Research paper

Corrosion Fatigue Behavior of 316LN Stainless Steel Hollow Specimen in High-Temperature Pressurized Water

  • Jibo TAN ,
  • Xiang WANG ,
  • Xinqiang WU ,
  • En-Hou HAN
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  • CAS Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China

Received date: 2020-04-27

  Revised date: 2020-09-08

  Online published: 2020-09-17

Supported by

National Key Research and Development Program of China(2017YFB0702103);National Natural Science Foundation of China(51671201)

Abstract

Environmentally assisted fatigue is an important factor in the design, safety review, and life management of key components used in nuclear power plants. Piping systems, valves, and small-bore pipes are sensitive to fatigue damage in nuclear power plants. In this work, a kind of hollow specimen for fatigue testing was designed. High-temperature pressurized water flows through the inside of the specimen, and the outside of the specimen is exposed to air. The corrosion fatigue behavior of 316LN stainless steel was investigated in high-temperature pressurized water using the hollow specimens. The experimental results show that the fatigue strength of 316LN stainless steel was reduced in a high-temperature pressurized water environment, and its fatigue life decreased with decreasing strain rate. The fatigue lives obtained by hollow and standard round bar specimens were comparable, which indicate that it is reasonable and feasible to use the hollow specimen to study the environmentally assisted fatigue performance of nuclear-grade structural materials in a high-temperature pressurized water environment. At low strain rate conditions, the fatigue crack initiation region is a typical fan-shaped pattern with quasi-cleavage cracking characteristics. The fatigue crack growth region is characterized by fatigue striation, and the environmental effects are highly significant in the stage of fatigue crack initiation. The fatigue damage mechanism of 316LN stainless steel in a high-temperature pressurized water environment is also discussed.

Cite this article

Jibo TAN , Xiang WANG , Xinqiang WU , En-Hou HAN . Corrosion Fatigue Behavior of 316LN Stainless Steel Hollow Specimen in High-Temperature Pressurized Water[J]. Acta Metall Sin, 2021 , 57(3) : 309 -316 . DOI: 10.11900/0412.1961.2020.00134

References

1 Carey J. Materials reliability program, fatigue issues assessment (MRP-138) [R]. Electric Power Research Institute, 2005
2 Chopra O K, Shack W J. Effect of LWR coolant environments on the fatigue life of reactor materials [R]. NUREG/CR-6909, ANL-06/08, 2007
3 Chopra O K, Stevens G. Effect of LWR coolant environments on the fatigue life of reactor materials [R]. NUREG/CR-6909, 2018
4 Japan Nuclear Energy Safety Organization. Environmental fatigue evaluation method for nuclear power plants [R]. JNES-SS-1005, 2011
5 Kondo T, Nakajima H, Nagasaki R. Metallographic investigation on the cladding failure in the pressure vessel of a BWR [J]. Nucl. Eng. Des., 1971, 16: 205
6 US Nuclear Regulatory Commission (NRC). Regulatory Guide 1.207, Guidelines for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors [Z]. Washington DC, USA: Nuclear Regulatory Commission, 2007
7 Tan J B, Wu X Q, Han E H, et al. Strain-rate dependent fatigue behavior of 316LN stainless steel in high-temperature water [J]. J. Nucl. Mater., 2017, 489: 33
8 Wu X Q, Xu S, Han E H, et al. Corrosion fatigue of nuclear-grade stainless steel in high temperature water and its environmental fatigue design model [J]. Acta Metall. Sin., 2011, 47: 790
8 吴欣强, 徐 松, 韩恩厚等. 核级不锈钢高温水腐蚀疲劳机制及环境疲劳设计模型 [J]. 金属学报, 2011, 47: 790
9 American Society of Mechanical Engineers. ASME boiler and pressure vessel code section III (Z), New York, 2015
10 Faidy C. Status of French road map to improve environmental fatigue rules [A]. Proceedings of the ASME 2012 Pressure Vessels and Piping Conference [C]. Toronto, Ontario, Canada: ASME, 2012
11 American Society Mechanical Engineers. Fatigue design curves for light water reactor environments [Z]. ASME Code-Case N-761, 2010
12 American Society Mechanical Engineers. Fatigue evaluations including environmental effects [Z]. ASME Code-Case N-792, 2010
13 Tan J B, Zhang Z Y, Zheng H, et.al. Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants [J]. J. Nucl. Mater., 2020, 541, 152407
14 Cho H, Kim B K, Kim I S, et al. Low cycle fatigue behaviors of type 316LN austenitic stainless steel in 310℃ deaerated water-fatigue life and dislocation structure development [J]. Mater. Sci. Eng., 2008, A476: 248
15 Hong S G, Lee S B. Mechanism of dynamic strain aging and characterization of its effect on the low-cycle fatigue behavior in type 316L stainless steel [J]. J. Nucl. Mater., 2005, 340: 307
16 Hong S G, Lee S B. Dynamic strain aging under tensile and LCF loading conditions, and their comparison in cold worked 316L stainless steel [J]. J. Nucl. Mater., 2004, 328: 232
17 Kuang W J, Wu X Q, Han E H. Influence of dissolved oxygen concentration on the oxide film formed on 304 stainless steel in high temperature water [J]. Corros. Sci., 2012, 63: 259
18 Gavenda D J, Luebbers P R, Chopra O K. Crack initiation and crack growth behavior of carbon and low-alloy steels [R]. Orlando, FL: ASME, 1997: 243
19 Gao J, Tan J B, Wu X Q, et al. Effect of grain boundary engineering on corrosion fatigue behavior of 316LN stainless steel in borated and lithiated high-temperature water [J]. Corro. Sci., 2019, 152: 190
20 Chopra O K, Park H B. Mechanism of fatigue crack initiation in light water reactor coolant environments [R]. ANL/ET/CP-101178, 2000
21 Huin N, Tsutusmi K, Legras L, et al. Fatigue crack initiation of 304L stainless steel in simulated PWR primary environment: Relative effect of strain rate [A]. Proceedings of the ASME 2012 Pressure Vessels and Piping Conference [C]. Toronto, Ontario, Canada: ASME, 2012
22 Xu S, Wu X Q, Han E H, et al. Crack initiation mechanisms for low cycle fatigue of type 316Ti stainless steel in high temperature water [J]. Mater. Sci. Eng., 2008, A490: 16
23 Turnbull A. Modelling of crack chemistry in sensitized stainless steel in boiling water reactor environments [J]. Corros. Sci., 1997, 39: 789
24 Turnbull A. Modeling of the chemistry and electrochemistry in cracks—A review [J]. Corrosion, 2001, 57: 175
25 Dumerval M, Perrin S, Marchetti L, et al. Hydrogen absorption associated with the corrosion mechanism of 316L stainless steels in primary medium of pressurized water reactor (PWR) [J]. Corros. Sci., 2014, 85: 251
26 Jambon F, Marchetti L, Jomard F, et al. Mechanism of hydrogen absorption during the exposure of alloy 600-like single-crystals to PWR primary simulated media [J]. J. Nucl. Mater., 2011, 414: 386
27 Laird C, Smith G C. Crack propagation in high stress fatigue [J]. Philo. Mag., 1962, 7: 847
28 Zhang Z Y, Tan J B, Wu X Q, et al. Corrosion fatigue behavior and crack-tip characteristic of 316LN stainless steel in high-temperature pressurized water [J]. J. Nucl. Mater., 2019, 518: 21
29 Kanezaki T, Narazaki C, Mine Y, et al. Effects of hydrogen on fatigue crack growth behavior of austenitic stainless steels [J]. Int. J. Hydrogen Energy, 2008, 33: 2604
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