Li-ping, HUANG Xuan

2013, 34(6): 24-26,35.

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In this paper, the seismic analysis software of reactor structure is studied, a necessarry connect procedure and the loads combination and sensitivity analysis procedure are developed, and the commercially available software used in the seismic analysis, the imported software and the developed loads combination and sensitivity analysis procedure are combined to set up a software system. The modeling and analysis techniques used in 3D nonlinear seismic analysis of reactor structure are studied, and the necessity and feasibility of conducting the seismic analysis of reactor structure with 3D nonlinear timing method is explored. The reactor structure 3D nonlinear model is established, and the seismic analysis and calculation of reactor structure is conducted by the direct integral calculus method and using ANSYS software. The loads needed in the stress analysis assess and the acceleration timing required by CRDM seismic authenticate experiment are proposed.

Optimized Design for TWR Assembly by CFD Calculations

LU Jian-chao, LU Chuan, YAN Ming-yu

2013, 34(6): 27-30.

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High temperature difference in travelling wave reactor bundle was found in the previous work. It could not be used in bundle design. Various analysis focused on helical wrapped wires and assembly housing was carried out by CFD calculation which found that the helical wrapped wires could influence the temperature differences while the effect was not obvious. Adding the strips and fillets on the assembly housing could optimize the thermal characteristics greatly, which can be used in the TWR assembly design.

Research on 90° Elbow Structure Optimization for Main Coolant Piping of CPR1000 Nuclear Power Plant

LI Quan-bing, ZHANG Xing-hui, REN Hong-bing, ZHU Ling-ju, DUAN Yuan-gang

2013, 34(6): 31-35.

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For main coolant piping 90° elbow of CPR1000 nuclear power station, the theoretical thickness distribution was researched, and an optimized elbow structure was designed. The main difference between the optimized elbow and present design elbow is as follows: for the present design 90° elbow, the 120° of internal rounding area are thickened, then transitioned to the mid surface, while the external rounding area keeps normal thickness; for the optimized structure, based on the departure of internal and external circle center of elbow section, wall thickness is designed with regularly changing from higher thickness in the internal rounding area to less thickness in the external rounding area, and keep the internal section a regular circle. This optimized elbow structure can realize mechanization, and improve efficiency and machining precision.

Effect of Dynamic Structure Stiffness on Rotor Characteristics of Vertical Pump

OU Ming-xiong, WANG Yan, YAN Jian-hua, SHENG Jiang

2013, 34(6): 36-39.

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The support structure dynamic stiffness of rotor of vertical circulation pump used in AP1000 nuclear power plant was analyzed with finite element method, and the results demonstrated that the dynamic stiffness value decreased with frequencies and suddenly drop at resonant frequency. Based on this relation result, a contrast modal analysis about three situations which consider rigid bearing stiffness, sliding bearing stiffness and support structure dynamic stiffness respectively were done for the rotor model’s dynamic characteristic. The result demonstrated that the difference between 1st lateral vibration frequency of rotor are not obvious, the frequencies were 28.1,21.7 and 20.2Hz, nevertheless the higher vibration frequency difference between the three situations are obvious. The higher vibration frequencies decreased apparently with the support stiffness considered.

Nuclear Auxiliary Pipe Whip and Whip Restraint Stress Analysis

YUAN Feng, LU Yongbo, AI Honglei, YUAN Yanli

2013, 34(6): 40-42.

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Abstract:
According to RCC-P, it is obligatory to adopt effective measure to prevent important system and device of nuclear power plant from damage caused by the pipe whip which contains high energy liquid. To equip protective shield（such as whip restraint） to separate important components is a common effective measure to avoid pipe whip. In order to verify if the whip restraint can withstand the pipe whip without inducing failure or large displacement, stress analysis of whip restraint is performed based on energy equilibrium method, and the results show that the whip restraint can withstand the pipe whip and its stress can meet the requirement of RCC-M Vol.ZF.

Analysis of Nonlinear Response of Cracked Beams

CAI Feng-chun, ZHANG Yi-xiong, WANG Ming-li, GONG Jun-yong

2013, 34(6): 43-47,51.

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Based on the Lagrange’s equation, the equation of motion for pinned-pinned cracked beam with hollow section is derived. The equation of motion takes account of the geometric nonlinearity and opening and closing of the crack. Based on a numerical method, the superand sub-harmonic resonances are studied when the cracked supported pipes conveying fluid are excited by a harmonic force and the conclusions are the same as the existed researches. Then the effect of the geometric nonlinearity on response of beam with hollow section under harmonic force with large amplitude are further studied. These researches are useful for the cracks detection in the beam with hollow section.

Modeling Research on PWR Fuel Assembly Lateral Non-linear Characteristics

RU Jun, XIAO Zhong, PU Zeng-ping, YONG Jing, HUANG Chun-lan, GU Ming-fei, SU Min

2013, 34(6): 48-51.

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In PWR fuel assembly design, the fuel rod is usually clamped by grids to maintain its position laterally and axially. Fuel rod may slide relative to grid when fuel assembly bows, which brings non-linear characteristics to fuel assembly laterally. This paper includes the analysis of typical fuel assembly grid to fuel rod clipping system. Based on the analysis and the test, the characteristic curve of bending of the clamping system is considered as the combination of several sliding elements and spring elements, mechanical simulation of the clamping system is consequently established. Finally, the comparison of calculation results and test results are presented, thus the mechanical simulation is qualified. The model of clamping system is adopted by fuel assembly lateral model, which exhibit a good accordance to fuel assembly lateral pluck test results.

Numerical Research on Hydraulic Performance of ACPR1000⁺ Reactor

ZHANG Ming-qian, DUAN Yuan-gang, YU Xiaolei, ZHANG Ping

2013, 34(6): 52-54,60.

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The pressure drop and flow distribution were obtained by calculations for the scaled reactor hydraulic model. The computations were compared with the experiment conducted at the same conditions in order to determine the capabilities of the codes to model accurately the physical phenomena which occur in the ACPR1000⁺ reactor vessel. The results indicated that this numerical modeling was validated to allow an accurate evaluation of the pressure drop and flow distribution in nuclear reactor vessel for engineering applications.

Development and Abecedarian Verification of Reactor Core Physic and Thermal-Hydraulic Analysis Software Package DCNMC

ZHOU Xu-hua, CAI Qi, WANG Deng-ying, LI Fu, YI Xiong-ying

2013, 34(6): 55-60.

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With a view to the characteristics of the specific type nuclear reactor,five reactor core physic and thermal-hydraulic codes are integrated into the software package as calculation kernel. In the software package, the Dragon code is applied to calculate the homogenized few group constant for fuel assembly, the CITATION code, NGFM code and MCNP code are applied for core physic calculation with different methodology, the COBRA code is applied for core thermal-hydraulic calculation. In addition two homegrown codes named as DOCS and DCNMC are used as data transmission interface tool and calculation management tool. Then, a generic reactor core physic and thermal-hydraulic analysis software for the specific type nuclear reactor is developed. A calculation model on some reactor core is established. Then, the accuracy of the code system and model is verified, and the results indicate the accuracy meets the requirements.

Hydraulic Performance Analysis of Nuclear Axial Pump Based on CFD Method

JIANG Hong, HUANG Wei

2013, 34(6): 61-65.

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In this paper, a nuclear axial pump is simulated based on CFD method for its hydraulic performance. The k-ε Reynolds turbulent model is used to simulate the three-dimensional fluid field in this pump. The applied power is calculated based on the torque, and actual power based on the head and flowrate. The analysis results show good agreements with the test values. The maximum deviation for head, shaft power and efficiency is about 6.3%, 4.9% and 2.2%, respectively, and the minimum deviation is about 3.4%, 1.4% and 1.0%, respectively. Based on the analysis results, the methods used in this paper can be applied to design and optimize the nuclear pumps.

Research on Effect of Two-Phase Interfacial Friction Characteristics on Quench Front Velocity in Narrow Channel

ZENG Wei, ZHU Li, LIU Song-tao, YU Hong-xing, SUN Yu-fa

2013, 34(6): 66-69,74.

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This paper simulates the Saxena experiment by the RELAP5 code, and the quench front velocity is overestimated during the simulation. Then this paper constitutes the narrow channel drift flux interfacial friction model based on the Momentum conservation equation and rectangular narrow channel drift flux velocity founded by Griffith. This model reflects the effects on the interfacial friction characteristics from the narrow geometry, so the quench front velocity can be estimated well, meanwhile one of the key factors for the quench front velocity during the narrow channel reflood has been found, which is also the important basis for developing the narrow channel reflood model.

Study on Model of Onset of Nucleate Boiling under Natural Circulation in Narrow Rectangular Channel by Using Unascertained Mathematics

DUAN Jun, ZHOU Tao, LI Jing-jing, JU Zhong-yun, LIU Ping, HUANG Yan-ping, XIAO Ze-jun

2013, 34(6): 70-74.

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Onset of nucleate boiling（ONB） is the key point of the flow boiling heat transfer process under natural circulation. According to the experimental data, the paper obtains the unascertained mathematical model of ONB’s heat flux in narrow rectangular channel based on the theory of unascertained mathematics. The paper analyzes the results of the calculation, then gets three sections: Available, Checkable, Rejectable. The result shows that the unascertained mathematical model is good at dealing with the unascertained information, can enhance the understanding of the experimental data, obtain an accurate and complete description on ONB’s heat flux.

Analysis of Rolling Induced Additional Effect on Two-phase Flow

TIAN Dao-gui, SUN Li-cheng, YAN Chang-qi, LIU Guo-qiang

2013, 34(6): 75-78,91.

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Both the additional inertial force as well as the pressure drop induced by it on a two-phase flow system in tubes with vertical, horizontal and tilted layouts had been derived in detail. And correlations were given for calculating the additional force and pressure drop under the three layout conditions. Comparing the fluctuation amplitude of additional inertial force and gravity in flow direction, it was found that the gravity is a critical factor in determining the flow characteristics. The correlation for calculating the additional pressure drop in a vertical pipe were evaluated against the experimental data, showing that the correlation could well predict the variation of additional pressure drop in rolling motion.

Numerical Simulations on Flow Field Structure and Free Surface Behavior of ADS Windowless Target

HU Chen, SU Guan-yu, GU Han-yang, CHENG Xu

2013, 34(6): 79-82.

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Free surface flow is one of the most important parts in the studies of windowless target in Acceleration Driven System（ADS）. CFD simulations were performed in order to investigate the hydraulic behaviors of the free surface flow in the optimal structure of windowless target designed by Shanghai Jiao Tong University. Volume of fluid（VOF） method and large eddy simulation（LES） were applied in transient calculations on FLUENT platform at Re=35,000~80,000. The simulation results achieve the variation characteristics of free surface stability, free surface width, vortex structure, vortex stagnation length and pressure along conical channel with the increase of Re.

An Assessment Methodology of Thee-Layers Melt Configuration during IVR for AP1000

XIANG Qing-an, GUAN Zhong-hua, DENG Chun-rui, CHEN Bao-wen

2013, 34(6): 83-87.

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In-Vessel Retention Analysis Model（IVRAM）is formulated to assess the thermal response of the Two-Layers melt configuration and Thee-Layers melt configuration during IVR for AP1000 with anticipative depressurization and reactor cavity flooding in severe accident. The results of the calculations for the melt configuration show that the presence of the heavy metallic layer at the bottom of the lower head decreases the thickness of the top light metallic layer and consequently increases the risk of focusing effect, the focusing effect of the light metallic layer increases with the increase of the thick of heavy metallic layer.

Characteristics and Application Study of AP1000 NPPs Equipment Reliability Classification Method

GUAN Gao

2013, 34(6): 88-91.

Abstract(18) PDF(1)

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AP1000 nuclear power plant applies an integrated approach to establish equipment reliability classification, which includes probabilistic risk assessment technique, maintenance rule administrative, power production reliability classification and functional equipment group bounding method, and eventually classify equipment reliability into 4 levels. This classification process and result are very different from classical RCM and streamlined RCM. It studied the characteristic of AP1000 equipment reliability classification approach, considered that equipment reliability classification should effectively support maintenance strategy development and work process control, recommended to use a combined RCM method to establish the future equipment reliability program of AP1000 nuclear power plants.

Research on Transfer Rule of the Monitoring of Operator in Digital Main Control Room of Nuclear Power Plant

ZHANG Li, LI Lin-feng, LU Zhang-shen, CHEN Qing-qing, LI Peng-cheng, HUANG Wei-gang, DAI Zhong-hua, HUANG Yuan-zheng

2013, 34(6): 92-96.

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In the digital main control room of nuclear power plants, monitoring the operating status of the system of reactor is not only one of the most important tasks of the operators, but also the basis and premise of controlling the system of reactor running correctly. After analyzing, inducing, summarizing the data obtained, we found the operators’ monitor behavior could be classified as procedure transfer, abnormal transfer, and exchange transfer. The times of exchange transfer is 29% of the total transfer times, abnormal transfer is 14%, regulation transfer is 36%, and others are 21%.

Preliminary Analysis of Power Control Characteristics for ADS Equipment in China

SUN Zhang-yi, ZHAO Fu-yu

2013, 34(6): 97-101.

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A neutron and core thermal dynamic model is built to develop a computer calculation code CICC to simulate the transient behavior of core power. Simulation results show that the power control system works well when k_eff locates in 0.94 and 0.98 and serious power overshoot will appear when k_eff ＞ 0.98. Control effects could be largely improved after introducing a feed-forward signal into the PID controller, which could effectively eliminate the reactivity disturbance lower than 1×10^-3. The results also suggest that the reactivity will oscillate divergently if control rods in core when there is accelerator flow loss accident. Calculation result points out that the optimal match of Q and k_eff locates in the interval of（0.94, 0.96） and（1.2×10¹⁰ cm^-3 s^-1, 1.85×10¹⁰cm^-3 s^-1） respectively.

Simulation Experimental Research on Passive Residual Heat Removal Performance for Nuclear Seawater Desalination Reactor

NIE Chang-hua, XU Shi-jie, LIU Xun, ZHUO Wen-bin, LI Zhang-lin, ZHENG Hua, LI Peng-zhou, YU Qing-lin

2013, 34(6): 102-106.

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Simulation experiments were conducted on the integral simulation test facility with scaling ratio 1/45 for nuclear seawater desalination reactor, the performance of the passive residual heat removal from the core was presented in the paper. The experimental results demonstrated that residual heat can be removed from the core effectively after the reactor was shut off in emergency because some accidents take place, such as station blackout accidents. The paper also presented the RELAP5/MOD3.2 code simulation results for the experimental conditions, and the calculation results of the code is well accordance with the experimental results.

Conceptual Design of Passive Residual Heat Removal System for 10 MW Molten Salt Reactor

SUN Lu, SUN Li-cheng, YAN Chang-qi, FA Dan, ZHAO Xing-bin

2013, 34(6): 107-110.

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A conceptual design of the passive residual heat removal system which can meet the safety requirement was developed for 10 MW Molten Salt Reactor Experiment（MSRE） designed by ORNL. The constitution, main components and main design parameters of the system were presented. Thermal-hydraulic behaviors such as natural circulation and heat removal ability were numerically analyzed. The results show that, the system can meet the design requirements of heat removal and has enough safety margin. Furthermore, the heat removal rate of the system is approximate to the decay heat generation rate, making the temperature of the molten salt decrease steadily.

Effect of Guide Vane Outlet Position on Nuclear Reactor Coolant Pump under Gas-liquid Two-phase Condition

FU Qiang, WANG Xiu-li, YUAN Shou-qi, XI Yi, ZHU Rong-sheng

2013, 34(6): 111-114.

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To study the inner fluid field effect caused by guide blade outlet side position of reactor coolant pump in two-phase condition, three dimensional software Pro/E, non-structural mesh software ICEM, Reynolds time-averaged N-S equations, standard k-ε turbulent model, and CFD software CFX were used to calculate the three-dimensional turbulent flow of nuclear model pumps with different guide blade outlet side position. The void fraction, two-phase relative velocity distribution in the center section of vane and volute are analyzed. The results show that the effect of different guide blade outlet side position to the performance of reactor coolant pump is obvious. It revealed that the flow performance is best when guide blade outlet side is in the centre vertical plane of volute. Research results reflect the internal flow field characteristics of reactor coolant pump in a certain extent, and it also could provide the beneficial reference of the place of vane in the volute of nuclear reactor coolant pump.

Cavitation Characteristics for Impeller of Centrifugal Charging Pump of M310 Nuclear Power Plants

FU Qiang, YUAN Shou-qi, ZHU Rong-sheng, WANG Tao, JIANG Xu-song

2013, 34(6): 115-120.

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In order to study the first stage impeller cavitation characteristics of the centrifugal charging pump, Pro/E and ICEM software were respectively used as the tools for three-dimensional modeling and meshing of the hydraulic machinery, and a numerical calculation method was determined based on Reynolds time-averaged N-S equations, RNG k-ε two-equation turbulence model, SIMPLEC algorithm and cavitation model, and last unsteady numerical simulation and experimental comparative study of the first stage impeller cavitation characteristics of the centrifugal charging pump were finished. The results showed that: when initial cavitation happened, gas volume fraction less affected the impeller internal pressure; when critical NPSH reached, inlet cavitation changes near the blade has great effect on the pressure; simulation and experimental results of critical NPSH were respectively 7.35 m and 7.58 m, and error was about 2%, and they all satisfied the nuclear power specifications of ≤ 7.8 m.

Precise Prefabrication of Primary Pipes Automatic Welding in CPR1000 Nuclear Island Installation

ZHANG Ke, LI Jia-bin, WANG Dong

2013, 34(6): 121-124.

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The precise prefabrication of primary pipes is a prerequisite for the successful implementation of primary pipes automatic welding technology. Based on the geometric characteristics of the primary equipments（including RPV, SG and RCP） and primary pipes, it presents a pertinent completion size measurement scheme for calculating the processing data of primary pipes weld grooves, in order to digest the primary equipments manufacture error; and then through the development of the weld grooves processing program to ensure the machining accuracy of the weld grooves, and ultimately to ensure the precise prefabrication of primary pipes automatic welding in CPR1000 nuclear island installation.

Position Design for Instrumentation Penetration of Reactor Vessel Based on None-Alignment Analysis

XIA Xin, DU Hua, LI Ning, XU Bin, LI Yan, ZHAO Wei

2013, 34(6): 125-127.

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The position of the instrumentation penetrations of the reactor pressure vessel is the most important factor that influences the none-alignment between instrumentation penetrations and instrumentation tube of reactor internals. The more the none-alignment, the more wear the flux thimble will suffer, which will damage the flux detector and cause economic losses. This paper analyses the none-alignment between instrumentation penetrations and instrumentation tube, and combines the feedback of the experience in nuclear power plants having been built or being build, puts forward the reasonable design value for the instrumentation penetrations’ position.

Design Improvement and Open Item Analysis of Containment Sump Strainer in QINSHAN Phase II Extension Project

ZHANG Wei, GONG Zhao, ZHU Jing-mei, ZHU Ming-hua, QU Chang-ming

2013, 34(6): 128-131.

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Considering the disadvantage of the small filter area and insufficient filter capacity of the conditional design sump strainers, China National Nuclear Safety Administration has raised the new surveillance requirement for the design and corresponding analysis of sump strainer to ensure the safety function. This paper discusses the clogging and design improvement of the sump strainer in Qinshan Nuclear Power Plant Phase II Extension Project and analyzes the consistency between the design of sump strainers and the requirement of RG 1.82. The open item of the sump strainer has also been discussed.

Design and Application of Concrete Floor Modular Hoisting in Nuclear Power Plants

YU Xi-nian, YANG Ying-yu, WANG Jian-guo

2013, 34(6): 132-135.

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The whole modular hoisting of concrete slab is a major hoisting project for construction of nuclear power plants. Concerning the actual situation that the geometric center of concrete slab modules in nuclear power plants are deviated from gravity, a specific hoisting connection system is designed considering present hoisting points, which used welded H-beam lifting frame and a series of research is accomplished, which contained calculating the bary centre of concrete base module, establishing the mathematical model with ANSYS software to analyze the stress and displacement of the hoisting frame, hanging ears and pin wheel. The results show that the H-beam welding lifting frame structure is reliable and the lifting parameter design is reasonable. The overall hoisting task is completed by using mobile cranes, hydraulic jacking towers, hoisting beams, connecting hoisting frames, Loxic-leveling and other complex processes.

Design and Verification of Detective Device of Boron Meter

WANG Hongbo, DENG Sheng, WANG Can-hui, DAI Hang-yang, ZONG Xun-cheng, FU Guo-en

2013, 34(6): 136-137,147.

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A boron meter was designed according to reactor operation safety needs and site condition, and tests for the boron meter prototype were conducted. The results show that the design can meet the radiation protection requirements for the neutron source and monitor boron concentration of coolant in the primary system.

Effect of Thermal Aging on Failure Probability of a Nuclear Primary Pipe

LI Shu-xiao, LI Shi-lei, WANG Xi-tao, ZHANG Hai-long, WANG Yan-li, XUE Fei

2013, 34(6): 138-142.

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In this study, degradation of fracture toughness due to thermal aging was considered in the probabilistic fracture mechanics analysis of nuclear pressure pipes. A predicting program for pipe break probability based on thermal aging embrittlement was developed. On the basis of experimental data of a pipe material, evolution of the fracture toughness at 280℃ and 330℃ were estimated using the Argonne National Laboratory（ANL） procedure. Failure probability of a pipe with a circumferential inner surface crack was calculated in two cases: consideration and without consideration of thermal aging.The result show that failure probability is higher when thermal aging embrittlement was considered. Also, cumulative failure probability at 280℃and 330℃ were compared. The results indicate that pipe will suffer more serious thermal aging and have higher failure probability when operating at a higher temperature.

Analysis on NDE Problem of Weld Joint between AP1000 Steam Generator and Reactor Coolant Pump Casing

MAO Chang-sen, CHEN Fu-bin

2013, 34(6): 143-147.

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For the problems of unacceptable imperfection indications found in the weld joint between AP1000 Steam Generator and RCP Casing during Ultrasonic Examination（UT）, the comparison of UT inspection techniques have been demonstrated for different UT Procedures, and examination sensitivities difference have been analyzed between ASME Section Ⅺ and Section Ⅲ. It is proved that ASME Section Ⅺ inspection techniquesof in-service inspection were mixed with Section Ⅲ acceptance standardsof examination for fabrication, which is the immediate cause why imperfection indications were unacceptable. Furthermore, the re-UT was performed by several different degree probes with examination sensitivities established on an added hole reflector in the UT block. It is found that imperfection identity of UT indications were misidentified by original operators, and finally proved that this weld joint meet the requirement of acceptance standards of ASME Section Ⅲ.

Screening and Management of Dead-End Pipe in Qinshan NPP

CAO Xueming, LI Shiwei, XUE Fei

2013, 34(6): 148-152.

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This paper introduces the occurring conditions for dead-end pipe phenomenon and the corrosion failure mechanism in the pressurized water reactor nuclear power plant. According to the screening process for the dead-end pipes, the dead-end pipes in the Qinshan nuclear power plant of 300 MW Unit is screened to determine the dead-end pipes concerned. Management countermeasures are proposed. The results show that the dead-end pipe phenomenon is caused by the water saturation vaporization in the dead-end pipe, and the reason is that the water in the dead-end pipe is heated and the pressure can not be maintained higher than the saturation pressure. The dead-end pipe phenomenon can be control and alleviate by improving the relevant procedures, and can be eliminated completely by system modification.

Analysis and Improvement on Broken Coupling Bolts

YOU Lei, DENG Xiao-yun, CHEN Rong, LIANG Xia-xiang

2013, 34(6): 153-155.

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An extensive investigation on chemical element, fracture surface and so on was conducted into these broken bolts. The investigation showed that the broken bolts resulted from hydrogen-brittleness, high-hardness, high-strength and residual hydrogen. In order to improve the quality of bolts, the supplier changed the design model, improved the surface treatment method, and decreased the strength of bolts. The bolts after 2ndimprovement satisfied the requirements for safety operation.

Research on Abnormal Operation Status Detection Method for Nuclear Power Plants Based on Operation Data Analysis

YU Ren, KONG Jing-song, LUO De-sheng, ZHANG Huan-lin, YANG Huai-lei

2013, 34(6): 156-160.

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An abnormal operation status detection method based on dynamic Hopfield artificial neural network（ANN） is designed for nuclear power plants. By online training of the ANN, it can be ensured that the ANN can tail after the normal change of the dynamic characteristics of the NPP caused by the change of its operation state, so as to reduce the possibility of misdiagnosis. By observing the weighted mean square error of the ANN predictive output and the real output of the device, the abnormal change of the parameters can be detected in early time. Taking the primary loop pressure of a NPP as example, several tests are performed to validate the ability of the method to detect the operation parameter abnormal change. The results show that within the entire operation spectrum of the NPP, the method exhibits well faculty of the parameter abnormal change detection.

Study on Treating of Low-Level Radioactive Reactor Wastewater by Combined Membrane Process（UF-RO）

LU Yun-yun, CAO Qiru, CHEN Yun-ming, HUANG Lijuan, BAI Xiao-feng, LI Bing, FENG Liang

2013, 34(6): 161-164,172.

Abstract(26) PDF(0)

Abstract:
According to the characteristics of radionuclide exists in the low-level radioactive reactor waste water from HFETR, we use a new combined membrane process separation technology to study the efficient treating of low-lever radioactive reactor wastewater. First, the prepared the simulated wastewater contained Cs+, Sr²⁺, Co²⁺, Ni²⁺, and Fe³⁺. Then, we sequentially investigated the pressure, ion concentration, pH value and EDTA, which have effects on the desalination rate of membrane processing metal ions in wastewater. The results show that: in the condition of pH = 7, and added 0.15mol/L EDTA, the simulated wastewater separated by UF-RO, desalination rates of Cs+, Sr²⁺, Co²⁺, Ni²⁺and Fe³⁺are all above 95%; In the subsequent trials, adding 0.15mol/L EDTA into the radioactive residuary solution, and then treating by UF-RO-RO, the decontamination efficiency can reach 95.7%.

Preparation of SBR Rubber Based Flexible Shield Material

FU Ming, WANG Yong, LI Fang

2013, 34(6): 165-168.

Abstract(19) PDF(0)

Abstract:
SBR flexible shielding material is used for the first time in high benzene content forth benzene rubber with radiation resistance instead of silicon rubber that is not radiation resistant. By dense refine（or opening refine） process, the reasonable multiple group additive is adopted to improve the physical and chemical characteristics, and the mixing refine process is used to disperse the lead powder and boron carbide powder in forth benzene rubber uniformly. This kind of material can be arbitrarily plastic formation, with good shield characteristics, and high toughness and structure adaptability.

Venturi Purification Device and Its Application in Purification of Gaseous Waste of Nuclear Facilities

KONG Jing-song, YU Ren, YANG Huai-lei

2013, 34(6): 169-172.

Abstract(14) PDF(0)

Abstract:
The working principle of Venturi purification device and its purification of aerosol have been described. Then, taking the gaseous iodine as an example, the absorption process of insoluble gas pollutants is discussed, the calculation methods of the gas-liquid contact area, mass transfer rate and efficiency of mass transfer are educed, and the factors that affect the efficiency of mass transfer are analyzed.

2013 Vol. 34, No. 6