skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions

Abstract

As a first step toward full scale utilization of the Advanced Test Reactor (ATR) and associated post irradiation examination (PIE) equipment at INL, the NRC and INL staffs have formulated a test program that is designed to help to establish INL and the Nuclear Science User Facilities (NSUF) as a viable destination for irradiation and PIE of reactor structural materials. This initial research program utilized two materials (304 Stainless Steel weld Heat Affected Zone (HAZ) and sensitized 304L stainless steel) that have previously been irradiated at the Halden reactor in Norway and tested at Argonne National Laboratory (ANL) so that test results may be compared, thereby establishing a measure of comparability between Halden and ATR irradiations. A secondary objective of this test program was to characterize the quality of data produced using INL’s newly constructed irradiation assisted stress corrosion cracking (IASCC) test cells. The specimens were be tested in typical boiling water reactor (BWR) conditions to measure stress corrosion cracking (SCC) and fracture toughness of un-irradiated specimens and then IASCC and fracture toughness of specimens irradiated to a fluence equal to approximately 1.0 X 10 21 n/cm 2 (E > 1 MeV).

Authors:
 [1];  [1];  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science User Facilities (NSUF)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE); USNRC
OSTI Identifier:
1408502
Report Number(s):
INL/EXT-17-42276
DOE Contract Number:  
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 304L Weld HAZ; Fracture Toughness; IASCC; Irradiation; Sensitized 304 SS

Citation Formats

Jackson, J. H., Teysseyre, S. P., and Heighes, M. P. Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions. United States: N. p., 2017. Web. doi:10.2172/1408502.
Jackson, J. H., Teysseyre, S. P., & Heighes, M. P. Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions. United States. https://doi.org/10.2172/1408502
Jackson, J. H., Teysseyre, S. P., and Heighes, M. P. Mon . "Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions". United States. https://doi.org/10.2172/1408502. https://www.osti.gov/servlets/purl/1408502.
@article{osti_1408502,
title = {Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions},
author = {Jackson, J. H. and Teysseyre, S. P. and Heighes, M. P.},
abstractNote = {As a first step toward full scale utilization of the Advanced Test Reactor (ATR) and associated post irradiation examination (PIE) equipment at INL, the NRC and INL staffs have formulated a test program that is designed to help to establish INL and the Nuclear Science User Facilities (NSUF) as a viable destination for irradiation and PIE of reactor structural materials. This initial research program utilized two materials (304 Stainless Steel weld Heat Affected Zone (HAZ) and sensitized 304L stainless steel) that have previously been irradiated at the Halden reactor in Norway and tested at Argonne National Laboratory (ANL) so that test results may be compared, thereby establishing a measure of comparability between Halden and ATR irradiations. A secondary objective of this test program was to characterize the quality of data produced using INL’s newly constructed irradiation assisted stress corrosion cracking (IASCC) test cells. The specimens were be tested in typical boiling water reactor (BWR) conditions to measure stress corrosion cracking (SCC) and fracture toughness of un-irradiated specimens and then IASCC and fracture toughness of specimens irradiated to a fluence equal to approximately 1.0 X 1021 n/cm2 (E > 1 MeV).},
doi = {10.2172/1408502},
url = {https://www.osti.gov/biblio/1408502}, journal = {},
number = ,
volume = ,
place = {United States},
year = {2017},
month = {6}
}