NSUF 17-1081: Microstructure and microchemical characterization of neutron and proton irradiated Alloy D9 using APT and TEM
The objective of this study is to understand the effect of irradiation on the microstructure and microchemistry of advanced austenitic alloy D9. Advanced austenitic alloys with improved radiation tolerance over conventional 300-series stainless steels are of interest for fuel cladding and wrappers, fuel assembly grids and straps, and other applications that require a formable, weldable material that can maintain structural integrity over 10-50 displacements per atom (dpa). One such alloy, D9 (nominally Fe-15Cr-15Ni stabilized with TiC precipitates), has 40-50 dpa of additional swelling resistance over 300-series austenitic steels. However, most studies on D9 have thus far focused on void swelling, ignoring other microstructural aspects that also contribute to radiation tolerance. For example, grain boundary radiation-induced segregation (RIS) is an indicator of radiation tolerance and susceptibility to grain boundary attack. One of the most recent known irradiation studies of D9 exhibited grain boundary Cr depletion more than a factor of 5 lower than in typical 300 series stainless steels, suggesting that D9 also has improved tolerance to RIS. Other microstructural features, such as dislocation loops and precipitates, also warrant investigation; to the PIs’ best knowledge, the effect of irradiation on the morphology of TiC precipitates in D9 has never been investigated. This project will conduct a comprehensive irradiated microstructure characterization of D9, in order to fill the knowledge gap on the evolution of loops, precipitates, and RIS.
This work will focus first on high value D9 specimens that were irradiated in the Advanced Test Reactor to 3 dpa at 400°C and 500°C (specimens 048-331, 029-331). Specimens of the identical alloy heat irradiated with 2.0 MeV protons at 500°C to damage doses of 3 and 7 dpa, are also available to the PIs and will be included in this study. The opportunity to study neutron and ion irradiated specimens from the same process heat is rare and adds great value to this project, as we eliminate the concern of heat-to-heat variability. The irradiation conditions planned for study are relevant to both small modular (~400°C) and advanced reactors (500°C). We will investigate D9 specimens with transmission electron microscopy (TEM) and atom probe tomography (APT) from all four irradiation conditions using focused ion beam (FIB) sample preparation. We will then conduct high-resolution scanning TEM (STEM) for dislocation loop and precipitate morphology characterization. APT studies will be carried out for RIS measurement along grain boundaries and to elucidate composition and morphology evolution of the TiC precipitates after irradiation. The project will focus on ascertaining the dependence of microstructure and microchemistry on irradiation temperature, fluence, and irradiating particle type/dose rate.
Results of this work will shed new light on the mechanisms and meaning of “radiation tolerance” in austenitic stainless steels. Subsequently, this work can be used to inform alloy development strategies for future nuclear-grade austenitic steels. The project outcome is the most systematic dataset to date on the irradiated microstructure of D9. More broadly, this work will enable science-based design of irradiation-tolerant austenitic alloys.
Additional Info
| Field | Value |
|---|---|
| Abstract | The objective of this study is to understand the effect of irradiation on the microstructure and microchemistry of advanced austenitic alloy D9. Advanced austenitic alloys with improved radiation tolerance over conventional 300-series stainless steels are of interest for fuel cladding and wrappers, fuel assembly grids and straps, and other applications that require a formable, weldable material that can maintain structural integrity over 10-50 displacements per atom (dpa). One such alloy, D9 (nominally Fe-15Cr-15Ni stabilized with TiC precipitates), has 40-50 dpa of additional swelling resistance over 300-series austenitic steels. However, most studies on D9 have thus far focused on void swelling, ignoring other microstructural aspects that also contribute to radiation tolerance. For example, grain boundary radiation-induced segregation (RIS) is an indicator of radiation tolerance and susceptibility to grain boundary attack. One of the most recent known irradiation studies of D9 exhibited grain boundary Cr depletion more than a factor of 5 lower than in typical 300 series stainless steels, suggesting that D9 also has improved tolerance to RIS. Other microstructural features, such as dislocation loops and precipitates, also warrant investigation; to the PIs’ best knowledge, the effect of irradiation on the morphology of TiC precipitates in D9 has never been investigated. This project will conduct a comprehensive irradiated microstructure characterization of D9, in order to fill the knowledge gap on the evolution of loops, precipitates, and RIS. This work will focus first on high value D9 specimens that were irradiated in the Advanced Test Reactor to 3 dpa at 400°C and 500°C (specimens 048-331, 029-331). Specimens of the identical alloy heat irradiated with 2.0 MeV protons at 500°C to damage doses of 3 and 7 dpa, are also available to the PIs and will be included in this study. The opportunity to study neutron and ion irradiated specimens from the same process heat is rare and adds great value to this project, as we eliminate the concern of heat-to-heat variability. The irradiation conditions planned for study are relevant to both small modular (~400°C) and advanced reactors (500°C). We will investigate D9 specimens with transmission electron microscopy (TEM) and atom probe tomography (APT) from all four irradiation conditions using focused ion beam (FIB) sample preparation. We will then conduct high-resolution scanning TEM (STEM) for dislocation loop and precipitate morphology characterization. APT studies will be carried out for RIS measurement along grain boundaries and to elucidate composition and morphology evolution of the TiC precipitates after irradiation. The project will focus on ascertaining the dependence of microstructure and microchemistry on irradiation temperature, fluence, and irradiating particle type/dose rate. Results of this work will shed new light on the mechanisms and meaning of “radiation tolerance” in austenitic stainless steels. Subsequently, this work can be used to inform alloy development strategies for future nuclear-grade austenitic steels. The project outcome is the most systematic dataset to date on the irradiated microstructure of D9. More broadly, this work will enable science-based design of irradiation-tolerant austenitic alloys. |
| Award Announced Date | 2017-09-20T12:35:45.867 |
| Awarded Institution | None |
| Facility | None |
| Facility Tech Lead | Yaqiao Wu |
| Irradiation Facility | None |
| PI | Mukesh Bachhav |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 1081 |