CORROSION FATIGUE MECHANISM OF NUCLEAR-GRADE LOW ALLOY STEEL IN HIGH TEMPERATURE PRESSURIZED WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL

doi:10.11900/0412.1961.2014.00421

Acta Metall Sin

2015, Vol. 51

Issue (3): 298-306 DOI: 10.11900/0412.1961.2014.00421

Current Issue | Archive | Adv Search

CORROSION FATIGUE MECHANISM OF NUCLEAR-GRADE LOW ALLOY STEEL IN HIGH TEMPERATURE PRESSURIZED WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL

WU Xinqiang^1,²(

), TAN Jibo^1,², XU Song^1,², HAN En-Hou^1,², KE Wei^1,²

1 Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016
2 Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016

Cite this article:

WU Xinqiang, TAN Jibo, XU Song, HAN En-Hou, KE Wei. CORROSION FATIGUE MECHANISM OF NUCLEAR-GRADE LOW ALLOY STEEL IN HIGH TEMPERATURE PRESSURIZED WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL. Acta Metall Sin, 2015, 51(3): 298-306.

Download:

HTML

PDF(6558KB)
Export: BibTeX | EndNote (RIS)

Abstract

The service degradation and life assessment of key components in light water reactor nuclear power plants (NPPs) mainly depend on the accumulation of service property data of component materials, understanding of environmental degradation mechanism, and construction of evaluation models or methods. The current ASME design fatigue code does not take full account of the interactions of environmental, loading and material's factors. In the present work, based on the corrosion fatigue tests in simulated NPPs' high temperature pressurized water, the environmental fatigue behavior and dominant mechanism of nuclear-grade low alloy steel have been investigated. A design fatigue model was constructed by taking environmentally assisted fatigue effects into account and the corresponding design curves were given for the convenience of engineering applications. The process for environmental fatigue safety assessment of NPPs' components was proposed, based on which some tentative assessment cases have been given.

Key words: nuclear-grade low alloy steel high temperature pressurized water corrosion fatigue design model environmental fatigue safety assessment

ZTFLH:

TG172

Fund: Supported by National Basic Research Program of China (No.2011CB610506) and National Science and Technology Major Project (No.2011ZX06004-009)

	Service

	E-mail this article
	Add to citation manager
	E-mail Alert
	RSS
	（）
	Articles by authors
	Xinqiang WU
	Jibo TAN
	Song XU
	En-Hou HAN
	Wei KE

URL:

https://www.ams.org.cn/EN/10.11900/0412.1961.2014.00421 OR https://www.ams.org.cn/EN/Y2015/V51/I3/298

Fig.1 Fatigue lifes of domestic nuclear-grade low alloy steel SA508-III in high temperature high pressure water (N₂₅—fatigue life, e_a—strain amplitude, w_DO—dissovled oxygen)

Fig.2 Fatigue crack morphologies on specimen surfaces of nuclear-grade low alloy steels
(a) 561 K, air, 0.1 %·s^-1 (b) 561 K, water, 0.1 %·s^-1 (c) 561 K, water, 0.01 %·s^-1 (d) 561 K, water, 0.001 %·s^-1

Fig.3 Low (a, c, e) and high (b, d, f) magnified fatigue fracture morphologies of nuclear-grade low alloy steels^[3]
(a, b) 561 K, air, 0.1 %·s^-1 (c, d) 561 K, water, 0.1 %·s^-1 (e, f) 561 K, water, 0.001 %·s^-1

Fig.4 Best-fit curve of nuclear-grade low alloy steels (SA508-III and A533B) in air (a) and residual of strain amplitude (b) (

ε ˙

—strain rate)

Fig.5 Factors affecting fatigue life of nuclear-grade low alloy steels (SA508-III and A533B) in high temperature high pressure water (w_S—sulfur content, T—temperature, t_a— thermal aging time)

(a) strain rate (b) dissolved oxygen (c) sulfur content in steels (d) temperature (e) thermal aging time

Fig.6 Environmental fatigue design curves of nuclear-grade low alloy steels (SA508-III and A533B)

Fig.7 Schematic flow chart for environmental fatigue life assessment of nuclear power plants' components (S_alt—stress amplitude, N_i_air—fatigue life predicted by ASME code design curve, N_i_env—fatigue life predicted by present fatigue model in LWR environment, U_air—cumulative damage factor in air (room temperature), U_env—cumulative damage factor in LWR environment)

S_alt/ MPa	T / ℃	$ε ˙$ / %s^-¹	Design cycle / cyc	N_ASME	U_air	N_env	U_env
567.2	200	0.028	120	1024	0.1172	92	1.304
500.6		0.026		1429	0.0630	142	0.634
444.1	200	0.026	142	1967	0.0722	231	0.615
268.8	200	0.002	555	9272	0.0599	2191	0.253
201.9	200	0.001	10	23830	0.0004	7149	1.40×10^-³
143.8	200	0.001	120	81350	0.0015	88833	1.35×10^-³
132.4	200	0.001	98	115630	0.0008	231742	4.23×10^-⁴
121.1	200	0.001	10	159810	0.0001	703538	1.42×10^-⁵
120.2	288	0.001	10	163810	0.0001	777252	1.29×10^-⁵
95.5	288	0.001	222	444850	0.0005	1000000	2.22×10^-⁴
92.6	200	0.001	666	523970	0.0013	1000000	6.66×10^-⁴
91.9	288	0.001	120	560450	0.0002	1000000	1.20×10^-⁴
					Total 0.317		Total 2.81

Table 1 Fatigue damage evaluation for safe end of a boiling water reactor feedwater pipe^[29~31]

Table 2 Typical design basis cycles, anticipated cycles for 40 a and 60 a^[31]

Table 3 Life extension of a pressurized water reactor inlet nozzle^[31]

[1]	Cahn R W, Haasen P, Kramer E J. Materials Science and Technology 10B: Nuclear Materials. Germany: VCH, 1994: 38
[2]	Wu X Q, Han E H, Ke W, Katada Y. Nucl Eng Des, 2007; 237: 1452
[3]	Wu X Q, Katada Y. Corros Sci, 2005; 47: 1415
[4]	Raman S G S, Kitsunai Y. Eng Fail Anal, 2004; 11: 293
[5]	Chopra O K, Shack W J. ASME PVP, 2003; 453: 71
[6]	Chopra O K, Shack W J. Nucl Eng Des, 1998; 184: 49
[7]	Higuchi M, Iida K, Asada Y. ASTM STP, 1997; 1298: 216
[8]	Regulatory Guide 1.207. Washington D C,USA: Nuclear Regulatory Commission, Office of Nuclear Regulatory Research Guide 1.207. Washington D C, USA: Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2007: 5
[9]	Srikantiah G. TR-107263, EPRI Report, 1996
[10]	Chopra O K, Shack W J. NUREG/CR-6909. Washington D C, USA: Nuclear Regulatory Commission, 2007
[11]	JNES-SS-1005 Report. Nuclear Energy System Safety Division, Japan Nuclear Energy Safety Organization, 2011
[12]	2004 ASME Boiler Pressure Vessel Code Section III. New York: The American Society of Mechanical Engineers , 2004
[13]	Kuang W J, Wu X Q, Han E H. Chin Pat, 200810012594.2, 2010
	(匡文军, 吴欣强, 韩恩厚. 中国专利, 200810012594.2, 2010)
[14]	Kuang W J, Wu X Q, Han E H. Chin Pat, 200810230396.3, 2011
	(匡文军, 吴欣强, 韩恩厚. 中国专利, 200810230396.3, 2011)
[15]	Kuang W J, Wu X Q, Han E H. Chin Pat, 200910011110.7, 2012
	(匡文军, 吴欣强, 韩恩厚. 中国专利, 200910011110.7, 2012)
[16]	Xu S, Wu X Q, Han E H, Ke W. Chin Pat, 201010240911.3, 2013
	(徐松, 吴欣强, 韩恩厚, 柯伟. 中国专利, 201010240911.3, 2013)
[17]	Xu S, Wu X Q, Han E H, Ke W. Chin Pat, 201010240899.6, 2013
	(徐松, 吴欣强, 韩恩厚, 柯伟. 中国专利, 201010240899.6, 2013)
[18]	Hänninen H, Cullen W, Kemppainen M. Corrosion, 1990; 46: 563
[19]	Kuniya J, Anzai H, Masaoka I. Corrosion, 1992; 48: 419
[20]	Atkinson J D, Yu J. Fatigue Fract Eng Mater Struct, 1997; 20: 1
[21]	Chopra O K,Shack W J. J Pressure Vessel Technol Trans ASME, 1999; 121: 49
[22]	Ford F P. Corrosion, 1996; 52: 375
[23]	Maiya P S. J Pressure Vessel Technol Trans ASME, 1987; 109: 116
[24]	Wu X Q, Kim I S. Mater Sci Eng, 2003; A348: 309
[25]	Scully J C. Corros Sci, 1980; 20: 997
[26]	Wu X Q, Xu S, Han E H, Ke W. Acta Metall Sin, 2011; 47: 790
	(吴欣强, 徐松, 韩恩厚, 柯伟. 金属学报, 2011; 47: 790)
[27]	Langer B F. ASME J Basic Eng, 1962; 84: 389
[28]	Miner M A. J App Mech, 1945; 12: 159
[29]	Ware A G, Morton D K, Nitzel M E. NUREG/CR-6260. Washington D C, USA : Nuclear Regulatory Commission, 1995
[30]	Chopra O K, Shack W J. NUREG/CR-6583. Washington D C, USA : Nuclear Regulatory Commission, 1996
[31]	Majumdar S, Chopra O K, Shack W J. NUREG/CR-5999. Washington D C, USA: Nuclear Regulatory Commission, 1993

[1]	TAN Jibo, WANG Xiang, WU Xinqiang, HAN En-Hou. Corrosion Fatigue Behavior of 316LN Stainless Steel Hollow Specimen in High-Temperature Pressurized Water[J]. 金属学报, 2021, 57(3): 309-316.
[2]	WU Xinqiang XU Song HAN En-Hou KE Wei. CORROSION FATIGUE OF NUCLEAR--GRADE STAINLESS STEEL IN HIGH TEMPERATURE WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL[J]. 金属学报, 2011, 47(7): 790-796.
[3]	HAN En--Hou. RESEARCH TRENDS ON MICRO AND NANO--SCALE MATERIALS DEGRADATION IN NUCLEAR POWER PLANT[J]. 金属学报, 2011, 47(7): 769-776.
[4]	FANG Bingyan; HAN Enhou; WANG Jianqiu; KE Wei. Influence of strain rate on near-neutral pH environmentally assisted cracking of pipeline steels[J]. 金属学报, 2005, 41(11): 1174-1182 .
[5]	YANG Jihong; JIA Weiping; LI Shouxin. Observation Of The Dislocation Structure In [013] Copper Single Crystal During Corrosion Fatigue In Nacl Aqueous Solution[J]. 金属学报, 2004, 40(1): 99-102 .
[6]	WEI Xuejun;LI Jin;LIU Su'e;KE Wei (State Key Laboratory of Corrosion and Protection; Institute of Corrosion and Protection of Metals; The Chinese Academy of Sciences; Shenyang 110015). ANALXSIS OF FATIGUE FRACTURE WITH CYCLIC OVERLOADING FOR A537 STEEL IN 3.5%NaCl SOLUTION AT AN APPLIED CATHODIC POTENTIAL[J]. 金属学报, 1998, 34(2): 146-150.
[7]	XIE Jianhui;WU Yinshun; ZHU Rizhang (Universily of Science and Technology Beijing;Beijing 100083). INTERGRANULAR CORROSION BEHAVIOUR OF IMPLANT STAINLESS STEEL DURING CORROSION FATIGUE[J]. 金属学报, 1997, 33(3): 304-308.
[8]	WANG Jianqiu; LI Jing; WANG Zhengfu; ZHU Ziyong; KE Wei (Corrosion science laboratory; Institute of Corrosion and Protection of Metals; Chinese Academy of Sciences; Shenyang 110015); ZANG Qishan; WANG Zhongguang (Key laboratory of fatigue and fracture; Institute of Metal Research; Chinese Academy of Sciences;Shenyang 110015). CRACK INITIATION OF SUPERPURE STAINLESS STEEL DURING LOW CYCLE CORROSION FATIGUE IN 3.5% NaCl SOLUTION[J]. 金属学报, 1995, 31(15): 103-108.
[9]	XING Zhiqiang; CHENG Yongmei(Beijing Polytechnic University;100022);CHU Wuyang(University of Science and Technology Beijing; 100083)(Manuscript received 94-03-10;in revised form 94-07-19). LOW-FREQUENCY FATIGUE OF TWO Ti_3Al INTERMETALLICS IN DISTILLED WATER[J]. 金属学报, 1995, 31(1): 34-39.
[10]	WANG Zhengfu;KE Wei Corrosion Science Laboratory; Institute of Corrosion and Protection of Metals; Academia Sinica; ShenyangLINJin;Institute of Corrosion and Protection of Metals;Academia Sinica;Shenyang 110015. INFLUENCES OF APPLIED POTENTIALS AND LOADING WAVEFORM ON FATIGUE CRACK GROWTH FOR STEEL A537[J]. 金属学报, 1993, 29(6): 82-87.
[11]	ZHOU Xiangyang;LI Jin;LIU Guanglei;KE Wei Corrosion Science Laboratory; Academia Sinica; Institute of Corrosion and Protection of Metals; Academic Sinica; ShenyangWEI Xuejun;Institute of Corrosion and Protection of Metals Academia Sinica Shenyang110015. AN APPLICATION OF SPI TECHNIQUE FOR STUDING STRAIN DISTRIBUTION AT CF CRACK TIP[J]. 金属学报, 1993, 29(6): 77-81.
[12]	HAN Yumei;ZHENG Yuli;KE Wei Corrosion Science Laboratory;Institute of Corrosion and Protection of Metals; Academia Sinica; ShenyangHAN Enhou;Institute of Corrosion and Prolection of Metals;Academia Sinica;Shenyang110015. EFFECTS OF STRESS RATIO AND FREQUENCY ON CORROSION FATIGUE CRACK GROWTH IN LOW ALLOY STEEL[J]. 金属学报, 1993, 29(5): 31-36.
[13]	LIU Xiaokun;WANG Jianjun;LU Minxu;JIN Shi;FU Xiangjiong Xi'an Petroleum Institute Northwestern Polytechnical University; Xi'an. CORROSION FATIGUE CRACK PROPAGATION OF MARTENSITIC AND BAINITIC GC-4 ULTRA-HIGH STRENGTH STEEL[J]. 金属学报, 1993, 29(12): 27-33.
[14]	LU Minxu;ZHENG Xiulin Northwestern Polytechnical University; Xi'an. OVERLOADING BEHAVIOUR OF CORROSION FATIGUE CRACK INITIATION FOR 300M STEEL[J]. 金属学报, 1993, 29(11): 66-73.
[15]	CHEN Liangshi;GAO Xuejie;FENG Tao;LIU Minzhi Corrosion Science Laboratory; Institute of Corrosion and Protection of Metals; Academia Sinica; Shenyang Correspondent professor; Institute of Corrosion and Protetion of Metals; Academia Sinica; Shenyang 110015. SCF FEATURE OF AUSTENITIC STAINLESS STEEL IN BOILING MgCl_2 SOLUTION[J]. 金属学报, 1992, 28(5): 73-78.

No Suggested Reading articles found!

Viewed

Full text

Abstract

Cited

Shared

Discussed