核动力工程

Analysis of Full-Core Calculation of RMC

QIU Yishu, SHE Ding, FAN Xiao, WANG Kan, LI Zeguang, LIANG Jingang

2013, 34(S1): 1-4,23.

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Abstract:
With the Inspur TS1000 HPC Server of Tsinghua University,a lot of calculations have been done based on the NEA Data Bank full-core benchmark model and EDF 3D pressurized water reactor（PWR） full-core calculations through large-scale paralleling.The performance of MCNP which is the widely used general Monte Carlo code and RMC which is used for reactor analysis and developed by Tsinghua University is compared systematically.It is found that MCNP is unable to calculate the local power density of full-core reactors at the required accuracy because its limits on the parallel model,the performance of tallying and so on,while RMC is well-suited owing to its improvement on those limits.Thus,it can be concluded that the performance of RMC is better than MCNP in terms of detailed power density distributions calculation in full-core reactors.

Resonance Calculation and Speedup Optimization for 2D Arbitrary Geometry Subgroup

HE Lei, WU Hongchun, CAO Liangzhi

2013, 34(S1): 5-9.

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Resonance calculation is the footstone of the reactor lattice design,core design and fuel management.The subgroup resonance calculation method which bases on subgroup parameter,uses the transport equation as a solver and solves the subgroup flux for effective resonance self-shield cross section,finally achieves 2D complicated geometrical resonance calculation.As the transport solver is used iteratively for every resonance energy group,the calculation efficiency is lower than that of equivalence theory.This study is based on the subgroup resonance method and subgroup resonance calculation code designed by ourselves.The multi-group library,transport calculation source,and multi resonance nuclides iteration are proposed and completed.The verification of benchmarks showed that the scheme can increase the efficiency while keeping the same accuracy and insure the engineering practicability.

Development and Validation of Nuclear Cross Section Processing Code for Reactor Analysis-RXSP

YU Jiankai, LI Songyang, WANG Kan, WANG Guanbo, YU Ganglin

2013, 34(S1): 10-13.

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The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor physics calculations.The code RXSP（Reactor Cross Section Processing Code） is developed by REAL（Reactor Engineering Analysis Laboratory） of Department of Engineering Physics in Tsinghua University.The key methods such as fast Doppler broadening,thermal libraries interpolation,and OpenMP parallel acceleration,have been achieved within RXSP.This code is able to process the original data of ENDF/B（Evaluated Nuclear Data File/B） efficiently and accurately to produce the continuous energy point cross section data which is used for RMC.By comparing with NJOY,the microscopic and integral verifications show that RXSP has the same accuracy as NJOY while RXSP has much less processing time to meet the demand in the reactor physics-thermal-hydraulic coupling calculations requiring the frequent updates of a large number of materials and temperature points.In addition,RXSP make it available to process the resonance parameters of the R-matrix Limited format.

High Order Harmonic Wave Filtering Method Used In Prompt Neutron Attenuation Constant Calculation Of ADS Sub-Critical Reactor

XIE Jinsen, YU Tao, ZUO Guoping, HE Lihua, LI Xiaohua

2013, 34(S1): 14-17.

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MUSE（Multiplication with an external source） series experiments carried out by the EU show that the Pulsed Neutron Source（PNS） method is appropriate for keffmeasurement in deep sub-critical conditions.In PNS method,the accuracy of prompt neutron attenuation constant α plays a key role for accurately keff measurement.In this paper,the analysis of α constant measurement on fast-thermal coupled subcritical facility Venus-1# is performed.By using high order harmonic wave filtering technique,the time interval for α fitting is obtained and the fitted α values are spatially independent with locations of detectors.Furthermore,the comparison of prompt multiplication factors kpderived from fitted α and calculated by MCNP（Monte Carlo N Particles） is made,which shows a good conformation.Results in this research indicate that,the high order harmonic wave filtering method can effectively avoid the problem that the measured α values depend on the locations of detectors,and the α values obtained from which can be used for keff off-line monitoring in ADS sub-critical reactors.

Study on Acceleration of Three-Dimensional Method of Characteristics by GPU

ZHANG Zhizhu, LI Qing, WANG Kan

2013, 34(S1): 18-23.

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The three-dimensional method of characteristics（MOC） can solve neutron transport equation for arbitrary geometry accurately.However,the MOC has some drawbacks: the convergence speed is slow and very time consuming.As a result,the research of acceleration of MOC is carried out.Compared with the CPU computing,the GPU computing,which is one of the most promising high performance computing,can achieve higher computing speed but with lower cost.And the development of general computing on GPU can be simplified with CUDA.To reduce the computing time and increase the computing efficiency,the study of three-dimensional MOC is performed and applied to the three-dimensional MOC code TCM.The computing results confirm the excellent acceleration of the code running on GPU.

Development and V＆V of Advanced Neutronics Code System SARCS-4.0

CHAI Xiaoming, MA Yongqiang, WANG Yuwei, LU Wei, YAO Dong

2013, 34(S1): 24-26.

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An advanced neutronics code system SARCS-4.0,which can simulate the core composed by arbitrary square assemblies,calculate Uranium-Plutonium and Thorium-Uranium cycle,treat many burnable poisons（such as B,Gd,Er,Hf,Ag,In,Eu,Sm.etc） and several control rod materials,is introduced in this paper.The SCWR reactor core is calculated by SARCS-4.0.The results show that SARCS-4.0 is accurate enough to design the reactor core.

Study on Prediction Models Development for Calculation of Reflooding Phase of Tight Lattice

WU Dan, YU Hongxing, YU Junchong

2013, 34(S1): 27-31.

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Reflooding is the most important and complex process in a LOCA transient.Compared with ordinary PWRs,the reflooding phase of tight lattice has the characteristics of higher peak cladding temperatures and slower quench front rates.The existing best estimate codes could not be used for the reflooding calculations of tight lattice.Based on RELAP5,we developed a modified reflooding model,and did validations using experimental results and other researchers’ results.It is proved that the modified code could be used to model the reflooding phase of tight lattice.Further investigations are needed on the development of interfacial drag and interfacial heat transfer models of the dispersed flow region and the region just downstream of the quench front.

Approach for Simulating Severe Accident of UO₂-Zr Plate by SCDAP/RELAP5

ZHANG Zhuohua, PENG Shinian, HUANG Shanfang, YU Junchong

2013, 34(S1): 32-36.

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Abstract:
As a common mechanistic code for safety analysis of severe accident,SCDAP/RELAP5 can simulate many types of core components phenomenon during severe accidents.Comparison of simulation model of fuel behavior under severe accident between fuel rod and ATR plate is described in this paper and the approach for simulating severe accident of UO₂-Zr plate is concluded by combining structure properties of UO₂-Zr.It is concluded that the basic analysis of severe accident of UO₂-Zr plate could be achieved by S/R code from the code simulation.However,new core structure,new model of fuel behavior and combination of existing model should be developed in S/R code to simulate the precise core behavior of reactor assembled with UO₂-Zr plate under severe accidents.

Research of Solidification Model for the Melt Based on SIMPLE Algorithm

ZHANG Yapei, TIAN Wenxi, QIU Suizheng, SU Guanghui

2013, 34(S1): 37-41.

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The cooling characteristics of the melt have important influence on the process of IVR in a severe accident.An enthalpy formulation based on fixed grid methodology was developed for the numerical solution of convection-diffusion controlled mushy zone phase-change solidification of the melt.The process of the solidification was simulated by solving the N-S equations and the enthalpy formulation equation using the SIMPLE algorithm.The phase-change model could simulate the solidification and mushy zone well,and it was verified by the use of analytical solutions and benchmark problems.

Contrastive Simulation Research on PM₁ Particles Deposition in Two Narrow Rectangular Channels

RU Xiaolong, ZHOU Tao, LIN Daping, YANG Xu

2013, 34(S1): 42-46.

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This paper intends to make numerical simulation to study the regularity of the deposition of PM1 particles under the turbulent condition at different temperatures in two narrow vertical rectangular channels.The channel 1 length is 1000mm,and the sectional dimension is 20 mm×20mm.Channel 2 sectional dimension is 10 mm×30 mm,and channel 2 length is as long as channel 1.The gas phase uses the standard k– model and the PM1 particles use DPM（Discrete Phase Model）.The study found that,the turbulent diffusion,the thermophoresis and the secondary flow together make the PM1 particles gather in the channel near-wall regions,the turbulent diffusion make PM1 particles concentrate in the channel near-wall regions,and the important reason of PM1 particles concentration is the secondary flow in the corner.In the temperature field in near-wall regions,it can be found that the thermophoretic force is the most important factor for the PM1to deposit on the cold wall.With the rise of the main flow temperature,the random motion and diffusion of the PM1 particles tend to be intensified due to the significant intensification of the PM1 particles Brownian motion,which is not conducive to form the stable gathering areas of PM1 particles in the near-wall regions,and it tends to weaken the deposition of PM1particles on the cold wall.

Analysis of Hydrogen Source Term and Effectiveness of Hydrogen Control in Thousand Megawatt PWR Severe Accident

ZOU Jie, TONG Lili, CAO Xuewu, GU Jian, XUE Junfeng, JIANG Yu, HAO Lulu, CHOU Suchen, LIU Li

2013, 34(S1): 47-50.

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The integrated severe accident analysis model of 100 MW PWR NPP is built to analyze the hydrogen risk under severe accidents.Large break loss of coolant accident（LB-LOCA）,medium break loss of coolant accident（MB-LOCA）,small break loss of coolant accident（SB-LOCA）,station blackout（SBO）,steam generator tube rupture（SGTR） and main steam line break（MSLB） are chosen as typical severe accident sequences to analyze the hydrogen source.Considering the hydrogen quantity of 100% zirconium water reaction,the LB-LOCA is selected as a representative sequence to evaluate the hydrogen mitigation system.The results show that a certain number of PARs could remove hydrogen and oxygen effectively,and protect the containment integrity against hydrogen de agration or detonation.

Study on Time-Average Friction Characteristics of Low Flow Rate Single-Phase Flow in a Rectangular Channel in Rolling Motion

TAN Sichao, WANG Zhanwei, LAN Shu, ZHANG Hong

2013, 34(S1): 51-54,60.

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Studies of time-average friction characteristics of low flow rate single-phase flow in a rectangular channel under rolling motion condition were carried out.The test fluid is deionized water and the equivalent diameter of the rectangular channel is 5.39 mm,the range of the Reynolds number is 800-20000 and the amplitude of the flow rate exceeds ±30%.Two methods were employed to calculate the time-average friction coefficient and the results showed that the time-average friction coefficients obtained by the two methods are different in laminar flow.Different methods represent different physical meanings.When the time-average friction coefficient is calculated by integrating the instantaneous friction coefficient which is acquired by Darcy equation in a period,this time-average value represents time-average viscous dissipation.On the other hand,when the time-average coefficient is computed by Darcy equation which employs time-average pressure drop and velocity,this time-average friction coefficient represents time-average pressure drop.

Study on Coupled Neutronic and Thermodynamic Instabilities in a Double Rectangular Channel Natural Circulation System under Rolling Motion

ZHOU Linglan, ZHANG Hong, TAN Zhanglu, DONG Huaping

2013, 34(S1): 55-60.

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The ocean condition thermal-hydraulic analysis code RELAP5/MC is coupled with a 3D neutron transport code TDOT-T by Parallel Virtual Machine（PVM） and coupled neutronic and thermodynamic instabilities in a double parallel channel natural circulation system under rolling motion have been studied.Results indicate that the system has two oscillation modes: the in-phase oscillation caused by rolling motion and the out-of-phase oscillation caused by DWO.The neutronic feedback stabilizes the system in type 1 DWO region and suppresses the in-phase oscillation in two-phase region,but has little effect on the type 2 DWO and the in-phase oscillation in single-phase region.The calculated results have been analyzed based on the nonlinear theory and it is found that the nonlinear of system is enhanced with nuclear feedback.As the result of the system flow oscillation caused by the rolling motion superimposes with the DWO,and this extremely complicated phenomenon includes the synchronization and chaotic phenomena in the coupling of nonlinear oscillators.

Study on Transient Characteristics of SCWR under Typical Accidents by TACOS Code

ZHU Dahuan, TIAN Wenxi, QIU Suizheng, SU Guanghui

2013, 34(S1): 61-65.

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TACOS（Transient Analysis Code for SCWRs） code has been developed to analyze the transient characteristics for SCWR system.In the present study,the transient analysis of various categories events for SCWR with mixed spectrum core（SCWR-M） are performed using the TACOS code,and the comparative study of three different SCWR concepts is conducted to compare the transient thermal-hydraulic characteristics and safety performance.The total loss of flow accident,one pump seizure,reactivity reduced accident,turbine trip without bypass are calculated for SCWR-M.These transients cover the three main categories events of SCWR,which are flow rate abnormality,reactivity abnormality and pressure abnormality.The comparative study of the three concepts with different flow path designs indicated that the design of multi-pass core could have more complicated transient flow mode,and would face greater challenge for the mitigation of total loss of flow accident.

Experimental Study on A Swirl-vane Steam Separator with Air-Water

XIONG Zhenqin, WANG Minglu, LI Yazhou, LU Mingchao, ZU Hongbiao, ZHANG Kai

2013, 34(S1): 66-68.

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Small scale swirl-vane steam separator is experimentally studied by high speed camera,using air-water modeling fluid.The experimental result shows that the thickness of the water film on the inner tube is vital to the separation.The separation efficiency is mainly influenced by the water flow rate.It rises gradually due to the increase of the water film thickness and the reduction of the water droplets with the increasing of water flow rate.The incensement declines when the water flow rate is larger than 0.3m³/h,and furthermore it decreases.The reason is that the water film is distorted to be droplets by the high speed air due to reducing flow area for air.

Application Research of Fuzzy Petri Net Expert System in Nuclear Power Plant Fault Diagnosis

PENG Qiao, YU Ren

2013, 34(S1): 69-72.

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The Object-Oriented Fuzzy Petri Net is used as a knowledge representation method for the design of the knowledge base,to improve the traditional Expert System’s ability of knowledge expressing and reasoning efficiency.Based on this,the inference engine of Expert System was designed.A simulation experiment was made with a Nuclear Power Plant Simulator,and the results show that,the Expert System constructed in this paper can diagnose the faults accurately.

Demonstration of Flywheel Scheme and Sensitivity Analysis of Bearing Water Film Stiffness on Rotor Dynamic Characteristics of RCP

ZHAO Xuecen, DENG Liping, LIU Lizhi, YANG Song

2013, 34(S1): 73-76,79.

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In order to ensure that the main coolant pump（RCP） meets the requirements of half flow time after reactor shutdown and power accident,the moment of inertia of the RCP,whether need to add the flywheel and the RCP rotor’s dynamic performance are analyzed in this paper.Demonstration of flywheel scheme and sensitivity analysis of bearing water film stiffness on the rotor dynamic characteristics are presented.

Design and Implementation of Nuclear Reactor Software Platform

FENG Bo, LU Wei, FENG Jintao, FAN Jiajie, YUAN Guanghui

2013, 34(S1): 77-79.

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This paper proposes a platform for nuclear reactor design software.The technology for virtualization and hardware platform for multiple servers is applied to facility the operation of the platform on the popular operation systems.A graphic library for the reactor core display and operation is developed to provide user friendly man-machine interface and thus enhance the use efficiency of the software.The platform has been applied in the Special software for Core Operation and Reservation.The application and research show that,this platform can solve the disadvantage of the traditional nuclear reactor software,and enhance the efficiency of nuclear design.

Study on Optimization of Shutdown and Critical Model for VVER-1000

LU Zongjian, LIU Tongxian, WANG Jinyu, WU Lei, YU Yingrui

2013, 34(S1): 80-83,98.

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On the premise of safety,economy is one of the important objectives of nuclear power plants.For VVER-1000,there are several cases of operation modes when the plant is off the grid,such as the critical operation after refueling,the hot shutdown and critical operation,and the critical operation after test.For the purpose to reduce the time of the operations,the consumption of materials,and the amount of waste water produced,the optimization researches are carried out on these cases.The qualitative and quantitative analysis is carried out,on the configurations of the control rod and boric acid,which are the important factors to influence the shutdown and critical operations.The general steps and the basic principles of the optimizations are described.Three cases are optimized.

Research on Theoretical Calculation Method for Refined Modeling of Fast Neutron Flux in Irradiation Surveillance Capsule

DENG Lilin, LU: Huanwen, TAN Yi, XIAO Feng, WEI Shuping

2013, 34(S1): 84-86.

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To deal with the problem of the notable calculation error of Irradiation surveillance capsule fast neutron flux compared with the measured values,the calculation model has been improved and a refined calculation model has been built by detailed inner configuration of assemblies in core.Comparison has been made between conventional compendious model results,refined model results and measured values.It is obviously that the calculation error of the refined model has been greatly reduced.

Research and Design of Reactor Neutron Source

LIU Jiajia, XIAO Feng, LU: Huanwen

2013, 34(S1): 87-90.

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This paper introduces the research of neutron source design of Ling’ao NPP phase 2 by Monte Carlo code MCNP.The results are in well accordance with the FARMATOME’s design.According to the research,the strength of neutron source of Ling’ao NPP phase 2 can be lowed by changing the location of neutron source,and lower strength will bring considerable economical benefit.

Study on Pin Power Reconstruction Method Based on 3-D Multi-Group Analytic Nodal Method for Hexagonal-z Geometry

SUN Wei, LI Qing, NI Dongyang, WANG Kan

2013, 34(S1): 91-94.

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Based on the theory that the hexagonal intra-nodal flux distribution can be expanded to a 3D-form in a series of analytic basic functions for each group,the analytic function expansion nodal method can be directly used for pin power reconstruction.Based on the proposed model,a module of pin power reconstruction has been developed for HANDF-E code.Compared with Monte Carlo code MCMG,the numerical results of VVER-440 benchmark and 3-D four-group thermal reactor benchmark show that HANDF-E can predict accurately the pin powers.

Calculation Research of Hydrogen Production Amount in Containment after LOCAL in PWR Nuclear Power Plants

HU Jianjun

2013, 34(S1): 95-98.

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The ORIGEN2 code is adopted to calculate the amount of hydrogen production in the core and sump region after LOCAL in PWR nuclear power plants,to reduce the conservatism for the design evaluation of the combustible gas control in the containment.The calculation model of radiolytic decomposition coolant and other related calculation model are used to calculate the amount of hydrogen production after LOCA in a 600MW PWR nuclear power plant,and the results show that over conservatism of the original evaluation,and there still exists abundant time to prepare and startup the hydrogen recombiners in the containment after LOCAL.

Preliminary Study on 24-Month Fuel Management Strategy

WANG Dan, WANG Jinyu

2013, 34(S1): 99-102.

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Most of the PWRs in China have changed or plan to change their cycle length to 18 month,but a longer fuel cycle（like 24 month） is the trend in the future.Taking the Qinshan Phase II nuclear power plant as a research target,this paper designs two types of fuel management strategies with 56 and 60 fresh assemblies.And the enrichment of the fresh fuel assemblies increases to 4.95%..Through the optimization of the assembly and poison locations,both strategies meet the demand of 24 month cycle length,and all are good at economics and safety

Research of 60-Year Design Life for Domestic PWR Reactor Pressure Vessel Based on Irradiation Embrittlement

QIU Tian, LUO Ying, MA Shuli, ZHOU Gaobin, LI Zhangxiang

2013, 34(S1): 103-108,115.

Abstract(16) PDF(1)

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Reactor pressure vessel is one of the most important equipments of PWR nuclear power plants.The life of nuclear power plants depends on the life of RPV.Based on the research of irradiation embrittlement of RPV materials,a comparison among the design of material,structure design,irradiation surveillance of M310,CNP1000,AP1000 and EPR is made in this paper,and also a discussion for 60-year design life is taken.Then several measures meeting 60-year design life of RPV are put forward.

Analysis and Control of Welding Deformation of Nuclear Power Reactor Plant CRDM Anti-Seismic Ring

HE Peifeng, WANG Qingtian, ZHANG Yi, MU Dianpeng, LI Yan, LI Ning

2013, 34(S1): 109-111,119.

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CRDM anti-seismic ring is one of the most important parts of pressurized water reactor vessel closure head upper assembly,and this paper gives the example of anti-seismic ring welding deformation,analyzes the principle and mechanism of the deformation,and puts forward the control measures,such as structure optimizing,welding process control,and machining craft optimizing;Proved by the practical application,these measures could effectively control the deformation of anti-seismic ring.

Research on Lift Pole Thread Fatigue of Magnetic Lifting Control Rod Drive Mechanism

TANG Xiangdong, YANG Bo, CHEN Xinan, YU Zhiwei, WANG Dejun

2013, 34(S1): 112-115.

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Control Rod Drive Mechanism（CRDM） is a servomechanism of reactor control and protection system,and it is the only movable equipment in reactor body.Its reliability affects the reactor safety and operation directly.Especially in accident conditions,CRDM must release the control rod to insert into the core immediately.This paper presents the thread fatigue analysis of the CRDM lift pole through electromagnetic analysis,mechanical analysis and fatigue analysis method,and presents a fatigue analysis method for thread bearing impact load in Magnetic Lifting CRDM.This method can be used in the analysis and design of the similar type of CRDM.

Design Improvement for Spacer Column Assembly

ZHANG Yi, LI Na, HE Peifeng, LI Ning, RAO Qiqi, MU Dianpeng

2013, 34(S1): 116-119.

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The origin for welding deformations induced by unbefitting alignment,of spacer column assembly and thermocouple placed in improper position in Reactor Vessel Internals,is analyzed,and the design improvement of the front edge and the conduit nozzle is proposed,to effectively avoid the reoccurring of similar problems and provide the technical reference for the follow-up design of the spacer column assembly.

Research on Foreign Body In-Service Cleaning Technology for Pressure Vessel Lower Head

HONG Long, HUANG Xindong, WANG Bingyan, DENG Jing, REN He

2013, 34(S1): 120-122,127.

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To satisfy the need of cleaning solid impurity in the reactor pressure vessel lower head,foreign body pumping technology is studied.After designing a simplified force model of solid granule in fluid,relationship between the liquid velocity and both mass and density is elicited via analysis.Based on this,draining circuit and mechanical structure is engineered,and analysis and experiment are also conducted.The result shows that the equipment performs well,which proved the correctness of theoretical research.

Experimental and Simulation Study on Cooling Heat Transfer Characteristics in Narrow Annuli Fluid Channel

HE Ronghui, SUN Zhongning

2013, 34(S1): 123-127.

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Under the condition of 6 different equivalent diameters,using water as working media,the experiments of cooling heat transfer in annular channel is studied.The results show that the force convention heat transfer in narrow annuli is different with that in conventional channel,in former case the turbulent zone significantly ahead of schedule.When the equivalent diameter is 0.94 mm,only turbulent flow region exists in the channel.Narrow annuli can enhance or inhibit the heat transfer.Simulation and experimental results are in good agreement.The change of inner diameter core pipe has effect on the narrow annuli heat transfer.

Effects of Power Supply Frequency Change on Performance of Canned RCP Motor and Related Analysis of Design Adjustment

JIANG Xiaomao

2013, 34(S1): 128-131.

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In this paper,the effect of power supply frequency change on the performance of canned RCP motor and the design adjustment for canned motor when the bigger change of power supply frequency occurs are analyzed.A 60 Hz motor design is changed into 50 Hz motor design,and their performance are calculated by the finite element analysis method.At the same time,the comparison analysis of the performance of these two motor is performed.

Study on Flow-Induced Vibration in Pipe Conveying Fluid with Orifice Plate

LIU Xianghong, LUO Yushan, WANG Haijun

2013, 34(S1): 132-135,144.

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Taking the vibration and noise in Reactor Cavity and Spent Fuel Pit Cooling and Treatment pipe line in nuclear power plant as an example,based on the actual experimental engineering parameters,the experimental studies on fluctuating pressure excitation and pipe vibration resulted from flow disturbance induced by single local resistance element-orifice plate are conducted under the condition of the different flow rate and the same back pressure.The pipe flow field and pressure field is numerically simulated,especially the flow station of orifice plate.The simulation results and experimental data are compared and analyzed.The research results show that the energy spectrum increases with the increasing of the degree of throttle of orifice plate.As the increasing of flow rate and fluid disturbance,the spectrum breadth of pressure fluctuate increases without the disturbance of other excitation source.

Numerical Analysis of Dynamic and Heat Transfer Characteristics of Droplet Sprayed in Pressurizer

DENG Feng, HE Jingsong, HUANG Yan, LI Huanming

2013, 34(S1): 136-140.

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This paper established the dynamic model of the single droplet and the model of unsteady heat transfer between the droplet and steam,and compiled the computer program according to the characteristics formulation by Fortran language.And the dynamic and heat transfer parameters of single droplet are calculated according to various initial velocities,dimensions and injection angles.Calculation results indicate that the hang time of the droplet is determined by the vertical distance between the spray head and the steam-liquid interface.The most heat is accomplished shortly after the droplet leaving the spray head.The initial velocity and the injection angle hardly affect the droplet-steam heat transfer quantity,and heat transfer quantity between the droplet and steam is decided by the droplet dimension.

Study on Welding of Dissimilar Metals of Safety End and Pressurizer Nozzle

HUANG Junlin, LIU Hongbin

2013, 34(S1): 141-144.

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The welding of safety end and pressurizer nozzle is the welding of dissimilar metals.This paper studies the effect of some factors（such as the choose of filling material,the performance of filling material,the characteristics of welding procedure of nickel-base alloy,joint type,qualification of welding procedure and production welding） on the quality of the weld.The study indicates that,the nickel-base alloy weld molten pool is viscous,and it is suggested to agitate the molten pool gently during the welding;the welding of nickel-base alloy may generate hot cracks and gas cavity easily,so the chemical constitution of the nickel-base alloy filling materials shall be controlled strictly.

Non-Contact Precision Measurement Method for Fuel-Assembly Header

YONG Jing, LIU Zhaodong, ZHENG Hongtao, FENG Linna

2013, 34(S1): 145-147,151.

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Based on the analysis of geometric characteristics and location characteristics of the positioning pin holes of the fuel-assembly header,and combined with the visual measurement characteristics of the fuel-assembly header,non-contact precision measurement method can be proposed.The method uses an optical camera to capture the edge image of positioning pin holes.Then the error points of the edge points are removed,and the diameters and their relative position of the pin holes by coordinate values arc fitting of the edge points can be obtained.The detection accuracy of the method is up to 5μm,and in this way can reduce the human measurement error and avoid damage ferrule caused by the contact measurement.Actual operation results show the higher detection accuracy and stability of this method.

Analysis of Fuel Assembly Hold-Down System

ZHANG Lin, PU Cengping, FENG Linna

2013, 34(S1): 148-151.

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Hold-down system is an important part of the fuel assembly,whose performance has effect on the integrity and security of the fuel assembly,and it is a concern in the nuclear power plant safety evaluation.The loads of leaf spring hold-down system in life are calculated in this paper.The reason of insufficient hold-down force is analyzed and some correction measures are proposed.

Methods for Calculation of HCLPF Value for Structures and Equipments in Nuclear Power Plants

CAI Fengchun, YE Xianhui, LIU Wenjin

2013, 34(S1): 152-156.

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Calculation of the HCLPF value of structure,system and component（SSC） in nuclear power plants（NPPs） is an important process for seismic probability safety assessment（SPSA） and seismic margin assessment（SMA）.In this paper,three methods for calculation of the HCLPF value of SSC are presented,such as fragility analysis,conservative deterministic failure margin（CDFM）,and HCLPF value based on test.The recent progresses in calculation of the HCLPF value of SSC are investigated.Finally some suggestions for the calculation of the HCLPF value of SSC are proposed.

Study on 3D Stress and Fatigue Analysis of RPV Nozzle Based on APDL

YANG Wen, ZHANG Yixiong

2013, 34(S1): 157-161.

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The ANSYS APDL is programmed to optimize the process of RPV nozzle stress and fatigue analysis,the distribution of stress and fatigue usage factor can be obtained rapidly,and the evaluation is based on applicable RCC-M criteria.The results and evaluation show that the stress and fatigue usage factor in the RPV nozzle meet all applicable RCC-M criteria.

Research on Erection Techniques and Special Fixture for Main Equipments

DONG Zhengping, WENG Songfeng

2013, 34(S1): 162-163,171.

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Depending on the research of the technical characteristics and installation requirements of the main equipments in a nuclear project,as well as the disposition of the nuclear power plant,the installation techniques of the main equipments's was optimized and improved experience.Based on that,a systemic installation techniques which could be used in M310 nuclear power plant is finally carried out,and all the devices used in the installation were designed,which indicated that both the techniques and the devices needed in the installation firstly localization.

Application of Contact Analysis in Nuclear Reactor Structural Design

WANG Yaxi

2013, 34(S1): 164-167.

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The contact problems between the plate and shell often exist in nuclear reactor structure.In order to develop appropriately the reactor structure dynamic finite element model,it is important to correctly stimulate the stiffness characteristics between contact pairs.This paper focuses on the numerical calculation methods and the selection of reasonable contact arithmetic and parameter by comparing with experimental data.The nonlinear contact stiffness between contact objects in reactor structure is studied,and the numerical calculation methods are consummated.The study results verify that it can represent experimental behaviors by adopting Augmented Lagrange Multiplier method and plane stress model to analyze the contact stiffness between the plate and shell.

Analysis of Shutdown Operation Reliability of Reactor Trip Breaker

LI Hongwei, SUN Yu, ZHENG Xiao

2013, 34(S1): 168-171.

Abstract(27) PDF(0)

Abstract:
Both anti-operation probability and mal-operation probability with different structures used in shutdown reactor trip breaker are analyzed by failure-tree method.In the analysis,mal-operation probability,which is crucial for reactor safety,is especially considered to analyze the effect of shutdown operation reliability of reactor trip breaker.Results indicate that,the structure of 2/4 form reactor trip breaker for the second type is prior to other structures compared with the performance of mal-operation probability and reliability,and the improvement to decrease mal-operation probability should aim at the purpose of decreasing mal-operation probability of driving signals.

Discussion on Improved Design of RCP Electrical system Based on Frequency-Conversion Technology

HE Liang, JIE Ming

2013, 34(S1): 172-174.

Abstract(11) PDF(0)

Abstract:
The RCP electrical systems supply the power,control and protection to the RCP.The paper introduced the demand of the RCP and the actuality of the RCP electrical system,improved the RCP electrical system by the frequency-conversion technology,and analyzed the feasibility of the scheme.The analyzed result shows that the improved system design scheme is applicable to the demand of the RCP.

Security Policies for Design of Reactor Instrument ＆Control System

WU Zhiqiang, LIU Chaohui, HE Li, YANG Yang, MA Quan

2013, 34(S1): 175-178.

Abstract(17) PDF(0)

Abstract:
The security problem caused by ‘Stuxnet’ of Iran was described in this paper.A set of security design policies and methods that can strengthen the weak area of Reactor I＆C system security design are also suggested in order to avoid serious accidents.

Relationship between Architecture of Reactor Protection System and Reliability

XIAO Peng, ZHOU Jixiang, LIU Hongchun

2013, 34(S1): 179-183.

Abstract(24) PDF(0)

Abstract:
Based on the design criterion of the reactor protection system,this paper provides the analysis of measurements which improve the system reliability qualitatively,and gets the quantitative calculated results for two typical logic processing parts of reactor trip system in the way of fault tree analysis.It can be used as a reference when designing the structure of reactor protection system in the future.

Research on Reactor Control Rod Position Indication Information Monitoring Technology Based on SOPC

ZHENG Xiao, CAI Chen, SUN Yu, LIU Mingxing

2013, 34(S1): 184-187.

Abstract(13) PDF(0)

Abstract:
A new monitoring technology for control rod position is developed by utilizing the FPGA（Field-Programmable Gate Array） platform and using the SOPC（System On Programmable Chip） technology.In this SOPC system,Nios II CPU,VGA（Video Graphics Array）display controller and CAN（Controller Area Network） bus controller are integrated in one FPGA chip.Thus the SOPC hardware platform with comprehensive functionality is constructed.Based on this SOPC platform,the real time data of control rod position indication system can be vividly displayed on the LCD and stored in the external nonvolatile RAM as history records.Therefore,the operator can obtain the overall operation status of control rod position indication system quickly and conveniently.The developed prototype has proved the feasibility of this technology.

Research of Condensation and Evaporation CFD Model for Passive Containment Cooling System

HUANG Daishun, JIANG Xiaoyu, YU Hongxing

2013, 34(S1): 188-191,195.

Abstract(19) PDF(0)

Abstract:
Convection,vapor condensation and film evaporation is the key heat transfer mechanism in passive containment cooling system design adopted in AP1000 to accomplish the safety function.In this paper,a simplified model of the AP1000 PCCS has been developed and studied with considering condensation and evaporation models developed under a CFD platform of the CASTEM code and validated using the condensation tests and film evaporation tests.The research results have shown that the condensation and evaporation model can reflect the heat and mass transfer characteristics of the passive containment cooling system.

Analysis of Effect of Hydraulic Characteristic of Reactor Coolant Pump on Large-Break Loss-of-Coolant Accident

DING Shuhua, QIAN Libo, WU Dan

2013, 34(S1): 192-195.

Abstract(16) PDF(0)

Abstract:
The best estimate code WCOBRA/TRAC was utilized to model AP1000 Large-Break Loss-of-Coolant Accident.The effect of four different types of pump characteristics curves on the system pressure,break mass flow rate and peak clad temperature was analyzed.It was found that for LB LOCA accident,the difference on the peak clad temperature during the blow-down and the reflooding phase could be up to 150℃ with different pump curves.The safety margin of AP1000 could be improved through optimizing or improving the pump characteristics.

Investigation of Validating Over PowerΔT Protection

CHEN Hongxia, ZHANG Shu

2013, 34(S1): 196-200.

Abstract(14) PDF(0)

Abstract:
Overpower ΔT protection channel is respectively designed to protect the core against fuel fusion（limiting linear power density）.This paper studies the main steam line break event at full power to validate the overpower ΔT protection for M310 nuclear power plant.The influences of initial conditions,break spectrum,reactivity feedback coefficient and control rod regulation are investigated.A set of analysis method to validate the overpower ΔT protection is formulated based on the discussion.

Analytical Study on Coupling of CATHARE and TRIO_U Code for Nuclear Reactor Thermal-Hydraulic Analysis

PENG Qian, YU Hongxing, Simone VANDROUX, Fabien PERDU, LI Songyu, YANG Wen

2013, 34(S1): 201-205.

Abstract(17) PDF(0)

Abstract:
The overlapping coupling method was used for coupling the system code CATHARE（developed by CEA,EDF and Framatome） and the 3-D code TRIO_U（developed by CEA）,which were taken to analyze the steady state of a simplified model built in this paper.Several test calculations were taken before the coupling calculations,including the code error test,the coupling platform error test,verifying the result of analytical case,system calculation,TRIO_U calculation and calculations of other codes.The test calculation results are in well accordance with the results of coupling calculations through the coupling platform.Three different source domains（heat source,momentum source and heat transfer source） are also calculated using the coupling method,and compared with the results of CATHARE.The research indicated that the overlapping coupling method can be used to simulate the simplified model for the whole reactor system built in this paper.

Optimization Analysis of Overtemperature ΔT Set-point

WANG Yanping, XU Liangjian, SHEN Caifen

2013, 34(S1): 206-209.

Abstract(21) PDF(0)

Abstract:
For the plant operation which the safety margin is enough,it is positive to optimize the reactor trip signal set-point which could increase the efficiency of nuclear power plant.Overtemperature ΔT set-point optimization analysis is processed based on the accident of uncontrolled RCCA（Rod Cluster Control Assembly） bank withdrawal at power.Thermal-hydraulic sub-channel analysis code and transient analysis code are introduced into the research.The new set-point will influence the reactor trip time,DNBR and the limit of reactivity insertion rate.The result shows that the optimized overtemperature ΔT set-point leads to increase the operational margin and improve the economy efficiency when the safety margin remains large enough.The analysis provides the way to improve the reactor economy through optimizing the reactor trip signal set-point.

Development of Severe Accident Simulator Based on Event Sequence in SAMG

ZHAO Xin, LIU Dong, WANG Jiachang, HE Tengjiao

2013, 34(S1): 210-213.

Abstract(14) PDF(0)

Abstract:
A type of severe accident simulator has been developed based on event sequence in SAMG.This paper summarizes and illustrates the technology of graphic platform,scene build and data management in the simulator.An application in simulating an accident in Qinshan second nuclear power plant is described in this paper.The result shows that this simulator can build the accident scene quickly and control the simulation process of event sequence.

Study on Application of Cloud Computing Technology in Nuclear Power Plant Design Platform

PENG Hui, XIAO Anhong, YANG Dawei, WANG Zheng, GUAN Hui

2013, 34(S1): 214-217.

Abstract(14) PDF(0)

Abstract:
Considering the current problems in the use of hardware and software resources in the design of the nuclear power plant project,this paper proposed the use of cloud computing technology to build a private cloud architecture-based platform for the construction of nuclear power plant design ideas and implementing programs.Program of nuclear power plant design platform architecture based private cloud architecture elaborate the key technology and realization,resource allocation,operation management and other aspects.

Information Resources Guarantee under Environment of Scientific Research

LIU Shanyong

2013, 34(S1): 218-220.

Abstract(13) PDF(0)

Abstract:
Based on the reality of a State Key Laboratory,the paper studies the enviroment of science and technology information resources guarantee and the changes of user behavior under modern scientifical research environment,sums up misunderstandings that scientific and technolgical departments should avoided and provides a new pattern of information resources guarantee.At the same time,it discusses the sustainable development of science and technology information resources construction and puts forward the implenentation patterns of the core competencies of information department and knowledge services.

Study on CANDLE Burnup Concept

TANG Huapeng, YAN Mingyu, LU Chuan, FENG Linna, CHEN Bin, LIANG Tao

2013, 34(S1): 221-224.

Abstract(24) PDF(0)

Abstract:
This paper introduced the mathematic model and numerical solution of the CANDLE burnup theory and explained the deducing of the CANDLE burnup equation and calculation of the CANDLE burnup velocity.We approved the theoretical feasibility of CANDLE burnup,which brings high utilization efficiency of the nuclear fuels,with excellent comprehensive performance parameters,through the coupled neutronic-thermal-hydraulic equilibrium analysis of the 1000 MW TWR core.

2013 Vol. 34, No. S1