ML24260A103
| ML24260A103 | |
| Person / Time | |
|---|---|
| Issue date: | 07/30/2024 |
| From: | Andrew Bielen NRC/RES/DSA/FSCB |
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| Shared Package | |
| ML24260A098 | List: |
| References | |
| Download: ML24260A103 (1) | |
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NRCs Advanced Reactor Applications NRC & ORNL ATF & non-LWR Seminar July 30, 2024 Dr. Andrew Bielen Fuel & Source Term Code Development Branch Division of Systems Analysis Office of Nuclear Regulatory Research
Overview
- Regulatory basis for recent research on non-LWRs & HALEU
- NRCs computational tools & codes with source term applications
- Licensing support case studies non-LWR Source Term Demonstration Project (Volume 3) non-LWR Fuel Cycle Demonstration Project (Volume 5) 2 7/30/2024 NRCs Advanced Reactor Applications
NRR & NMSS Sponsored Research for ATF/HBU/EE & HALEU Supporting advanced reactor technologies aimed at:
Code capabilities, assessments, validation, licensing support, and guidance development 3
Focus Areas RES is actively engaged in the latest advances in technology and relies on innovative research to ensure NRC readiness for new spent fuel activities ATF/HBU/EE, HALEU & non-LWRs NRR-2021-016, Regulatory Research Supporting Licensing of Burnup and Enrichment Extensions in Near-Term Accident Tolerant Fuel NRR-2021-017, Regulatory Research Supporting Licensing of Near-Term Accident Tolerant Fuel (ATF)
NMSS-2020-004, Regulatory Research Supporting Licensing of Near-Term Accident Tolerant Fuel (ATF)
NMSS-2020-005, Regulatory Research Supporting Licensing of Increased Enrichment (IE) and High Burnup (HBU) LWR Fuels NRR-2024-012, Modern BWR Assessment and Analysis Methods Enhancement for Improved Simulation Capabilities NRR-2023-015, Technical Analyses and Scoping Studies Supporting the Impacts of Increased Enrichment on Criticality Safety NRR-2024-002, Independent Confirmatory Analysis for non-LWR Construction Permit Application Reviews NMSS-2022-005, Recommendations for Fuel and Irradiation Parameters for HALEU and High Burnup LWR Spent Fuel NMSS-2022-006, Recommendations for using sensitivity/uncertainty (S/U) analyses in validation of computational methods for criticality safety of HALEU systems NMSS-2022-007, Updating Fission Product Bias Estimates in NUREG/CR-7109 7/30/2024 NRCs Advanced Reactor Applications
Computer Codes in the Regulatory Framework 4
7/30/2024 NRCs Advanced Reactor Applications
NRC Neutronics and Source Term Safety Assessment Codes 5
Modeling and simulation suite for neutronics-related nuclear safety analysis and design NRCs fuel performance code for calculating the thermal-mechanical response of steady-state and accident conditions NRCs reactor core simulator for determining the core behavior in nominal, transient and accident conditions Engineering-level code for simulating the response of the reactor core, primary coolant system, containment, and surrounding buildings to a severe accident 7/30/2024 NRCs Advanced Reactor Applications SCALE, MELCOR, FAST, and PARCS are established, assessed, validated and well-known NRC tools used to support safety assessment.
SCALE & PARCS in the Regulatory Framework 6
7/30/2024 NRCs Advanced Reactor Applications
Severe Accident Code Development & Regulatory Applications 7
7/30/2024 NRCs Advanced Reactor Applications
8 CABRI & PERFROI (France)
NSRR/TOKAI (Japan)
Halden (Norway)
STUDSVIK (Sweden)
ANL (U.S.)
RIA & LOCA data and analysis RIA & LOCA data and analysis Fuel Thermal &
Mechanical Analysis Cladding integrity, PCI, FFRD Cladding oxidation, embrittlement and reorientation NRR review - advanced alloys (M5, ZIRLO) and fuels (additives, dopants, and burnable absorbers)
NMSS Spent fuel cask storage & accident analysis Regulatory Guide and Staff Technical Positions revision Accident Tolerant Fuel NMSS Spent fuel initial conditions Spent fuel cask storage & accident analysis Extended storage NRR Fuel and gap inventory release fractions for HBU fuel Evaluation of fuel fragmentation and dispersal ORNL (U.S.)
HBU fuel mechanical properties + RG support In-house fuel analysis (NRC)
Rulemaking 10 CFR Part 50.46(c)
FIDES (NEA)
Fuel Thermal &
Mechanical Analysis, cladding creep, RIA NRR RIA for HBU RG 1.236 Advanced Reactors (Non-LWRs)
Fuel Performance Code Development & Regulatory Applications 7/30/2024
SCALE & MELCOR non-LWR Source Term Demonstration Project 9
7/30/2024 NRCs Advanced Reactor Applications
10 NRCs Strategy for non-LWR reviews
- Phase 1 - Vision & Strategy
- Phase 2 - Implementation Action Plans
- Strategy #2 - Computer Codes and Review Tools
- Identifies computer code & development activities
- Identifies key phenomena
- Assess available experimental data & needs IAP Strategy #2 Computer Codes and Tools Volume #1 Systems Analysis Volume #2 Fuel Performance Volume #3 Source Term, Consequence Volume #4 Licensing &
Dose Volume #5 Nuclear Fuel Cycle 7/30/2024 NRCs Advanced Reactor Applications
11 Volume 3 Project Objectives Demonstrate use of SCALE and MELCOR for non-LWRs Show how codes can be used to understand accident behavior
- Provide insights for regulatory applications
- Facilitate dialogue on staffs approach for source term Demonstrate ability to identify accident characteristics and uncertainties affecting source term Develop publicly available input models for representative designs
- Heat pipe reactor (HPR)
- Pebble-bed gas-cooled reactor (HTGR)
- Pebble-bed molten-salt-cooled (FHR)
- Molten-salt-fueled reactor (MSR)
- Sodium-cooled fast reactor (SFR)
For More Information on Volume 3 7/30/2024 NRCs Advanced Reactor Applications
12 Reference Plant Models HTGR HPR FHR MSR SFR Hermes I & II Natrium Xe-100 Reference Plants Example Applications Aurora Terrestrial 7/30/2024 NRCs Advanced Reactor Applications
Recent Application for non-LWR Licensing Support 13 7/30/2024 NRCs Advanced Reactor Applications
Licensing Support Artificial Intelligence Artificial Intelligence Artificial Intelligence Developed Efficient Strategy and Plan Developed Reference Plant Models Developed Expertise & Show Readiness NRC Integrated Action Plan (IAP) Strategy 2: Modernizing our Tools Public Workshops: SCALE/MELCOR Non-LWR Source Term Demonstration Project 2021 2022 Heat Pipe Reactor Gas Cooled Reactor Molten Salt Cooled Reactor Molten Salt Fueled Reactor Sodium Fast Reactor Modify Reference Plant Model Support Hermes Review 14 7/30/2024
Reactor Characteristics o 236 MWth reactor / TRISO-pebble fueled / Flibe cooled o Atmospheric pressures / Online refueling o Direct Reactor Auxiliary Cooling System (DRACS) 3 trains / Capacity - 2.36 MW/train Molten-salt-cooled Pebble-bed Rx - UCB Mark 1 15 Accidents Modeled
- SBO -Station blackout
- LOCA -Loss-of-coolant accident 7/30/2024
Existing SCALE reactor physics models used
- No new phenomenological models needed Developed methods to handle unique aspects of FHR TRISO particle and pebble bed geometry Varying burnup of different pebbles in core Added a generic equation of state utility for thermal hydraulic analysis in advanced reactor working fluids Added fission product transport and retention models for molten salts Enhanced fission product release modeling for TRISO Point kinetics enhancements for reactivity insertion 16 FHR Reference Plant - UCB Mark 1 Code Improvements 7/30/2024
FHR Reference Plant SCALE Results and Conclusions MG analyses in SCALE used to generate core-average fuel compositions, power profiles, decay heat, reactivity feedback coefficients, tritium production rates, and effects of Xe.
Core-average Inventories Reactivity Feedback Coefficients 17 7/30/2024
Anticipated Transient without Scram (ATWS)
Fuel heat-up was limited by reactivity feedback and the passive decay heat removal system Station Blackout (SBO)
With failure of the passive decay heat removal system, coolant boiling occurred over the course of several days Loss of Coolant Accident (LOCA)
With one train of decay removal system operating, coolant boiling was possibly averted.
With failure of the passive decay heat removal system, fuel damage occurred.
FHR Reference Plant MELCOR Results and Conclusions End of the Xenon transient and a return to power.
ATWS with variable DRACS 18 LOCA SBO 7/30/2024 NRCs Advanced Reactor Applications
Hermes
- CP application submitted September 29, 2021
- NRC performed independent analysis using SCALE and MELCOR
- Better understand plant response to DBA scenarios
- Compare with Kairos PSAR analysis
- Scenarios o Reactor heat-up (e.g., loss of forced circulation),
o Insertion of excess reactivity (e.g., accidental control rod withdrawal) o Analysis described in SE and presented to ACRS o Analysis benefitted from previous SCALE and MELCOR experience modeling UCB-Mark 1 FHR Reference Plant 19 7/30/2024 NRCs Advanced Reactor Applications
Blue: Flibe Red: Fuel Pebble Black: Moderator Pebble MG Monte Carlo transport using SCALE/KENO-VI, fuel isotopics calculated with SCALE/ORIGEN Random pebble geometry approximated by regular lattice Equilibrium fuel isotopics generated iteratively via 2D slice models with SCALE/TRITON Axially-dependent fuel isotopics inserted into 3D core model for reactivity and power shape evaluations Does not currently include shutdown (in-bed) elements - on list for further development Relative Power Kairos PSAR SCALE Axial (-)
1.2 1.19 Radial (-)
1.2 1.76 Peak Pebble (-)
1.8 2.09 Parameter Kairos PSAR SCALE*
Fuel Doppler (pcm/K)
-4.1
-4.30 +/- 0.27 Moderator (pcm/K)
-0.4
-0.47 +/- 0.13 Coolant (pcm/K)
-1.6
-1.62 +/- 0.02 Void (pcm/% void, @3% void)
-53
-46.6 +/- 4.0 Reflector (pcm/K)
+2.0
+1.92 +/- 0.23 (pcm) 605 576 +/- 10 Hermes - SCALE Model
- - includes Monte Carlo uncertainty
- calculated assuming temperature distributions provided by MELCOR 20 7/30/2024 NRCs Advanced Reactor Applications
MELCOR results as compared with PSAR (upper right)
Insertion of Excess Reactivity Loss of Forced Circulation Hermes - SCALE/MELCOR Results 21 Withdrawal of control element inserts 3.02$ over 100 seconds Reactor trips on high power Concurrent trip of primary and intermediate coolant pumps Reactor trips on overtemperature 7/30/2024
SCALE & MELCOR non-LWR Fuel Cycle Demonstration Project 22 7/30/2024 NRCs Advanced Reactor Applications
LWR Nuclear Fuel Cycle Regulations for the Nuclear Fuel Cycle Protects onsite workers, public and the environment against radiological and non-radiological hazards that arise from fuel cycle operations.
Radiation hazards Radiological hazards Non-radiological (chemical) hazards Applicable Regulations Uranium Recovery / Milling - 10 CFR Part 20 Uranium Conversion - 10 CFR Parts 30, 40, 70, 73 and 76 Uranium Enrichment - 10 CFR Parts 30, 40, 70, 73 and 76 Fuel Fabrication - 10 CFR Parts 30, 40, 70, 73 and 76 Reactor Utilization - 10 CFR Parts 50 & 74 Spent Fuel Pool Storage - 10 CFR Parts 50.68 Spent Fuel Storage (Dry) - 10 CFR Parts 63, 71, and 72 23 7/30/2024 NRCs Advanced Reactor Applications
Project Scope - Non-LWR Fuel Cycle Enrichment UF6 enrichment UF6 Transportation Fuel Fabrication Fresh Fuel Transportation Fuel Utilization (including on-site spent fuel storage)
- Not envisioned to change from current methods.
Uranium Mining & Milling
- Successfully completed and leveraged from the Volume 3 - Source Term& Consequence work Power Production
- Large amount of uncertainties for non-LWR concepts & lack of information Spent Fuel Off-site Storage & Transportation
- Large amount of uncertainties for non-LWR concepts & lack of information Spent Fuel Final Disposal Stages in scope for Volume 5 Stages out of scope for Volume 5 24 7/30/2024 NRCs Advanced Reactor Applications
25 Non-LWR Designs Considered 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
Representative Fuel Cycle Designs Completed 5 non-LWR fuel cycle designs for -
HPR - INL Design A HTGR - PBMR-400 FHR - UCB Mark 1 MSR - MSRE SFR - ABTR Identifies potential processes & methods, for example:
What shipping package could transport HALEU-enriched UF6? What are the hazards associated?
How is spent SFR fuel moved? What are the hazards associated?
How is fissile salt manufactured for MSRs? What are the various kinds of fissile salt that may be used? What are the hazards?
Used as the Initial and Boundary Conditions for the SCALE &
MELCOR Analyses 26 ML24004A270 7/30/2024 NRCs Advanced Reactor Applications
SCALE Analyses for Selected Accidents from the HTGR Fuel Cycle Criticality-related Analyses Spent Fuel Pebble Inventory & Fission Product Release In-Facility UF6 Release Water ingress into the DN30-X during UF6 transport
- Shown to be subcritical Spent Fuel Storage Tank Release
- Spent fuel tank holding 620,000 pebbles simulated
- Approx. 500 pebbles discharged daily / 1284 days to fill SFT
- Total decay heat and inventory of SFT determined UF6 Cylinder Rupture
- UF6 cylinders are overfilled & heated resulting in rupture and release 27 7/30/2024 NRCs Advanced Reactor Applications
SCALE Analyses for Selected Accidents from the SFR Fuel Cycle Radiative Source Term & Dose Analysis Criticality-related Analyses Spent Nuclear Fuel Assembly Drop
- ABTR HALEU fuel assembly drop during fuel handling
- SCALE used for neutron & gamma source term generation Misfeed of Material during Reprocessing
- During batch electro-processing, there is a misfeed of material leading to a fissile material buildup.
- SCALE is used to determine fuel inventory and perform criticality analyses.
28 7/30/2024 NRCs Advanced Reactor Applications
MELCOR Analyses for a Dropped SFR Assembly Fuel Assembly Damage & Radionuclide Transport 29 Release to Environment Fuel Assembly Damage Dropped Fuel Assembly Spent Nuclear Fuel Assembly Drop
- Insufficiently cooled FA is removed, loaded within a IBC, dropped, and results in no active cooling (failed).
- MELCOR is used to determine FA damage, failure, and inventory release to the environment.
7/30/2024 NRCs Advanced Reactor Applications
SCALE Analyses for Selected Accidents from the MSR Fuel Cycle Release of Fission Products during Operations 30 Operational Dose Rate & Radiative Source Term Generation
- MSBR used in the Volume 5 MSR analyses; a larger thermal spectrum reactor than the MSRE
- CSAS-Shift used to generate neutron fission source used by MAVRIC-Shift for 3D dose analysis
- Did not account for fission product decay & activation Doses were dominated by gamma radiation.
7/30/2024 NRCs Advanced Reactor Applications
SCALE Benchmarking & Validation Activities 31 HTGRs MSRs SFRs SCALE Validation in Four Major Areas (Spent Fuel Inventory, Reactor Physics, Radiation Shielding, and Criticality Safety) 7/30/2024 NRCs Advanced Reactor Applications
32 Upcoming SCALE and MELCOR Training Opportunities SCALE user workshops Public workshop on heat pipe reactor analysis MELCOR user workshop
- October 14-18, 2024
- SCALE/ORIGEN Standalone Fuel Depletion, Activation, and Source Term Analysis
- October 21-25, 2024
- Source Terms for Advanced Reactor Spent Fuel Applications
- October 28-November 1, 2024
- SCALE Criticality Safety Calculations June 9-13, 2025: week-long MELCOR workshop 2025: Demonstration of SCALE and MELCOR for horizontal heat pipe reactor using TRISO compact fuel For more information, contact Lucas Kyriazidis, SCALE Project Manager For more Information, contact Shawn Campbell, MELCOR Project Manager 7/30/2024
Conclusions
- Significant capabilities, assessment, and validation activities have been completed in NRCs accident progression codes for non-LWRs
- SCALE/MELCOR Capabilities highlighted in public workshops under NRCs non-LWR Source Term &
Fuel Cycle Demonstration Projects
- Leveraged Volume 3 reference plant models to support NRR/DANUs Hermes I Construction Permit Review
- Novel technologies require strategic planning
- Confirmatory analyses support efficient and technically sound regulatory decision-making 33 RES is ready to support NMSS & NRR 7/30/2024 NRCs Advanced Reactor Applications
Questions & Comments 34 7/30/2024 NRCs Advanced Reactor Applications