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Standard Review Plan for Transportation Packages for MOX Spent Nuclear Fuel
Standard Review Plan for Transportation Packages for MOX Spent Nuclear Fuel
The NRC contracted with LLNL to compile this supplement to NUREG-1617 to incorporate additional
information specific to mixed uranium-plutonium oxide (MOX) fuel. This supplement provides details
on package review guidance resulting from significant differences between spent nuclear fuel from
irradiated LEU fuel and that from irradiated MOX fuel. The information presented is not to be
construed as having the force and effect of NRC regulations (except where regulations are cited), or as
Criticality Risks During Transportation of Spent Nuclear Fuel
Criticality Risks During Transportation of Spent Nuclear Fuel
This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies. The report takes into account the results of a 2007 peer review of the initial version of this probabilistic risk assessment, which was published as EPRI Technical Report 1013449 in December 2006.
NRC SFST ISG-8: Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks
NRC SFST ISG-8: Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks
Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of
Radioactive Material, and 10 CFR Part 72, Licensing Requirements for the Independent
Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater
Than Class C Waste, require that spent nuclear fuel (SNF) remain subcritical in transportation
and storage, respectively. Unirradiated reactor fuel has a well-specified nuclide composition
that provides a straightforward and bounding approach to the criticality safety analysis of
Transportation of Commercial Spent Nuclear Fuel Regulatory Issues Resolution
Transportation of Commercial Spent Nuclear Fuel Regulatory Issues Resolution
The U.S. industry’s limited efforts at licensing transportation packages characterized as “highcapacity,”
or containing “high-burnup” (>45 GWd/MTU) commercial spent nuclear fuel
(CSNF), or both, have not been successful considering existing spent-fuel inventories that will
have to be eventually transported. A holistic framework is proposed for resolving several CSNF
transportation issues. The framework considers transportation risks, spent-fuel and cask-design
SCALE-4 Analysis of Pressurized Water REactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5
SCALE-4 Analysis of Pressurized Water REactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5
The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
(AFR) criticality safety analyses be validated against experimental measurements. If credit for the
negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark AFR criticality analysis methods using selected critical
configurations from commercial pressurized-water reactors (PWR).
PWR Axial Burnup Profile Analysis
PWR Axial Burnup Profile Analysis
Burn-up Credit Criticality Safety Benchmark - Phase VII, UO2 Fuel: Study of Spent Fuel Compositions for Long-term Disposal
Burn-up Credit Criticality Safety Benchmark - Phase VII, UO2 Fuel: Study of Spent Fuel Compositions for Long-term Disposal
Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation--Calvert Cliffs, Takahama, and Three Mile Island Reactors
Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation--Calvert Cliffs, Takahama, and Three Mile Island Reactors
This report is part of a report series designed to document benchmark-quality radiochemical isotopic
assay data against which computer code accuracy can be quantified to establish the uncertainty and bias
associated with the code predictions. The experimental data included in the report series were acquired
from domestic and international programs and include spent fuel samples that cover a large burnup range.
The measurements analyzed in the current report, for which experimental data is publicly available,
Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask
Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask
The Interim Staff Guidance on burnup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission’s Spent Fuel Project Office, recommends a burnup measurement for each assembly to confirm the reactor record and compliance with the assembly burnup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained.