NSUF 19-2883: Effect of interstitial elements on the irradiation response of HT9 tempered ferritic/martensitic steels
There is a worldwide interest in nuclear energy due to the increase in the world’s population and the desire to reduced greenhouse gasses from the burning of fossil fuels. Therefore, fission reactors are being investigated to meet the need for clean sources of energy. However, next generation advanced reactors are expected to operate at extreme conditions such as high temperatures and neutron damages as well as corrosive environments. It has been discovered that the ferritic steels having a bcc structure have higher swelling tolerance along with a lower thermal expansion compared to fcc austenitic steels. Therefore, tempered martensitic steels with bcc structure are one of the best candidates for next generation reactors because of their high defect sinks of submicron size lath structure, smaller dislocation bias and higher self-diffusion coefficient. Still, once they reach the steady state swelling regime, they swell with the rate of 0.2%/dpa. Radiation damage resistance in metals is directly correlated with the alloy composition, therefore microstructure. Initially, interstitial elements were reported to deteriorate radiation resistance of the alloys; however, later, swelling resistance has been reported to be improved with the interstitial content. It is still unknown the effect of interstitials on the radiation response of the materials. In this project, tempered ferritic/martensitic HT9 steels produced in various nitrogen contents will be investigated by in-situ irradiations. Ex-situ irradiations up to ~10 dpa at 300 °C have been conducted at Los Alamos National Laboratory (LANL) to investigate the radiation induced hardening, dislocation loop and radiation induced precipitate formation. Therefore, in order to investigate the evolution of the defects in the deformed materials at the early stages of the irradiation, in-situ ion irradiations will be performed. 3-mm foils for transmission electron microscopy (TEM) will be prepared by electropolishing and initial TEM characterization will be performed at LANL and Sabanci University. In-situ irradiations at the Intermediate Voltage Electron Microscopy (IVEM) will be performed using heavy ion irradiations (Kr) at 300 °C, 450 °C and 600 °C up to ~10 dpa. Further detailed characterization will be performed at LANL and Sabanci University. Ultimately, both ex-situ irradiations and low dose in-situ irradiations will reveal the effect of interstitial elements on the radiation resistance of the materials and defect evolution mechanisms. This research will provide a fundamental understanding on the effect of interstitial content on dislocation loop evolution (size, density, type, etc.) and radiation induced particle (G-phase) formation. Application of this understanding will deliver insights on the optimization of the compositions of the cladding materials for next generation advanced reactors. We ask for 10 days of beam time at IVEM as we have 6 to 9 low and high N alloys which will be irradiated to ~10 dpa at various temperatures. The expected period to run this project is 6 months starting from September 2019.
Additional Info
| Field | Value |
|---|---|
| Abstract | There is a worldwide interest in nuclear energy due to the increase in the world’s population and the desire to reduced greenhouse gasses from the burning of fossil fuels. Therefore, fission reactors are being investigated to meet the need for clean sources of energy. However, next generation advanced reactors are expected to operate at extreme conditions such as high temperatures and neutron damages as well as corrosive environments. It has been discovered that the ferritic steels having a bcc structure have higher swelling tolerance along with a lower thermal expansion compared to fcc austenitic steels. Therefore, tempered martensitic steels with bcc structure are one of the best candidates for next generation reactors because of their high defect sinks of submicron size lath structure, smaller dislocation bias and higher self-diffusion coefficient. Still, once they reach the steady state swelling regime, they swell with the rate of 0.2%/dpa. Radiation damage resistance in metals is directly correlated with the alloy composition, therefore microstructure. Initially, interstitial elements were reported to deteriorate radiation resistance of the alloys; however, later, swelling resistance has been reported to be improved with the interstitial content. It is still unknown the effect of interstitials on the radiation response of the materials. In this project, tempered ferritic/martensitic HT9 steels produced in various nitrogen contents will be investigated by in-situ irradiations. Ex-situ irradiations up to ~10 dpa at 300 °C have been conducted at Los Alamos National Laboratory (LANL) to investigate the radiation induced hardening, dislocation loop and radiation induced precipitate formation. Therefore, in order to investigate the evolution of the defects in the deformed materials at the early stages of the irradiation, in-situ ion irradiations will be performed. 3-mm foils for transmission electron microscopy (TEM) will be prepared by electropolishing and initial TEM characterization will be performed at LANL and Sabanci University. In-situ irradiations at the Intermediate Voltage Electron Microscopy (IVEM) will be performed using heavy ion irradiations (Kr) at 300 °C, 450 °C and 600 °C up to ~10 dpa. Further detailed characterization will be performed at LANL and Sabanci University. Ultimately, both ex-situ irradiations and low dose in-situ irradiations will reveal the effect of interstitial elements on the radiation resistance of the materials and defect evolution mechanisms. This research will provide a fundamental understanding on the effect of interstitial content on dislocation loop evolution (size, density, type, etc.) and radiation induced particle (G-phase) formation. Application of this understanding will deliver insights on the optimization of the compositions of the cladding materials for next generation advanced reactors. We ask for 10 days of beam time at IVEM as we have 6 to 9 low and high N alloys which will be irradiated to ~10 dpa at various temperatures. The expected period to run this project is 6 months starting from September 2019. |
| Award Announced Date | 2019-09-17T14:45:07.18 |
| Awarded Institution | None |
| Facility | None |
| Facility Tech Lead | Wei-Ying Chen |
| Irradiation Facility | None |
| PI | Eda Aydogan |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 2883 |