No abstract

Irradiation stress corrosion cracking (IASCC) is a critical degradation mechanism for reactor internal components that contributes to the safety and economic operation of reactors. As nuclear power plants age and irradiation dose increases, IASCC becomes an increasingly important issue because of its potential impact on the integrity of reactor internal components. In this study, slow strain rate tensile tests were conducted on irradiated tensile specimens at strain rates between 3 to 7x10-7 s-1 to evaluate cracking susceptibility of austenitic stainless steels in simulated light water reactor (LWR) environments. Significant increases in yield strength were observed for all irradiated specimens and a dose dependence of irradiation hardening was obtained at temperatures relevant to LWRs. Ductility and strain hardening capability were also found decrease rapidly with the increase of dose. After the tests, the specimens were examined using a scanning electron microscopy to characterize fracture morphology. The area fractions of non-ductility fracture were used to evaluate the IASCC susceptibility along with the tensile properties. IASCC susceptibility was also compared for several stainless steels irradiated in the Halden and BOR-60 reactors. A possible neutron spectrum effect was discussed.

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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. Paperwork Reduction Act Statement This NUREG does not contain information collection requirements and, therefore, is not subject to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a current valid OMB control number. vi sensitized at 600 o C for 10.5 and 24 hours, the degree of sensitization has a negligible effect on IASCC resistance. Also, the GBE treatment (the fraction of coincident-site-lattice boundaries ~ 60%) did not improve IASCC susceptibility in the test materials and environment. Conversely, increasing the neutron fluence significantly decreased both the IASCC resistance and the fracture toughness. The research described in this report provides relevant test da...

2016-11-02T18:49:00