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00204 STRESS CORROSION CRACKING OF TYPE 304 STAINLESS STEEL IRRADIATED TO VERY HIGH DOSE

Picture for 00204 STRESS CORROSION CRACKING OF TYPE 304
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Product Number: 51300-00204-SG
ISBN: 00204 2000 CP
Author: H. M. Chung, W. E. Ruther, R. V. Strain, and W. J. Shack
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Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs], accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our data bases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and micro structural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to =50 dpa at =370°C. Slow-strain-rate tensile tests were conducted at 289°C in air and in water at several levels of electrochemical potential (ECP), and micro structural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, b u t was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating t h a t the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.

Key words: core internals, stainless steel, irradiation damage, intergranular stress corrosion cracking, electrochemical potential, grain-boundary Cr depletion, dislocation channeling
Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs], accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our data bases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and micro structural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to =50 dpa at =370°C. Slow-strain-rate tensile tests were conducted at 289°C in air and in water at several levels of electrochemical potential (ECP), and micro structural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, b u t was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating t h a t the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.

Key words: core internals, stainless steel, irradiation damage, intergranular stress corrosion cracking, electrochemical potential, grain-boundary Cr depletion, dislocation channeling
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