An investigation into stress corrosion cracking susceptibility of 304 stainless steel in simulated boiling water reactor start-up environments
Creators
- 1. Dept. of Engineering and System Science, National Tsing-Hua University, Taiwan (China)
- 2. Nuclear Science and Technology Development Center, National Tsing-Hua University, Taiwan (China)
Description
Numerous incidents of stress corrosion cracking (SCC) in the pressure boundary components of boiling water reactors (BWRs) have been observed in recent decades. Hydrogen water chemistry (HWC) technology has been widely adopted worldwide for mitigating the SCC problem in the structural components of BWRs. Conventional work temperature of HWC is between approximately 230°C and the normal operating temperature, depending on the HWC system design. The reactor coolant, which typically contains high levels of dissolved oxygen because of the intrusion of atmospheric air during a cold shutdown, can remain relatively oxidizing during start-up operations. Laboratory data have indicated that under oxidizing conditions, the rates of intergranular SCC (IGSCC) at intermediate temperatures during start-up are actually higher than those during normal operations in nuclear power plants. In this research, we evaluated the effect of operating temperatures during plant startup processes on the SCC response of austenitic 304 stainless steel (SS) in a pure-water environment. The SCC behavior was investigated using slow strain rate tensile (SSRT) tests at a strain rate of 3.0 × 10-7/s in pure water (with a dissolved-oxygen concentration of 300 ppb) at 150, 200, 250, and 288°C. After the SSRT tests, the samples were observed using a detailed characterization of the microstructure and chemical composition of the sensitized 304 SS surface region. The preliminary results revealed that the solution - annealed specimens exhibited a higher SCC resistance than that of the sensitized samples at the same operating temperature. The elongation of the sensitized samples operating at 150, 200, 250, and 288°C were 56.4%, 34.7%, 22.9%, and 31.4%, respectively. The sensitized samples at 250°C yielded significantly larger amounts of IGSCC and total SCC among all of the operating temperatures. Among the operating conditions investigated in this study, an operating temperature would pose a detrimental effect on the samples in the aspects of mechanical property and degree of SCC. (author)
Additional details
Publishing Information
- Imprint Title
- Proceedings of nuclear plant chemistry conference 2014 Sapporo (NPC 2014)
- Imprint Pagination
- 2471 p.
- Journal Page Range
- 11 p.
Conference
- Title
- Nuclear plant chemistry conference 2014
- Acronym
- NPC 2014
- Dates
- 26-31 Oct 2014
- Place
- Sapporo, Hokkaido (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 47089543
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BWR TYPE REACTORS; CRACKS; INTERGRANULAR CORROSION; REACTOR START-UP; SCANNING ELECTRON MICROSCOPY; STAINLESS STEEL-304; STRAIN RATE; STRESS CORROSION; TENSILE PROPERTIES; WATER CHEMISTRY
- Descriptors DEC
- ALLOYS; AUSTENITIC STEELS; CARBON ADDITIONS; CHEMICAL REACTIONS; CHEMISTRY; CHROMIUM ALLOYS; CHROMIUM-NICKEL STEELS; CORROSION; CORROSION RESISTANT ALLOYS; ELECTRON MICROSCOPY; ENRICHED URANIUM REACTORS; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; HIGH ALLOY STEELS; IRON ALLOYS; IRON BASE ALLOYS; MATERIALS; MECHANICAL PROPERTIES; MICROSCOPY; NICKEL ALLOYS; POWER REACTORS; REACTORS; STAINLESS STEELS; START-UP; STEEL-CR19NI10; STEELS; THERMAL REACTORS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- Available as USB Flash Memory Data in PDF format, Folder Name: Poster3(PWR), Paper ID: 10028NPC2014proceedings.pdf; 8 refs., 8 figs., 6 tabs.