核动力工程

Development of Transport-Burnup Code MCORGS

SHI Xue-ming, ZHANG Ben-ai

2010, 31(3): 1-4.

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A new code MCORGS which links MCNP and ORIGENS is developed to do the transport-burnup calculation.MCORGS can be used to compute both criticality problem and subcriticality problem with an external neutron source.It has no limitation on neutron spectrum or geometry shape.MCORGS is more versatile than MCCOOR which is dedicated to compute thermal reactors with regular shape.Visual FORTRAN is used to develop MCNP,ORIGENS and MCORGS,so it can be run in any WINDOWS operation system.MCORGS is easy to use for it automatically produces lots of information in input files.There are different treatments in absolute flux calculation and cross section update between MCORGS and MCCOOR.Results of a VVER-1000 benchmark show that MCORGS can get more accurate results in less time than MCCOOR.ADS benchmark is also calculated to validate its ability to deal with fast sub critical reactor with external source.

Angular Multigrid Acceleration Method and Lyusternik-Wagner Extrapolation Acceleration Technique for Numerical Calculation of Neutron Transport Equation

ZHANG Hong-bo, WU Hong-chun, CAO Liang-zhi, SUN Xing-guang

2010, 31(3): 5-8,18.

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The angular multigrid（ANMG） acceleration method and the Lyusternik-Wagner extrapolation acceleration technique are applied to accelerate the neutron transport equation solver DNTR,which was developed by using SN nodal method in triangular-z meshes.Some numerical results demonstrate that the ANMG acceleration method applied to DNTR is effective,even in strong anisotropic scattering and high scattering ratio situations,especially combined with the Lyusternik-Wagner extrapolation acceleration technique.

Study on Coupling of NLSANMT and RELAP5/mod3.2

CHEN Yu-qing, CAI Qi, YU Lei

2010, 31(3): 9-13.

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Based on the thermal-hydraulic code RELAP5/mod3.2,the coupling model of three-dimension space-time neutron kinetics and one-dimension thermal-hydraulic computation is established by the explicit method,and the three-dimension transient physical analysis model based on the show that the developed NLSANMT/RELAP5（mod3.2） program system is with better ability for core transient analysis,compared with RELAP5/mod3.2.nonlinear iterative Semi-Analytical Nodal code NLSANMT is transplanted.Proof-checking and computation of feedback coefficients and steady parameters of nuclear power reactor,and the simulation analysis of the uncontrolled accidental draw-out of one bundle of control rods prove the validity of the interface.The results show that the developed NLSANMT/RELAP5(mod3.2) program system is with better ability for core transient analysis, compared with RELAP5/mod3.2.

Calculation of Ex-Core Detector 3D Space Response Function via Monte Carlo Adjoint Transport Method

ZHOU Xu-hua, LI Fu, WANG Deng-ying

2010, 31(3): 14-18.

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The readings of ex-core detectors contain the information of in-core power distribution,and the relationship between ex-core detector readings and in-core power distribution is the detector space response function.The key prerequisite of using ex-core detector readings to reconstruct the in-core power distribution is to calculate the space response function accurately.The space response function of one pressured water reactor is calculated via Monte Carlo adjoint transport method in this paper by using the MCNP code.The numerical result shows that there are some difference from that obtained via S_N adjoint transport on two-dimensional geometry,and indicates the necessity to calculate the ex-core detector space response function on three-dimensional geometry.Monte Carlo adjoint transport method is convenient to calculate the three-dimensional spatial response function.

Discrete Direction Probability Method for Solving Neutron Transport Equation in Complex Geometry

LIU Guo-ming, WU Hong-chun, CAO Liang-zhi, CHEN Qi-chang

2010, 31(3): 19-22,29.

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For solving the neutron transport equation in complex geometry,the discrete direction probability method（DDPM） is developed.In the DDPM,the calculation system is divided into subsystems which can be subdivided into fine meshes.The ray tracing is done for each type of subsystem.Within a subsystem,fine meshes are coupled by the discrete direction collision probability,while subsystems are coupled by the discrete direction transmission probability.The numerical results demonstrate the high efficiency and accuracy of this method.

Linear Source Approximation in Three Dimensional Characteristics Method

CHAI Xiao-ming, YAO Dong, WANG Kan

2010, 31(3): 23-29.

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A method of Characteristics with Linear Source Approximation（LSA） is proposed in this paper,and the corresponding code TCM__L is developed.Numerical results show that the LSA MOC can get more accurate results than the Flat Source Approximation（FSA） MOC and SLC MOC,and the LSA MOC code can use less storage and shorter computing time when using large meshes.

Waiting Time Analysis for Asymptotic-Period Measurement of Reactivity

YIN Yan-peng, QIU Dong, ZHENG Chun

2010, 31(3): 30-33,39.

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Evaluation of waiting time is a key problem for asymptotic-period measurement of reactivity.Based on the point reactor kinetics equation,the neutron multiplication function and error function for two period-measuring methods are deduced.The two methods include the double power method and the exponential-simulating method.The conclusion is drawn that waiting time is concerned with the external neutron source,delayed neutron parameters and reactivity.The exponential-simulating method can get the earlier asymptotic period than the double power method.

Neutronic Calculation of Cold Neutron Source

TANG Feng-ping, HU Chun-ming, YANG Cheng-de, LIU Xiao-bo

2010, 31(3): 34-35.

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Cold Neutron Source（CNS）is an experiment facility which moderate the hot neutron to cold neutron.Based on the Monte Carle 3D transport calculation program,and using the continuity point cross section of ENDF-BV,the core and in-pile parts of the CNS was combined together,the complete MCNP calculation model was built,and the CNS neutronic calculation was performed.

Prediction of Critical Heat Flux in Vertical Tube under Flow Oscillation Condition

ZHAO Da-wei, SU Guang-hui, QIU Sui-zheng

2010, 31(3): 36-39.

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The reduction limit of critical heat flux（CHF） in vertical tube under flow oscillation condition,obtained by a temperature response model,was modified using the CHF experimental data of Umekawa and Ozawa et al..Based on the modified reduction limit of CHF,a new correlation was proposed to predict the CHF under flow oscillation condition in vertical tubes at low velocity and pressure conditions.The comparisons between the prediction of the new correlation and experimental data show a reasonable agreement with a mean error of 11.5%,which is smaller than the error of prediction using Kim correlation,21.04%.

Research on Fluctuation Avoidance of Passive Moderator Residual Heat Removal System of TACR

TIAN Ye, JIA Bao-shan, YAN Lin

2010, 31(3): 40-44.

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To avoid the fluctuation of the Passive Moderator Residual Heat Removal System（PMRHR） of TACR,this paper improves the system structure and uses CATHENA（Canadian Algorithm for Thermal hydraulic Network Analysis） code to simulate its cooling capability under the condition of lost of flow accident.The results show that the improved system can keep the reactor safe and reduce the instability in light water loop.

Finite Element Simulation for Effective Thermal Conductivity Coefficient of Dispersion Fuel Element

JIANG Xin, DING Shu-rong, HUO Yong-zhong

2010, 31(3): 45-49.

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The behavior of thermal conduction of dispersion fuel elements depends on the heat generation of the fissionable particles,the volume fraction and size and array mode of the particles,the geometrical configuration of the element and the thermal circumstance in the reactor core.Following the idea of micromechanics of particle reinforced composites,a representative volume element or a unit cell is chosen and the finite element analysis is applied to study the effective thermal conductivity coefficient of the fuel element.The effects of temperature,burnup and volume fraction of the fissionable particles on the coefficient are investigated in details.The results are compared with several analytical equations and indicate that the Maxwell’s model provides the best agreement with the FEM data.

Mechanism Analysis of Effect of Rolling Motion on Heat Transfer

HUANG Zhen, GAO Pu-zhen, TAN Si-chao, SHE Ying-juan

2010, 31(3): 50-54.

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Theoretical analysis on the fluid flow in a vertical channel under rolling motion is performed to establish the mathematical model.The velocity distribution in the channel is derived from the model.The fluid flow near the channel wall is essentially investigated.The analysis shows that,the channel flow under rolling motion is different from the conventional unsteady flow.The boundary layer near the channel wall is destroyed and the cause is figured out based on the flow analysis.Furthermore,the mechanism of the effect of rolling motion on heat transfer is propounded.That is the rolling motion enhances the disturbance in the fluid flow near the heating channel wall.

3D Numerical Analysis of Flow Field in Condenser Throat

CAO Li-hua, ZHOU Yun-long, CHEN Yang, MENG Fang-qun, LI Yong

2010, 31(3): 55-59.

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The condenser throat with the inner low-pressure heater and exhaust of turbine driven feed pump of some steam turbine is 3-D numerically simulated using the k-ε model combined with wall-function method and the SIMPLEC algorithm,and the non-uniformity of flow field and the change of energy loss coefficient are analyzed.Results indicated that,the low speed areas are formed at four angles of the condenser throat outlet section.For the inner low-pressure heater,the low speed eddy zone is formed just below it.The non-uniformity of flow field is increased for the existence of exhaust of turbine driven feed pump,because two partial low speed areas are formed at the outlet section near the side of the exhaust of turbine driven feed pump.The velocity distribution nearby the exhaust of turbine driven feed pump is only affected,however far away from it,the velocity distribution of throat flow field is hardly affected but the values are increased.At rated conditions,the energy loss coefficient with exhaust of turbine driven feed pump is increased compared with that without exhaust of turbine driven feed pump.

Study on Comprehensive Evaluation Model for Nuclear Power Plant Control Room Layout

ZHU Yi-ming, LIU Yuan, FAN Hui-xian

2010, 31(3): 60-63,68.

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A comprehensive evaluation model for layout of the main control room of nuclear power plants was proposed.Firstly the design scope and principle for the layout of the main control room were defined based on the standards,and then the index system for the comprehensive evaluation was established.Finally,comprehensive evaluation was carried out for the layout design by applying the fuzzy comprehensive evaluation method in the index system.

Common Cause Failure Analysis of Feedwater System in Marine Nuclear Reactors Based on GO Methodology

REN Xin, ZHAO Xin-wen, CAI Qi, GUO Qiang

2010, 31(3): 64-68.

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Feedwater system is divided into two parts: feedwater making and feedwater supplying.Two procedures are carried out in analyzing the feedwater system according to different working condition.In the first stage,the effect of common cause component group 2 to the part of feedwater supplying is analyzed,and in the second stage,the effect of common cause component group 1 and 2 to the whole feedwater system is comprehensively analyzed.This paper respectively analyzes the reliability of feedwater system in two stages with GO Methodology.Firstly,the model of system and construct the GO chart of feedwater system is constructed by analyzing the system,and then the system unavailability in two stages is calculated based on the GO algorithm dealing with common cause failures and the contribution of common cause failure to it.The analysis results show that the common cause failure greatly affects the system reliability and the GO algorithm is the effective and practical method to analyze the system with common cause failure.

Reliability Analysis of Nuclear Power Plant Bus Systems Arrangement Based on GO Methodology

LI Zhe, LU Zong-xiang, LIU Jing-quan

2010, 31(3): 69-73,77.

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In this paper the GO method is used for analyzing the reliability of NPP bus system arrangement.Focusing on the typical one and a half breakers bus system,the detailed GO chart of typical work/failure（maintenance） two or three-state-component system is given,and the qualitative and quantitative analysis is conducted.Compared with the FTA results,the correctness and advantage of the GO methodology is verified.

Dynamic Grey Correlation Analysis of Human Error In Nuclear Power Plants

LI Meng, YUAN Ce-feng

2010, 31(3): 74-77.

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Based on the techniques of THERP and CREAM,the dynamic gray correlation analysis control and prevention of human error factors in management.method,which uses a dynamic way and considers the consequences of human error,is applied.From 2006 to 2008,human events in six nuclear power plants in China are collected,classified and analyzed.Corresponding correlation coefficients are measured.The paper proposed that nuclear power plants should pay more attention to the control and prevention of human error factors in management.

Study on Pressure Fluctuation in Main Pump of 300 MW Nuclear Power Plants

CHEN Xiang-yang, YUAN Dan-qing, YANG Min-guan, YUAN Shou-qi

2010, 31(3): 78-82.

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From the safety of reactor coolant pump,the numerical investigation of internal unsteady flow in reactor coolant pump of one 300MW nuclear power plant in China that is based on the calculation Navier-Stokes equation and standard k-ε turbulence model is carried out in this paper.Through the analysis of pressure fluctuation which locates at impeller,guide vane and inlet parts,it is concluded that the frequency of pressure fluctuation in the reactor coolant pump is exclusively governed by the blade-passing frequency.For the spherical shell pump,the maximum amplitude of pressure fluctuation decreases from inlet to guide vane but increases from hub to impeller edge except for the vicinity of impeller edge in the impeller part which follows the former mode.The comprehensive analysis of maximum amplitude of pressure fluctuation shows that the hydraulic parts could reduce the effects of pressure fluctuation,increase only a few vibrations to the whole unit and these are in favor of improving the safety performance of main pump unit.

Experimental Research and Analysis on Location of Stratified Interface in Density Lock

YU Pei, LU Hai-jin, YAN Chang-qi

2010, 31(3): 83-87.

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A theoretical model is built to calculate the location of fluid stratified interface in density lock,the calculated values and experimental observations are compared and the difference between them are analyzed.The effect of disturbance velocity and temperature difference on the interface location is discussed,respectively.The result shows that the larger disturbance velocity responds to the lower interface location,and the higher temperature difference responds to the higher interface location.In condition of certain temperature and diameter,the experiential relationships among the interface location,disturbance velocity and Richardson number are fitted,respectively.

Development of Test System for Governor of Emergency Diesel Based on Virtual Instrument Technology

DU Yu, SHI Hai-ning, YAO Jian-lin, CHEN Yao-ling, DING Jun-chao

2010, 31(3): 88-91.

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In Daya Bay nuclear power plant,the governor pre-debug of emergency diesel is inefficient,and the result could not be verified.To solve these problems,a governor testing system was developed by adopting the virtual instrument technology.The system developed with LabVIEW and consisted of notebook,signal conditional unit and data acquisition cards,can realize the test of the governor static and dynamic characteristics,including the simulation of input signal,real-time measurement and preservation of the output signal,and data analysis.The successful application in Daya Bay Nuclear Power Plant shows that this governor test system is stable and efficient.

Application of Mechanical Draft Cooling Tower in Inland Nuclear Power Plants

PAN Wen-gao, LI Chao-ming, HU Bin, LI Xiao-yan

2010, 31(3): 92-95,101.

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This paper described the nuclear power plant related design criterion classification of mechanical draft cooling tower used in inland nuclear power plant ultimate heat-sink systems,analyzed and summarized the special design and construction requirements of nuclear-classified mechanical draft cooling tower,and brought forward the qualification procedure of equipments in such cooling tower,and finally proposed a reference and guide for the design and development of domestic nuclear-classified mechanical draft cooling tower.

Application of Three-Dimensional Design and Cooperation Design Platform in Nuclear Power Institutes

WANG Yong

2010, 31(3): 96-101.

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Based on the experiences of the establishment,promotion and application of SNERDI three-dimensional design and cooperation design platform,this paper describes the method to establish the cooperation design platform satisfing the nuclear engineering design demands and the nuclear power quality assurance requirement completely and controllable for the nuclear design institutes using the traditional design methods.

Optimization of Preventive Maintenance Cycle Based on Experimental Feedback in Nuclear Power Plants

SHI Jie

2010, 31(3): 102-105,112.

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The preventive replacement method based on the experimental feedback was introduced.In this method,the initial preventive replacement cycle was acquired by expert votes.The preventive replacement cycle combined with the operation experience of the equipment was gained by means of Bayesian theorem.The Optimized preventive replacement cycle can be acquired by comparing the two probabilities that no fault occurs within the cycle.This method was tested on the switches which were used in Daya Bay Nuclear Power Plant and the results indicated its validity.

Change of Main Steam Pressure after Outage

LIU Zhen, PAN Ze-fei, YANG Shao-jie

2010, 31(3): 106-109.

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Steam Generator outlet steam pressure is an important parameter in nuclear power plant.Within monitoring this parameter,we find the main steam pressure turn lower after outage,then recover gradually along operating.In this paper,we describe and analyze this phenomenon at first,then give some means for improving the main steam pressure.

Analysis and Treatment for High Voltage Bushings of Electric Auxiliary Boiler of Nuclear Power Plant

TANG Fang-xuan, ZHOU Qian-sheng, CENG Li-min, ZHANG Jian

2010, 31(3): 110-112.

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The structure and theory of the two high pressure discharged steam electrode boilers installed in QINSHAN Nuclear Power Plant II is presented,and the causes for the High Voltage Bushings malfunctions since the boiler put into operation are also analyzed in this paper.The suggestions for the improvement of replacement and maintenance of the Bushings are given.

Simulation of Dynamic Response of Nuclear Power Plant Based on User-Defined Model in PSASP

ZHAO Jie, LIU Di-chen, XIONG Li, CHEN Qi, DU Zhi, LEI Qing-sheng

2010, 31(3): 113-117,142.

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Based on the energy transformation regularity in physical process of pressurized water reactors（PWR）,PWR NPP models are established in PSASP（Power System Analysis Software Package）,which are applicable for calculating the dynamic process of PWR NPP and power system transient stabilization.The power dynamic characteristics of PWR NPP is simulated and analyzed,including the PWR self-stability,self-regulation and power step responses under power regulation system.The results indicate that the PWR NPP can afford certain exterior disturbances and 10%P_n step under temperature negative feedbacks.The regulate speed of PWR power can reach 5%P_n/min under the power regulation system,which meets the requirement of peak regulation in Power Grid.

Effect of Brazing Temperature on Tensile Strength and Microstructure for 304 Stainless Steel Plate-Fin Structure

JIANG Wen-chun, GONG Jian-ming, TU Shan-dong

2010, 31(3): 118-121.

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The vacuum brazing of 304 stainless steel plate fin structure was carried out,and the tensile tests were performed.The scanning electron microscope（SEM） and Energy-dispersive X-ray spectrometer（EDS）tests were carried out to reveal the microstructure and chemical components in the brazed joint.The effect of brazing temperature on tensile strength and microstructure for 304 stainless steel plate-fin structures was discussed.The results show that the tensile strength is increased with the brazing temperature increasing from 1025 to 1100℃.With the brazing temperature increasing,the boron is diffused more adequately,which increases the strength.The microstructure becomes more uniform with the brazing temperature increasing.Hence the brazing temperature is determined at 1100℃.When the brazing temperature is very low（1030℃）,more brittle phases are generated in the brazing joint,which decreases the strength.When the brazing temperature is increased to 1100℃,full solid solution is generated in the middle of joint,and the microstructure in the fillet becomes more uniform.

Interface Characteristic during Bonding of Be and HR-1 Stainless Steel by Hot Pressing

LI Hui, KANG Ren-mu, ZHOU Shang-qi, KONG Ji-lan, ZHANG Peng-cheng

2010, 31(3): 122-127.

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Be and HR-I stainless steel was diffusion bonded by hot pressing.The microstructure,composition and phase distribution and mechanical properties of the joints were analyzed using the optical microscopy,scanning electron microscopy（SEM）,scanning auger microspectroy（SAM） and x-ray diffraction（XRD）,and the effect of different interlayer materials Cu and Al and Ag-Cu alloy was also discussed.The results show that it is not easy to joint Be and HR-1 stainless when Al was used as the interlayer material,but good joints can be obtained using Cu and AgCu alloy as the interlayer materials,because the formation of Be-Fe brittle intermetallic compounds was prevented.Ag-Cu alloy is the best interlayer materials among them,as it can reduce the mutual diffusion between Be and Fe.

Analysis of Advances of AP1000 Gaseous Radwaste System

DONG Bo, GAO Ming-shi

2010, 31(3): 128-131.

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The concept of passiveness,which has been generally used in the third-generation nuclear power technology AP1000,makes a great innovation in the design of the security systems,and the safety and economy have been greatly enhanced.However,this concept is used not only in safety systems,but also in the design of non-safety systems.Through the detailed introduction of the Gaseous Radwaste System,and the contrast with the traditional pressurized water reactors,it could be seen that the use of passiveness makes great innovation compared with tranditional technology.The operation does not depend on the active components,and the treatment of gaseous radwaste has been greatly improved from compress-store-decay to adsorb-decay by activated carbon.This innovation simplifies the process and improves the safety of operation.

Analysis of Economy Characteristics and Improvement Ways for Chinese Nuclear Power

LI Yong

2010, 31(3): 132-135.

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For the nuclear power industry,due to characteristics of its own,its economy is quite different from that of the traditional fossil-fuel power.This paper studied the basic characteristics of the nuclear power economy and the status of economy of domestic nuclear power,and analyzed the main ways to improve the nuclear power economy.

Analysis of Demand on Backup Documents for Feasibility Study for Nuclear Power Plant Construction Projects

LIU Yong-qing, ZHOU Ping, GE Zheng-fa, XU Peng-fei

2010, 31(3): 136-142.

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This paper analyzed the process for the approval of the 24 backup documents during the feasibility study stage of Fang Jiashan nuclear power plant,and defined the 27 backup documents and derivative documents that should be obtained during feasibility study stage for nuclear power construction projects,and finally expound the 8 representative documents in detail.The analysis showed that the process for the approval of all the backup documents under the Approval System prolonged the project construction period.It recommended that the backup documents can be obtained from the nine departments,such as those involved in the economic planning,urban planning,land use planning and nuclear safety,as well as environmental protection.

2010 Vol. 31, No. 3