NSUF 16-633: Microstructural Characterization of Post-Irradiation Alloys for Integrated FHR Technology Development
The Fluoride salt-cooled High-temperature Reactor (FHR) is being actively considered as the next generation nuclear reactor because it offers, among other benefits, a high degree of passive safety and high thermal efficiency. Integrating the successful experience of the Molten Salt Reactor Experiment (MSRE) program at the Oak Ridge National Laboratory (ORNL) with current advanced nuclear technologies, the primary coolant in the FHR will most likely be 7Li enriched FLiBe (2LiF-BeF2) salt because of its desirable neutronic and heat transfer properties. In support of the materials development for FHR, the in-reactor corrosion tests of candidate alloys, nickel-based Hastelloy N® and 316L stainless steel, in molten FLiBe were successfully conducted in nuclear research reactor at the Massachusetts Institute of Technology for 1000 hours with temperature controlled at 700±3°C. The neutron fluences achieved for this irradiation test are approximately 8.8x10^19 n/cm2 thermal and 4.4x10^20 n/cm2 fast (E>0.1 MeV). To understand the effect of in situ neutron irradiation on the alloys’ corrosion, out-of-reactor corrosion tests in identical crucibles and salt were accomplished at the University of Wisconsin-Madison. The corrosion behavior of both alloys in molten FLiBe, without irradiation effect, was well understood through a serial of microstructural analyses. The intergranular grain boundary corrosion attack, massive Cr depletion, and phase transformation as well as various carbides formations evidenced the corrosion for the alloys in molten FLiBe salt. The in-reactor corrosion is different from the out-of-reactor corrosion because of the tritium generation in molten salt and radiation-induced structural defects within alloys. Our preliminary post-irradiation examinations (PIE) including visual observation and weight change indicate significant difference in the corrosion rate of in-reactor tested samples. However, the microstructures of all post-irradiation alloy samples have not been analyzed due to the limited access to the campus instruments for radioactive materials. Due to this issue, we propose to apply for using the PIE facilities at Idaho National Laboratory (INL) to analyze four in-reactor corrosion tested alloy samples: Hastelloy N® in nickel-lined and in graphite crucibles, and 316L stainless steel in 316L stainless steel-lined and in graphite crucibles. These four samples will be sectioned to ~2mmx2mmx1mm at the MIT Nuclear Reactor Laboratory to reduce overall radioactivity. We propose to use the instruments at INL’s Materials and Fuels Complex (MFC) for the proposed PIE: (1) x-ray diffraction (XRD, Bruker D8 Discover) for one day to identify the phases in near surface layer; (2) FEI QUANTA 3D FEG dual beam focused ion beam (FIB) in conjunction with scanning electron microscope (SEM), energy dispersive spectrometry (EDS) and electron backscatter diffraction (EBSD), for six days in total to observe surficial and cross-sectional microstructure (1 day), to analyze chemical compositions (1 day), to map grain boundaries (2 days), and to mill TEM samples (2 days) as well; (3) FEI Technai TF-30-FEG scanning transmission electron microscope (STEM) for two days to characterize radiation-induced defects, nano-sized carbide phases and grain boundary precipitates.
Additional Info
| Field | Value |
|---|---|
| Abstract | The Fluoride salt-cooled High-temperature Reactor (FHR) is being actively considered as the next generation nuclear reactor because it offers, among other benefits, a high degree of passive safety and high thermal efficiency. Integrating the successful experience of the Molten Salt Reactor Experiment (MSRE) program at the Oak Ridge National Laboratory (ORNL) with current advanced nuclear technologies, the primary coolant in the FHR will most likely be 7Li enriched FLiBe (2LiF-BeF2) salt because of its desirable neutronic and heat transfer properties. In support of the materials development for FHR, the in-reactor corrosion tests of candidate alloys, nickel-based Hastelloy N® and 316L stainless steel, in molten FLiBe were successfully conducted in nuclear research reactor at the Massachusetts Institute of Technology for 1000 hours with temperature controlled at 700±3°C. The neutron fluences achieved for this irradiation test are approximately 8.8x10^19 n/cm2 thermal and 4.4x10^20 n/cm2 fast (E>0.1 MeV). To understand the effect of in situ neutron irradiation on the alloys’ corrosion, out-of-reactor corrosion tests in identical crucibles and salt were accomplished at the University of Wisconsin-Madison. The corrosion behavior of both alloys in molten FLiBe, without irradiation effect, was well understood through a serial of microstructural analyses. The intergranular grain boundary corrosion attack, massive Cr depletion, and phase transformation as well as various carbides formations evidenced the corrosion for the alloys in molten FLiBe salt. The in-reactor corrosion is different from the out-of-reactor corrosion because of the tritium generation in molten salt and radiation-induced structural defects within alloys. Our preliminary post-irradiation examinations (PIE) including visual observation and weight change indicate significant difference in the corrosion rate of in-reactor tested samples. However, the microstructures of all post-irradiation alloy samples have not been analyzed due to the limited access to the campus instruments for radioactive materials. Due to this issue, we propose to apply for using the PIE facilities at Idaho National Laboratory (INL) to analyze four in-reactor corrosion tested alloy samples: Hastelloy N® in nickel-lined and in graphite crucibles, and 316L stainless steel in 316L stainless steel-lined and in graphite crucibles. These four samples will be sectioned to ~2mmx2mmx1mm at the MIT Nuclear Reactor Laboratory to reduce overall radioactivity. We propose to use the instruments at INL’s Materials and Fuels Complex (MFC) for the proposed PIE: (1) x-ray diffraction (XRD, Bruker D8 Discover) for one day to identify the phases in near surface layer; (2) FEI QUANTA 3D FEG dual beam focused ion beam (FIB) in conjunction with scanning electron microscope (SEM), energy dispersive spectrometry (EDS) and electron backscatter diffraction (EBSD), for six days in total to observe surficial and cross-sectional microstructure (1 day), to analyze chemical compositions (1 day), to map grain boundaries (2 days), and to mill TEM samples (2 days) as well; (3) FEI Technai TF-30-FEG scanning transmission electron microscope (STEM) for two days to characterize radiation-induced defects, nano-sized carbide phases and grain boundary precipitates. |
| Award Announced Date | 2016-04-11T00:00:00 |
| Awarded Institution | None |
| Facility | None |
| Facility Tech Lead | Yaqiao Wu |
| Irradiation Facility | None |
| PI | Guiqiu Zheng |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 633 |