ML23025A031
| ML23025A031 | |
| Person / Time | |
|---|---|
| Issue date: | 01/26/2023 |
| From: | Jason Schaperow NRC/NRR/DANU/UTB1 |
| To: | Hayes M NRC/NRR/DANU/UTB1 |
| References | |
| Download: ML23025A031 (1) | |
Text
January 26, 2023 MEMORANDUM TO: Michelle W. Hayes, Chief Advanced Reactor Technical Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation FROM: Jason H. Schaperow, Senior Reactor Systems Engineer Advanced Reactor Technical Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Signed by Schaperow, Jason Office of Nuclear Reactor Regulation on 01/26/23
SUBJECT:
SUMMARY
OF SEPTEMBER 13 AND 20, 2022, PUBLIC WORKSHOPS ON THE SCALE/MELCOR SOURCE TERM DEMONSTRATION PROJECT - MOLTEN-SALT-FUELED REACTOR AND SODIUM-COOLED FAST REACTOR To prepare for safety reviews of non-light water reactor (non-LWR) license applications, the U.S. Nuclear Regulatory Commission (NRC) is performing the SCALE/MELCOR source term demonstration project. The project involves modifying and performing demonstration calculations with the SCALE and MELCOR codes for the following objectives:
Understand severe accident behavior in non-LWRs, Provide insights for developing regulatory guidance, Develop publicly available input models for representative designs, and Facilitate dialogue on the staffs approach for determining source term.
Following completion of code modifications and demonstration calculations for a molten-salt-fueled reactor and a sodium-cooled fast reactor, the NRC held public workshops on September 13, 2022, and September 20, 2022, respectively. During each workshop, NRC, Sandia National Laboratories, and Oak Ridge National Laboratory staff presented SCALE and MELCOR modeling methods and results for simulating core fission product inventory and decay heat during normal operation and core heat up and fission product release to the environment during an accident. About 100 people from the NRC, U.S. nuclear industry, international organizations, and other stakeholders (e.g., Union of Concerned Scientists) attended each workshop.
CONTACT: Jason Schaperow, NRR/DANU 301-415-6093
M. Hayes 2 In 2021, the NRC completed code modifications and demonstration calculations for a heat pipe reactor, a high-temperature gas-cooled reactor, and a pebble-bed salt-cooled reactor, and held public workshops on June 29, 2021, July 20, 2021, September 14, 2021, respectively. The meeting summaries are available at Agencywide Documents Access and Management System (ADAMS) accession numbers ML21202A380, ML21236A285, and ML21288A039, respectively.
The NRC put links to the video recordings, slides, and related reports for the 5 workshops at the NRC public webpage on advanced reactor source term. The slides for each workshop also are available at the following ADAMS accession numbers:
Heat pipe reactor - ML21179C060 High-temperature gas-cooled reactor - ML21200A179 Pebble-bed salt-cooled reactor - ML21256A231 Molten-salt-fueled reactor - ML22353A101 Sodium-cooled fast reactor - ML22353A109
ML23025A031 NRC-001 Office NRR/DANU/UTB1/SRSE NRR/DANU/UTB1/C NRR/DANU/UTB1/SRSE Name JSchaperow MHayes JSchaperow Date 1/25/2023 1/25/2023 1/26/2023