核动力工程

Study on Fuel Loading Pattern Optimization for A Pressurized Water Reactor Using Particles Warm Method

LIU Shi-chang, CAI Jie-jin

2013, 34(5): 1-5.

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Abstract:
A mathematical model for the loading pattern（LP） optimization for a pressurized water reactor（PWR） was established based on Daya Bay Nuclear Power Plant.The discreet,multi-objective particle swarm optimization algorithm was integrated with the core physics calculation code "Donjon" based on finite element method,and assembles group constant calculation code "Dragon",developing an optimizing code for fuel arrangement.This code is applied to optimize the first cycle loading of Daya Bay Nuclear Power Plant.The result shows that,compared with the reference LP,the core effective multiplication factor increases by 9.16%,while the local power peaking factor is lower than 1.4,satisfying the safety requirement.

Two-Point Reactor Core Model for Pressurized Water Reactor and Its Application

WU Yu-zhong, CHEN Shi-he, LIU Yang, ZHAO Fu-yu

2013, 34(5): 6-11.

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Based on the nodal method,a two-point core model for pressurized water reactor is developed in this paper.Furthermore,the model is formulated through the linearization of the nonlinear model,and then a Linear Time-Invariant model is obtained.To simulate the control rod system more accurately,a two-point control rod system is built.Referring to the mechanical shim control strategy in AP1000,the average temperature control system and the axial offset control system simulation program are built on the Matlab/Simulink platform and are applied to simulate load transient of step power.The simulation results show that the power derivation remains in the target band,and the steady state deviations between the actual values and target values for the reactor power and mean coolant temperature are in the dead bands.Therefore,the developed model is available for the simulation and design of the power control system and the power distribution control system.

Neutronics Analysis Code System for Hybrid Reactor

ZU Tie-jun, WU Hong-chun, ZHENG You-qi, CAO Liang-zhi

2013, 34(5): 12-15,36.

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The configuration of the fusion driven subcritical hybrid reactor is complex,and the neutron energy spectrum varies among wide energy range.In this paper,a neutronics analysis code system for the hybrid reactor,named NAPTH,is developed based on the conventional two-step calculation scheme widely used in the design of fission reactor.The lattice calculations are carried out by DRAGON4,and the core calculations are executed by multigroup function of MCNP.The validation results show that the results of IAEA-ADS benchmark have good agreement with results from other countries; and as for the pressure tube type hybrid reactor calculations,the code system is proven reliable and exhibits good calculation efficiency.The code system is suitable for the pressure tube type hybrid reactor calculations.

Research on Calculation of PWR Coolant ¹⁶N and ¹⁷N Activity Source Terms Based on Transport Calculation Method

HU Jian-jun, TANG Bin, YANG Bin

2013, 34(5): 16-19.

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The two-dimensional Sn transport calculation code DORT and ENDF/BVI library are used to calculate the neutron flux distribution in PWR reactor pressure vessel,and the special activity source terms calculation code is developed and used to calculate the reactor coolant ¹⁶N and ¹⁷N activity source terms.It is used for the validation of the effect of core radial and axial power distributions.Results of sensitivity analysis indicate that it has little effect on the ¹⁶N and ¹⁷N activity source terms calculation results for low or high leakage core load mode of PWR NPP.

Neutronics Analysis for Core Assembly in Inverted Hydride-Fueled Pressurized Water Reactor

CENG Zheng-kui, YU Tao, XIE Jin-sen, LIU Jie, QIN Mian

2013, 34(5): 20-24.

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In order to obtain the size of assembly cell with a better physical performance,a preliminary neutronics study was carried out for IPWR core using MCNP program.Firstly,the paper have studied how the k_eff changed with different H/HM（Hydrogen atoms/Heavy metal atoms） in different soluble boron concentration,and a size arrangement of assembly cell lattices is obtain.By comparing the physical performance of different size of assembly cell lattices,some assembly cell lattices which have a better physical performance are obtained.The result shows that:when the H/HM is 6.5 and water channel diameter ranges from 9 to 11mm,the assemblies have a better physical performance and an inherent safety.

Effect of Fuel Assembly When Changing from AFA 2G to AFA 3G on Seismic Loads of Reactor Internal

LIU Wen-jin, CENG Zhong-xiu, YE Xian-hui, WU Wan-jun

2013, 34(5): 25-29.

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Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established.Using ANSYS software,seismic nonlinear time-history analysis is completed and the effects on seismic loads of reactor system are obtained.The result shows that when the fuel assembly changing from AFA2G to AFA 3G,it is necessary to reevaluate the fuel assembly itself,but not the reactor internal.

Analysis of Fast Fracture of Core Region of RPV

ZHANG Li-ping, ZHENG Lian-gang, LU Yue-chuan, LIU Wen-jin

2013, 34(5): 30-32.

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The core region of reactor pressure vessel is the main part irradiated by the core,since irradiation raises the nil ductility transition temperature.This increases the sensibility to fast fracture.To prevent the vessel fast fracture of vessel,the fast fracture mechanics analysis is necessary in the design of reactor vessel.In the paper,fast fracture mechanics analysis was performed using the first method and the second method for RPV according to RCC-M.The analysis results met the allowable limits.

Comparison and Validation of Dynamic Characteristic Analytical Method for Tubular Heat Exchanger

HUANG Qing, XU Ding-geng, CHEN Meng, SHEN Rui

2013, 34(5): 33-36.

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In this study,the natural frequencies of Normal Residual Heat Removal Heat Exchangers are evaluated based on the beam and shell-beam finite element models.The corresponding results are compared and some discrepancies are observed.These discrepancies are analyzed in terms of the analysis of a cylindrical shell and the unreasonable treatment of boundary conditions is accordingly pointed out.The experimental data of the natural frequencies of heat exchangers used for Qishan Phase I Nuclear Power Plant are compared with the computational results from the shell-beam models for corresponding heat exchangers of C-2 program.The experimental and numerical results agree quite well,which implies that the shell-beam finite element simplification is applicable to the heat exchangers.The results indicate that the procedures introduced in this article apply to the dynamic analysis of other similar heat exchangers.

Analysis of Probabilistic Fracture Mechanics for PTS of Reactor Pressure Vessel

LIU Zhi-wei, QIAO Hong-wei, ZHANG Yong

2013, 34(5): 37-40,44.

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In over-cooling transient of reactors,the difference in temperature along the RPV wall will cause a large thermal stress,and then will generate a high tensile stress.Coupled with the internal big pressure,the defects in the inner surface of the RPV will propagate rapidly and even through the thickness.The probabilistic fracture mechanics theories were used in this paper.The uncertainty factors of the parameters were considered.The random numbers were generated using the Monte Carlo method on the basis of the random distribution of the parameters firstly,and then the limit function was established according to the fracture mechanics assessment criteria.Finally the failure probability was obtained by statistic calculation of the reliability data.

Comparative Research of Finite Element Methods for Perforated Structures of Nuclear Power Plant Primary Equipments

XIONG Guang-ming, DENG Xiao-yun, JIN Ting

2013, 34(5): 41-44.

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Many perforated structures are used for nuclear power plant primary equipment,and they are complex,and have various forms.In order to explore the analysis and evaluation method,this paper used finite element method and equivalent analytic method to do the comparative analysis of perforated structures.The paper considered the main influence factors（including perforated forms,arrangements,and etc.）,obtaining the systematic analysis methods of perforated structures.

Study on Application of Green’s Function Method in Thermal Stress Rapid Calculation

ZHANG Gui-he, DUAN Yuan-gang, XU Xiao, CHEN Rong

2013, 34(5): 45-47.

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This paper presents a quick and accuracy thermal stress calculation method,the Green’s Function Method,which is a combination of finite element method and numerical algorithm method.Thermal stress calculation of Safe Injection Nozzle of Reactor Coolant Line of PWR plant is performed with Green’s function method for heatup and cooldown thermal transients as a demonstration example,and the result is compared with finite element method to verify the rationality and accuracy of this method.The advantage and disadvantage of the Green’s function method and the finite element method are also compared.

Stationary Stochastic Seismic Response Analysis of Coupled Structures Interconnected by Hysteretic Dampers Subjected to Non-uniform Random Seismic Excitations with Pseudo Excitation Principle

HUANG Qian, ZANG Feng-gang, ZHANG Yi-xiong

2013, 34(5): 48-51,60.

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To calculate the stationary stochastic seismic response of coupled structures interconnected by hysteretic dampers subjected to non-uniform random seismic excitations,a pseudo excitation method based on the multi-support and multi-component was proposed.As the unknown quantity is not the dynamic relative displacement but the absolute displacement,the proposed method can be suitable for all the seismic excitation including uniform seismic excitation.Numerical research shows that:results from the proposed method,traditional pseudo excitation method based on the dynamic relative displacement and the Monte-Carlo numerical simulation method agree well,verifying the reasonableness of the proposed method.Further,parametric study of hysteretic damper about the yielding force and the yielding displacement is also developed.

Evaluation of Influence of Diffuse Earthquakes Based on Near-Source Ground Motion Acceleration Recordings

WU Jian, YU Yan-xiang, PAN Hua

2013, 34(5): 52-56.

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On the basis of near-source saturation characteristics of ground motion,the peak ground motion acceleration and the acceleration response spectra of diffuse earthquakes can be determined by the statistic method.The ground motion samples are chosen by the saturation ranges which are determined comprehensively by near-source attenuation characteristics,size of sources and range of meizoseismal area.In order to obtain sufficient recording samples,a transformation method is proposed to transform the response spectra on the soil site to that on the equivalent rock site.Finally we obtain the design response spectra and discrete characteristics for diffuse earthquakes of which magnitude is 5.0,5.5 and 6.0 respectively.

Analysis of Effect of Damping on Seismic Loads of Primary Equipment

YE Xian-hui, QI Huan-huan, ZHANG Yi-xiong, LIU Wen-jin, WANG Ming-li, GONG Jun-yong

2013, 34(5): 57-60.

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Based on the primary nonlinear system of the nuclear plant,a primary seismic analysis was completed with ANSYS software.The results show that the load is conservative using directly Rayleigh damping.If the calculation method is permission,using the modal damping can decrease the conservativeness of seismic loads and improve the economy of the plant.

Thermodynamic Analysis of Chemical Vapor Depositing Nb Coating on Fuel Particles

PAN Xiao-qiang, YANG Jing, ZHANG Liang, LI Tongye, LIU Ting-wei, WANG Xiao-min

2013, 34(5): 61-64.

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Nb coating is the chemical vapor deposited on the surface of fuel particles in the fluidized bed.Thermodynamic calculation of chemical vapor deposition process was studied with HSC Chemistry 6.0 code.According to the data,the effect of processing parameters on the composition of the resultant was discussed.The result shows that:higher deposition temperature is in favor of the generation of Nb and the content of by-products is reduced; with the increasing of the H₂/NbCl₅ mole ratio,the Nb content in the resultant increases; some Ar addition improves the productivity of Nb and presses the formation of by-products.

Simulation Study on Flow Uniformity in Casting Nozzle during Preparation of UO₂ Kernel

LIU Ma-lin, HAO Shao-chang, LIU Bing, DENG Zhang-sheng

2013, 34(5): 65-70.

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The computational fluid dynamics（CFD） simulation was adopted to study the internal flow field in nozzle chamber,which was involved in the casting nozzle in the preparation process of UO₂ kernel particles.In particular,the flow flux uniformity analysis of each nozzle,which is the important determinant factor of UO₂ kernel size uniformity,was focused on.Four possible factors that may affect the flow flux uniformity of nozzle in the actual production process were investigated,including the inflow direction,the nozzle distribution,the nozzle chamber levelness and the nozzle orifice size.From the simulation results we can be found that,relative to the other three influencing factors,the nozzle orifice size is the key factor and is one of the main reasons for giving rise to the non-uniformity of particle size.The simulation results suggested that the size of the nozzle orifice should be particularly concerned about in the actual production processes.Furtherly,the mathematical analysis of the flow equation was given,and the theoretical validation of the simulation results was drawn out.The conclusions of this study can be used to develop targeted experimental studies further and also provide a reference to improve the uniformity of particle size.

Optimization of Carbonization Process in Manufacture of Fuel Elements for HTGR

LU Zhen-ming, ZHANG Jie, ZHOU Xiang-wen, LIU Bing, TANG Ya-ping

2013, 34(5): 71-75.

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The decomposition process of phenolic resin which is a binder of spherical fuel elements matrix material for high temperature gas-cooled reactor（HTGR） was studied by means of TG-IR and DSC at the temperature of 20～800 ℃.The thermal expansion characteristics of graphite ball samples were investigated through thermal mechanical analysis（TMA）.The results showed that the phase transition of the phenolic resin occurred at the temperature of 50～70℃ and the decomposition had two continuous exothermic stages.Accordingly,the graphite ball samples were expanded and shrink successively during carbonization.The properties of the matrix material made through four-stages heating process completely satisfied the design requirements and the production efficiency of carbonization was increased by 71%.Studies showed that the heating process established according to the sample’s size varied with carbonization temperature under dynamic conditions is more reasonable and the volume change caused by condensation is the critical factor to heating rate.

Study on High-Cycle Fatigue Behavior of 6XN Stainless Steel and Alloy 825

XIONG Ru, QIAO Ying-jie, ZHAO Yu-xiang, TANG Rui, YI Wei, XIAO Ze-jun

2013, 34(5): 76-79,83.

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The high-cycle fatigue（HCF） experiments of 6XN stainless steel and alloy 825 were conducted under bending and rotating loads at room temperature（RT） as well as at 550℃ in air.The results indicate that the fatigue limited stress of 6XN at RT is higher than that of 825,which consistent with the order of their tensile strength.The oxidation rate of the specimen increased at 550℃,therefore the fatigue life of the specimen decreased,among them 6XN was more sensitive to high temperature with the larger decreasing tendency which make the fatigue limited stresses of the two alloys more closer at 550℃.While 825 is more sensitive to the stress cycles,both materials have good resistance to high cycle fatigue when comparing their experimental data with the calculated value from the empirical formula.The fracture morphology presents the areas of crack initiation,crack growth and fracture,and the fracture area has much dimples.

Study on Feasibility of Replacing 321 with 316LN Stainless Steel for Main Reactor Coolant Pipe Material

LUO Yi-jun

2013, 34(5): 80-83.

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The metallurgical,physical and mechanical performance,and the corrosion and welding properties 0f 00Cr17Ni12Mo2（controlled Nitrogen,ANSI316LN） and 0Cr18Ni10Ti（ANSI321SS） for main pipe material were analyzed comparatively in this paper.The feasibility of 316LN pipe material manufacturing was studied too.The analysis results showed that under the operation condition of the nuclear reactor,the general properties of 316LN are better than that of 321SS.Therefore,316LN could be used for main pipe material,replacing 321SS.

Study on Performance of Novel Overlaying Welding Materials on the Sealing Face of RPV

DANG Ying, WEN Yao-ceng, QIU Shao-yu, CHEN Yong, LUO Qiang, LI Chuan-qian

2013, 34(5): 84-88.

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Compared with E308L,the hardness,microscopic structure and intergranular corrosion resistance of E308LMoT0-3 and E309LMoT0-3,which are selected as the novel overlaying welding materials on the sealing face of RPV,were investigated.Meanwhile,the pitting potential and the pit corrosion and crevice corrosion resistance of the materials were also tested under the condition of water deviating.The results show that the corrosion resistance of the E308LMoT0-3 and E309LMoT0-3 are better than that of E308L.And the results could also provide reference for the material selection of the remedy and design of the overlaying welding on the sealing face of RPV.

Limitations and Shortcomings of Duplex Stainless Steel Choice in RCC-M

SHEN Yu-sheng, XIAO Jian

2013, 34(5): 89-91,99.

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The performance and property of duplex stainless steels is introduced,and the limitations and shortcomings of duplex stainless steel choice in RCC-M are analyzed based on the application in sea water pumps and a large amount of comparative analysis data.

Degradation of Zirconium Alloy Pressure Tube in Qinshan CANDU-6 Heavy Water Reactor and Its Mitigation

ZHAO Wei-dong, SHI Xiu-qiang

2013, 34(5): 92-95.

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Pressure tubes are critical components in heavy water reactors,and its ageing and degradation directly has effect on the service life of the reactor.The material of the pressure tube of Qinshan Phase Ⅲ CANDU-6 reactors are zirconium alloy（Zr-2.5Nb）,and the dimensions and mechanical properties of pressure tubes change under long-time service conditions of high temperature,high pressure,and high radiation.Also there is risk that delayed hydride cracking（DHC） may happen on pressure tubes due to the pickup of deuterium by zirconium alloy.To improve the radial creep rates of pressure tubes,reduce the residual stress and risk of DHC,and follow up the degradation extent of pressure tubes to provide important information and evidences for life assessment,Qinshan Phase Ⅲ implemented the improvements of manufacture and installation process,control of service chemical conditions and periodic inspections,and will implement preventative maintenance.

Reliability Residual-Life Prediction Method for Thermal Aging Based on Performance Degradation

REN Shu-hong, XUE Fei, YU Weiwei, TI Wen-xin, LIU Xiao-tian

2013, 34(5): 96-99.

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The paper makes the study of the nuclear power plant main pipeline.The residual-life of the main pipeline that failed due to thermal aging has been studied by the use of performance degradation theory and Bayesian updating methods.Firstly,the thermal aging impact property degradation process of the main pipeline austenitic stainless steel has been analyzed by the accelerated thermal aging test data.Then,the thermal aging residual-life prediction model based on the impact property degradation data is built by Bayesian updating methods.Finally,these models are applied in practical situations.It is shown that the proposed methods are feasible and the prediction accuracy meets the needs of the project.Also,it provides a foundation for the scientific management of aging management of the main pipeline.

Efficiency Contrast Test for Initial and One Refueling Cycle Used Catalytic Plates of Passive Autocatalytic Recombiner

GUO Xiang-li, YU Tao, XU Rui-yin, WANG Hai-wei, ZHU Jian-bin, JIANG Xiang, LUO Ya-bin

2013, 34(5): 100-103.

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A sort of Passive Autocatalytic Recombiners（PAR）,including function,component,site arrangement,in-service test method and period,used in Qinshan Second Nuclear Power Plant Unit 3 and 4,were introduced.Efficiency comparison tests of four groups of initial catalytic plates and that used for one refueling cycle（1C） were conducted.Rust and oil stain issues,which were prone to occur during the commissioning and operation process,impacting on the hydrogen reduction efficiency of catalytic plates were put to test and analysis.The results show that,compared with the initial samples,the hydrogen depletion rate of the catalytic plates after 1C is slightly decreased with the influence of oil stain and dust.A small amount of rust and oil stains does not have a significant impact on the efficiency of the catalytic plates.However,rust is liable to cause the catalytic plate to be perforated and to be damaged.Flue gas generated in the test of the oil stain plates is easy to interfere with the dehydrogenation process.

PWR Core Power Control and Simulation Based on Multi-Model Strategy

WU Yu-zhong, CHEN Shi-he, LI Gang, ZHAO Fu-yu

2013, 34(5): 104-107.

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The Pressurized Water Reactor（PWR） core nonlinear model is built and the linearized models of the core at five power levels are respectively calculated as local models of the core to approximately substitute the core model.The full-state feedback method with a full-state observer of a local model is utilized to design a controller with robustness of every local model,which is treated as a local controller of the core for controlling the nonlinear core within a corresponding range of power level.The simulation results show that the core multi-model control system can satisfactorily control the core power.

Analysis and Research of PWR Instrument Commissioning Based on Simulink

LUAN Zhen-hua, LIU Daog-uang, CHOU Shao-shuai, YANG Zong-wei, FENG Guang-yu

2013, 34(5): 108-111.

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Based on Simulink platform,a mathematical model of the lead lag link,differential unit,arithmetic logic module is built.Considering the specific problems encountered in the debug field work,this model is applied in the analysis of key modules and controller and in the resolving of eddy current problems.The test process dynamic control characteristic is simulated,to analyze the trend of the actual response,and conduct the simulation study and propose the concrete solutions.The actual debugging process proved that the use of simulation technology to find the problem,optimize the control data,and adjust the control strategy is very important for the early detection of problems and to speed the test process,shorten the debug duration and increase the debug quality.

Process Management during Construction of Digital I＆C System in Tianwan Nuclear Power Plant

DUAN Peng, XU Jie, SHI Qing-wei, REN Chun-xiang

2013, 34(5): 112-114.

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The full digital I＆C system was used in Tianwan Nuclear Power Station Phase I（Unit 1＆2）.This paper describes the reflection and implementation of the experiences learned from the contractual negotiation and execution in Phase I,and the practice and optimization obtained during the system operation in the construction of Tianwan Nuclear Power Station Phase II（Unit 3＆4） for I＆C System.

Application of AFAL Methodology in Substantiation of Instrument Calibration Intervals Extension in Nuclear Power Plant

ZHOU Ping, QIU Chun-hui, CHU Qi-bao

2013, 34(5): 115-117,131.

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This paper introduces the AFAL analysis methodology for nuclear power plants and its application in the demonstration of the extension of instrument calibration interval.The paper discusses the technological means and implementation.By the analysis of instrument drift,the feasibility for the extension of the instrument calibration interval is demonstrated.On the premise of nuclear safety,the interval of fuel reload overhauling in the nuclear power plants can be prolonged.

Research on Heat and Mass Transfer Model for Passive Containment Cooling System

JIANG Xiao-yu, YU Hong-xing, SUN Yu-fa, HUANG Dai-shun

2013, 34(5): 118-123.

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Different with the traditional dry style containment design without external cooling,the PCCS design increased the temperature difference between the wall and the containment atmosphere significantly,and also the absolute temperature of the containment surfaces will be lower,affecting properties relevant in the condensation process.A research on the heat and mass transfer model has been done in this paper,especially the improvement on the condensation and evaporation model in the presence of noncondensable gases.Firstly,the Peterson’s diffusion layer model was proved to equivalent to the stagnant film model adopted by CONTAIN code using the Clausius-Clapeyron equation,then a factor which can be used to stagnant film model was derived from the comparison between the Y.Liao’s generalized diffusion layer model and the Peterson’s diffusion layer model.Finally,the model in CONTAIN code used to compute the condensation and evaporation mass flux was modified using the factor,and the Wisconsin condensation tests and Westinghouse film evaporation on heated plate tests were simulated which had proved the improved model can predict more closer value of the heat and mass transfer coefficient to experimental value than original model.

Natural Circulation Characteristics in a Symmetrical Two-Circuit Loop under Inclined Condition

YANG Xing-tuan, ZHU Hong-ye, GONG Hou-jun, LIU Zhi-yong, JIANG Sheng-yao

2013, 34(5): 124-127.

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Natural circulation characteristics of an integrated natural circulation reactor were investigated theoretically and experimentally.Experiments were carried out on a test loop that had a symmetric two-circuit configuration.The inclination condition was achieved by a marine condition simulation platform.A series of data,such as the total flow rate,branch flow rates,inlet and outlet temperatures of electrical heater and heat exchangers,were obtained.Both experimental and analytical results showed that under inclination condition the total flow rate was reduced,whereas one branch flow rate was increased; meanwhile the temperature difference between hot leg and cold legs was increased.It is revealed that density difference existed between the up plenum and down plenum,which caused an additional driven force perpendicular to the original one and introduces an outer circulation.Under the co-action of inner and outer circulation,one branch circulation was enhanced and the other was depressed.This unbalance between branch circulations can be alleviated by shortening the distance between heat exchangers and enlarging the distance between the heat exchanger and heating section.

Experimental Study on Over Reading Coefficient in Wet Steam Flow Measurement

PENG Xing-jian, HE Can-yang, YUAN De-wen, HUANG Yan-ping, BAI Xue-song, ZHANG Yan

2013, 34(5): 128-131.

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Experiment of wet steam flow measurement was performed based on existing experimental loop.Different geometry type of Venturi was used as the test object with different steam quality and mass flow rate within the operation range.The effect of pressure,LM parameter,mass flow rate and throat radius on the over reading factor was analyzed.

Numerical Study on Over Reading Coefficient in Wet Steam Flow Measurement

BAI Xue-song, YUAN De-wen, YAN Xiao, PENG Xing-jian

2013, 34(5): 132-134,138.

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This paper investigated the flow process of wet steam in Venturi under interested conditions with CFD simulation software.The effect of pressure,mass flow rate,throat radius on over reading factor was analyzed.This paper aims to improve the wet steam over reading model and the prediction accuracy in wet steam.The results prove that the mass flow has a small effect on over reading coefficient,while the effect that throat radius has on over reading coefficient increases as the pressure rises.

Discussion on Operation Occasion of Pre-Service Inspection and Main Primary System Hydrostatic Test of PWR

SUN Hai-tao, ZHANG Yue, WANG Chen, SHENG Chao-yang, JIA Pan-pan

2013, 34(5): 135-138.

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By comprehensive analysis of Guide 103/07 of Nuclear Safety of the People’s Republic of China（HAD 103/07）,In-Service Inspection Rules for the Mechanical Components of PWR Nuclear Islands（RSE-M）,Section XI of American Society of Mechanical Engineers Boiler and Pressure Vessel Code（ASME）and French 1999 decree about in-service supervision items for main primary and second loops of PWR Nuclear Plant,and etc,as well as combined with the present domestic engineering practice in pre-service inspection,besides the analysis of the affects respectively caused by that pre-service inspection executes before hydrostatic test and hydrostatic test executes before pre-service inspection,technique views were formed for operation occasion between pre-service inspection and hydrostatic test for main primary system.

Research on pH Automatic Adjustment System in Radioactive Decontamination Waste Liquid Evaporation Treatment System

YU Ren, KONG Jing-song, XIANG Xin-min, GUO Wei-qun

2013, 34(5): 139-141,153.

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During the operation process of the waste liquid evaporation treatment system,the pH value of the radioactive decontamination waste liquid will change constantly,and lead to the violent fluctuation of the liquid level in the evaporator.In the paper,the causes of this phenomenon are analyzed,and the technical solution for this problem which introduces a waste liquid pH value online monitoring and automatic adjustment system into the waste liquid loading pipeline is designed.By adding suitable amount of acid or alkali solution automatically into the loading waste liquid,its pH valve can be stabilized within the given range.

Design of Maintenance Pattern for Valves in Radioactive Waste Processing Facilities

KONG Jing-song, YU Ren, MENG Kai

2013, 34(5): 142-144.

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Based on the types and characteristics of valves in radioactive waste treatment facilities,a set of valve maintenance pattern are designed,and the valve maintenance time,steps,content and method are defined clearly,to ensure the safety of nuclear facilities through regular maintenance.

Research on Monitoring and Management Information Integration Technique in Waste Treatment and Management

KONG Jing-song, YU Ren, MAO Wei

2013, 34(5): 145-148.

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The integration of the waste treatment process and the device status monitoring information and management information is a key problem required to be solved in the information integration of the waste treatment and management.The main content of the monitoring and management information integration is discussed in the paper.The data exchange techniques,which are based on the OPC,FTP and data push technology,are applied to the different monitoring system respectively,according to their development platform,to realize the integration of the waste treatment process and device status monitoring information and management information in a waste treatment center.

Analysis of Radwaste Treatment Localization Scheme at Nuclear Power Plants

LIU Pei, LIU Yu, YAO Bing, DAI Bo

2013, 34(5): 149-153.

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According to principle of radwaste minimization in nuclear power plants,off-reactor radwaste treatment scheme is a very potential choice for multi reactor and specialization operation nuclear power plant.The off-reactor radwaste treatment scheme of AP1000 is introduced and compared to traditional CPR1000radwaste treatment methods.Then,the optimization scheme of off-reactor radwaste treatment is provided,which is hoped to be a reference to newly designed plant radwaste treatment facility or operation plant radwaste treatment system improvement.

Analysis of Falling Characteristics of Small Absorber Ball in HTR Shutdown Device

CHEN Feng, HE Xue-dong, LI Tian-jin, CAO Li, LUO Jie, SUN Bo, HUANG Zhi-yong

2013, 34(5): 154-156.

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The falling process of small absorber ball in the drop-ball device of HTR absorber ball shutdown system has been studied in this paper.Glass ball with the identical dimension of the actual absorber ball diameter（φ6 mm）,and drop-ball device made of plexiglass tube with the identical inside diameter（φ60mm）of the actual absorber ball falling canal in the core have been applied.Experiments of glass ball falling due to gravity in the air environment have been accomplished.The calculation results coming from Beverloo empirical formula,which has been used to calculate mass flow rate of particle at orifice owing to gravity,agrees well with experimental results in this research with the deviation of 4% to 12%.The angle of repose of the measured glass ball in a ball storage vessels made of plexiglass is about 23.Experimental results showed that glass ball in the vertically above center orifice firstly begin to flow,and then the circumambient particles supply to the orifice center in the ball storage vessels designed with the cone angle of the funnel.Glass balls gradually decline freely and smoothly from ball storage vessels.

Design of Test Device for RPN Neutron Source Component

WANG Hong-bo, ZHUO Wen-bin, WAN Hai-sheng, SONG Jian, YANG Tai-bo, WEI Dong

2013, 34(5): 157-159.

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A special neutron source component test unit was designed based on RPN source test needs,and related tests were conducted.The findings showed that the test unit not only could provide ideal test data for the source test instrument commissioning personnel,but also could maximally meet the requirements of radiation protection for the neutron source operators.

Research of Electrochemistry Behavior of Corrosion Inhibitor for KAA of Tianwan NPP

LIU Jin-hua, GONG Bin, JIANG E, MA Wei-gang, HAN Bin, WEN Ju-hua, XIE Yin-yan

2013, 34(5): 160-164.

Abstract(14) PDF(0)

Abstract:
Based on the corrosion issue of cooling water system in Tianwan nuclear power plant,the corrosion electrochemistry of inhibition on copper and stainless steel was studied using electrochemical method.Results show that inhibitor have excellent corrosion inhibition efficiency on copper in low chloride concentration solution and simulated solution.The optimum inhibitor is the compound consisting of TTA and sodium orthophosphate.At the same time,the inhibitor elevated the breakdown potential of stainless steel and contributed to the enhancement of corrosion resistance.

2013 Vol. 34, No. 5