核动力工程

Study on Breeding of Fast Zone Assembly in SCWR-M

SAN Dong-yong, YU Tao, XIE Jin-sen, PENG Hong-hua, ZENG Zheng-kui

2013, 34(3): 1-4.

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In this paper,a new breeding design of the fast zone in Supercritcial water-cooled Mixed-spectrum Reactor（SCWR-M） is proposed on the basis of the neutron energy spectrum characteristics,that it is harder than the thermal reactor’s and softer than the fast breeding reactor’s.Then,to increase the number of neutron leakage in the fission layers of fuel rods,the layers of fuel rod are increased and the conversion ratio of the fast spectrum assembly in SCWR-M is improved.Finally,the breeding ratio of the fast spectrum assembly in SCWR-M achieves 1.06569 by increasing the fuel rod proliferation layer thickness,correspondingly to reduce its fission layer thickness.

Study on Life Extension of AP1000 Nuclear Power Plant CORE

LIU Zhan

2013, 34(3): 5-7,23.

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At the end of AP1000 plant core life,the critical boron concentration required to maintain full thermal power conditions reaches approximately 7 ppm.To achieve full power conditions,it is necessary to perform a core life extension during end of life（EOL） conditions.The paper,in the base case,analyzes the methods to control the turbine throttle valve and decrease the average coolant temperature,making the plant gain an additional 17 days of rated full power operation.Furthermore,the effect of the moderator temperature coefficient（MTC） and the number of high pressure feedwater heaters out of service have also considered in a sensitivity analysis.The result shows that,the more negative the MTC,the smaller the decreasing rate of the core coolant average temperature and the longer the core life.However,in two sensitivity cases,the plant will separately gain an additional 12 days and 54 days of rated full power operation.

Study on Mismatch Phenomenon of Power Change and Reactivity

XIN Feng, YOU Guo-shun, ZHANG Chun-ming

2013, 34(3): 8-10.

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The mismatch phenomenon of reactor power change and reactivity is found in the analysis of main steam line break（MSLB） accident.the point-reactor kinetics proved that the traditional inhour equation can not describe this phenomenon effectively.The origin for the mismatch of reactor power change and reactivity is analyzed.The reactor period equation with dynamic reactivity would include the differential of the reactivity,which can describe such cases well.This equation also showed the physical meaning of such cases.

Study on Super Equivalence Method

LI Man-cang, WANG Kan, YAO Dong

2013, 34(3): 11-16.

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Based on the general equivalence theory（GET） and the superhomogenization method（SPH）,the super equivalence method（SPE） has been proposed to preserve the surface current,the reaction rates and the lattice reactivity,using the implicit equivalence factors.SPE has been implemented into the Monte Carlo method based homogenization code MCMC.The C5G7 benchmark problem has been calculated to verify the SPE method.The results show that the SPE method provides improved accuracy and adaptation compared to the traditional GET and SPH methods.

Application of Runge-Kutta Method in 3-D Coupled Netronics and Thermal-Hydraulics Transient Analysis

ZHAO Wen-bo, YAO Dong, WANG Kan

2013, 34(3): 17-23.

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The coupled neutronics and thermal-hydraulics model is studied with the module coupling.Nodal Green’s Function Method（NGFM） based on the second boundary condition is used for solving neutronics module.The sub-channel code COBRA-Ⅳ is used for solving thermal-hydraulic module.Diagonally Implicit Runge-Kutta（DIRK） method is applied in the temporal discretization of coupled nonlinear equations.Automatic time step size control is achieved through Richardson Extrapolation or embedded lower order Runge-Kutta formulas.Hermite interpolation is used for continuous time variable solutions（dense output）.The solving procedure is based on block-iterative method,and the NCC code is developed.The numerical results show that the solution of NCC agrees well with reference solutions,and the DIRK（2,2）-E formula is more accurate and efficient than others in this paper.

Study on Direct Extraction of Uranium from Uranium Oxides Powder Using Supercritical CO₂

LI Jia, ZHU Chang-gui, LIU Xiao-long, REN Meng, LIU Jin-hong

2013, 34(3): 24-27,62.

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The factors with effects on direct extraction of uranium from uranium oxides powder is studied using supercritical CO₂/TBP-HNO₃ extraction system.The results show that the extraction efficiency of uranium（E_U） is increasing obviously with an increased HNO₃/TBP,and E U is the best at the ratio value of 1.4;Under certain HNO₃/TBP,E_U is direct proportional to the volume of organic phase.The extraction reaction can be divided into three phases: react slowly in initial phase,fast in stable phase and slowly in final phase due to powder condition;As temperature increasing,E_U has maximum value 91.4% at 60℃;As pressure increasing,E_U has maximum value 94.7% at 12 MPa;According to extraction equipment condition,the best separation of extraction products can be achieved in 8 MPa（separation vessel 1） and 5 MPa（separation vessel 2）.

Study on Preparation of B₄C-Zircaloy Annular Burnable Absorber

LI Yuan-yuan, DAI Sheng-ping, YANG Jing, PEI Qi-lin, QIAO Hui-wu

2013, 34(3): 28-31.

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B₄C-zircaloy annular burnable absorber pellet was prepared by powder metallurgy method,using Zr-2 hydride powder and B₄C powder as raw materials,compacted by rubber isostatic pressing method,dehydrogenated in vacuum furnace.It is sintered at 800~1100℃ about an hour.The theoretical density and open porosity were tested by immersion method.The effect of the B₄C content and sintering temperature on the sintering performance of the B₄C-Zircaloy were studied.The microstructure of green body and sintered body were analyzed by SEM.B₄C hinders the densification of sintered body,the relative density of the sintered body were increasing with the decreasing of B₄C content and increasing of sintering temperature.The particle of B₄C distribute uniformly.When the B₄C content was 0.5% and the sintering temperature above 1000℃,the relative density of the sintered body reaches 95%,and the open porosity is lower than 1%.

Study on Cladding Oxidation Model Improvement in SCDAP/RELAP5

YU Hong-xing, HE Xiao-qiang, SU Guang-hui, DONG Zheng-ping

2013, 34(3): 32-36.

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At present,the parabolic rate correlation oxidation model is used in SCDAP/RELAP5 code for simulating cladding oxidation during reflooding in the early phase of severe accident,but it can not simulate oxidation well when the cladding temperature is high or the steam flowrate of cladding surface is small.Besides,this model does not analyze the oxygen distribution in the cladding and can not describe the failure of cladding suitably.In this study,those differences and relationships between parabolic rate model and diffusion model are analyzed,and the diffusion oxidation model is employed in SCDAP/RELAP5.Then,the effect of those two models on reflooding in the early phase of severe accident is investigated.It is shown that the diffusion model can simulate the reflooding phenomena better in the early phase of severe accident,and the parabolic rate model is a simplified form of diffusion model based on some specified assumptions.

Effect of TIG Welding on Corrosion Resistance of Zirconium R60702

LING Kun, WANG Zheng-dong

2013, 34(3): 37-41.

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Protection and the nonequilibrium crystallization of TIG welding on Zr R60702 were applied to prepare the specimens of welded joints for metallographic observation and corrosion potential determination.Microstructures of the specimens were analyzed with a high power metallographic microscope.Polarization curves of the specimens were measured for evaluation of the corrosion properties.The microstructure and second phase distribution of Zr R60702 were studied on their influence to the corrosion properties.The results indicate that the corrosion resistance properties of the HAZ is lower.

Effect of Hydrogen on Irradiation Stability of Dispersed Oxide Particles （Y₂O₃） in 12Cr-ODS Steel

GUO Li-na, HU Ben-fu, JIA Cheng-chang, H. Takahashi, H. Kinoshita, S. Watanabe

2013, 34(3): 42-45.

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Irradiation damage behavior was investigated by hydrogen ion-electron ion（H⁺+e^-） dual-beam irradiation.The morphology change was observed in-site,and the effect of（H⁺+e^-） dual-beam irradiation on the stability of oxide particle（Y₂O₃） in a new 12Cr-ODS ferrite steel,which was produced by using ethylene diamine tetraacetic acid（EDTA） complex method,was studied in this work.The results show that after electron ion irradiation with a dose of 15 dpa followed by an（H⁺+e^-） dual-beam irradiation dose of 15 dpa,the size of Y₂O₃ particles decreased or dissolved in this particular condition,and the radiation-induced point defects and the chemical reaction around the interface are the main reasons for the instability of the oxides.

Effect of Burn-off on Oxidation Rate of Nuclear Graphite

WANG Peng, YU Su-yuan

2013, 34(3): 46-49.

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The air oxidation performance of IG-110 grade nuclear graphite was investigated at 600-750℃ using thermal-gravimetric method,and the effect of burn-off on the oxidation rate was analyzed.Applying the Random Pore Model,the microstructrual parameters of nuclear graphite IG-110,H-451,NBG-18 and NBG-10 were determined,and the burn-off factor formulas were fitted.Results indicated that the maximum oxidation rate of IG-110 graphite occurred at ~35% burn-off,and the microstructrual parameters increased with the mean grain size of different types of nuclear graphite.

Research on Reflood Model of Narrow Channel Based on RELAP5

ZENG Wei, YU Hong-xing, SUN Yu-fa, LI Feng

2013, 34(3): 50-57.

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This paper evaluates the applicability of RELAP5 reflood models for the narrow channel based on the Saxena experiment.The calculated results show that RELAP5 overestimates the quench front velocity and underestimates the wall temperature.Combined the post-DNB thermal-hydraulic characteristics in the narrow channel,the interphase drag,single vaper,film boiling and critical heat flux model were modified.The calculated results based on the new reflood model agree better with the Saxena experiment.

Experimental Investigation of Subcooled Flow Boiling Incipience in Volumetrically Heated Packed Bed

ZHANG Xiao-ning, SUN Zhong-ning, MENG Xian-ke, XU Guang-zhan

2013, 34(3): 58-62.

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Based on the innovative water-cooled pebble-bed nuclear reactor,an experimental investigation was performed to study the characteristics of subcooled flow boiling incipience in packed bed with internal heat generation.Oxidized carbon steel balls with 8 mm diameter were used to form a random packed bed,which was volumetrically heated by induction.Effects of interstice mass flux G,inlet temperature T_inand vertical position on the heat flux of ONB was investigated.Results indicate that there is a turning point on the relation curve of wall temperature T_w and heat flux q,which is the subcooled flow boiling incipience in volumetrically heated packed bed.The heat flux of ONB increases with the increasing of mass flux and decreases with the increasing of inlet temperature,respectively.The heat flux of ONB is smaller as the measurement position is closer to the exit of packed bed.A dimensionless relationship was developed to predict the q_ONB in packed bed with internal heat generation,and the deviation between predicted and experimental data is in the range of ±20%.

Numerical Simulation of Fluid Flow and Heat Transfer in Pebble Bed Reactor

XU Guang-zhan, SUN Zhong-ning, MENG Xian-ke, ZHANG Xiao-ning

2013, 34(3): 63-66.

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The DEM approach is used to simulate the random distribution of fuel spheres.The purpose of this work is to study the flow fluid and heat transfer by means of CFD techniques.Analysis shows that the pressure is lower,but the velocity,eddy and temperature are higher in the clearance.The pressure,eddy and heat transfer coefficient are increased with the porosity reduction,but velocity is reduced with the porosity reduction.

Research on Void Fraction Models for Vertically Upward Two-Phase Cross-Flow in Horizontal Tube Bundle

JIANG Nai-bin, ZANG Feng-gang, ZHANG Yi-xiong, GUAN Hui

2013, 34(3): 67-70.

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Compared with the void fraction models of two-phase flow in tubes,the research on void fraction models for vertically upward two-phase cross-flow in horizontal tube bundle are not enough.With experimental data,the existing void fraction models of vertically upward two-phase cross-flow in horizontal tube bundle are compared.Moreover,two existing void fraction models are modified.The modified void fraction models are validated by some other experimental data.The result shows the modified void fraction models give better prediction than the original models.

Transient Flow Characteristics of Nuclear Reactor Coolant Pump in Recessive Cavitation Transition Process

WANG Xiu-li, YUAN Shou-qi, ZHU Rong-sheng, YU Zhi-jun

2013, 34(3): 71-76.

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The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX,and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed.The flow field analysis shows that,in the recessive cavitation transition process,the speed diversification at the inlet is relative to the bubble increasing,and makes the speed near the blade entrance increase when the bubble phase region becomes larger.The bubble generation and collapse will affect the the speed fluctuation near the entrance.The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase,and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet.Pump asymmetric structure causes the asymmetry of the flow,velocity and outlet pressure distribution within every impeller flow passage,which cause the asymmetry of the transient radial force.From the dimensionless t/T =0.6,the bubble phase starts to have impact on the impeller transient radial force,and results in the irregular fluctuations.

Study on Flow Characteristics of Integral Reactor Test Facility under Rolling Motion without Heating

GONG Hou-jun, YANG Xing-tuan, HUANG Yan-ping, JIANG Sheng-yao

2013, 34(3): 77-81.

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To investigate the flow characteristics of the full power natural circulation integral reactor under ocean condition,a series of experiments under rolling motion without heating were performed using test facility and followed theoretical analysis using flow model in non-inertial reference system.The loop flow rate changes cyclically with the rolling angle and no water flow into heat exchangers from heating channel through rise section.The amplitude of flow rate oscillations depends on facility configuration,flow resistance,position of rolling axis and rolling angle acceleration,and the amplitude can be reduced by decreasing the length of distribution channel.The phase lag between flow rate oscillation and rolling angle correlates with test facility dimension,and the quantity in direct proportion to rolling frequency and in inverse proportion to flow resistance coefficient.

Study on Homogeneous Flow Model for Air-Water Two-Phase Flow in Mini-Rectangular Channel

XING Dian-chuan, YAN Chang-qi, SUN Li-cheng, JIN Guang-yuan, WANG Guang-fei

2013, 34(3): 82-86.

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The applicability of homogenous flow model to vertical mini-rectangular channel is evaluated by using air and de-ion water as the working fluid.The results show that the homogenous flow model with McAdams viscosity and the Chen modified homogeneous flow model could predict the pressure drop for the rectangular channel having the narrow gap of 1.41mm with the Mean Absolute Error（MAE） of 10.92% and 12.20%,respectively.The homogenous flow model with McAdams viscosity fails to predict the pressure drop for the channel having the gap of 3mm,except that the two-phase Reynolds number is greater than 6000 in which the MAE is 10.04%.The scale of homogenous flow model applied for rectangular duct having the narrow gap of 3mm is proposed by analyzing the experimental data.A new correlation is proposed for the channel with the gap of 3mm when the gas and liquid flow rate is low,and shows good agreement with the experimental data.

Study on Density Wave Oscillation in Parallel Channels by Section Form

HUANG Jun, HUANG Yanping, WANG Yanlin

2013, 34(3): 87-90,95.

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Based on 170 density wave oscillation experimental data from parallel round tube and narrow rectangular channel,the experiment method,identification method of oscillation and analysis method of experimental data have be uniformed,and the oscillation boundary of round tube and narrow rectangular channel have be analyzed.The investigation results show that the oscillation boundary is not affected by the channel section forms with identical equivalent diameter with pressure 1.0~19.2 MPa,mass flux 101.9~1200.0 kg.m^-2.s^-1 and inlet sub cooling 18.0~85.2℃.

Numerical Simulation of Steam Condensation Inside Vertical Tube in Presence of Noncondensable Gases

HUANG Zheng

2013, 34(3): 91-95.

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Numerical model is established based on mass,momentum,energy and composition equations by the analysis of liquid film and steam-air boundary layer.The influence of air on condensation heat transfer and the effect of interfacial shear stress and suction have been simulated.Calculated results are compared with Kuhn’s experimental results and the agreement is satisfactory.Thus the model is validated.

Numerical Simulation of Multiphase Flow and Boiling using Lattice Boltzmann Method

LI Long-jian, HU An-jie, CENG Jian-bang, HUANG Yan-ping

2013, 34(3): 96-100.

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A new lattice Boltzmann method was used to simulate the multiphase flow and boiling in the pool and horizontal pipe in this peper.Numerical results showed the process of bubbles growing up and breaking away from the bottem with the effect of buoyancy and interaction force,and collision and aggregation can be found during the rising of bubbles.Also,the effect of horizontal acceleration on pool boiling was studied.The above results proved the ability of the present model to simulate the phase change problem.

Parameter Analysis for Pebble Dynamics of Two-Dimensional TwoRegion Pebble-Bed Reactor using Phenomenological Method

WANG Wen-jun, YANG Xing-tuan, JIANG Sheng-yao, TANG Zhen-wei, LIU Zhi-yong

2013, 34(3): 101-104.

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In order to investigate the parameter effects on the pebble dynamics in the high-temperature reactor core where the fuel pebbles and graphite pebbles form two regions,based on the two-dimensional pebble-dynamics experiments,the discrete element method（DEM） was used to simulate the pebble-dynamics and phenomenological method was used to analyze the two-region arrangement,the size and shape of the central pebble column,the dispersion in the mixing region and the stagnant regions at different friction coefficient and different cone angle.The results show that with the increasing of the friction coefficient,the ball flow tends to more uniform,the stagnant zone increases and the mixing zone decreases.While the cone angle increases,the flow seems more even,the stagnant zone and the slow zone decrease.

Experiment of Pebble Flow in Two-Region Pebble-Bed Reactor Core Vessel

TANG Zhen-wei, YANG Xing-tuan, SUN Yu-liang, WANG Wen-jun, LIU Zhi-yong

2013, 34(3): 105-108,112.

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The design of two-region pebble-bed reactor core represents one of the main technology orientation of high temperature gas-cooled reactor（HTGR）.The 2-meter experiment vessel is established to carry out the study on the fundamental characteristics of pebble dynamics which include the formation of two-region,mixing zone and stagnant zone,and etc.The result indicates that a steady two-region can be obtained under the experiment condition with a strip of mixing zone exist in the boundary where two kinds of pebbles scatter together.Mixing zone can be reduced effectively with the installing guide plates on the surface.There exists a visible stagnant zone at the bottom of the vessel which will be highly concerned.The experiment summarized the basic pebble behaviour in the reactor core which is randomicity and dispersibility as well as certainty in general.

Study on Nuclear Reactor Power Control Based on T-S Neuro-Fuzzy Network

LIAO Long-tao, CHEN Zhi

2013, 34(3): 109-112.

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Nuclear power plant runs at a large operation range and keeps a set of parameters of the reactor stable by controlling the nuclear power.In this paper,in order to improve the effect of nuclear power control,an adaptive fuzzy controller based on T-S neuro-fuzzy network is studied.The advantage of the controller is illuminated by the simulation results under multiple operation conditions and comparison with the classical PID controller.

Application of Ethernet-Based Distributed Measurement and Control System in Thermal Hydraulic Tests

KUANG Shu-qing, WANG Hai-ying, ZHONG Yan-min, ZHENG Hua

2013, 34(3): 113-115.

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The application of a new,industrial Ethernet-based distributed I/O measurement and control system is introduced.The system architecture,hardware configuration,and software design are respectively described in detail,as well as the system feature analysis.Proved by tests,this system has the advantages of high reliability,strong anti-interference capacity,high precision and easy to expand.Good control effects have been obtained in practical applications.

Reconstruction of Water Level of Steam Generator Based on Elman Neural Network

DUAN Meng-qiang, ZHENG Ying-feng, CHEN Yu-qing, WANG Gui-li

2013, 34(3): 116-119.

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Falsehood or forfeiture occurs frequently on the water level indication meters of the steam generator,which grievously affects the judgment of the operator on the nuclear-power plant.Elman Neural Network is a typical dynamic neural network.It can reflect the characteristics of dynamic system directly,especially when dealing with the complicated non-linear object.A method that rebuilding the water level of steam generator with Elman Neural Network in the circumstance of the crevasse of primary stream pipeline appears is described in the paper.In comparison with the emulation results,it shows that Elman Neural Network can reconstruct the water level of steam generator exactly.The outcome can meet the practical needs and give guidance on the safe motion of the marine nuclear-power plant.

Best Estimate Plus Uncertainty Method and Its Application

RAN Xu, WU Dan, CHEN Bing-de, ZHANG Yu

2013, 34(3): 120-123.

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Best estimate plus uncertainty method could be used to eliminate the over-conservatism in the current method and gain more safety margin.In the current paper,uncertainty method which has considered the effect of the initial conditions,boundary conditions,and model uncertainties has been developed based on the best estimate code RELAP5.This method was used to analyze the NEPTUN reflood experiment.The results showed that the new method was reasonable.

Experimental Study on Emergency Ventilation to Reduce Nuclear Accident Consequence

LIN Xiao-ling, NIU Wen-liang, CHEN Ying-feng

2013, 34(3): 124-125,163.

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Based on the model of nuclear ship equipment,the tracer diffusion experiments were conducted,with the related systems and facilities simulated in the actual conditions.SF₆ was used as tracer,simulating the radioactive source term.The tracer was released in the reactor cabin,and the concentration of cabins were monitored separately in two different conditions,i.e.,with and without the operation of emergency ventilator.The results were compared and analyzed.The study shows that emergency ventilating could reduce greatly the radioactivity in the ship.The radioactivity in the reactor cabin was reduced over 99% after the ventilator worked 30 minutes,and the radioactivity in other cabins were almost reduced to zero after 5 hours.In comparison,the reduction of radioactivity were not obvious without the operation of the ventilator.

Development of Non-destruction Test Device by γ-Ray Scanning

ZHANG Jian-bing

2013, 34(3): 126-128.

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This paper describes the development of non-destruction test device by γ-ray scanning and solving the problem of non-destruction test in ultra-thick iron castings.firstly,the work pieces to be tested are divided into several regions.The device collects the regional data in order,and we can judge the exist of defect and the size of the defect after the analyst compared the abnormal dat（a Number of mutants: the value is bigger usually）and uniform numeral.Defects position of the ultra-thick iron casting is determined according to the corresponding number of abnormal data,and we have tested ultra-thick iron castings such as shielding door by the device.The error of test results is within ±5 mm.

Application and Modification of On-Line Monitoring System for Containment Leakage in Qinshan Second Nuclear Power Plant

WANG Hai-wei, SHI Chao-jie, YANG Gang, CHEN Can, GUO Xiang-li, JIANG Xiang, ZHANG Hui

2013, 34(3): 129-132.

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The on-line monitoring system for containment leakage,including the frame and the calculation model,used in Qinshan Second Nuclear Power Plant,was introduced.Different typical false alarms of the system,which were found during the using process,were analyzed in detail.The analysis is useful for judging and eliminating in time the false alarms when meeting the similar cases.The modification function of manual intervention and the implementation effect were introduced.The results show that,the false alarms can be eliminated as long as the exception data could be found and deleted in time,and this method can ensure the safe reliability of the nuclear power plant.

Seismic Response Analysis of Circulation Pump in AP1000 Nuclear Power Plant

OU Ming-xiong, ZHAO Li, SHI Wei-dong, TIAN Fei

2013, 34(3): 133-136.

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To investigate the large double-base vertical mix-flow circulation pump used in AP1000 nuclear power plant,using the security shut-down earthquake load as the design seismic load,the dynamic characteristics and seismic response of sea water circulation pump under seismic load are analyzed,and the operability during SSE load is evaluated.The results demonstrated that the first natural frequency is about 17.5 Hz,and its low-order mode shapes are almost lateral bending in pump case.The combined results of response spectrum and static analysis of pump demonstrate that,the max combined stress on pump case is 173 MPa,and the max.deformation is 4.4 mm.

Study of Venturi Geometry Parameter Design for the Better Wet Steam Measurement

WANG Hai-ping, YUAN De-wen, HE Can-yang, XIE Tao, CAO Zhi-yong, WANG Jun-jie

2013, 34(3): 137-139.

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This paper was to investigate the flow process of wet steam in Venturi under interested conditions with CFD simulation software based on the thermal paramete of oil field industry.The influence of pressure,mass flow rate,and Venturi geometry parameter on Over Reading factor was analyzed.The better Venturi geometry parameter was obtained.This experiment aims to improve the wet steam Over Reading model and the development of advanced wet steam flowmeter.

Thermal Research of CITP-ⅡTritium Production Irradiation Device

ZHANG Zhi-hua, MI Xiang-miao, DUAN Shi-lin, DENG Yong-jun, QIAN Da-zhi, LI Run-dong

2013, 34(3): 140-143.

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As the core component of CITP-Ⅱ,the online tritium production irradiation device is the pivotal equipment in the research of tritium breeders’ tritium production and release.This paper has studied the effect of breeder heat power,hotspot,grads of temperature distribution,utmost temperature,uneven factor,the components of gap gas and carrier gas to the balance temperature.The key thermal parameters and its change trend are obtained.

Dropping Accident Analysis of Liquid Storage Container using Finite Element Method

NIE Jun-feng, ZHANG Hai-quan, LI Hong-ke, WANG Xin, ZHANG Zheng-ming

2013, 34(3): 144-147.

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Considering the fluid-structure interaction effects in the analysis of liquid storage container dropping,coupling Euler-Lagrange（CEL） method is used to simulate the inertia effect of fluid and the lateral hydraulic pressure to the container,and the fluid-structure interaction effects on the deformation and dynamic response of the container during the process of dropping is considered.The numerical result shows that the method can provide a more accurate evaluation of vessel safety and structural design.Therefore,the CEL mentioned in the paper also provides a reference design evaluation method for the same structure.

Stress and Fatigue Analysis for Lower Joint of Control Rod Drive Mechanisms Seal House

SHAO Xue-jiao, ZHANG Li-ping, DU Juan, XIE Hai

2013, 34(3): 148-151.

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Two kinds of seal houses for control rod drive mechanisms which have different thickness of the lower seal ring was analyzed for its stress and fatigue by finite elemet method.In the fatigue compution,all the transitions were grouped into several groups,and then the elastoplastic strain correction factor was modified by analyzing thermal and mechanical load separately referring the rules of RCC-M 2002.The results show that the structure with thicker seal ring behaves more safely than the other one except in the second condition.Meanwhile,the amplify of the primary and secondary stress as well as fatigue usage factor can be reduced by regrouping the transients.The precision of fatigue usage factor can be elevated using modified Ke when the amplify of the primary and secondary stress is large to some extent produced by both thermal and mechanical loads.

Transient Analysis for CPR1000 Reactor Main Pump and Electromotor

XIAO San-ping

2013, 34(3): 152-155.

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The operation characteristics of electromotor are simulated with RETRAN-02 when the electric network frequency falls based on the electro mechanics principle.The hydraulics characteristics of CPR1000 reactor main coolant pump and the resistance characteristics of the reactor coolant system are also simulated.The operation transient of the reactor coolant pump and electromotor is studied in detail when the electric network frequency falls.In the end,the transient behavior of CPR1000 reactor coolant pump and electromotor is analyzed when the electric network frequency falls at the rate of 4 Hz/s.

Study on Chemical Cleaning Formula for Steam Generator Secondary Side during Tianwan NPP Cooling Down

XIA Xiao-jiao, WEN Ju-hua, MA Wei-gang, XIE Yin-yan, HU Zhi-hui

2013, 34(3): 156-158.

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The effectiveness and safety of steam generator chemical cleaning process during reactor plant cooling down was validated in the test.Autoclave was used to simulate the process parameters of steam generator chemical cleaning process during reactor plant cooling down.The effectiveness and safety test of chemical cleaning was performed with Russia formula and A3B1 formula.Experimental results showed that A3B1 formula was more effective and safer than Russia formula for steam generator chemical cleaning.Experimental results also showed that chemical cleaning with composition A3B1 can effectively remove the sludge from SG secondary side of Tianwan NPP.

Root Cause Analysis of Ventilation System Motor Shaft Fracture in Nuclear Power Plants

YANG Li-fei, CHE Yin-hui, LÜ Qun-xian, GUAN Jian-jun

2013, 34(3): 159-163.

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The motor shaft rupture incident of Ling’ao nuclear power plant was analyzed by the international popular technique of root cause analysis.Firstly,all the probable failure modes of the motor rupture cause were analyzed with the efficient analytical techniques of Metallographic,Scanning Electron Microscope,Spectral Element Analysis,Stress Analysis and the Feedback of the Maintenance and Vibration test.And then the non-related failure modes were rejected,and the root cause and the contributor cause of the motor rupture were evidenced.Finally,the efficient maintenance and management advice are given to ensure the similar motor in a safe and reliable condition.

Analysis of Interfaces in AP1000 EOP Procedures

YE Cheng, ZHENG Ming-guang, WANG Yong, XUE Shan-hu, XU Yong-hua

2013, 34(3): 164-167,174.

Abstract(18) PDF(0)

Abstract:
AP1000 Emergency Operation Procedures（EOP） are based on the observed symptoms to recover the safety functions.Performing EOP procedures in time can reduce the severe accident risk as much as possible.However,the EOP entry interfaces are complex,and not limited to setpoints in the protection and safety monitoring system.Analyzing all EOP entry interfaces in AP1000 EOP procedures is very helpful for operators to understand the criteria of safety shutdown and engineered safety features（ESF） actuation,and at same time,it is also helpful for the optimization of EOP procedures.

Study Scheme of Reactor Decommissioning Simulation Technology

DAI Bo, ZHANG Yong-ling, ZHOU Bin, YAN Jing, WU Wei

2013, 34(3): 168-171.

Abstract(14) PDF(0)

Abstract:
Based on the functional requirement of nuclear reactor decommissioning simulation system,the overall study scheme of the decommissioning simulation technology is designed,including the overall structure design,software development platform scheme,hardware development scheme and etc.Five key technologies,which are the data processing technology,the technology for visual display of radiation field,the computation of personal exposure doses based on decommissioning program,the technology of virtual cutting and dismantling,and the technology of collision detection,are analyzed and extracted from the overall scheme,and its design scheme are provided in general.

Feasible Research on VLLW Disposal in Control Area of Nuclear Installation

KONG Jing-song, GUO Wei-qun

2013, 34(3): 172-174.

Abstract(18) PDF(0)

Abstract:
Based on the basic requirements on the VLLW landfill disposal specified by the national codes and standards,a on-site disposal of VLLW in the control area of nuclear installation was proposed.A detail analysis of the advantages and disadvantages about the disposal method and the problem to be solved were described.Results showed that the on-site disposal of VLLW in the control area of nuclear installation was feasible in practice.

Design of Detection Platform for Radioactive Waste Package

KONG Jing-song, MENG Kai

2013, 34(3): 175-178.

Abstract(17) PDF(0)

Abstract:
Detection of radioactive waste standard package is an important part in the radioactive waste management.Considering the features of the radioactive waste package and special security requirement,the advanced detection technique and processing technology from China and international are used as the reference,and a set of package detection platform with high control precision and wide continuous range of adjustability is finally designed.The force analysis and practical application of platform indicate that the platform is with fine operability.

2013 Vol. 34, No. 3