ML24152A290
| ML24152A290 | |
| Person / Time | |
|---|---|
| Issue date: | 03/14/2024 |
| From: | Lucas Kyriazidis NRC/RES/DSA/FSCB |
| To: | |
| Lucas Kyriazidis 301-415-7834 | |
| References | |
| Download: ML24152A290 (8) | |
Text
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024 Bethesda North Marriott Hotel and Conference Center Rockville, Maryland
- nrcric2024 www.nrc.gov ADAPTING TO A
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE Application of Computational Tools for Advanced Nuclear Technologies FAST, SCALE, and MELCOR have been used to support NRC research, licensing, and oversight activities for more than four decades. The NRC continues to update and improve our computer codes and analysis methodologies due to the recent interest in advanced nuclear technologies such as accident tolerant fuel (ATF) small modular reactors (SMRs) and advanced, non-light water reactors (non-LWRs).
This exhibit describes new features and capabilities that have been added to FAST, SCALE, and MELCOR to accommodate these new technologies. Examples of how these updated computational tools are applied for these advanced nuclear technologies are also provided. These include SCALE/MELCOR modeling of the Hermes nonpower test reactor for the recently approved construction permit application, development of regulatory source term for HALEU/HBU/ATF fuels, and the NRCs non-LWR demonstration projects.
Office of Nuclear Regulatory Research Division of Systems Analysis Fuel & Source Term Code Development Branch
- This digital exhibit does not necessarily represent the views of the NRC.
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE FAST Code Development and Applications What Is It?
FAST (Fuel Analysis under Steady-State &
Transients) calculates the thermal-mechanical response of nuclear fuel under steady-state and accident conditions.
Who Uses It?
FAST is used by more than 75 domestic and international organizations, including other regulatory bodies, technical scientific organizations, and utilities, for safety and core reload applications.
How Is It Used?
FAST is used to support licensing reviews by assessing specified acceptable fuel design limits, evaluating vendor fuel codes and methods, and providing initial conditions for design-basis accident analysis. It is also used to perform spent fuel analyses.
How Has It Been Assessed?
FAST is built on more than 30 years of assessment stemming from the FRAPCON/FRAPTRAN codes, as well as experience with fuel vendor codes and data.
It offers more than 200 assessment cases that cover the UO2/zirconium fuel system, and new cases added for metallic fuels.
0 500 1000 1500 2000 2500 0
500 1000 1500 2000 2500 Predicted Temperature, K Measured Temperature, K
=4.7%
UO2 1980s-Present Halden Reactor Project 2004-Present Studsvik Cladding Integrity Project (Phase I - V) 2021-2024 Second Framework for Irradiation Experiments (FIDES-II) 2000-Present Cabri International Project 2021-Present OECD/NEAs QUENCH-ATF Program Ongoing 2002-Present Lead Test Assemblies
& Lead Test Rods Programs DOEs Advanced Gas Reactor Program (AGR) 2016-Present DOEs Sibling Rod Program
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE SCALE Code Development and Applications What Is It?
The SCALE code system is a modeling and simulation suite for nuclear safety analysis and design. It is a modernized code with a long history of application in the regulatory process.
Who Uses It?
SCALE is used by the NRC and in 61 countries (about 11,000 users and 33 regulatory bodies).
How Is It Used?
SCALE is used to support licensing activities (e.g., analysis of spent fuel pool criticality, generating nuclear physics and decay heat parameters for design-basis accident analysis, and review of consolidated interim storage facilities, burnup credit).
How Has It Been Assessed?
SCALE has been validated against numerous critical experiments that cover a range of fuel and moderator materials and geometries, and against measured PWR and BWR spent fuel isotopic composition and decay heat measurements.
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE MELCOR Code Development and Applications What Is It?
MELCOR is an engineering-level code that simulates the response of the reactor core, primary coolant system, containment, and surrounding buildings to a severe accident.
Who Uses It?
MELCOR is used by domestic universities and national laboratories and around 30 international organizations. It is distributed as part of the NRCs Cooperative Severe Accident Research Program (CSARP).
How Is It Used?
MELCOR is used to support severe accident and source term activities at the NRC, including the development of regulatory source terms; support for probabilistic risk assessment models and site risk studies; containment analysis; and forensic investigations of the Fukushima accident.
How Has It Been Assessed?
MELCOR has been validated against numerous international standard problems, benchmarks, separate effects (e.g., VERCORS) and integral experiments (e.g., Phebus FPT), and reactor accidents (e.g., TMI-2, Fukushima).
Phenomena modeled 1988-2010 Phébus-Fission Products &
Source Term Program 2006-2019 Behavior of Iodine Project (BIP) 2005-2016 Experimental Program for Iodine Chemistry Under Radiation (EPICUR) 2011-2019 Source Term Evaluation and Mitigation (STEM)
Project 2013-2018 Benchmark Study of the Accident at Fukushima (BSAF)
Project 2019-2023 2020-2024 Management and Uncertainties of Severe Accidents (MUSA)
Experiments on Source Term for delayed Releases (ESTER)
Reduction of Severe Accident Uncertainties (ROSAU) 2022-2024 Thermodynamic Characterization Of Fuel debris and Fission (TCOFF-2) 2023-2026 Fukushima Accident Information Collection &
Evaluation (FACE)
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE Non-LWR Demonstration Projects & Licensing Blue: FLiBe Red: Fuel Pebble Black: Moderator Pebble Planning Implementation Applications INL Design A (Heat pipe reactor)
UCB Mark 1 (Fluoride salt-cooled high temperature reactor)
PBMR-400 (High temperature gas-cooled reactor)
ABTR (Sodium-cooled fast reactor)
MSRE (Molten salt reactor)
Fluoride Salt Cooled Pebble Bed Reactor MELCOR Model IAP Strategy 2 Computer Codes and Tools Overview &
Technical Approach System Analysis (Vol. 1)
Fuel Performance (Vol. 2)
Source Term Consequences (Vol. 3)
Licensing &
Dose (Vol. 4)
Nuclear Fuel Cycle (Vol. 5)
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE Fission Gas Release Rod Internal Pressure Developing FFRD Models Based on Experimental Programs Application of FFRD Insights to Full Plant Simulations Using NRC Tools Experimental Basis Melt Progression Fission Product Release Phenomena Identification & Ranking Tables Accident Analysis DBA Source Term Scenario # 1 Cs Diffusivity Scenario # n Oxidation/Gas Generation Fission Product Transport Synthesize results Process of Developing Regulatory Source Term Evolution of Technical Basis for Regulatory Guide (RG) 1.183 SAND2023-01313 Peer Review RG 1.183 Revision 2 Regulatory Applications of Tools for ATF/HBU/EE Front-End (Criticality Safety)
In-Reactor & Power Production Shielding, Dose, and Decay Heat Neutronics Fuel Cycle Research Fuel Fragmentation, Relocation, and Dispersal (FFRD)
Regulatory Source Term NUREG-1465 SAND2011-0128 Cr-Coated ST FeCrAl ST RG 1.183 Revision 0 RG 1.183 Revision 1
RIC 2024 Hybrid U.S. Nuclear Regulatory Commission 36th Annual Regulatory Information Conference MARCH 12-14, 2024
- nrcric2024 www.nrc.gov ADAPTING TO A CHANGING LANDSCAPE Additional Information For questions or comments about material in this presentation, please contact Lucas Kyriazidis & Dr. Shawn Campbell in the NRC Office of Nuclear Regulatory Research, at Lucas.Kyriazidis@nrc.gov &
Shawn.Campbell@nrc.gov.
For code documentation, a selected list of publications, and contact information, please visit the following websites:
Non-LWR Demonstration Project