ML24291A150
| ML24291A150 | |
| Person / Time | |
|---|---|
| Issue date: | 10/17/2024 |
| From: | Lucas Kyriazidis NRC/RES/DSA/FSCB |
| To: | |
| References | |
| Download: ML24291A150 (1) | |
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Updates on NRC ATF/HE/HBU Back-end Activities Lucas Kyriazidis Office of Nuclear Regulatory Research Division of Systems Analysis Fuel & Source Term Code Development Branch Extended Storage Collaboration Program Fall 2024 October 21, 2024 1
Motivation & Objectives Purpose Provide a snapshot of recently completed research related to spent fuel storage and transportation, covering topics related to:
Accident Tolerant Fuels Extended Enrichment Higher Burnup Fuels (> 45 GWd/MTU)
Highlighted Today New HBU PWR Radiochemical Assay Data & Use Assessing the Impacts of ATF/HBU/EE for Spent Fuel Applications PIRT on Spent Fuel Storage and Transportation for ATF/HBU/EE Assessing Alternative Criteria for Gross Ruptures in Spent Nuclear Fuels 2
NRC is performing innovative research to ensure readiness for more effective
& efficient regulatory decision-making for spent fuel activities.
Recently Completed & Upcoming SNF Research 2020 - 2023 2024
- Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses Feb. 2024 NUREG/CR-7302
- Fuel Assembly and Irradiation Parametric Study for EE and HBU LWR Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages*
Apr. 2024 NUREG/CR-7306
- Phenomena Identification and Ranking Table (PIRT) for Fuel and Cladding Property Changes Relevant to SNF Storage & Transportation of ATF Concepts*
Oct. 2024 Available later this month.
- Assessment of Gross Breaches in Nuclear Fuel Oct. 2024 ML24284A100
- Assessment of Alternative Criteria for Gross Ruptures in Spent Nuclear Fuel*
Oct. 2024 ML24284A091
- Development of Decay Heat Sensitivity Analysis Capability in SCALE/ORIGEN Nov. 2024 TBD
- Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses Dec. 2024 TBD
- Sensitivity / Uncertainty Methods for Nuclear Criticality Safety Validation Dec. 2024 TBD
- Determination of Bias and Bias Uncertainty for Criticality Safety Computational Methods Dec. 2024 TBD
- Assessment of Existing Transportation Packages for Use with HALEU Sept. 2020 ML21040A518
Feb. 2021 ML21088A336
Mar. 2021 ML21088A354
- Ext. Enrichment Accident Tolerant LWR Fuel Isotopic and Lattice Parameter Trends Mar. 2021 ML21088A254
- Impacts of LEU+ and ATF on Fresh Fuel Storage Criticality Safety Feb. 2022 ML22098A137
- Impacts of LEU+ and HBU Fuel on Decay Heat and Radiation Source Term Apr. 2022 ML22159A191
- Validating Actinides and Fission Products for Burnup Credit Criticality Safety Analyses - Nuclide Compositions Prediction with Extended Validation Basis*
Sept. 2023 NUREG/CR-7303
- Highlighted in Todays Presentation 3
New High Burnup PWR Radiochemical Assay Data NUREG/CR-7303 - Validating Actinides and Fission Products for Burnup Credit Criticality Safety Analyses 4
New High Burnup PWR Radiochemical Assay Data Review of SFCOMPO revealed gaps in existing databases (e.g., HBU samples,
>60 GWd/MTU)
SFCOMPO is the largest international database of open data for spent nuclear fuel.
NRC sponsored high-precision RCA measurements on 8 HBU PWR samples (2018 - 2023)
- Leveraged remaining fuel segments, at ORNL, following destructive examinations for DOEs Sibling Rod Program Analytical protocols & experimentally-measured data is publicly available in ML23094A047.
5 8 new PWR samples
NUREG/CR-7303 - Validating Actinides and Fission Products for Burnup Credit Criticality Safety Analyses Used the 8 high-precision RCA measurements on HBU PWR samples to expand the validation basis for criticality safety analyses Burnup range 34 - 67 GWd/MTU Enrichments < 5 wt. % U-235 Data is critical in validating computer codes for estimating nuclide inventories in spent fuel ML23094A047 Burnup credit accounts for the reduced reactivity due to depletion of fissile material and buildup of fission products and actinides Provides an approach for estimating the bias and bias uncertainty from the calculated nuclide inventories from fuel depletion codes Extended the validation basis up to 80 GWd/MTU Supported by incorporating new measurement data from 129 spent fuel samples New data reduced the overall keff bias uncertainty NUREG/CR-7303 6
Future RCA Measurements US industry is pursuing many new initiatives & fuel designs Near-term and long-term accident tolerant fuel designs, Current fuel designs with higher burnups and increased U-235 loading, New operating conditions (e.g., uprated powers)
Future efforts to review and add to SFCOMPO for new RCA data for validating fuel depletion codes for these new fuel designs & initiatives.
Identify where new experimental data would be beneficial 7
Assessing the Impacts of ATF / HBU / EE for Spent Nuclear Fuel Applications 8
NUREG/CR-7306 - Parametric Study for EE/HBU for Dry Storage Casks and Transportation Packages NUREG/CR-7306 SCALE 6.3 used to perform many case studies to determine the effects of various fuel assembly, irradiation, and decay parameters has on dose rates &
criticality.
- Fuel burnups up to 75 GWd/MTU,
- Enrichments up to 8 wt. % U-235,
- Cooling times up to 100 years,
- Specific power between 15 - 50 MW/MTU,
- Effects of soluble boron during operations (600 - 1800 ppm),
- Many other parameters investigated (e.g., burnable absorbers, fuel density, fuel temperature, coolant void, etc.).
Provide general insights on the impacts of extended enriched and higher burnup fuels have on radiation shielding and criticality safety analyses 9
NUREG/CR-7306 - Parametric Study for EE/HBU for Dry Storage Casks and Transportation Packages Neutron dose rates as a function of fuel burnup and fuel enrichment for PWR fuel assemblies; dry storage casks (left) and transportation packages (right).
10 Insights generated for both dry storage casks and transportation packages. Increasing burnup and enrichment has a competing effect on neutron dose rate.
Burnup - Neutron Dose Rate Enrichment - Neutron Dose Rate
PIRT on Spent Fuel Storage and Transportation for ATF/HBU/EE 11
PIRT on Spent Fuel Storage and Transportation for ATF/HBU/EE PNNL-30451 - Literature Review for SFST Applications of Accident Tolerant Fuel Concepts - Current State-of-Knowledge Revision 0 (Sept. 2020) performed a literature review for spent fuel cladding technologies for ATF Revision 1 (Mar. 2024) updated the literature review for new information and expand to include EE/HBU fuels Supporting documentation for the PIRT May 2024, NRC sponsored a PIRT on SFST applications for ATF/HBU/EE Provided PNNL-30451 Revision 1 (literature review) to PIRT members Identify phenomena that are of high importance with low knowledge levels Findings to be used to inform development & decisions on future test plans ML24071A255 12 A systematic approach to help identify and prioritize important to safety phenomena.
PIRT on Spent Fuel Storage and Transportation for ATF/HBU/EE Held May 29 - 30, 2024 at PNNL (Richland, WA)
Panel members reviewed the current state-of-knowledge on ATF/HBU/EE for SFST applications in PNNL-30451 Panel members from industry, academia, US National Laboratories PIRT documented in PNNL-30451 Revision 2, Appendix A Ranked each phenomena based on importance (H, M, L) & knowledge level (H, M, L)
Identified phenomena relevant for SFST applications for:
10 for Cr-coated Zr-alloy claddings (No H/L) 8 for FeCrAl claddings (5 H/L) 2 for high enrichment fuels (No H/L) 9 for high burnup fuels (No H/L) 13 To be released in NRC Public ADAMS later this month.
Example PIRT Findings for Cr-coated Zr-alloy Claddings 14
Example PIRT Findings for FeCrAl Claddings 15
Assessment of Alternative Criteria for Gross Ruptures in Spent Nuclear Fuel 16
Assessment of Alternative Criteria for Gross Ruptures in Spent Nuclear Fuel Research to identify an alternative definition and new metrics for classifying gross ruptures in spent nuclear fuel.
Grossly ruptured SNF must be confined in a damaged fuel can or an acceptable alternative, prior to loading into a dry storage system.
Identifying grossly rupture fuel using the > 1 mm metric is difficult, resulting in an overly conservative approach to fuel classification.
Suspected of gross rupture assembly classified as grossly ruptured load into a DFC 17 ML24284A091
Background - ISG-1 Revision 2 - Classifying the Condition of SNF for Interim Storage and Transportation on Function 10 CFR 72.122(h) - Spent fuel cladding must be protected during storage against degradation that leads to gross ruptures, but gross rupture is not defined.
ISG-1 Rev. 2 provides some clarifications for gross breaches Breached Fuel Rod Gross Breach Non-Gross Breach Metrics Reactor operating records show heavy metal isotopes in the reactor coolant system (RCS)
Visual examination indicates breach is
>1 mm (width)
Definition A breach in cladding that is larger than a pinhole leak or hairline crack.
Visually measuring breach size is difficult, resulting in overly classifying fuel as grossly ruptured, and increases number of assemblies stored in a damaged fuel can (DFC).
18
Summary of Definitions and Metrics for Evaluating Gross Ruptures ISG-1 Revision 2 EPRI PNNL Definition of Gross Rupture Potential release of fuel particulates greater than the average size fuel fragment Allowing fuel material release fragments that require additional precautions to prevent fuel release during handling Potential release of fuel particulate greater than average fuel fragment or cause significant loss of geometry Metrics
- Sipping to identify breached fuel rod, by detection of FG.
Gross Rupture, if:
- Presence of heavy metal in the RCS
- Breach size greater than 1 mm
- Sipping to identify breached fuel rod, by detection of FG.
Gross Rupture, if:
- Presence of heavy metal in the RCS
- Sipping to identify breached fuel rod, by detection of FG.
Gross Rupture, if:
- Reactor or fuel handling accident occurred 19 These efforts support a performance-based approach for identifying gross ruptures in spent nuclear fuel.