ML25043A203
| ML25043A203 | |
| Person / Time | |
|---|---|
| Issue date: | 02/12/2025 |
| From: | James Corson NRC/RES/DSA/FSCB |
| To: | |
| References | |
| Download: ML25043A203 (1) | |
Text
NRCs simulation capabilities supporting materials performance for TRISO-particle 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 TRISO-particle fuel forms
- Overview of data availability, gaps, and where additional data would be beneficial 2
Non-LWR Fuel Performance Analyses NRCs Fuel Analysis under Steady-state and Transients (FAST) 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 TRISO Particles
- TRISO - Tri-structural ISOtropic particle fuel; embedded in graphite pebbles or compacts
- Multi-later spherical fuel particle; consisting of kernel, buffer, pyrolytic carbon and cladding
- Kernel is the fissionable fuel; typically, UO2 or UCO Phenomena Relevant to Safety
- Fission product migration to and attack of SiC layer
- Mechanical stress analysis of pressure / kernel swelling of structural layers
- Impact of temperature and irradiation on material properties
- Impact of manufacturing defects on fission product migration High-Temp. Gas Cooled Reactor HTGRs
- Pebble-bed core, fueled with TRISO-pebbles, moderated with graphite and cooled with helium.
- Operated HTGRs in the US Peach Bottom Unit 1 and Fort St. Vrain
- Potential Future Designs - X-energy FHRs
- Pebble-bed core, fueled with TRISO-pebbles, moderated with graphite, and cooled with liquid salts.
- Potential Future Designs - Kairos Power Molten Salt-Cooled Reactor
TRISO Fuel Modeling with FAST
- New Standalone 1D code for TRISO fuel performance
- Leverages the framework of NRCs fuel performance code FAST
- Focuses on uranium oxycarbide (UCO) kernels surrounded by buffer, inner pyrocarbon (IPyC), silicon carbide (SiC), and outer pyrocarbon (OPyC) layers
- FAST-TRISO includes the following capabilities
- Heat transfer from the kernel to the particle surface
- Stresses in PyC and SiC layers
- Fission product transport from the kernel through the layers
- Monte Carlo analysis for layer failure probabilities NRCs fuel performance code FAST has been extended to model the steady-state response of individual TRISO particles. FAST-TRISO models the particles temperatures, pressures, and deformation.
ML21175A152 ML21175A151
Ongoing Code Development & Validation Efforts in FAST-TRISO NRC staff is maintaining awareness as more irradiation and integral data becomes available for code development and validation purposes.
Code development
- Mechanical model recently extended to include PyC swelling and creep
- Currently developing correlations for stress concentrations due to PyC cracking and debonding and aspherical particles (using Abaqus)
Code assessment
- Results in good agreement with CRP-6 fuel performance cases 1-8 in IAEA-TECDOC-1674
- Work comparing to AGR fission product release and failure data ongoing CRP-6 Case 8 CRP-6 Case 6 CRP-6 Case 4d
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 spent fuel, ex-reactor, modeling enhancements have focused on LWR fuels, these capabilities can be leveraged to support non-LWR spent fuel modeling.
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Applying FAST to TRISO Spent Fuel Storage and Transportation
- FAST-TRISO was developed with in-reactor behavior in mind
- However, the code addresses phenomena that are also important during storage and transportation conditions (e.g., PyC creep and SiC fracture, fission gas release and gas pressurization, fission product diffusion through particle layers)
- The code also provides initial conditions at start of storage or transportation Using FAST-TRISO for storage and transportation has some challenges
- Many models are only valid at higher temperatures (> 600 C) than what would be expected during normal storage conditions Extrapolating to lower temperatures is possible, but it is hard to trust the results without some data for validation Fortunately, many of the phenomena modeled by FAST-TRISO occur very slowly at low temperature (e.g., PyC creep, fission product diffusion)
- 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