ML23242A342
| ML23242A342 | |
| Person / Time | |
|---|---|
| Issue date: | 06/26/2023 |
| From: | Hart M NRC/NRR/DANU/UTB2 |
| To: | Specter H - No Known Affiliation |
| References | |
| Download: ML23242A342 (1) | |
Text
From:
Michelle Hart To:
Subject:
RE: Source terms for advanced reactors Date:
Monday, June 26, 2023 4:56:00 PM
Dear Mr. Specter,
This email is in response to your email dated June 14, 2023, requesting further information related to slides presented in the February 17, 2022, ACRS future plant designs subcommittee meeting on integration of source term activities in support of advanced reactor initiatives.
The figures you mentioned provide sample results from a MELCOR demonstration calculation done to assess the codes capability to model non-LWR accident phenomena.
They are not intended to provide accident source terms for use in licensing decisions. The NRC staff did MELCOR demonstration calculations for five types of non-LWRs: heat pipe reactor, high-temperature gas-cooled reactor, fluoride-salt-cooled high-temperature reactor, molten-salt fueled reactor, and sodium-cooled fast reactor. These calculations were done to demonstrate that new modeling added to MELCOR (e.g., fission product release from TRISO fuel) to simulate accidents for these non-LWRs was working properly. The MELCOR demonstration calculations were presented at public workshops in 2021 and 2022.The workshop material as well as additional information on nuclear power reactor source term can be found at the following webpage: https://www.nrc.gov/reactors/new-reactors/advanced/nuclear-power-reactor-source-term.html. There is no significance to the magnitude of the releases in the MELCOR demonstration calculations. The scenarios were chosen to show that the new modeling added to MELCOR was functional.
With respect to your questions on radionuclide retention, non-LWRs may develop accident-specific mechanistic source terms. Mechanistic source term development is based on the identification of sources of radionuclide materials at risk of release and the barriers to release, with estimation of the retention within each barrier and transport rates from each barrier to the next, eventually to the environment. This would result in the mechanistic source term defined as a release to the environment, as compared to the source term release to a containment structure as has been used for LWR design basis accident analysis. You are correct that radionuclide retention mechanisms and source term phenomena in general are specific to the technology and accident conditions. For example, reactors that use TRISO fuel are designed to retain fission products within the TRISO layers around the fuel kernel, which would not be a fission product retention barrier available for metallic fuel. The MELCOR code is a system-level code that has detailed fission product aerosol deposition modeling that handles thermophoresis, diffusiophoresis, settling, agglomeration, Brownian motion, and turbulent deposition, etc. The same MELCOR aerosol deposition modeling applies to both LWRs and non-LWRs.
Thank you for your questions on advanced nuclear reactor source term development. I hope you find the information available on our website useful.
Michelle Hart Senior Reactor Engineer Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission