Corrosion Fatigue Behavior of 316LN Stainless Steel Hollow Specimen in High-Temperature Pressurized Water
TAN Jibo, WANG Xiang, WU Xinqiang(), HAN En-Hou
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
Cite this article:
TAN Jibo, WANG Xiang, WU Xinqiang, HAN En-Hou. Corrosion Fatigue Behavior of 316LN Stainless Steel Hollow Specimen in High-Temperature Pressurized Water. Acta Metall Sin, 2021, 57(3): 309-316.
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.
Fig.1 Schematic of shape and size of 316LN stainless steel hollow specimen (unit: mm)
Fig.2 Schematic of high-temperature pressurized circulating water corrosion fatigue device (DO—dissolved oxygen, Con—conductivity)
Fig.3 Installation of 316LN stainless steel hollow specimen
Parameter
Value
Unit
Strain amplitude
0.4%-0.9%
Strain ratio
0.2
Strain rate
(0.4-0.004) × 10-2
s-1
Temperature
320
oC
Pressure
12
MPa
DO
<5 × 10-9 (by weight)
Li
2.2 × 10-6 (by weight)
B
1200 × 10-6 (by weight)
pH
6.5-7.0
Flow rate
10 (0.142)
L·h-1 (m·s-1)
Table 1 Corrosion fatigue test parameters
Fig.4 OM image of microstructure of 316LN stainless steel
Fig.5 Corrosion fatigue data of 316LN stainless steel hollow specimen and standard round bar specimen[13] in high-temperature pressurized water
Fig.6 The effect of strain rate on fatigue life of 316LN stainless steel hollow specimen
Fig.7 Relationships between peak load and number of cycles for 316LN stainless steel fatigue tested in high-temperature pressurized water
Fig.8 Macromorphologies of fatigue fractures of 316LN stainless steel hollow specimens at different strain rates
Fig.9 Morphologies of fatigue crack initiation sites of 316LN stainless steel hollow specimens at different strain rates (a, c, e), and corresponding enlarged views of marked positions (b, d, f )
Fig.10 Fatigue striation characteristics at different fracture locations (distance from crack initiation site) of 316LN stainless steel hollow specimens at different strain rates of 0.4 × 10-2 s-1 (a, d, g), 0.04 × 10-2 s-1 (b, e, h), and 0.004 × 10-2 s-1 (c, f, i) (The line in the figure dencotes the fatigue crack propagation direction, and the number is the average fatigue striation spacing in the line length area, unit: μm/cyc)
Fig.11 Morphology of secondary cracks (a) and oxides (b) on inner surface of 316LN stainless steel hollow specimen
Fig.12 The fatigue crack growth rate (da/dN) at different distances to crack initiation sites for 316LN stainless steel in high-temperature pressurized water
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
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