NSUF 17-880: Mechanical characterization of neutron irradiated FSW ODS alloys
One of the goals of the Fuel Cycle R&D program is the development of accident tolerant fuel (ATF) and cladding materials for light water reactors (LWR), which exhibit increased oxidation resistance during a loss of coolant accident. Alternate cladding materials with significantly reduced oxidation kinetics in high-temperature steam environments (when compared to the zirconium alloys) that could reduce heating and hydrogen generation rates, are being investigated. Advanced nuclear reactors require high performance materials that can serve under harsh service conditions, such as higher temperatures, higher radiation doses and corrosive environment. In recent years, oxide dispersion strengthened alloys (ODS) have shown promise as high performance nuclear materials, due to their good elevated temperature properties and radiation damage resistance. Potential applications for this class of material include fast reactor fuel cladding, LWR ATF cladding, and selected LWR core support structures, all of which have the potential to require some form of welding technique to join structures either out-of-core or in-core. Thus, the evaluation of a suitable welding technique is of significant importance. Conventional fusion welding techniques applied to ODS alloys lead to excessive level of porosity, agglomeration and non-uniform distribution of fine oxide particles. Friction stir welding (FSW) can play a vital role in improving the weld quality of ODS alloys and have real implications for their use in nuclear applications. As a DOE-funded project, efforts were made to successfully optimize FSW technique with regard to joining ODS alloys, MA956 (Fe-Cr-Al-Y2O3) and MA754 (Ni-Cr-Al-Y2O3). Microstructures and mechanical properties of the parent and processed materials (i.e., friction stir welded) were characterized by the PI of this RTE proposal earlier, and the results showed FSW did not significantly deteriorate ODS alloys, unlike fusion welding. Microstructural investigation of neutron-irradiated materials (1 dpa) through our NSUF Irradiation Experiment revealed encouraging results, and the particles responsible for the superior properties appeared stable under irradiation. The proposed project aims to evaluate the mechanical properties at room temperature and 320°C (reactor core materials relevant temperature) to understand the effects of radiation damage (1 dpa) on FSW ODS alloys, and to successfully develop appropriate structure-property correlations. Thus, the proposed work is to test and analyze 16 irradiated (1 dpa) SPT specimens (number of specimens to be tested: Microhardness - 8; SPT - 16) present in the library. The project performance is expected to take place during May-July 2017, and will yield results for two journal articles. Neutron irradiation studies have not been performed on FSW ODS alloys to date to the best of our knowledge. Hence, our team can begin to fill the void in the literature by successfully completing the proposed work. In addition, the proposed work could open up new opportunities for friction stir welding (for ODS alloys), and could be developed as an ‘enabling technology’ for nuclear applications.
Additional Info
| Field | Value |
|---|---|
| Abstract | One of the goals of the Fuel Cycle R&D program is the development of accident tolerant fuel (ATF) and cladding materials for light water reactors (LWR), which exhibit increased oxidation resistance during a loss of coolant accident. Alternate cladding materials with significantly reduced oxidation kinetics in high-temperature steam environments (when compared to the zirconium alloys) that could reduce heating and hydrogen generation rates, are being investigated. Advanced nuclear reactors require high performance materials that can serve under harsh service conditions, such as higher temperatures, higher radiation doses and corrosive environment. In recent years, oxide dispersion strengthened alloys (ODS) have shown promise as high performance nuclear materials, due to their good elevated temperature properties and radiation damage resistance. Potential applications for this class of material include fast reactor fuel cladding, LWR ATF cladding, and selected LWR core support structures, all of which have the potential to require some form of welding technique to join structures either out-of-core or in-core. Thus, the evaluation of a suitable welding technique is of significant importance. Conventional fusion welding techniques applied to ODS alloys lead to excessive level of porosity, agglomeration and non-uniform distribution of fine oxide particles. Friction stir welding (FSW) can play a vital role in improving the weld quality of ODS alloys and have real implications for their use in nuclear applications. As a DOE-funded project, efforts were made to successfully optimize FSW technique with regard to joining ODS alloys, MA956 (Fe-Cr-Al-Y2O3) and MA754 (Ni-Cr-Al-Y2O3). Microstructures and mechanical properties of the parent and processed materials (i.e., friction stir welded) were characterized by the PI of this RTE proposal earlier, and the results showed FSW did not significantly deteriorate ODS alloys, unlike fusion welding. Microstructural investigation of neutron-irradiated materials (1 dpa) through our NSUF Irradiation Experiment revealed encouraging results, and the particles responsible for the superior properties appeared stable under irradiation. The proposed project aims to evaluate the mechanical properties at room temperature and 320°C (reactor core materials relevant temperature) to understand the effects of radiation damage (1 dpa) on FSW ODS alloys, and to successfully develop appropriate structure-property correlations. Thus, the proposed work is to test and analyze 16 irradiated (1 dpa) SPT specimens (number of specimens to be tested: Microhardness - 8; SPT - 16) present in the library. The project performance is expected to take place during May-July 2017, and will yield results for two journal articles. Neutron irradiation studies have not been performed on FSW ODS alloys to date to the best of our knowledge. Hence, our team can begin to fill the void in the literature by successfully completing the proposed work. In addition, the proposed work could open up new opportunities for friction stir welding (for ODS alloys), and could be developed as an ‘enabling technology’ for nuclear applications. |
| Award Announced Date | 2017-04-26T10:03:12.02 |
| Awarded Institution | None |
| Facility | None |
| Facility Tech Lead | Stuart Maloy |
| Irradiation Facility | None |
| PI | Ramprashad Prabhakaran |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 880 |