ML25043A240
| ML25043A240 | |
| Person / Time | |
|---|---|
| Issue date: | 02/12/2025 |
| From: | Andrew Bielen Office of Nuclear Regulatory Research |
| To: | |
| References | |
| Download: ML25043A240 (1) | |
Text
NRCs simulation capabilities supporting criticality, reactor physics, decay heat, and shielding for TRISO-particle fueled non-LWRs Andy Bielen, 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 supporting nuclear fuel safety for TRISO-particle fuel designs
- Decay Heat
- Neutron Multiplication & Criticality
- Shielding and Radiation Protection
- Overview of data availability, gaps, and where additional data would be beneficial 2
Nuclear Physics Considerations for TRISO/SFR Spent Fuel Safety NRC Regulations limit radiation dose under all phases of the fuel cycle:
Direct radiation dose Radioactive material releases Inadvertent criticality Computer codes used to determine:
Irradiated fuel composition for nuclides that contribute to:
Direct radiation dose and dose from radioactive material releases Decay heat Determination of criticality safety (keff)
Radiation dose and keff Codes must be validated against measured irradiated fuel data 3
Decay Heat Shielding and Radiation Protection Neutron Multiplication and Criticality 10 CFR 20 - Radiation Protection 10 CFR 50/52 - Power Plants 10 CFR 70 - Fuel Cycle Facilities 10 CFR 71 - Transportation 10 CFR 72 - SNF Storage
Non-LWR Source Term & Fuel Cycle Demonstration Projects NRCs comprehensive neutronics package
- Cross-section processing
- Decay heat analyses
- Criticality safety
- Radiation shielding
- Radionuclide inventory & depletion generation
- Reactor core physics NRCs comprehensive severe accident progression and source term code
- Accident progression
- Thermal-hydraulic response
- Core heat-up, degradation, and relocation
- Fission product release and transport behavior Source Term Fuel Cycle Non-LWR demonstration projects improve and validate SCALE & MELCOR for simulating non-LWRs for severe accident progression and fuel cycle analyses.
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Decay Heat, Criticality Safety, and Radiation Shielding / Dose Decay Heat Criticality Safety Shielding &
Dose Fuel Cycle SCALE/CSAS is used to perform criticality safety analyses. CSAS is a sequence that uses Monte Carlo transport codes KENO or Shift.
Used to determine the multiplication factor of any system.
SCALE/TRITON is used to generate specific ORIGEN reactor libraries; functionally bounds fuel enrichment and burnup.
SCALE/ORIGAMI is used to obtain the spent fuel inventories; uses ORIGEN to compute detailed irradiated and decayed isotopic compositions.
SCALE/MAVRIC is used to perform the shielding and dose analyses.
Uses the radiation source term & radionuclide inventories generated from SCALE/TRITON or SCALE/ORIGAMI.
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Non-LWR Reference Models INL Design A 5 MWth with a 5-year operating lifetime 1,134 heat pipes fueled with UO2 fuel (19.75 wt.% U-235)
Reactivity controlled via control drums PBMR-400
- 400 MWth reactor, graphite moderated
- Helium-cooled & TRISO-particle pebble-fueled at 10 wt.% U-235
- Fuel discharged at high burnup (90 GWd/MTIHM)
UCB Mk1 PB-FHR
- 236 MWth reactor at atmospheric pressures
- Online refueling MSRE
- 10 MWth reactor, graphite moderated at near atmospheric pressures
- Reactor fueled with liquid dissolved fuel in molten salt (34.5 wt. % U-235)
ABTR
- 250 MWth pool-type reactor, utilizing metallic U / HT-9 fuel rods
- Reactor fueled with U-Pu-Zr fuel slugs
- Liquid sodium coolant High-Temp. Gas Cooled Reactor Sodium-Cooled Fast Reactor Molten Salt-Cooled Reactor Molten Salt-Fueled Reactor Heat Pipe Reactor 6
Pebble Bed Reactor Workflows Source Term Fuel Cycle Fuel Depletion, Decay Heat, &
Radionuclide Inventory Generation Decay Heat Criticality Safety Shielding &
Dose 7
High-Temp. Gas Cooled Reactor PBMR-400
- 400 MWth reactor, graphite moderated
- Helium-cooled & TRISO-particle pebble-fueled at 10 wt.% U-235
- Fuel discharged at high burnup (90 GWd/MTIHM)
UCB Mk1 PB-FHR
- 236 MWth reactor at atmospheric pressures
- Online refueling Molten Salt-Cooled Reactor
Fuel Depletion, Decay Heat, and Nuclide Inventory Generation Source Term Fuel Depletion, Decay Heat, & Radionuclide Inventory Generation 8
UCB Mk1 PB-FHR
- 236 MWth reactor at atmospheric pressures
- Online refueling Molten Salt-Cooled Reactor SCALE/TRITON used for fuel depletion Continuous energy Monte Carlo physics or MG methods available (KENO or Shift)
MG methods utilize SCALEs double-het methods Equilibrium inventories generated via SLICE method SCALE Leap-In Method for Cores at Equilibrium Generates region-average fuel inventories Accounts for average behavior of pebbles as they transverse through the core Radionuclide inventories used to support downstream analyses.
MELCOR for severe accident progression & radionuclide transport ORIGAMI for decay heat analyses; utilizes the ORIGEN libraries from TRITON
High Temperature Gas-Cooled Reactors Fuel Cycle 9
Decay Heat, Criticality Safety, and Radiation Shielding / Dose Decay Heat Criticality Safety Shielding &
Dose Fuel Cycle 10
Decay Heat Analyses for TRISO-based Fuels Determine average spent fuel pebble inventory after discharge Leveraged from the non-LWR demonstration source term work (for HTGR)
TRITON & ORIGAMI used for generating inventories & performing decay-correction from ORIGEN reactor libraries Radionuclide inventories used to support downstream analyses.
MELCOR for severe accident progression & radionuclide transport MAVRIC for shielding & dose analyses 11 Vehicle / collision strike with a spent nuclear fuel storage tank loaded with spent TRISO-pebbles.
Once burnup limits are reached, pebble is moved into a spent fuel tank, with a capacity of holding ~620K pebbles.
Discharge rate - 483 pebbles / day; 1,284 days to fill spent fuel tank.
NRCs Computer Codes and Validation 12 SCALE Validation in Four Major Areas (Criticality Safety, Radiation Shielding, Reactor Physics, and Spent Fuel Inventory)
SCALE has been heavily validated for standard fuel designs in LWRs. SCALE 6.3 validation efforts are underway to validate SCALE for several advanced non-LWR systems.
Applications of non-LWR Demonstration Project - Kairos Hermes Construction Permit Application Support Blue: FLiBe Red: Fuel Pebble Black: Moderator Pebble Source Term Non-LWR demonstration project was instrumental in an effective and efficient review of a first of a kind non-LWR.
UCB Mk1 PB-FHR
- 236 MWth reactor at atmospheric pressures
- Online refueling Molten Salt-Cooled Reactor Kairos Hermes I
- 35 MWth reactor at atmospheric pressures
- Online refueling Generated a library of well-tested & demonstrated non-LWR reference plant models in SCALE & MELCOR.
Models can be heavily leveraged to support licensing reviews.
Leveraged the FHR model to supportrt the licensing review of Hermes I Similarities between the UCB Mk1 & Hermes I noted Leveraged existing models & insights from non-LWR demonstration project SCALE and MELCOR used for analyzing various scenarios (e.g.,
loss of forced circulation, accidental control rod withdrawal) 13
For More Information 14 Public workshop videos, slides, reports at advanced reactor source term webpage