超临界水堆候选材料腐蚀行为的研究
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摘要
超临界水堆(SCWR)相比现有轻水反应堆(LWR)具有更高的热效率与设备简化等诸多优势,被选为第四代核反应堆概念设计之一。然而,超临界水温度高、且对金属材料有极强的腐蚀性,SCWR核燃料包壳的选择面临着巨大的困难。堆压水堆中广泛使用的锆合金燃料包壳材料由于高温强度不足、且氧化速度过快,无法在超临界水堆中作为包壳材料使用。人们对用于超临界火电、航空发动机、压水堆核电站、聚变堆以及快堆的材料进行了筛选,得到了一系列候选材料,但其在超临界水堆堆环境的适用性,尤其是耐腐蚀性能,需要进行深入的研究。
本文完成了奥氏体不锈钢800H、HR3C、316Ti和镍基合金718在650。C/25MPa超临界水中的腐蚀实验。实验结果表明,316Ti在超临界水中均匀腐蚀严重,表面氧化膜发生剥落现象,而HR3C、718、800H在3000小时实验后的腐蚀增重均小于50mg/dm2,在抗均匀腐蚀性能方面性能优异。
本文还研究了表面处理(盐浴复合处理、电镀Cr和磁控溅射Cr)对9Cr、12Cr和改进型12Cr铁素体/马氏体(F/M)钢在550。C/25MPa超临界水中的抗腐蚀性能的影响。结果表明,经过1000小时的腐蚀实验后,未经任何表面处理的F/M钢在超临界水中表现出很高的腐蚀速率,而盐浴复合处理并不能改善F/M钢的抗腐蚀性能,电镀Cr和磁控溅射Cr处理则能大大降低F/M钢的腐蚀速率。其中经过磁控溅射Cr处理的9CrF/M钢、磁控溅射Cr和电镀Cr处理的改进型12CrF/M钢这三种试样的腐蚀增重均低于50mg/dm2。
本文通过应变速率为1×10-6s-1的慢应变速率拉伸实验研究了奥氏体不锈钢AL-6XN(未辐照和辐照过)、316Ti、HR3C、TP347HFG和镍基合金718在550。C、600。C、650℃/25MPa超临界水中的应力腐蚀开裂敏感性。在这五种候选材料中,AL-6XN表现出最好的塑性,并且在550。C时的应力应变曲线中观察到了动态应变时效现象,主要原因是溶质原子和位错之间的相互作用。随着温度从550。C升高到650℃,316Ti由于材料的软化屈服强度和抗拉强度下降同时延伸率上升,而HR3C和TP347HFG则因为发生应力松弛强度和延伸率均同时下降。镍基合金718则表现出高于之前四种奥氏体不锈钢的屈服强度与抗拉强度,但是延伸率却是所有材料里最低的。在本次实验中,AL_6XN在650℃、HR3C和TP347HFG在550。C、718在550。C和650。C超临界水中的断口表面均观察到有沿晶断裂。
The supercritical water cooled reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWR). The supercritical water is extremely corrosive to the metal materials. Zr alloy which is widely used as the fuel cladding material in the current LWR is no longer applicable for the SCWR due to the high corrosion rate and loss of strength. Evaluation of several candidate materials for the SCWR has been conducted on the major fire tube materials used in supercritical fossil fired plant, structural materials used in current nuclear reactor and materials for aeroengine blades. For the application of a structural material to a fuel cladding in the SCWR, the material should be evaluated in terms of its corrosion and stress corrosion cracking susceptibility.
General corrosion behaviors of austenitic stainless steel 800H, HR3C,316Ti and Ni-base alloy 718 were investigated in the supercritical water at 650℃/25MPa for 3000h in this paper. The experimental results showed that the weight gain of HR3C,718,800H is less than 50mg/dm2.
Effects of surface treatment on corrosion resistance of 9Cr,12Cr and modified 12Cr ferritic/martensitic (F/M) steel were evaluated in supercritical water at 550℃/25MPa for 1000h. The results showed that the QPQ complex salt bath treatment could not improve the corrosion resistance of F/M steel in the supercritical water environment while Cr coatings prepared by both electro-plating and the magnetron sputtering can greatly reduce the corrosion weight gain rate. The weight gains of modified 12Cr F/M steel treated by Cr electro-plating or magnetron sputtering, and the 9Cr F/M steel treated by magnetron sputtering are less than 50mg/dm2.
Slow stress rate tests (SSRT) were used in this paper to investigate the stress corrosion cracking (SCC) behaviors of austenitic stainless steel AL-6XN (both unirradiated and irradiated),316Ti, HR3C, TP347HFG and Ni-base alloy 718 in the supercritical water at temperature of 500, 600 and 650℃and at pressure of 25MPa, with the strain rate of 1×10-6s-1. AL-6XN showed the best ductility among the five candidate materials. The dynamic strain aging (DSA) phenomenon related to the interaction of solute atoms with dislocations was observed in the stress-strain curve of AL-6XN at 550℃. As the test temperature increased from 550℃to 650℃, the yield strength and tensile strength of 316Ti decreased and the elongation increased while both the strength and elongation of HR3C and TP347HFG decreased. Ni-base alloy 718 showed the highest yield strength and tensile strength but the lowest elongation. Intergranular cracks were observed on the fractured surface of the AL-6XN (650℃), HR3C (550℃), TP347HFG (550℃) and 718 (both 550℃and 650℃)in the tests.
本文完成了奥氏体不锈钢800H、HR3C、316Ti和镍基合金718在650。C/25MPa超临界水中的腐蚀实验。实验结果表明,316Ti在超临界水中均匀腐蚀严重,表面氧化膜发生剥落现象,而HR3C、718、800H在3000小时实验后的腐蚀增重均小于50mg/dm2,在抗均匀腐蚀性能方面性能优异。
本文还研究了表面处理(盐浴复合处理、电镀Cr和磁控溅射Cr)对9Cr、12Cr和改进型12Cr铁素体/马氏体(F/M)钢在550。C/25MPa超临界水中的抗腐蚀性能的影响。结果表明,经过1000小时的腐蚀实验后,未经任何表面处理的F/M钢在超临界水中表现出很高的腐蚀速率,而盐浴复合处理并不能改善F/M钢的抗腐蚀性能,电镀Cr和磁控溅射Cr处理则能大大降低F/M钢的腐蚀速率。其中经过磁控溅射Cr处理的9CrF/M钢、磁控溅射Cr和电镀Cr处理的改进型12CrF/M钢这三种试样的腐蚀增重均低于50mg/dm2。
本文通过应变速率为1×10-6s-1的慢应变速率拉伸实验研究了奥氏体不锈钢AL-6XN(未辐照和辐照过)、316Ti、HR3C、TP347HFG和镍基合金718在550。C、600。C、650℃/25MPa超临界水中的应力腐蚀开裂敏感性。在这五种候选材料中,AL-6XN表现出最好的塑性,并且在550。C时的应力应变曲线中观察到了动态应变时效现象,主要原因是溶质原子和位错之间的相互作用。随着温度从550。C升高到650℃,316Ti由于材料的软化屈服强度和抗拉强度下降同时延伸率上升,而HR3C和TP347HFG则因为发生应力松弛强度和延伸率均同时下降。镍基合金718则表现出高于之前四种奥氏体不锈钢的屈服强度与抗拉强度,但是延伸率却是所有材料里最低的。在本次实验中,AL_6XN在650℃、HR3C和TP347HFG在550。C、718在550。C和650。C超临界水中的断口表面均观察到有沿晶断裂。
The supercritical water cooled reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWR). The supercritical water is extremely corrosive to the metal materials. Zr alloy which is widely used as the fuel cladding material in the current LWR is no longer applicable for the SCWR due to the high corrosion rate and loss of strength. Evaluation of several candidate materials for the SCWR has been conducted on the major fire tube materials used in supercritical fossil fired plant, structural materials used in current nuclear reactor and materials for aeroengine blades. For the application of a structural material to a fuel cladding in the SCWR, the material should be evaluated in terms of its corrosion and stress corrosion cracking susceptibility.
General corrosion behaviors of austenitic stainless steel 800H, HR3C,316Ti and Ni-base alloy 718 were investigated in the supercritical water at 650℃/25MPa for 3000h in this paper. The experimental results showed that the weight gain of HR3C,718,800H is less than 50mg/dm2.
Effects of surface treatment on corrosion resistance of 9Cr,12Cr and modified 12Cr ferritic/martensitic (F/M) steel were evaluated in supercritical water at 550℃/25MPa for 1000h. The results showed that the QPQ complex salt bath treatment could not improve the corrosion resistance of F/M steel in the supercritical water environment while Cr coatings prepared by both electro-plating and the magnetron sputtering can greatly reduce the corrosion weight gain rate. The weight gains of modified 12Cr F/M steel treated by Cr electro-plating or magnetron sputtering, and the 9Cr F/M steel treated by magnetron sputtering are less than 50mg/dm2.
Slow stress rate tests (SSRT) were used in this paper to investigate the stress corrosion cracking (SCC) behaviors of austenitic stainless steel AL-6XN (both unirradiated and irradiated),316Ti, HR3C, TP347HFG and Ni-base alloy 718 in the supercritical water at temperature of 500, 600 and 650℃and at pressure of 25MPa, with the strain rate of 1×10-6s-1. AL-6XN showed the best ductility among the five candidate materials. The dynamic strain aging (DSA) phenomenon related to the interaction of solute atoms with dislocations was observed in the stress-strain curve of AL-6XN at 550℃. As the test temperature increased from 550℃to 650℃, the yield strength and tensile strength of 316Ti decreased and the elongation increased while both the strength and elongation of HR3C and TP347HFG decreased. Ni-base alloy 718 showed the highest yield strength and tensile strength but the lowest elongation. Intergranular cracks were observed on the fractured surface of the AL-6XN (650℃), HR3C (550℃), TP347HFG (550℃) and 718 (both 550℃and 650℃)in the tests.
引文
[1]Ryskamp.J.M. A technology roadmap for generation IV nuclear energy systems[C]. US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, GIF-002-00, December,2002
[2]程旭,刘晓晶.超临界水冷堆国内外研发现状与趋势[J].原子能科学技术,2008,42(2)：167-172
[3]Arthur T. M., Jeong Y. H.. Advanced corrosion resistant Zr alloys for high burnup and generation IV applications, Department of Energy International Nuclear Energy Research Initiative DOE/ROK Project Number:www.nuclear.gov,2003-020-K
[4]Arthur T. M., Yilmazbayhan A.. Zirconium alloys for supercritical water reactor applications challenges and possibilities[J]. Journal of Nuclear Materials,2007,371:61-75
[5]X. J. Liu, X. Cheng. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly[J]. Annals of Nuclear Energy,2009,36 (1):28-36
[6]Zhang L. F., Bao Y. C., Tang R.. Materials selection for supercritical water cooled reactors[C]. The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety, 2010, Shanghai, China
[7]DING Z Y, SAVAGE P E. Catalytic oxidation in supercritical water[J]. Ind Eng Chem Res,1996, 35:3257-3279
[8]Stellwag B.. The mechanism of oxide film formation on austenitic stainless steels in high temperature water[J]. Corrosion Science,1998,40:337-370
[9]刘秀晨,安成强.金属腐蚀学[M].北京：国防工业出版社,2002：267-273
[10]Nelson J L, Andreson P L. Environmental Degradatioft of Materials in Nuclear Power Systems--Water Reactors[C]. Proceedings of 5th lmernation M Conference Ⅱ, Moaterrey, August 1991, Huston:NACE1992:10
[11]杨武.辐照促进应力腐蚀破裂研究的进展[J].材料保护,1994,27(2)：1-4
[12]董月香.几种应力腐蚀试验方法的比较[J].大型铸锻件,2010,5：P45-47
[13]Chen Y, Kruizenga A, Ren X, et al. Progress in Understanding Corrosion in Supercritical Water Systems [C].3rd Int Symposium on SCWR, March,2007, Shanghai,China.
[14]Yun Chen, Kumar Sridharan, Todd Allen. Corrosion Behavior of Ferritic-Martensitic Steel T91 in Supercritical Water[J]. Corrosion Science,2006,48(9):2843-2854.
[15]Yun Chen, Kumar Sridharan, Todd R. et al. Microstructural Examination of Oxide Layers Formed on An Oxide Dispersion Strengthened Ferritic Steel Exposed to Supercritical Water [J]. Journal of Nuclear Materials,2006,359 (1-2):50-58.
[16]Gupta G, Ampornrat P, Ren X, et al. Role of Grain Boundary Engineering in the SCC Behavior of Ferritic-Martensitic Alloy HT-9[J]. Journal of Nuclear Materials,2007,361:160-173.
[17]Cho H S, Kimura A, Ukai S, et al. Corrosion Properties of Oxide Dispersion Strengthened Steels in Super-Critical Water Environment[J]. Journal of Nuclear Materials,2004,329-333 (part 1): 387-391.
[18]张丽,王俭秋,关辉等.超临界水氧化技术及其环境中材料的腐蚀研究现状[J].腐蚀科学与防护技术,2001,13(5)：24-28
[19]韩恩厚.超临界水环境中材料的腐蚀研究现状[J].腐蚀科学与防护技术,1999,11(1)：53-56.
[20]朱发文.超临界水冷堆堆芯候选材料腐蚀性能研究[学位论文].上海交通大学,2010
[21]鲍一晨.超临界水冷堆堆内候选材料腐蚀行为与机理研究[学位论文].上海交通大学,2011
[22]Was G S, Ampornrat P, Gupta G, et al. Corrosion and Stress Corrosion Cracking in Supercritical Water [J]. Journal of Nuclear Materials,2007,371 (1-3):176-201.
[23]Xinqiang Wu, Zhaoen Zhang, Mingcheng Sun.Analyses of oxide films grown on Alloy 625 in oxidizing supercritical water[J]. The Journal of Supercritical Fluids,2008,47(2):309-317
[24]鲍一晨,张乐福,朱发文,等.304NG在超临界水中的腐蚀增重随温度的异常关系[J].原子能科学技术,2010,44(9)：1093-1098
[25]Cho H. S., Kimura A.. Corrosion resistance of high-Cr oxide dispersion strengthened ferritic steels in supercritical pressurized water[J]. Journal of Nuclear Materials,2007,367-370: 1180-1184
[26]Kristian Mathis, Dusan Prchal, Radek Novotny, et al.. Acoustic emission monitoring of slow strain rate tensile tests of 304L stainless steel in supercritical water environment[J]. Corrosion Science,2011,53:59-63
[27]S. Teysseyre, Z. Jiao, E. West, GS. Was. Effect of irradiation on stress corrosion cracking in supercritical water [J]. Journal of Nuclear Materials,2007,371:107-117
[28]E.A. West, G.S. Was. IGSCC of grain boundary engineered 316L and 690 in supercritical water[J]. Journal of Nuclear Materials,2009,392:264-271
[29]Seong Sik Hwang, Byung Hak Lee, Jung Gu Kim, et al.. SCC and corrosion evaluations of the F/M steels for a supercritical water reactor[J]. Journal of Nuclear Materials,2008,372:177-181
[30]N. Muthukumar, J.H. Lee, A. Kimura. SCC behavior of austenitic and martensitic steels in supercritical pressurized water[J]. Journal of Nuclear Materials,2011,417(1-3):1221-1224
[31]R. Novotny, P. Hner, J. Sieg, et al.. Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions[J]. Journal of Nuclear Materials,2010,409(2):117-123
[32]Pantip Ampornrat, Gaurav Gupta, Gary S. Was. Tensile and stress corrosion cracking behavior of ferritic-martensitic steels in supercritical water[J]. Journal of Nuclear Materials,395:30-36
[33]L. Fournier, D. Delafosse, C. Bosch, Th. Magnin, Stress Corrosion Cracking of Nickel Base Super alloys in Aerated Supercritical Water[C], Corrosion 2001, Paper No.01361,NACE, Houston, TX,2001
[34]Y. Tsuchiya, F. Kano, N. Saito, et al.. SCC and Irradiation Properties of Metals under Supercritical water Cooled Power Reactor Conditions[C]. GENES4/ANP2003, paper 1096,September, Kyoto, Japan,2003
[35]R. Fujisawa, K. Nishimura, T. Kishida, et al.. Cracking Susceptibility of Ni Base Alloys and 316 Stainless Steel in Less Oxidizing or Reducing SCW[C], Corrosion 2005, NACE International, Houston, TX, Paper 05395,2005
[36]H. Abe, Y. Watanabe, K. Sue.Stress Corrosion Cracking Mechanisms of 316 Stainless Steels in Sub-Critical Liquid Water and Supercritical Water from a-Dielectric Constant of Water-Point of View[C]. ICAPP'04, paper4226, June 2004, Pittsburgh, PA USA
[37]Was G. S., Ampornrat P., Gupta G., et al.. Corrosion and stress corrosion cracking in supercritical water[J]. Journal of Nuclear Materials,2007,371(1-3):176-201
[38]Tan L., T.R. Allen a, Y. Yang b. Corrosion behavior of alloy 800H (Fe-21Cr-32Ni) in supercritical water [J]. Corrosion Science,2011,53(2):703-711
[39]杨富,章应霖,任永宁,等.新型耐热钢焊接[M].北京：中国电力出版社,2006：143
[40]沈琦,刘鸿国,唐丽英.超超临界机组新型不锈钢Super304H、HR3C运行5400h后的性能试验[J].电力建设,2009,30(9)：62-67
[41]Kunimitsu S, Iwamoto T, Hotta A, et al.. Effect of Mo and Si on Laves phase precipitation in 10% Cr-steels[C]. Proc Int Conf on Stainless Steels, Chiba ISIJ,1991:627-632
[42]朱发文,张乐福,唐睿,等.奥氏体不锈钢AL_6XN在超临界水中的腐蚀[J].腐蚀与防护2010,31(8)：595-599
[43]刘金华,李长香.基于超临界火电站锅炉用材的超临界水冷却反应堆堆芯候选材料分析[J].核动力工程,2009,30(2)：55-58
[44]Qiang Zhang, Rui Tang, Cong Li, et al.. Corrosion behavior of Ni-base alloys in supercritical water[J]. Nuclear engineering and technology,2009,41(1):107-112
[45]董建新INCONEL718高温合金的发展[J].兵器材料科学与工程,1996,19(2)：46-47
[46]Tan L., Ren X., Sridharan K., Allen T. R.. Corrosion behavior of Ni-base alloys for advanced high temperature water-cooled nuclear plants[J]. Corrosion Science,2008,50:3056-3062
[47]杨亚峰,马荣,燕青芝,等.12Cr铁素体/马氏体钢在650℃/25MPa超临界水中的腐蚀行为[J].材料导报B：研究篇,2011,25(5)：100-102
[48]Tan L., Ren X., Allen T. R.. Corrosion behavior of 9-12% Cr ferritic-martensitic steels in supercritical water[J]. Corrosion Science,2010,52:1520-1528
[49]朱发文,张乐福,唐睿,等.铁素体-马氏体钢P92在超临界水中的腐蚀性能研究[J].原子能科学技术,2010,44(8)：979-983
[50]李惠友,罗德福.QPQ技术的现状和展望[J].金属热处理,2004,29(1)：39-43
[51]杨武保.磁控溅射镀膜技术最新进展及发展趋势预测[J].石油机械,2005,33(6)：73-76
[52]徐万劲.磁控溅射技术进展及应用(上)[J].现代仪器,2005,11(5)：1-5
[53]王斐霏,任瑞铭,陈春焕,等.电镀铬层纳米结构及其热稳定性研究[J].中国表面工程,2007,20(3)：30-33
[54]关矞心,李岩,董超芳,等.高温水环境下温度对316 L不锈钢应力腐蚀开裂的影响[J].北京科技大学学报,2009,31(9)：1122-1126
[55]孟丽君,庞淦文,邢辉,等.AL-6XN超级奥氏体钢高温拉伸时的动态应变时效.[J]材料热处理学报,2010,31(4)：50-54
[56]Katada Y, Nagata N. The effect of temperature on fatigue crack growth behavior of a low-alloy pressure vessel steel in a simulated BWR environmen[J]. Corrosion Science,1985,25:693
[57]Atkinson J D, Yu J. Role of dynamic strain ageing in the environment assisted cracking observed in pressure vessel steels[J]. Fatigue Fract Eng Mater Struct,1997,20:1
[58]钱匡武,彭开萍,陈文哲,等.金属动态应变时效现象中的“锯齿屈服”[J].福建工程学院学报,2003,1(1)：4-8
[59]Shankar V, Valsan M, Bhanu K, et al.. Effects of temperature and strain rate on tensile properties and activation energy for dynamic strain aging in alloy 625[J]. Metallurgical and Materials Transactions A,2004,35(10):3129-3139
[60]Charnock W. The apparent activation energy for serrated yielding[J]. Philosophical Magazine, 1969,20:209-2
[2]程旭,刘晓晶.超临界水冷堆国内外研发现状与趋势[J].原子能科学技术,2008,42(2)：167-172
[3]Arthur T. M., Jeong Y. H.. Advanced corrosion resistant Zr alloys for high burnup and generation IV applications, Department of Energy International Nuclear Energy Research Initiative DOE/ROK Project Number:www.nuclear.gov,2003-020-K
[4]Arthur T. M., Yilmazbayhan A.. Zirconium alloys for supercritical water reactor applications challenges and possibilities[J]. Journal of Nuclear Materials,2007,371:61-75
[5]X. J. Liu, X. Cheng. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly[J]. Annals of Nuclear Energy,2009,36 (1):28-36
[6]Zhang L. F., Bao Y. C., Tang R.. Materials selection for supercritical water cooled reactors[C]. The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety, 2010, Shanghai, China
[7]DING Z Y, SAVAGE P E. Catalytic oxidation in supercritical water[J]. Ind Eng Chem Res,1996, 35:3257-3279
[8]Stellwag B.. The mechanism of oxide film formation on austenitic stainless steels in high temperature water[J]. Corrosion Science,1998,40:337-370
[9]刘秀晨,安成强.金属腐蚀学[M].北京：国防工业出版社,2002：267-273
[10]Nelson J L, Andreson P L. Environmental Degradatioft of Materials in Nuclear Power Systems--Water Reactors[C]. Proceedings of 5th lmernation M Conference Ⅱ, Moaterrey, August 1991, Huston:NACE1992:10
[11]杨武.辐照促进应力腐蚀破裂研究的进展[J].材料保护,1994,27(2)：1-4
[12]董月香.几种应力腐蚀试验方法的比较[J].大型铸锻件,2010,5：P45-47
[13]Chen Y, Kruizenga A, Ren X, et al. Progress in Understanding Corrosion in Supercritical Water Systems [C].3rd Int Symposium on SCWR, March,2007, Shanghai,China.
[14]Yun Chen, Kumar Sridharan, Todd Allen. Corrosion Behavior of Ferritic-Martensitic Steel T91 in Supercritical Water[J]. Corrosion Science,2006,48(9):2843-2854.
[15]Yun Chen, Kumar Sridharan, Todd R. et al. Microstructural Examination of Oxide Layers Formed on An Oxide Dispersion Strengthened Ferritic Steel Exposed to Supercritical Water [J]. Journal of Nuclear Materials,2006,359 (1-2):50-58.
[16]Gupta G, Ampornrat P, Ren X, et al. Role of Grain Boundary Engineering in the SCC Behavior of Ferritic-Martensitic Alloy HT-9[J]. Journal of Nuclear Materials,2007,361:160-173.
[17]Cho H S, Kimura A, Ukai S, et al. Corrosion Properties of Oxide Dispersion Strengthened Steels in Super-Critical Water Environment[J]. Journal of Nuclear Materials,2004,329-333 (part 1): 387-391.
[18]张丽,王俭秋,关辉等.超临界水氧化技术及其环境中材料的腐蚀研究现状[J].腐蚀科学与防护技术,2001,13(5)：24-28
[19]韩恩厚.超临界水环境中材料的腐蚀研究现状[J].腐蚀科学与防护技术,1999,11(1)：53-56.
[20]朱发文.超临界水冷堆堆芯候选材料腐蚀性能研究[学位论文].上海交通大学,2010
[21]鲍一晨.超临界水冷堆堆内候选材料腐蚀行为与机理研究[学位论文].上海交通大学,2011
[22]Was G S, Ampornrat P, Gupta G, et al. Corrosion and Stress Corrosion Cracking in Supercritical Water [J]. Journal of Nuclear Materials,2007,371 (1-3):176-201.
[23]Xinqiang Wu, Zhaoen Zhang, Mingcheng Sun.Analyses of oxide films grown on Alloy 625 in oxidizing supercritical water[J]. The Journal of Supercritical Fluids,2008,47(2):309-317
[24]鲍一晨,张乐福,朱发文,等.304NG在超临界水中的腐蚀增重随温度的异常关系[J].原子能科学技术,2010,44(9)：1093-1098
[25]Cho H. S., Kimura A.. Corrosion resistance of high-Cr oxide dispersion strengthened ferritic steels in supercritical pressurized water[J]. Journal of Nuclear Materials,2007,367-370: 1180-1184
[26]Kristian Mathis, Dusan Prchal, Radek Novotny, et al.. Acoustic emission monitoring of slow strain rate tensile tests of 304L stainless steel in supercritical water environment[J]. Corrosion Science,2011,53:59-63
[27]S. Teysseyre, Z. Jiao, E. West, GS. Was. Effect of irradiation on stress corrosion cracking in supercritical water [J]. Journal of Nuclear Materials,2007,371:107-117
[28]E.A. West, G.S. Was. IGSCC of grain boundary engineered 316L and 690 in supercritical water[J]. Journal of Nuclear Materials,2009,392:264-271
[29]Seong Sik Hwang, Byung Hak Lee, Jung Gu Kim, et al.. SCC and corrosion evaluations of the F/M steels for a supercritical water reactor[J]. Journal of Nuclear Materials,2008,372:177-181
[30]N. Muthukumar, J.H. Lee, A. Kimura. SCC behavior of austenitic and martensitic steels in supercritical pressurized water[J]. Journal of Nuclear Materials,2011,417(1-3):1221-1224
[31]R. Novotny, P. Hner, J. Sieg, et al.. Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions[J]. Journal of Nuclear Materials,2010,409(2):117-123
[32]Pantip Ampornrat, Gaurav Gupta, Gary S. Was. Tensile and stress corrosion cracking behavior of ferritic-martensitic steels in supercritical water[J]. Journal of Nuclear Materials,395:30-36
[33]L. Fournier, D. Delafosse, C. Bosch, Th. Magnin, Stress Corrosion Cracking of Nickel Base Super alloys in Aerated Supercritical Water[C], Corrosion 2001, Paper No.01361,NACE, Houston, TX,2001
[34]Y. Tsuchiya, F. Kano, N. Saito, et al.. SCC and Irradiation Properties of Metals under Supercritical water Cooled Power Reactor Conditions[C]. GENES4/ANP2003, paper 1096,September, Kyoto, Japan,2003
[35]R. Fujisawa, K. Nishimura, T. Kishida, et al.. Cracking Susceptibility of Ni Base Alloys and 316 Stainless Steel in Less Oxidizing or Reducing SCW[C], Corrosion 2005, NACE International, Houston, TX, Paper 05395,2005
[36]H. Abe, Y. Watanabe, K. Sue.Stress Corrosion Cracking Mechanisms of 316 Stainless Steels in Sub-Critical Liquid Water and Supercritical Water from a-Dielectric Constant of Water-Point of View[C]. ICAPP'04, paper4226, June 2004, Pittsburgh, PA USA
[37]Was G. S., Ampornrat P., Gupta G., et al.. Corrosion and stress corrosion cracking in supercritical water[J]. Journal of Nuclear Materials,2007,371(1-3):176-201
[38]Tan L., T.R. Allen a, Y. Yang b. Corrosion behavior of alloy 800H (Fe-21Cr-32Ni) in supercritical water [J]. Corrosion Science,2011,53(2):703-711
[39]杨富,章应霖,任永宁,等.新型耐热钢焊接[M].北京：中国电力出版社,2006：143
[40]沈琦,刘鸿国,唐丽英.超超临界机组新型不锈钢Super304H、HR3C运行5400h后的性能试验[J].电力建设,2009,30(9)：62-67
[41]Kunimitsu S, Iwamoto T, Hotta A, et al.. Effect of Mo and Si on Laves phase precipitation in 10% Cr-steels[C]. Proc Int Conf on Stainless Steels, Chiba ISIJ,1991:627-632
[42]朱发文,张乐福,唐睿,等.奥氏体不锈钢AL_6XN在超临界水中的腐蚀[J].腐蚀与防护2010,31(8)：595-599
[43]刘金华,李长香.基于超临界火电站锅炉用材的超临界水冷却反应堆堆芯候选材料分析[J].核动力工程,2009,30(2)：55-58
[44]Qiang Zhang, Rui Tang, Cong Li, et al.. Corrosion behavior of Ni-base alloys in supercritical water[J]. Nuclear engineering and technology,2009,41(1):107-112
[45]董建新INCONEL718高温合金的发展[J].兵器材料科学与工程,1996,19(2)：46-47
[46]Tan L., Ren X., Sridharan K., Allen T. R.. Corrosion behavior of Ni-base alloys for advanced high temperature water-cooled nuclear plants[J]. Corrosion Science,2008,50:3056-3062
[47]杨亚峰,马荣,燕青芝,等.12Cr铁素体/马氏体钢在650℃/25MPa超临界水中的腐蚀行为[J].材料导报B：研究篇,2011,25(5)：100-102
[48]Tan L., Ren X., Allen T. R.. Corrosion behavior of 9-12% Cr ferritic-martensitic steels in supercritical water[J]. Corrosion Science,2010,52:1520-1528
[49]朱发文,张乐福,唐睿,等.铁素体-马氏体钢P92在超临界水中的腐蚀性能研究[J].原子能科学技术,2010,44(8)：979-983
[50]李惠友,罗德福.QPQ技术的现状和展望[J].金属热处理,2004,29(1)：39-43
[51]杨武保.磁控溅射镀膜技术最新进展及发展趋势预测[J].石油机械,2005,33(6)：73-76
[52]徐万劲.磁控溅射技术进展及应用(上)[J].现代仪器,2005,11(5)：1-5
[53]王斐霏,任瑞铭,陈春焕,等.电镀铬层纳米结构及其热稳定性研究[J].中国表面工程,2007,20(3)：30-33
[54]关矞心,李岩,董超芳,等.高温水环境下温度对316 L不锈钢应力腐蚀开裂的影响[J].北京科技大学学报,2009,31(9)：1122-1126
[55]孟丽君,庞淦文,邢辉,等.AL-6XN超级奥氏体钢高温拉伸时的动态应变时效.[J]材料热处理学报,2010,31(4)：50-54
[56]Katada Y, Nagata N. The effect of temperature on fatigue crack growth behavior of a low-alloy pressure vessel steel in a simulated BWR environmen[J]. Corrosion Science,1985,25:693
[57]Atkinson J D, Yu J. Role of dynamic strain ageing in the environment assisted cracking observed in pressure vessel steels[J]. Fatigue Fract Eng Mater Struct,1997,20:1
[58]钱匡武,彭开萍,陈文哲,等.金属动态应变时效现象中的“锯齿屈服”[J].福建工程学院学报,2003,1(1)：4-8
[59]Shankar V, Valsan M, Bhanu K, et al.. Effects of temperature and strain rate on tensile properties and activation energy for dynamic strain aging in alloy 625[J]. Metallurgical and Materials Transactions A,2004,35(10):3129-3139
[60]Charnock W. The apparent activation energy for serrated yielding[J]. Philosophical Magazine, 1969,20:209-2
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