NSUF 24-5016: Detailed characterization of in-service IASCC in 316 and 347 stainless steel baffle-former bolts
The objective of this project is to assess the mechanisms for initiation and development of irradiation-assisted stress corrosion cracking (IASCC) in austenitic stainless steel internal components harvested from commercial pressurized water reactors (PWRs). Core internal structural components of PWRs are subjected to multiple environmental extremes such as high radiation doses, high temperatures, corrosive environments, and mechanical stresses. This combination of environmental extremes makes internal alloy components susceptible to IASCC, which may lead to in-service failure and inability to extend the lifetime of existing PWRs. Thus, it is important to understand how the microstructures evolve with irradiation and during service for evaluation of lifetime extension of existing reactors. We must resolve the critical need to understand the mechanisms for IASCC initiation and development and determine potential avenues for mitigating this highly detrimental phenomenon. We will test the hypothesis that IASCC susceptibility increases with radiation dose rate for the same time in the reactor, IASCC susceptibility strongly depends on the alloy composition, and IASCC occurs more readily in the presence of hydrogen from transmutation or radiolysis. The stress-corrosion cracks initiate due to strain localization brought on by radiation-induced dislocation channeling at oxidized grain boundaries in the presence of Si grain boundary segregation. Thus, it stands to reason low radiation damage rates resulting in less RIS and strain localization and materials, like 347 SS, with reduced RIS susceptibility due to carbide formation may be less susceptible to IASCC. The scientific outcome of this project will be to grow the fundamental understanding of IASCC in PWR internal component materials while also providing in-service comparison for research reactor and ion-beam irradiation campaigns. This will facilitate the broader engineering impact of establishing safety margins, design-basis risk evaluation, and model validation for the lifetime extension of the current PWR fleet as well as advancing the knowledge needed for advanced reactor designs. This Super RTE will assess the in-service IASCC susceptibility of PWR baffle former bolts (BFBs) which have received high radiation dose, exposure to PWR coolant water, and mechanical strain. This proposal will expand beyond the current DOE-NE Light Water Reactor Sustainability (LWRS) Program, which performed characterization through mechanical testing and microstructure evaluation on two 316 SS BFBs. Here, we will do further characterization on those bolts plus three 347 SS BFBs from the NSUF Nuclear Fuel and Materials Library. In addition to transmission electron microscopy, atom probe tomography, and traditional 2D electron backscatter diffraction (EBSD) characterization, we will perform 3D-EBSD with focused ion beam-based time-of-flight secondary ion mass spectrometry on both the 316 and 347 bolts. This work will provide valuable insights into the mechanisms of IASCC in service in the current fleet of PWR components and provide a basis for use in advanced reactor designs.
추가 정보
| 필드 | 값 |
|---|---|
| Award Announced Date | 2024-08-15T09:36:25.553 |
| Awarded Institution | Oak Ridge National Laboratory |
| Facility Tech Lead | Alina Montrose, Catou Cmar, Kory Linton |
| Irradiation Facility | |
| PI | Timothy Lach |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | None |