ML25044A015
| ML25044A015 | |
| Person / Time | |
|---|---|
| Issue date: | 02/13/2025 |
| From: | James Corson NRC/RES/DSA/FSCB |
| To: | |
| References | |
| Download: ML25044A015 (1) | |
Text
NRCs simulation capabilities supporting fuel & cladding performance modeling for metallic fueled non-LWRs James Corson, Ph.D.
Office of Nuclear Regulatory Research Division of Systems Analysis Fuel & Source Term Code Development Branch Storage and Transportation of TRISO and Metal Spent Nuclear Fuels
Objectives
- NRCs simulation capabilities for modeling metallic fuels
- Fission Gas Release
- Fuel Swelling
- Cladding Deformation
- Overview of data availability, gaps, and where additional data would be beneficial 2
Non-LWR Fuel Performance Analyses NRCs comprehensive fuel performance code
- Models the thermal-mechanical response of nuclear fuel
- Is used for normal operations, anticipated operational occurrences, accident conditions, and spent fuel storage
- Is used for LWR & non-LWR fuel types Non-LWR demonstration project for fuel performance has improved and validated FAST for simulating non-LWR fuel designs, including metallic fuel designs for use in SFRs.
3 ML20030A177 Non-LWR demonstration project for fuel performance
- Developed new models for TRISO and metallic fuel designs
- Performed assessments & validation activities with available experimental data
Non-LWR Fuel Designs & Phenomena Relevant to Safety
- Pool-type reactors, utilizing metallic fuel designs.
- Fueled with metallic slugs of U-Zr or U-Pu-Zr
- High-temperature steel claddings (e.g., HT-9)
- Fuel pins designed with adequate smear density to accommodate U-Pu-Zr fuel swelling
- Fuel also designed with large plenum to accommodate fission gas release
- Metallic fuel rods operated successfully in Experimental Breeder Reactor II (EBR-II)
Phenomena Relevant to Safety
- Impact of temperature and irradiation on material properties
- Radial redistribution of fuel constituents and impacts on local melting / eutectic temperatures, power distribution, fuel swelling, and fuel-cladding chemical interaction
- Fission product migration, diffusion, and fission gas release
Metallic Fuel Models in FAST Existing U-10Zr fuel, HT-9 cladding models are empirical, based primarily on EBR-II experience
- Anisotropic fuel swelling fitted to experimental data
- Fission gas release fitted to experimental data
- Material properties for HT-9 cladding Thermal conductivity, specific heat capacity, melting temperature, thermal expansion, emissivity, density, Youngs modulus, creep, etc.
Future work needed for fuel failure models and to extend beyond the existing database
- Fuel clad chemical interaction (FCCI) model
- Cladding overpressure failure models
- Release of fission products other than noble gases (e.g., cesium, iodine) from the fuel Already covered by MELCOR for accident conditions FGR data from Pahl et al., JNM 188 (1992) 3 The existing framework of FAST has been leveraged for modeling metallic fuel forms. New material property and phenomenological models have been implemented for modeling in-reactor metal fuels.
Preliminary FAST Assessment on FGR - Metal Fuels
- FAST Initial Assessment on Metallic Fuel - 2018
- Included constant swelling and FGR rates
- Reassessment using new FGR model in progress
- Improved models can reduce uncertainties Geelhood & Porter, Top Fuel 2018 Validation and assessment activities are leveraging historic SFR fuel performance data which includes historic EBR-II.
LWR Spent Fuel Performance Analyses FAST has been used to support LWR spent fuel analyses, by determining
- Initial conditions to support cask analyses (e.g., end-of-life fuel characteristics)
- Cladding oxide thickness and hydrogen content
- HBU mechanical properties
- Rod internal pressure
- Initial conditions for creep rupture analyses
- Recent updates in FAST enhanced LWR spent fuel analyses
- Ability to change ex-reactor boundary conditions
- New backend, ex-reactor spent fuel models
- cladding creep models
- helium production and release model
- pellet swelling model
- New ex-reactor cladding creep rupture criteria While ex-reactor modeling enhancements have focused on LWR fuels, these capabilities can be leveraged to support non-LWR ex-reactor spent fuel modeling.
7
Applying FAST to Metallic Fuel Storage and Transportation
- To date, code development and assessment has focused on in-reactor behavior
- However, the code addresses phenomena that are also important during storage and transportation conditions (e.g., fission gas release, cladding mechanical deformation and integrity)
- The code also provides initial conditions (e.g., rod internal pressure, moles of fission gas available for release) at start of storage or transportation Using FAST for metallic fuel storage and transportation has some challenges
- Very little data available for metallic fuel behavior under storage conditions Such data would be useful for validating models in FAST
- Representative temperatures during storage or transportation conditions must be provided as input to the code Can be provided from other codes that can calculate expected temperatures Can also be taken from imposed limits