Stress corrosion cracking susceptibility of irradiated Type 304 stainless steels

Conference · Sat Aug 01 04:00:00 EDT 1992

OSTI ID:10108292

Chung, H M; Ruther, W E; Sanecki, J E; Kassner, T F

Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 {times} 10{sup 2l} n{center_dot}cm{sup {minus}2} (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of the CP absorber tubes or the CP control blade sheath. IGSCC susceptibilities of the BWR components could not be correlated to segregation impurities on grain boundaries. However for comparable fluence levels, Cr on grain-boundaries.

• Research Organization: Argonne National Lab., IL (United States). Materials and Components Technology Div.
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 10108292
• Report Number(s): ANL/MCT/CP--74743; CONF-9206282--2; ON: DE93003005
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100
36 MATERIALS SCIENCE
360105
360106
BWR TYPE REACTORS
CONTROL ELEMENTS
CORROSION AND EROSION
FAST NEUTRONS
INTERGRANULAR CORROSION
NEUTRON ABSORBERS
PHYSICAL RADIATION EFFECTS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
RADIATION EFFECTS
STAINLESS STEEL-304
STRESS CORROSION
TEMPERATURE RANGE 0400-1000 K