核动力工程

Quasi One Dimension Modified Computing Method for Multi-Cycle Fuel Management Calculation

LIU Yuhua, WU Hongchun, ZHOU Yongqiang

2007, 28(1): 1-4,8.

Abstract(4) PDF(4)

Abstract:
A quasi one dimension modified computing method is proposed in this paper.Batch relative powers are modified according to the real batch positions in this method.To consider the influence of different boundary positions,the peripheral assembly can be divided into many batches.Compared with the traditional zero dimension point reactor model,the quasi one dimension modified computing method can more exactly describe the multi-cycle real core reload calculations.The verified calculations have been executed for Qin-ShanⅠ（NPP） 1～6 cycle in this method.Compared with the calculated results given by the three dimensioncore code RSIM,the maximum cycle burnup difference is 2.89% and the maximum discharge burnup difference is 1.30%.The maximum discharge burnup difference is approximately one third of that of the zero dimension point reactor model.

Solution of Point Reactor Neutron Kinetics Equation Using Precise Time Integration Algorithm

ZHANG Xixi, HE Bin, FU Guangzhi

2007, 28(1): 5-8.

Abstract(7) PDF(1)

Abstract:
The precise time integration algorithm is an explicit numerical integration algorithm with absolute stability,which was applied to solve the point reactor neutron kinetics equation in this paper,with the numerical examples provided.The results indicated that the precise time integration algorithm could solve the stiff problem of the point reactor neutron kinetics equation commendably in long time-step condition with additional traits such as compact computing process,fast velocity and high precision,which verifies the practicability and the availability of the method in the safety and simulation analysis of the reactor.

Development and Benchmark Calculation of Multigroup Monte Carlo Code MCMG

JIANG Xiaofeng, LUAN Xingfeng, DENG Li

2007, 28(1): 9-12,17.

Abstract(7) PDF(1)

Abstract:
A multigroup Monte Carlo Code MCMG is developed based on the continuous-energy Monte Carlo Code MCNP.The continuous-energy point-cross sections are substituted by the homogenized multi-group cross sections as the function of burnup by the lattice code WIMS,which speedup the computation procedures and supply the capability of burnup simulations for the Monte Carlo Code.Because of the phenomenon that the negative self-scattering cross sections cause the negative probability sample,a nonnegative corrected method is put forward and validated by the benchmark problems.

Research on Capability of Accelerator Driven Coupled Fast/Thermal Spectrum System

JIANG Xiaofeng, XIE Zhongsheng

2007, 28(1): 13-17.

Abstract(6) PDF(1)

Abstract:
The advantages of simultaneous transmutation of actinide nuclides and fission products in the accelerator driven coupled fast/thermal spectrum system（ADFTS） were investigated.Through the neutronics analysis of ADFTS,the energy amplifying principle was interpreted.The concepts of neutron amplification factor and neutron coupling factor of ADFTS were brought forward,and the calculating method of neutron amplification factor was introduced.And then,the breeding capability of accelerator driven system（ADS） was studied.The study results showed that ADS had higher capability in breeding fuel from fertile material over the traditional reactor.

Simplified Spherical Harmonics Method for Neutron Time-Space Kinetics Calculation in Complex Environment

CAO Liangzhi, WU Hongchun, YAO Dong

2007, 28(1): 18-21.

Abstract(7) PDF(1)

Abstract:
The simplified spherical harmonics（SPN） method is utilized to discretize the angular variables of neutron transport time-space kinetics equation.The finite element method,fully implicit scheme and direct analytical time integration method are used to deal with the spatial,time variables and delayed neutron precursor equation,respectively.According to the model,a computer program is developed to solve multi-dimensional time-space kinetics equation in unstructured-meshes.The numerical results show that this method can be used to perform the kinetics calculation in complex environment.

Experimental Study of Mechanical Properties on Spacer in NHR

JIANG Yueyuan, SHI Jibin, XU Yong

2007, 28(1): 22-25,31.

Abstract(6) PDF(1)

Abstract:
The spacer of NHR-200 is composed mainly of the inner,outer and cornual strips which are ranged in egg-crate of 12×12-3.First,the pre-distortion of three kinds of three-arc springs on reactor working condition and their related clipping-force ranges are analyzed in this paper.Secondly,the mechanical experiments of 1:1 prototype,such as the load-distortion experiments,which the load and distortion are respectively measured by strain gauge and displacement sensor,of three kinds of springs,rigid supports and the spacers in two different directions are carried out on a special experimental facility.The experimental results show that the spacer can completely meet the design demands of mechanical properties of the fuel assemblies in NHR-200.

Analysis of Fatigue Life of Thick-Wall Cylinder with Inner Surface Crack Subjected to High Pressure

SONG Shuncheng, HUANG Jianwen, XU Xiang

2007, 28(1): 26-31.

Abstract(5) PDF(1)

Abstract:
Based on the basic formulation of K_I and using Schmitt-Keim coefficient and considering the effects of the pressure medium,the K_I of two thick-wall simulated tubes with initial cracks had been calcu-lated and the calculative results were agreed with the tests.The Paris equation was used as the model of crack growth rate and its coefficients were given in this paper.The fatigue lives of the thick-wall simulated tubes were calculated and the relative curves of fatigue life with pressure were obtained.For generality,the formulations for conservative calculations of the fatigue life of thick-wall tube with inner surface crack were pro-posed and can be used to estimate the fatigue life of thick-wall tube.

Thermal-Mechanical Analyses of Helium-Cooled Test Blanket Module

LI Minghai, SHI Guangmei, WANG Xiaoyu, HU Gang, ZHANG Guoshu

2007, 28(1): 32-35.

Abstract(6) PDF(1)

Abstract:
Test Blanket Module（TBM） is a key component of the International Thermonuclear Experimental Reactor（ITER）.Its design involves multi-disciplinary optimization analysis.The design description of blanket concept based on a solid breeder,a helium cooling,high temperature structural materials and beryllium multiplier is introduced.To validate the blanket concept,the thermal-mechanical analysis has been per-formed for the proposed design blanket models.The analysis results show that the highest temperature and thermal stress in each unit of the blanket do not exceed the limits.Therefore,the proposed blanket concept is safe and reliable under normal operation conditions.

Status and Existing Problems of Critical Heat Flux Lookup Table

DAN Jianqiang, ZHU Yulong, LI Changying, TAN Shunqiang, D.C.Groeneveld

2007, 28(1): 36-40.

Abstract(6) PDF(1)

Abstract:
CHF Look Up Table（LUT） has been widely applied in codes of nuclear reactor ther-further in 2005 LUT are discussed:prediction in low flow region;effect of diameter on CHF;prediction of CHF at near the critical pressure.

Prediction of Critical Heat Flux by Using Artificial Neural Network

WU Junmei, SU Guanghui

2007, 28(1): 41-44.

Abstract(5) PDF(1)

Abstract:
Three artificial neural networks（ANNs） are trained based on three types of databases to predict critical heat flux（CHF） in the present paper.The input parameters of the ANNs are the system pressure,mass flow rate and equilibrium quality/inlet subcooling,and the output is CHF.The detail effects of system pressure,mass flow rate,equilibrium quality and inlet subcooling on CHF are analyzed based on the trained ANNs.The ANNs are applied successfully for the predicting of CHF.The predicted results agree very well with experimental data.The analyzed results show that the ANN with the highest accuracy for predicting CHF is the one based on the type II database in the three types: inlet,local and outlet conditions.

Comparison of Two Different Fuel-Coolant Interaction Systems

CHEN Linghai, LUO Rui, WANG Zhou, YANG Xianyong

2007, 28(1): 45-48.

Abstract(4) PDF(1)

Abstract:
In the fuel-coolant interactions,water and Na are the two most commonly used coolants.It is proved that the water can produce violent interactions with molten metal and results in steam explosion.With the development of the MFBR technique,the problem of reactor safety must be solved urgently.Water-FCI phenomenon and Na-FCI phenomenon are compared by analyzing experimental results and numerical simulation,thus providing a basis for the analysis of MFBR safety.

Study on Coupling of Three-Dimension Space Time Neutron Kinetics Model and RELAP5 and Improvement of RELAP5

GUI Xuewen, LUO Bangqi, CAI Qi

2007, 28(1): 49-52,86.

Abstract(3) PDF(1)

Abstract:
A two-group three-dimension space-time neutron kinetics model is applied to the RELAP5 code,which replaces the point reactor kinetics model.A visual operation interface is designed to convenience interactive operation between operator and computer.The calculation results and practical applications indicate that the functions and precision of improved RELAP5 are enhanced and can be easily used.The improved RELAP5 has a good application perspective in nuclear power plant simulation.

Study on Application of Two-Fluid Model in Narrow Annular Channel

CHEN Jun, YANG Yanhua, ZHAO Hua

2007, 28(1): 53-57.

Abstract(6) PDF(1)

Abstract:
The Chexal-Harrison two-phase wall and inter-phase friction models developed by EPRI newly and the simple two-phase wall and inter-phase heat transfer models put forward by the paper are used to set up the two-fluid model which is fitted for boiling heat transfer and flow in narrow annular channel.On the base of the two-fluid model,a thermal hydraulic code-THYME is accomplished.Then the thermal hydraulic characteristic of narrow annular channel is analyzed by RELAP5/MOD3.2 code and THYME code.Com-pared with experimental data,RELAP5/MOD3.2 underestimates the outlet steam,and the results of THYME is agreed with the experimental data.

Simulation of Flow Excursion and Thermal Siphon under Natural Circulation Condition with Lower Pressure and Lower Quality

WANG Jianjun, YANG Xingtuan, JIANG Shengyao

2007, 28(1): 58-62.

Abstract(5) PDF(1)

Abstract:
In the experiments performed in HRTL-5,the system was in static instability under some conditions.In this paper,based on the drift flow model,the simulation of flow excursion phenomenon was achieved under HTRL-5 condition.In the same time,the mechanism analysis and the simulation of thermal siphon process were also performed.The result shows that the program can be used to to simulate the static instability of natural circulation system,and the simulation of the excursion phenomenon matches up to the experiment result basically.

Numerical Study of Turbulent Pulsating Flow in RectangularNarrow Channel with Periodically Mounted Longitudinal Vortex Generators on One Sidewall

WU Feng, WANG Qiuwang, WANG Ling, WANG Gang, HUANG Jun, HUANG Yanping

2007, 28(1): 63-67.

Abstract(4) PDF(1)

Abstract:
The turbulent pulsating flow and heat transfer in a rectangular narrow channel with periodically mounted Longitudinal Vortex Generators（LVGs） on one sidewall was numerically investigated by solving three-dimensional unsteady elliptical Navier-Stokes equations.The standard k～ε turbulent model was adopted.The dynamic response of fluid outlet temperature and the heat transfer of the LVGs were numerically analyzed in a pulsating period,and it was further investigated by changing the frequency of the pulsating flow.It was found that the pulsating flow could enhance the heat transfer of LVGs when the pulsating frequency increasing,and then the heat transfer of LVGs was improved.

Experimental Study of Flow Friction Characteristics of Integral Pin-fin Tubes

DING Ming, YAN Changqi, SUN Licheng

2007, 28(1): 68-71.

Abstract(8) PDF(1)

Abstract:
Friction characteristics of integral pin-fin tubes,through which lubricating-oil flowed vertically,were studied experimentally.Effects of the pitch,the height of fins and the machining direction on friction coefficient were analyzed.The experimental results showed that the friction coefficient of the integral pin-fin tube was obviously lager than that of smooth tube.Compared with other influential factors,the effect of the height of fins was dominant.Because the three-dimensional pin fin could disturb and destroy the boundary layer,when the Reynolds Number reached 200~300,the friction coefficient curve began to bend,that was,a turning point was appeared in the friction coefficient curve.

Pressure Drop Correlations of Two-Phase Bubble Flow in Rolling Tubes

CAO Xiaxin, YAN Changqi, SUN Zhongning

2007, 28(1): 72-77.

Abstract(6) PDF(1)

Abstract:
A series of experimental studies of frictional pressure drop for single water phase and gas-liquid two-phase bubble flow in two different rolling angles and three different rolling periods were carried out in the paper.Based on the experimental data,a new correlation for calculating single-phase frictional coefficient under rolling condition was presented,and the calculations not only were agreed well with the experimental data,but also could display the periodically dynamic characteristics of frictional coefficients.Applying the homogeneous flow model,two-phase frictional pressure drop of bubble flow in rolling tubes could be caculated,and the results showed that the relative error between the calculation and experimental data was less than ±25%.

Nucleate Boiling Characteristics of Water Jet Impingement on Large Superhydrophilic Surface

LIU Zhenhua, QIU Yuhao

2007, 28(1): 78-82.

Abstract(5) PDF(1)

Abstract:
The nucleate boiling heat transfer characteristics of a bar water jet impingement on a large flat superhydrophilic surface were experimentally investigated.The superhydrophilic heat transfer surface was formed by a TiO₂ coating process.The experimental results were compared with those from the common metal surface.In especial,the quantificational effects of the flow conditions,heating conditions and the coating methods on the critical heat flux（CHF） were systemically investigated.The experimental data showed that the nucleate boiling heat transfer from the superhydrophilic surface was significantly different from those from the common metal surface.The critical heat flux of boiling on the superhydrophilic surface was greatly increased by the decreasing of the solid-liquid contact angle.

Sliding Mode Fuzzy Control for a Once-through Stream Generator

ZHANG Guifeng, SHI Xiaocheng, SUN Tieli, XIONG Jinkui, ZHANG Hongguo

2007, 28(1): 83-86.

Abstract(4) PDF(1)

Abstract:
A once-through stream generator is important equipment in nuclear power plant,so its control level is high.A Sliding Mode Fuzzy Controller inherits the robustness property of Sliding Mode Control and the interpolation property of Fuzzy Logic Control.The robustness property of variable structure system results show that satisfying control results can be got.

Safety Consideration in Design of Closed Loop Control System for Miniature Neutron Source Reactor

CHEN Yu, GUO Chengzhan, WANG Liyu, SUN Huibin, ZHAO Haige

2007, 28(1): 87-89,104.

Abstract(6) PDF(1)

Abstract:
The configurations of the new closed-loop control system are introduced briefly.The system,served by two computers connected with LAN,is made to control Miniature Neutron Source Reactor（MNSR） which is built in Shenzhen University.The technique measures and design polestars for power control and safety protecting are described in the paper.The measures include the restricting of the built-in reactivity,using right working model with dual control bars,using control-bars dropping system in emergencies,setting the right factors for working/protection/emergencies,and taking the countermeasures for potential risks.The pivotal variables are checked redundantly and could be shared between two computers.This also enhanced the reliability of the system for the reactor.

Plasma Density Remote Control System of Experimental Advanced Superconductive Tokamak

ZHANG Mingxin, LUO Jiarong, LI Guiming, WANG Hua, ZHAO Dazheng, XU Congdong

2007, 28(1): 90-93.

Abstract(6) PDF(1)

Abstract:
In Tokamak experiments,experimental data and information on the density control are stored in the local computer system.Therefore,the researchers have to be in the control room for getting the data.Plasma Density Remote Control System（DRCS）,which is implemented by encapsulating the business logic on the client in the B/S module,conducts the complicated science computation and realizes the synchronization with the experimental process on the client.At the same time,Web Services and Data File Services are deployed for the data exchange.It is proved in the experiments that DRCS not only meets the requirements for the remote control,but also shows an enhanced capability on the data transmission.

Study on the Risk-Informed Regulation of NPP

WANG Chaogui

2007, 28(1): 94-98.

Abstract(5) PDF(1)

Abstract:
The risk-informed regulation is a modern type of NPP safety management mode using both deterministic and probabilistic approaches.It is necessary to entirely and systematically study the associated regulations,standards and practices in order to promote the developments of risk-informed regulations in China.This paper introduces the risk-informed regulation,gives out the basic principles,method and acceptance risk criteria of risk-informed decision-making,discusses the PSA requirements for risk-informed decision-making and makes some suggestions about the application of risk-informed regulations in Chinese NPP.

Primary Studies on Factors for Hydrogen Concentration Distribution in Containment under Severe Accidents

XIAO Jianjun, ZHOU Zhiwei, JING Yingqing

2007, 28(1): 99-104.

Abstract(6) PDF(1)

Abstract:
Three factors that may affect the hydrogen concentration distribution in the containment were investigated.Calculation results by FLUENT and GASFLOW were compared.The results indicate that RNG k-ε model can obtain better results in parameter fluctuation caused by turbulence,velocity field and hydro-gen concentration field than the other models.The results calculated by algebraic model in GASFLOW are different from the other results in the simulation of mass diffusion,momentum diffusion,and parameters fluctuation.The effects of steam concentration and mass flow rate at the leakage on hydrogen concentration distribution have been also investigated.The results indicate that they are two important factors having effects on hydrogen concentration distribution.Turbulence modeling and simulation of turbulent buoyant jets and plumes,which need further study,are of importance in hydrogen concentration distribution in the containment under severe accidents.

Nuclear Safety Culture Star-Class Assessment System Based BP Neural Network

JIAO Xiaoyou, SONG Shouxin, WU Junyong

2007, 28(1): 105-109,114.

Abstract(6) PDF(1)

Abstract:
In order to build the safety culture for nuclear power industry,tt is important to evaluate the safety culture scientifically.Considering the traits of safety culture in the nuclear power industry,24 safety culture assessment indexes are established from 4 aspects such as Safety consciousness,Safety attitude,Safety action and Safety actuality by using the SMART criteria.Safety culture star-class assessment criterion is presented and safety culture star-class assessment system is developed by using Visual Basic 6.0 and BP neural network.The system has a better generalization ability,and it can show exactly which phase the safety culture is in.Experimental results show that safety culture star-class assessment is practical and easy to perform.

Study on Intelligence Fault Diagnosis Method for Nuclear Power Plant Equipment Based on Rough Set and Fuzzy Neural Network

LIU Yongkuo, XIA Hong, XIE Chunli, CHEN Zhihui, CHEN Hongxia

2007, 28(1): 110-114.

Abstract(6) PDF(1)

Abstract:
Rough set theory and fuzzy neural network are combined,to take full advantages of the two of them.Based on the reduction technology to knowledge of Rough set method,and by drawing the simple rule from a large number of initial data,the fuzzy neural network was set up,which was with better topological structure,improved study speed,accurate judgment,strong fault-tolerant ability,and more practical.In order to test the validity of the method,the inverted U-tubes break accident of Steam Generator and etc are used as examples,and many simulation experiments are performed.The test result shows that it is feasible to incorporate the fault intelligence diagnosis method based on rough set and fuzzy neural network in the nuclear power plant equipment,and the method is simple and convenience,with small calculation amount and reliable result.

Design of Ventilation System of Reactor Building for CARR Engineering

LI Jianmin, RONG Feng

2007, 28(1): 115-119.

Abstract(6) PDF(1)

Abstract:
The ventilation rate of the reactor building is determined by the calculation results of radiation protection,which has referred to the demands of code,the experience of German FRMII reactor designing,isolation valves,of which the in-out-leakage rate is 0 under the reactor building tightness test pressure（12.5kPa）,are set on the air ducts which through the sealing boundary of operating hall.This measure can guarantee the radioactive matters against leaking into outside spaces through the ventilation ducts during accident conditions of the reactor.Direct-connected steel fans and integral stainless steel air cleaning equipments are chosen in the system designing.

Study of Hydrogen Induced α/β Phase Evolution in N18 Zirconium Alloy

LIU Yanzhang, ZHAO Wenjin, J-L Béchade, T.Guilbert

2007, 28(1): 120-123,134.

Abstract(6) PDF(1)

Abstract:
The transformation behavior of N18 alloy with and without hydriding was investigated using Setaram Multi HTC high-temperature and high-sensitivity calorimeter.The results showed that the α/β phase transformation temperatures decreased with the increase of the hydrogen content,and the lower transus temperature was more affected by hydrogen than the higher one.It was easy to simulate the phase evolution in N18 alloy using the JMAK model.

Virtual Simulation of Operation and Control of PWR Based on High Level Architecture

LIN Yajun, ZHANG Dafa, FANG Baoguo

2007, 28(1): 124-127,134.

Abstract(7) PDF(1)

Abstract:
Considering the function of PWR primary loop system,an architecture and federations design of primary loop system operation and control simulation based on High Level Architecture（HLA） is presented.MÄK’s VR-Link and RTI is used to construct a distributed network for simulation.The implementation of math model reckon federation,3-dimensional primary visual federation and 2-dimensional state federations are researched by the compilation of VC++ programs,MATLAB and Vega.The HLA is well extended and reused,so the system can be extended to implement the simulation of the whole nuclear power system.

Study of Hydrazine Deoxygenation in Neutral Aqueous Solution

WEN Juhua, WANG Zhengguang, QIU Tian, LI Yong

2007, 28(1): 128-130.

Abstract(4) PDF(1)

Abstract:
The efficiency of hydrazine deoxygenation added catalyst in the neutral aqueous solution is discussed.The test conditions are selected at temperatures of 10℃,20℃ and 30℃,the initial dissolved oxygen concentration of 10.0~10.8mg/L（10℃）,8.2~8.6mg/L（20℃） and 7.0~7.4mg/L（30℃）,and the hydrazine dosages are 6 and 3 times of the initial dissolved oxygen concentration.The results show that the dissolved oxygen concentration can be decreased to be 100μg/L when the catalyst dosage is within 600μg/L~50μg/L.

Disposition of Used Reactor Vessel Head

FAN Liming

2007, 28(1): 131-134.

Abstract(5) PDF(1)

Abstract:
In order to prevent the stress corrosion cracking of the housings in the reactor vessel heads,two reactor vessel heads in Daya Bay Nuclear Power Station have been replaced in 2003（unit 2） and reference for the future disposition of larger radioactive solid waste in China.2004（unit 1）.This paper will introduce the solution and process of the disposition of used heads,which include packing design,structure,used head cleaning,coating,packing,handling and transportation process.The result shows that,the design and site operation process is satisfactory,in the hope that this will create a

2007 Vol. 28, No. 1