Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

Technical Report · Tue Jun 01 04:00:00 EDT 2010

DOI:https://doi.org/10.2172/1224951· OSTI ID:1224951

Chen, Y. ^[1]; Chopra, O. K. ^[1]; Gruber, Eugene E. ^[1]; Shack, William J. ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States)

The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The effect of neutron irradiation on the fracture toughness of austenitic SSs was also evaluated at dose levels relevant to BWR internals.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1224951
• Report Number(s): NUREG/CR--7018; 118707
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
AGING
AUSTENITIC STEELS
BWR TYPE REACTORS
CHEMICAL COMPOSITION
COOLANTS
CRACK PROPAGATION
CRACKS
EMBRITTLEMENT
FAST NEUTRONS
FRACTURE PROPERTIES
FRACTURES
HEAT AFFECTED ZONE
IRRADIATION
NICKEL ALLOYS
PHYSICAL RADIATION EFFECTS
PROCESSING
RADIATION DOSES
REACTOR CORES
REACTOR INTERNALS
SHIELDED METAL-ARC WELDING
SIMULATION
STAINLESS STEELS
STRESS CORROSION
STRESSES
SUBMERGED ARC WELDING
TEMPERATURE RANGE 0400-1000 K
WATER CHEMISTRY
WELDED JOINTS