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Stress corrosion cracking of irradiated stainless steels in simulated PWR primary water

Book ·
OSTI ID:277812
;  [1];  [2];  [3]; ;  [4];  [5];  [6]
  1. Japan Power Engineering and Inspection Corp., Urayasu (Japan)
  2. Mitsubishi Heavy Industries, Ltd., Takasago (Japan)
  3. Mitsubishi Heavy Industries, Ltd., Kobe (Japan)
  4. Univ. of Tokyo (Japan)
  5. Sophia Univ., Tokyo (Japan)
  6. Westinghouse Electric Corp., Pittsburgh, PA (United States)
Slow Strain Rate Tensile (SSRT) tests on cold-worked 316 stainless steel (S.S.) and 304 S.S. irradiated to 3 x 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) were carried out in simulated PWR primary water chemistry at 290, 325 and 340 C in order to determine Stress Corrosion Cracking (SCC) susceptibility of irradiated S.S. in PWR primary water. Segregation of major elements and impurities at a grain boundary was measured using a transmission electron microscope equipped with a field emission electron gun in order to examine the correlation between SCC susceptibility and segregation. The threshold fluence to SCC susceptibility of irradiated S.S. in PWR primary water was estimated to be order of 10{sup 21} n/cm{sup 2} (E > 0.1 MeV) and decreased with an increase of testing temperature. The depletion of chromium and molybdenum and the enrichment of nickel and silicon at a grain boundary of irradiated S.S. were detected in the irradiated materials. This segregation of major elements and impurities at a grain boundary was considered to be one of possible causes of SCC susceptibility of irradiated S.S. in PWR primary water.
OSTI ID:
277812
Report Number(s):
CONF-960306--; ISBN 0-7918-1226-X
Country of Publication:
United States
Language:
English