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  4. 压水堆核电站反应堆压力容器钢断裂韧性研究进展

材料导报, 2013, 27(z1): 169-173.

压水堆核电站反应堆压力容器钢断裂韧性研究进展

李承亮 1, , 邓小云 2,

1.深圳中广核工程设计有限公司,深圳518124

2.深圳中广核工程设计有限公司,深圳518124

基金项目:   中广核工程有限公司科技创新项目资助(K-A2010.046)  

关键词： 反应堆压力容器钢 , 断裂韧性 , 韧-脆转变温度 , 参考温度T0 , ASME曲线法 , Master曲线法

Survey on the Fracture Toughness of Reactor Pressure Vessel Steel for PWR

LI Chengliang 1, , DENG Xiaoyun 2,

1.深圳中广核工程设计有限公司,深圳518124

2.深圳中广核工程设计有限公司,深圳518124

Keywords： reactor pressure vessel steel , fracture toughness , ductile-brittle transition temperature , reference temperature T0 , ASME curve method , Master curve method

反应堆压力容器(RPV)是压水堆核电站最为关键的设备之一,其恶劣的服役环境对RPV钢的材料性能提出了苛刻的要求.综述了RPV钢的低温脆化现象,详细介绍了RPV钢的无延性转变温度、修正无延性转变温度、线弹性断裂韧性要求、ASME曲线法及Master曲线法,最后展望了RPV结构完整性分析评价方法的发展方向.

参考文献

[1] EJ/T918-1994.压水堆核电厂反应堆压力容器压力-温度限值曲线制定准则[S].北京:中国标准出版社,1994:1.
[2] Ballesteros A .Integrity of reactor pressure vessels in nuclear power plants:Assessment of irradiation embrittlement effects in reactor pressure vessel steels[R].Vienna:International Atomic Energy Agency,2009.
[3] Regulatory guide 1.99-1988 radiation embrittlement to reactor pressure vessel materials[S].Washington:Office of Nuclear Regulatory Research,1988.
[4] 宁东;姚伟达.核电厂核级承压设备对铁素体材料的断裂韧性基本要求[J].核安全,2005(04):27.
[5] Odette G R;Lucas G F .Irradiation embrittlement of LWR pressure vessel steels[R].California:Electric Power Research Institute,1989.
[6] 周慧久.金属材料强度学[M].北京:科学出版社,1989
[7] ASME Ⅲ-2004.boiler & pressure vessel code,eivision 1 subsection NB[S].New York:The American Society of Mechanical Engineers,2004.
[8] Cao L W;Wu S J;Flewitt P E J .Comparison of ductile-tobrittle transition curve fitting approaches[J].International Journal of Pressure Vessels and Piping,2012,93-94:12.
[9] Diego Ferreno;Inaki Gorrochategui;Federico Gutiérrez-Solana .Degradation due to neutron embrittlement of nuclear vessel steels a critical review about the current experimental and analytical techniques to characterise the material,with particular emphasis on alternative met[R].Cantabria:Technological Centre of Components of Cantabria (CTC),2010.
[10] Oldfield W .Development of fracture toughness reference curves[J].ASME Journal of Engineering Materials and Technology,1980,102:107.
[11] Goran VUKELIC;Josip BRNIC .J-INTEGRAL AS POSSIBLE CRITERION IN MATERIAL FRACTURE TOUGHNESS ASSESSMENT[J].Engineering Review,2011(2):91-96.
[12] Construction and In-service Inspection Rules for Nuclear Island Components RCC-M-2002 design and construction rules for mechanical components of PWR nuclear islands[S].Paris:AFCEN,2002.
[13] E185 standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels[S].West Conshohocken:ASTM,2002.
[14] Rules and regulations title 10 code of federal regulations (CFR) Part 50.61,appendix G.fracture toughness requirements for protection against pressurized thermal shock events[S].Washington:Code of Federal Regulations,1986.
[15] E399 standard test method for standard test method for linear-elastic planestrain fracture toughness KIC of metallic materials[S].West Conshohocken:ASTM,2009.
[16] McCabe D E;Nastad R K .Investigation of the bases for use of the KIC curve[R].Columbus:Oak Ridge Nnational Laboratory,1991.
[17] Wallin K .The size effect in KIc results[J].Engineering Fracture Mechanics,1985,22(01):146.
[18] Corwin W R .Heavy-section steel irradiation program[R].Washington:Oak Ridge National Laboratory,1995.
[19] Fabry A .Survey of national regulatory requirements and identification of existing,planned and required standards at European level relevant with irradiation damage and mitigation methods[R].Stuttgart:European Network on Ageing Materials Evaluation and Studies,1995.
[20] 张亚平,王东辉,钟志民.RPV用钢美国常用断裂韧性KIC表达式的对比分析[J].压力容器,2011(03):16-21.
[21] 张晓中.压力容器用钢辐照脆化评估方法比较[J].核动力工程,2006(02):26-29.
[22] 姚伟达,宁冬,杨仁安,贺寅彪,谢永诚,张明,姚彦贵.ASME锅炉及压力容器规范中关于核电厂部件安全裕量的控制[J].核安全,2011(01):56-64.
[23] Jang Yunchan;Lee Youngsoog .A model to predict the fracture distance of cleavage fracture as a function of temperature[J].Engineering Fracture Mechanics,2011,78:2082.
[24] Scibetta, M.;Ferre?o, D.;Gorrochategui, I.;Lacalle, R.;Van Walle, E.;Martín, J.;Gutiérrez-Solana, F. .Characterisation of the fracture properties in the ductile to brittle transition region of the weld material of a reactor pressure vessel[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,2011(1/3):25-40.
[25] Lucas, G.E. .An evolution of understanding of reactor pressure vessel steel embrittlement[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,2010(1):59-69.
[26] Blasset S .Master curve approach to monitor fracture toughness of reactor pressure vessels in nuclear power plants[R].Vienna:International Atomic Energy Agency,2009.
[27] Brumovsky M .Guidelines for application of the master curve approach to reactor pressure vessel integrity in nuclear power plants[R].Vienna:International Atomic Energy Agency,2005.
[28] 杨文斗.核电厂压力容器安全评估的新方法--主曲线简介[J].核安全,2011(02):7-13,29.
[29] E1921 standard test method for determination of reference temperature,T0,for ferritic steels in the transition range[S].West Conshohocken:ASTM,2012.
[30] ASME code case N-629 use of fracture toughness test data to establish reference temperature for pressure retaining materials section Ⅺ division 1[S].New York:The American Society of Mechanical Engineers,1999.
[31] ASME code case N-631 use of fracture toughness test data to establish reference temperature for pressure retaining materials other than bolting for class 1 vessels section Ⅲ division 1[S].New York:The American Society of Mechanical Engineers,1999.
[32] McCabe D E;Merkle J G;Wallin K .Technical basis for the Master curve concept of fracture toughness evaluations in the transition range[R].New York:American Society for Testing and Materials,1999.
[33] Sokolov M A;McCabe D E;Davidov Y A et al.Use of precracked charpy and smder specimens to establish the master curve[R].New York:American Society for Testing and Materials,1997.
[34] Sham T L;Eno D R .Estimation uncertainty in the determination of the master curve reference temperature[R].New York:Lockheed Martin Corporation,2006.
[35] Rosier B A;Hardin T;Burgos B N et al.Analysis of capsule N from the R.E.Ginna reactor vessel radiation surveillance program[R].Pittsburgh:Westinghouse,2009.
[36] Burgos B N;Conermann J;Anderson S L .Analysis of capsule T from the dominion energy kewaunee power station reactor vessel radiation surveillance program[R].Pittsburgh:Westinghouse,2006.
[37] Lott R G;Kirk M T .Master curve strategies for RPV assessment[R].Pittsburgh:Westinghouse,1998.
[38] Ki-Hyoung Lee;Min-Chul Kim;Bong-Sang Lee;Dang-Moon Wee .Analysis of the master curve approach on the fracture toughness properties of SA508 Gr.4N Ni–Mo–Cr low alloy steels for reactor pressure vessels[J].Materials Science & Engineering, A. Structural Materials: Properties, Misrostructure and Processing,2010(15):3329-3334.
[39] Wu S J;Cao L W .Effect of intergranular failure on the critical fracture stress and the fracture toughness of degraded reactor pressure vessel steel[J].International Journal of Pressure Vessels and Piping,2013,101:23.
[40] S. Bhowmik;A. Chattopadhyay;T. Bose;S.K. Acharyya;P. Sahoo;J. Chattopadhyay;S. Dhar .Estimation of fracture toughness of 20MnMoNi55 steel in the ductile to brittle transition region using master curve method[J].Nuclear engineering and design,2011(8):2831-2838.
[41] Nanstad P K;Miller M K;Lewis W D .Reactor pressure vessel task of light water reactor sustainability program:Level 2 milestone-Results of examinations of surveillance specimens from commercial reactors[R].Columbus:Oak Ridge National Laboratory,2012.
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