NSUF 19-1676: Thermal diffusivity and microstructure analysis of in-core molten salt irradiated graphite
From the success of the Molten Salt Reactor Experiment (MSRE) operated at Oak Ridge National Laboratory (ONRL) during 1950s-1970, the Molten Salt Reactor (MSR) has been considering as one of the most promising candidates for Generation IV Nuclear Energy System because of its unique fuel cycle capabilities and safety characteristics. Nuclear graphite has been widely applied in various reactors as neutron moderator and reflector, and structural material owing to its low Z, high moderating ratio, high thermal conductivity, mechanical strength and etc. However, the porous structure nature of nuclear graphite causes molten salt penetration into its pores at high temperature, which resulting in local hot spots. If fuel salt or fission products penetrates into the graphite pores, the moderating rate dramatically changes due to the increase of neutron absorption in graphite. Therefore, the new graphite with better structure and properties than conventional materials are needed for improving the safety and efficiency of the MSR. In support of the materials development for advanced MSR, several different types of advanced nuclear graphite have been recently irradiated both in inert gas and in molten 2LiF-BeF2 (FLiBe) salt in the MIT Research Reactor (MITR) core at ~700°C with a neutron flux ~1.14×1014 n/cm2·s (E>0.1MeV) for 960 hours. Among these irradiated samples, NG-CT-50, NG-CT-10 and IG-110 (as reference irradiated at some conditions) are selected in this proposed research. Since the combined effects of radiation-induced chemistry and corrosion are complicate for the samples irradiated in FLiBe salt, the samples irradiated in gas are used as control to study the molten salt effects under neutron irradiation.After irradiation, all specimens were extracted from the molten salt and cleaned for post-irradiation examinations (PIEs). To analyze the irradiation effect and molten salt effect on these materials, x-ray diffraction (XRD) and laser flash analyzer (LFA) are needed. The XRD spectra on cleaned samples’ surface provides data to identify the new-formed crystals on surface resulting from the interactions between sample and the impurities in molten salt. The change of d(002)-spacing of graphite due to neutron irradiation can be calculated from the XRD data. In particular, the LFA with enclosed sample holders and inert gas protection at the INL-MFC is the desired system for measuring thermal diffusivity from 25 to 700°C because trace quantities of radioactive gases (such as tritium) and salt probably release at high temperature. The thermal diffusivity might occurs significantly change at the temperature near to the melting point of penetrated salt. This data is critically important to understand these materials behavior in the MSR environments.
Additional Info
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| Abstract | From the success of the Molten Salt Reactor Experiment (MSRE) operated at Oak Ridge National Laboratory (ONRL) during 1950s-1970, the Molten Salt Reactor (MSR) has been considering as one of the most promising candidates for Generation IV Nuclear Energy System because of its unique fuel cycle capabilities and safety characteristics. Nuclear graphite has been widely applied in various reactors as neutron moderator and reflector, and structural material owing to its low Z, high moderating ratio, high thermal conductivity, mechanical strength and etc. However, the porous structure nature of nuclear graphite causes molten salt penetration into its pores at high temperature, which resulting in local hot spots. If fuel salt or fission products penetrates into the graphite pores, the moderating rate dramatically changes due to the increase of neutron absorption in graphite. Therefore, the new graphite with better structure and properties than conventional materials are needed for improving the safety and efficiency of the MSR. In support of the materials development for advanced MSR, several different types of advanced nuclear graphite have been recently irradiated both in inert gas and in molten 2LiF-BeF2 (FLiBe) salt in the MIT Research Reactor (MITR) core at ~700°C with a neutron flux ~1.14×1014 n/cm2·s (E>0.1MeV) for 960 hours. Among these irradiated samples, NG-CT-50, NG-CT-10 and IG-110 (as reference irradiated at some conditions) are selected in this proposed research. Since the combined effects of radiation-induced chemistry and corrosion are complicate for the samples irradiated in FLiBe salt, the samples irradiated in gas are used as control to study the molten salt effects under neutron irradiation.After irradiation, all specimens were extracted from the molten salt and cleaned for post-irradiation examinations (PIEs). To analyze the irradiation effect and molten salt effect on these materials, x-ray diffraction (XRD) and laser flash analyzer (LFA) are needed. The XRD spectra on cleaned samples’ surface provides data to identify the new-formed crystals on surface resulting from the interactions between sample and the impurities in molten salt. The change of d(002)-spacing of graphite due to neutron irradiation can be calculated from the XRD data. In particular, the LFA with enclosed sample holders and inert gas protection at the INL-MFC is the desired system for measuring thermal diffusivity from 25 to 700°C because trace quantities of radioactive gases (such as tritium) and salt probably release at high temperature. The thermal diffusivity might occurs significantly change at the temperature near to the melting point of penetrated salt. This data is critically important to understand these materials behavior in the MSR environments. |
| Award Announced Date | 2019-02-08T00:00:00 |
| Awarded Institution | Center for Advanced Energy Studies |
| Facility | Microscopy and Characterization Suite |
| Facility Tech Lead | Alina Zackrone, Yaqiao Wu |
| Irradiation Facility | None |
| PI | Guiqiu Zheng |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 1676 |