核动力工程

Overview of Research and Development （Phase Ⅰ） on Key Technologies for Supercritical Water-Cooled Reactor

XIAO Ze-jun, LI Xiang, HUANG Yan-ping, TANG Rui, LUO Qi, ZANG Feng-gang, LI Qing, LI Peng-zhou, YI Wei

2013, 34(1): 1-4,14.

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Abstract:
The paper briefly introduces the overall objective, technical index and R＆D plan of Supercritical Water-Cooled Reactor （SCWR）, with detailed description on projects and subjects of R＆D （Phase I ） on Key Technologies for SCWR, and summarizes a lot of initiatives and original research achievements in the aspects of design R＆D, experimental research and material R＆D. The CSR1000 conceptual design was established with our own intellectual property.

Overview on Overall Design of CSR1000

LI Xiang, LI Qing, XIA Bang-yang, LI Man-chang, LIU Long-sheng

2013, 34(1): 5-8.

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This paper summarizes and illustrates the overall design issues of the CSR1000,including overall technical requirements,main technical selections,main technical parameters and important technical issues discussion and etc.

Core Preliminary Conceptual Design of Supercritical Water-Cooled Reactor CSR1000

XIA Bang-yang, YANG Ping, WANG Lian-jie, MA Yong-qiang, LI Qing, LI Xiang, LIU Jing-bo

2013, 34(1): 9-14.

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A 1000 MW supercritical water-cooled reactor（SCWR） design concept CSR1000 with independent intellectual property has been developed based on the existing technologies of ABWR,PWR and SCWR.In order to simplify the structural design and obtain more uniform moderation,the CSR1000 standard fuel assembly cluster is built of 4 small size square fuel assemblies,and each of them has only one water rod.Furthermore,a two-pass coolant flow scheme is adopted to increase the coolant flow stability and core averaged outlet temperature.The core consists of 157 fuel assembly clusters,and the out-in loading pattern is employed.After core conceptual design evaluation,these key parameters of the core such as cycle length,average discharge burnup,core coolant averaged outlet temperature,maximum fuel cladding temperature and maximum linear power density have been presented.

Preliminary Conceptual Design Study on Supercritical Water-Cooled Reactor CSR1000A with Annular Fuel

XIA Bang-yang, ZHAO Chuan-qi, CAO Liang-zhi, LI Qing, LI Xiang, LI Man-chang

2013, 34(1): 15-18.

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On the basis of the annular fuel for PWR,a new type annular fuel for SCWR has been designed in this paper,which has bigger size than that of PWRs,adopts UO₂ particles instead of fuel pellets and coats heat-barrier material on the surface of the inner cladding.Then a new 1000MWe SCWR design concept CSR1000A has been developed,whose core consists of 163 hexagonal fuel assemblies that are made up with 61 new annular fuels and assembly box.After CSR1000A core evaluation,these core key parameters such as discharge burnup,coolant averaged outlet temperature and maximum fuel cladding temperature are presented.

Primary Design and Calculation of a Breeder Supercritical Water Cooled Fast Reactor

YU Tao, LIU Zi-jing, XIE Jin-sen

2013, 34(1): 19-25.

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Abstract:
In this paper,an improved supercritical water cooled fast reactor has been established.At first,reasonable fuel assembly design is obtained by studying the influences of seed fuel pin diameter and blanket coolant channel diameter to conversion ratio（C_R）.Then,void reactivity coefficient and C_R are calculated for six different core arrangements.Finally,the effect of²³⁵U and ²³⁹Pu ratio on C_R and void reactivity coefficient are analyzed for the sixth core arrangement design.The results show that,negative void reactivity coefficient can be satisfied and C_R can be increased by reducing Hydrogen to Heavy-metal ratio（H/HM） and increasing the blanket assembly numbers by properly distribution;C_R is substantially increased and more negative void reactivity coefficient can be met by reducing Pu isotopes in fuel;when 20.8% Pu enriched MOX fuel and 0.2% enriched depleted Uranium fuel has been adopted as seed and blanket assembly respectively,the sixth core program reaches C_R=1.03128 and gives a negative void reactivity coefficient,which meets the primary requirements for SCFR breeding.Then,the Physical properties of the core,such as βeff,energy spectrum,neutron flux density and power density,have been calculated and analyzed.

Study on Reactivity Control Method for Supercritical Water-Cooled Reactor CSR1000

XIA Bangyang, YANG Ping, WANG Lianjie, LI Qing, LI Xiang

2013, 34(1): 26-30.

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The reactivity control of SCWR depends completely upon the burnable poisons and control rod clusters,so the burnable poisons placement and control rods strategy are very important for SCWR core design.In this paper,the burnable poison Er₂O₃ is chosen to homogeneously mix in the UO₂ pellets of all fuel rods,and the cross-plate control rods similar to that of BWRs are used.By the three-dimensional neutronics and thermal-hydraulic coupled calculations method,the control rods programming has been determined.In addition,the equilibrium core for CSR1000 has also been established by the three-dimensional equilibrium core search,and then an analysis of its key parameters shows that it can entirely satisfy the general design and safety requirements.

Performance Analysis on the Two-Row Hexagonal Fuel Assembly for Super Critical Water-Cooled Reactor

AN Ping, WANG Lian-jie, PAN Jun-jie, LU Wei, YAO Dong

2013, 34(1): 31-34.

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MCATHAS system,which is a code of coupled neutronics/thermal-hydraulics,is used for the calculating of the two two-row hexagonal fuel assemblies.Considering the mutual influence between the axial obvious evolution of density and temperature of the coolant and moderator and the relative power distribution,the results show the two-row hexagonal fuel assemblies keep the balance between uniform moderation and sufficient moderation.The radial power peaking factor of the D6-1 assembly in the paper is less than 1.10 under the condition of using the same enrichment fuel and without the burnable poison.In addition,due to the longer heated perimeter and the litter flow area of the assembly gap region,the conductivity of the assembly box needs to be smaller by adding some adiabatic material.

Supercritical Fuel Assembly Design Applicable for Cruciform Control Rod

ZHU Fa-wen, LEI Tao, CHENG Hua-yang, PANG Hua, PENG Yuan, RU Jun

2013, 34(1): 35-39.

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The supercritical water-cooled reactor（SCWR） has been selected as one of the most promising reactors for Generation Ⅳ nuclear reactors due to its higher thermal efficiency and more simplified structure compared to state-of-the-art LWRs.However,its higher outlet temperature and higher temperature difference between inlet and outlet bring much challenge to the design of SCWR fuel assembly.In this paper,the present status of supercritical fuel assembly design at home and abroad is studied and a kind of fuel assembly with two-flow structure applying for cruciform control rod is proposed.The results show that,the design basically meets the requirements of fuel assembly design,which has good performance.

Subchannel Analysis of SCWR CSR1000 Assembly

DU Dai-quan, XIAO Ze-jun, YAN Xiao, CENG Xiao-kang, HUANG Yan-ping

2013, 34(1): 40-44.

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Subchannel analysis models have been investigated for SCWR fuel assembly by using experimental data available and Computational Fluid Dynamics（CFD） code in the present paper,which applicability in SCWR analysis is also analyzed,and the analysis results is used to improve the ATHAS code.The steady state subchannel analysis is conducted on the CSR1000 fuel assembly using the improve ATHAS,to obtain the temperature distribution of coolant and cladding and pressure drop in assembly.The result shows that,smaller pitch will flatten the profile of coolant temperature and reduce MCST,but it also increases the pressure drop in the assembly.

Flow Instability Analysis of Supercritical Water-Cooled Reactor CSR1000 based on Frequency Domain

TIAN Wen-xi, TIAN Xiao-yan, FENG Jian, QIU Sui-zheng, SU Guang-hui, LU Jian-chao

2013, 34(1): 45-51.

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Flow instability of Supercritical Water-cooled Reactor CSR1000 was studied and mathematics model of core in supercritical water-cooled reactor CSR 1000 was established.A code named FREDO-CSR1000（Frequency domain analysis of CSR1000） and a code named TIMDO（Time-Domain Method） have been developed to analyze the flow instability of Supercritical Water-cooled Reactor CSR1000 after the codes was verified.The results show that the shape of stability map obtained by the two different methods are very similar,both of which are divided into two regions,respectively corresponding to two types of flow instability,namely the flow drift and the density wave oscillation instability.Besides,it is also found that the operation points of CSR1000 calculated by the frequency domain method and time domain method are both in the safety operation region.,which are far away from the unstable region.

Research on Overall Structure Design of CSR1000

ZHANG Hong-liang, LUO Ying, LI Xiang, FAN Heng, LIU Xiao, ZHOU Yu

2013, 34(1): 52-56.

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Based on China Supercritical-Water-Cooled Reactor 1000（CSR1000） research project,this paper introduces the structure design scheme of the two-pass reactor,of which the key technologies are studied or proposed,including the reactor structure material,sealing type,flow distribution,thermal stress analysis and flow-induced vibration response analysis.Basic research methods and solutions are provided,which are meaningful for the engineering practice.

Research on Reactor Internals Sealing Structure of Supercritical Water Cooled Reactor

LIU Xiao, FANG Cai-shun, WANG Liu-bing, ZHANG Hong-liang

2013, 34(1): 57-59,70.

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In order to realize the dual process design of China supercritical water cooled Reactor（CSR1000）,it is necessary to set sealing structure in more than one place inside the reactor.In view of the use experience of the sealing structure in the power station PWR and other industries,it is feasible for the use of O-ring and C-ring in the seal design of the CSR1000 reactor internals.The loading situation of the seal ring is simulated and some key parameters such as decrement and springback amount are calculated and analyzed by taking O-ring for example.

Study on Materials Selection for SCWR Reactor Vessel Internals

ZHOU Yu, ZHANG Hong-liang, LI Man-chang, TANG Rui, FAN Heng

2013, 34(1): 60-64.

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SCWR structural materials R＆D received worldwide attention as the foundation of reactor structure design.In this paper,materials selection rules and assessment system for SCWR vessel internals are provided,and a summary of current research progress of NPIC is introduced.

Analysis on Heat Transfer of Outlet Nozzle and Steam Cavity for SCWR RPV Using CFD

LI Yu-guang, WANG Xiao-bin, LUO Ying, YANG Min, LI Xiang, FU Qiang

2013, 34(1): 65-70.

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Based on Super-Critical Water Reactor（SCWR） conceptual configuration design,with the shell made by 508-III low alloy forging and the outlet nozzle and steam cavity made by Inconel 690 forging,this paper focuses on the application of Computational Fluid Dynamics（CFD） in the numerical simulation of SCWR outlet nozzle and steam cavity structural design.The temperature distribution of outlet nozzle and steam cavity on the Super-Critical condition has been obtained in this study.The simulation results indicate that the maximum temperature reaches 547K at welding line between 508-III shell and Inconel 690 nozzle,thereby it is feasible for conceptual configuration design and materials of SCWR outlet nozzle and steam cavity,and it will assist in validating the direction of configuration optimization and material selection for the design activities.

Design Scheme of Special Safety System for Supercritical-Water-Cooled Reactor

SUI Hai-ming, DAN Jian-qiang, HUANG Xue-kong, GOU Jun-li, YANG Hong

2013, 34(1): 71-74.

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This paper gives a brief description of the design requirement and design principle of Supercritical-Water-Cooled Reactor（SCWR） Nuclear Power Plant safety system,particularly describes the function and the system design scheme of the automatic depressing system（ADS）,high pressure feedwater tank（RMT）,passive residual heat removal system（ICS）,passive containment cooling system（PCCS） and gravity driving core cooling system（GDCS）.The operation of the safety system is analyzed after loss of feedwater flowrate accident,to verify the validity of the SCWR safety system described above.

Research on SCWR Turbine and Thermal System

HUANG Xue-kong, MA Ai-ping, SUN Qi, SUI Hai-ming, YANG Hong

2013, 34(1): 75-77,96.

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The turbine of SCWR is supercritical and impulse which rotors are made of materials maturely used in the turbines of supercritical thermal power plants.The turbine’s seal steam is from clean steam source.In order to realize the sliding pressure startup,the thermal system includes a startup and shut down system which supply sufficient flow for the reactor in startup and shut down phase.This paper describes the turbine and thermal system of SCWR.

Large-Break Accident Analysis of Supercritical Water-Cooled Reactor CSR1000

DANG Gao-jian, HUANG Dai-shun, LU Jian-chao, GAO Ying-xian, DAN Jian-qiang

2013, 34(1): 78-82.

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Large-break accident analysis of China Super-critical water-cooled Reactor with the related electirc power of 1000 MW（CSR1000） were carried out using APROS code to clarify its characteristics and to evaluate the capability of its safety system.At the cold-leg large-break,the reverse-flow in the core happens immediately due to the blowdown of coolant at the break.The reverse-flow drives the high temperature and low density fluid into the core,which results in the heat transfer deterioration in the core and the cladding temperature increasing quickly.After the opening of ADS valves,the positive-flow in the core has been covered,and the excessive core heat-up has been mitigated.The high feedwater tank can provide the coolant supply at the early period;therefore,the low pressure injection system gets more time to be actuated.After the blowdown phase,the core has been re-flooded gradually by low pressure safety injection system.The maximum cladding temperature of large break accident is lower than the criterion（1260℃） by about 340℃,which appears during the blowdown phase.

Analysis of Complete Loss of Forced Flow Accident in China Super-Critical Water Reactor

ZHANG Dan, LU Jian-chao, LIU Song-tao, DAN Jian-qiang

2013, 34(1): 83-86.

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China supercritical water reactor（CRS1000） was developed by NPIC,which has pressure vessel,double flow pass core,and directly cycle loop.As these characteristics,loss of feed water or loss of offsite power will cause the complete loss of forced flow accident,and the first pass core will encounter flow reversal course during the accident,which make the LOFA（Loss Of forced Flow Accident） become the more serious accident in CRS1000.The CRS1000 system model was created by APROS,and LOFA was analyzed by this model.The result shows that,during the short time of LOFA,the operation of HFT（High pressure Feed water Tank） will mitigate the accident,and in long time,the PRHR（Passive Residual Heat Removal system） will function and make the core at the safety condition.

Development of Coupled Neutronics/Thermal-Hydraulics CASIR Code System for SCWR Core Steady State

MA Yong-qiang, CHAI Xiao-ming, WANG Yu-wei, PAN Jun-jie, AN Ping

2013, 34(1): 87-91.

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A coupled neutronic and thermal-hydraulic code system CASIR（Core Analysis System for Innovation Reactor） for SCWR core steady state is developed due to the rapid variation of coolant density along the axial section in SCWR（supercritical water cooled reactor） core.The code system consists of the improved neutronic codes which are used for PWR（pressurized-water reactor） neutronic calculation and the sub-channel thermal-hydraulic code which is applicable for SCWR fuel assembly calculation,including the function of adjusting the flow rate distribution in the lower plenum of core.The coupled code system is tested on a two pass SCWR core of first cycle in the beginning of cycle by comparing the results with those calculated by Monte Carlo code,and it shows that the CASIR system is applicable to simulate the SCWR core preliminarily;main results of burn-up simulation are obtained by adjusting the core flow distribution in order to restrict the maximum cladding-surface temperature below the limit of design.All above show that the CASIR code system satisfies the requirements of SCWR core design and can be used for conceptual design of SCWR core with square assemblies.

Development of Safety Analysis Code for SCWR

WU Pan, DANG Gao-jian, GOU Jun-li, DAN Jian-qiang, JIANG Yang, ZHANG Bo, LI Xiang

2013, 34(1): 92-96.

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This paper introduces the numerical models,auxiliary models and calculation method of code SCTRAN,which is a safety analysis code for SCWR which is developed by ourselves.The simulation of blowdown experiment at supercritical pressure and the LOCA analysis of American SCWR are conducted by the code.By comparing with APROS code and RELAP5-3D code,we can find that the simulation results of SCTRAN agree approximately well with the public recognized code for SCWR.The results indicate that code SCTRAN is capable of carrying out safety analysis for SCWR and its results are reliable.

Preliminary Study on SCWR Fuel Rod Performance Analysis Code

XING Shuo, YAO Dong, YIN Chun-yu, PANG Hua, TU Xiao-lan

2013, 34(1): 97-100,120.

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A SCWR fuel rod performance analysis code is developed in this paper,based on a PWR fuel rod performance analysis code,by analyzing its model and calculation method.The new code is developed on the basis of intensive analyses of SCWR characteristics,focused on the coolant-fuel rod outside surface heat transfer and cladding properties model.Lastly,the performance analysis code was successfully tested on fuel rod of 1/8 HPLWR fuel assembly,then the paper compares the results with that from the sub-channel analysis code ATHAS,which shows that the error is small.

Experimental Study on Heat Transfer of Supercritical Water in Simple Channels

LI Yong-liang, ZENG Xiao-kang, HUANG Zhi-gang, YAN Xiao, HUANG Yan-ping, XIAO Ze-jun

2013, 34(1): 101-107.

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Heat transfer experiments of supercritical water in circular tube,circular annulus and rectangular annulus were carried out,which were abstracted as simple channels to represent the thermal hydraulic characteristics of several Supercritical Water Cooled Reactor（SCWR） fuel assembly design.It was found that,the effect of heat flux,mass flux and pressure on supercritical water heat transfer performance in different simple channels was similar in general.Under the same mass flux and pressure,there was a peak of heat transfer coefficient in pseudo-critical area,which would be decreased with the increasing of heat flux.Under the same heat flux and pressure,the heat transfer coefficient at the same bulk fluid enthalpy would be increased with the increasing of mass flux.The effect of pressure on supercritical water heat transfer performance in different simple channels was not strong except in the pseudo-critical area where the peak of heat transfer coefficient would occur.The heat transfer deterioration phenomenon was also found at different pressures and mass fluxes in different simple channels.

Method of Defining Pseudocritical Region Based on Secondary Phase Transition Theory

YAN Xiao, ZANG Jin-guang, CENG Xiao-kang, HUANG Yan-ping, XIAO Ze-jun

2013, 34(1): 108-113.

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Pseudocritical region is generally considered as a region where fluid property has a sharp variation near the pseudocritical point under supercritical conditions.The flow and heat transfer mechanism behaves uniquely and traditional subcritical formulas may not be applicable in this area.Many researchers have devoted to investigate its peculiarities.The pseudocritical region is similar to the two-phase subcritical saturation region.However,the pseudocritical region is a vague concept without specific physical definition,which can not borrow and learn the subcritical concepts.In this study,the pseudocritical region is considered as the secondary phase transition process and the point where the surface tension disappears represents the start of the phase change.Moreover,as the phase transition process satisfies the Ehrenfest equation,the end point of phase transition is determined in combination with the start point.The pseudocritical region based on this method could help and make contribution to the investigation of flow and heat transfer mechanism in this area.

Application of CFD Methods in Research of SCWR Thermo-Hydraulics

ZENG Xiao-kang, LI Yong-liang, YAN Xiao, XIAO Ze-jun, HUANG Yan-ping

2013, 34(1): 114-120.

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The CFD method has been an important tool in the research of SCWR thermo-hydraulics.Currently,the CFD methods uses commonly the subcritical turbulence models,which can not accurately simulate the gravity and thermal expansion acceleration effect,and CFD numerical method is not applicable when the heat flux is large.The paper summarizes the application status of the CFD methods in the research of SCWR thermo-hydraulics in RETH.

Numerical Analysis of Effect Factors on Heat Transfer and Flow Characteristics of Supercritical Water

LIU Lei, XIAO Ze-jun, YAN Xiao, CENG Xiao-kang, HUANG Yan-ping

2013, 34(1): 121-125,132.

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The effect of hydraulic equivalent diameter,thermal equivalent diameter,characteristic distance,and water properties on heat transfer and flow characteristics were numerically analyzed for supercritical water in circular tube and annular channel.The calculation was performed using FLUENT in r-z two dimensions.The appropriate mesh layout was determined based on mesh sensitivity analysis and SST model was taken as the best turbulence model for supercritical water in this work.The results show that the effect of thermal equivalent diameter can be neglected,while hydraulic equivalent diameter,characteristic distance and water properties are very important for heat transfer and flow characteristics of supercritical water.

Numerical Simulation of Heat Transfer Characteristics of Supercritical Water in Annular Channel with Helical Ribs

LIU Lei, XIAO Ze-jun, YAN Xiao, CENG Xiao-kang, HUANG Yan-ping

2013, 34(1): 126-132.

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One of the effective ways to improve the heat transfer of single-phase fluid is to rotate the fluid in the channel.In this paper,a numerical simulation was carried out to study the heat transfer characteristics of supercritical water in annular channel with helical ribs.Fluid-solid coupling model was established in thesimulation,taking V.G.Razumovskiy’s experiment as a reference.The results showed that the existing of helical ribs can improve the heat transfer in annular channel,but the effect is not obvious under normal condition.The effect of ribs number and pitch on heat transfer characteristics is important,which has to be considered in simulation and experiment.In addition,heat flux and mass flux have effects on circumferential non-uniformity of temperature.

Numerical Simulation of Heat Transfer Characteristics of Supercritical Water in 2×2 Bundles

ZANG Jin-guang, YAN Xiao, HUANG Shan-fang, HUANG Yan-ping, YU Jun-chong

2013, 34(1): 133-136,145.

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Complex channel lies between the typical channel and the typical cell,and is closer to the original rod bundles in channel shape and heated conditions.In this study,numerical simulations with CFD tools has been done to the 2×2 rod bundles.The heat transfer characteristics were analyzed and the impact of heat flux and mass flux was discussed as well.It was found that the increased mass flux or decreased heat flux will contribute to the improvement of the heat transfer capability.The wall temperature was non-uniform along the circumferential direction and the heat conduction of the solid wall may help impair this non-uniformity.

Study on Creep Property of Candidate Stainless Steels for SCWR

LIANG Bo, CHEN Le, TANG Rui, ZHANG Qiang

2013, 34(1): 137-139.

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The creep property of four austenitic stainless steels（316Ti,347,HR3C and 6XN） was investigated with test machine under four conditions: 550℃/90 MPa,600℃/85 MPa,650℃/70 MPa and 700℃/65 MPa.347 showed the best creep property under the 550℃/90 MPa condition while HR3C gradually got the upper hand above 600℃.316Ti displays the worst creep property under all the test conditions.Based on the test data analysis,the stress index followed an order of 316Ti ＞ 347 ＞ HR3C and the activation energy HR3C ＞ 347 ＞316Ti,which suggests the creep property order as HR3C ＞ 347 ＞316Ti.

Experimental Study on Creep/Fatigue Interaction Correlation

TAN Xiao-hui, MA Jian-zhong, LIU Yu-jie, DAI Zhen-yu

2013, 34(1): 140-145.

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Experimental study on creep/fatigue interaction of 316SS has been carried out.Three types of creep/fatigue interaction tests have been taken into consideration in the experiments,which include traditional creep/fatigue interaction subjected to cycling load with hold time and also creep/fatigue interaction by decoupling creep and fatigue loads.Based on the experimental data,the different types of creep/fatigue interaction have been compared and discussed.Total damage from creep and fatigue is evaluated according to linear life fraction rule.Test results show that in respect of specimen life,pre-fatigue/creep interaction is the worst,pre-creep/fatigue interaction is the best,and creep/fatigue interaction with hold time is in between.Fractograph（SEM） of the specimen is used to explain the test results.

Study on Low-Cycle Fatigue Property of Candidate Stainless Steels for SCWR

CHEN Le, TANG Rui, LIANG Bo, ZHANG Qiang, LIU Hong

2013, 34(1): 146-149,156.

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Low cyclic fatigue property of three austenitic stainless steels （316Ti, 347 and HR3C） as candidate materials for SCWR was investigated at room temperature（RT） and 650℃ under a strain amplitude of ±0.5%, and fracture morphology of all the samples was observed by scanning electron microscope（SEM）. The results showed that, at both temperatures the fatigue life of 347 was best and 316Ti worst. For each material, the area of hysteresis was nearly the same in the two temperatures. The elastic deformation was 0.1%-0.15% both at RT and 650℃ for the three materials with different fatigue lives, indicating it had no direct connection with fatigue life. There was different cyclic hardening/saturation behavior for each material. The maximum/minimum stress of either HR3C or 347 was quite different at the two temperatures, while of 316Ti was almost the same. The cyclic hardening behavior was more remarkable in 316Ti compared with 347 at 650℃. SEM observation found that the fatigue striation width was only 1.87 μm for 347, but up to 4.67 μm and 3.0 μm for 316Ti and HR3C respectively, which further demonstrated that 347 had the best fatigue property at 650℃.

Study on High-Cycle Fatigue Behavior of Candidate Stainless Steels for SCWR

XIONG Ru, ZHAO Yu-xiang, QIAO Ying-jie, ZHANG Qiang, WANG Hao, TANG Rui

2013, 34(1): 150-156.

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The fatigue experiments of commerce stainless steels including 347,316Ti and 310 were conducted under bending and rotating loadings.The environments were at room temperature（RT） as well as at 550℃ in air.The fracture morphology was observed by SEM,and the S-N curves were processed according to the experimental data.The results indicate the fatigue limited stresses for the 3 stainless steels were in the order of 347＜316Ti＜310,which consistent with the order of their tensile strength.Elevated temperature would accelerate the oxidation and therefore the fatigue life would decrease,among them 347 was more sensitive to temperature with the maximum decreasing tendency.All the 3 stainless steels have good resistance to high cycle fatigue when comparing their experimental data with the calculated value from the empirical formula.The fracture morphology presents areas of crack initiation,crack growth and fracture,the width of fatigue ripples is about 1μm,the fracture area has much dimples,and 347 presents much cavities of different sizes in dimples.

Study on Corrosion Behaviors of Stainless Steels and Refractory Alloys in Pseudo-Critical Zone

ZHANG Qiang, QIU Shao-yu, TANG Rui, YIN Kai-ju

2013, 34(1): 157-161.

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The general corrosion behavior of three super alloys and four austenitic stainless steels was investigated in 380℃ /25 MPa deionized water.Morphology and composition of the surface oxide scale after various exposure time were analyzed through FEG-SEM and EDS.For all the tested materials,the weight gain（WG） represents no apparent regularity while the weight loss（WL） follows typical logarithmic law.HR3C showed the least WL while 347 the most,an order of magnitude difference.All the exposed samples developed relatively enact and compact dual oxide layer,with the outer layer rich in O,depleted in Cr,Ni and lack of Mo.Several alloys（316Ti,718,825,800H and HR3C） displays pitting.The pits were rich in Nb andTi,depleted in Cr and Ni,and lack of Mn and Mo,based on the alloy composition.

Design and Application of Numeric Hardware Platform of Reactor Numeric Instrument and Control System

WU Zhi-qiang, GAO He, CENG Shao-li

2013, 34(1): 162-164.

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Reactor instrument and control system is an important part in nuclear power plants.Reactor protection system is 1E classified equipment of which hardware platform and developed technology are the key points.All 1E classified equipments were purchased from foreign countries in on-built or on operating power plants in China.The digital platform X86,researched and developed by NPIC,is composed of the controller,intelligent I/O and high speed bus,and satisfies the rules and functional requirement,and pass different types of tests.This platform could become one of the solutions as 1E digital platform for digital nuclear power plant.

2013 Vol. 34, No. 1