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An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions
An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions
Taking credit for the reduced reactivity of spent nuclear fuel (SNF) in criticality analyses is referred to as burnup credit (BUC). Criticality safety evaluations require validation of the computational methods with critical experiments that are as similar as possible to the safety analysis models, and for which the keff values are known. This poses a challenge for validation of BUC criticality analyses, as critical experiments with actinide and fission product (FP)
CSNF Loading Curve Sensitivity Analysis
CSNF Loading Curve Sensitivity Analysis
The purpose of this scientific analysis report, CSNF Loading Curve Sensitivity Analysis, is to establish the required minimum burnup as a function of initial enrichment for both pressurized water reactor (PWR) and boiling water reactor (BWR) commercial spent nuclear fuel (CSNF) that would allow permanent disposal of these waste forms in the geologic repository at Yucca Mountain. The relationship between the required minimum burnup and fuel assembly initial enrichment forms a loading curve.
Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned Fuel
Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned Fuel
There are more than 250 forms of U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program (NSNFP) has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. Fast Flux Test Facility (FFTF) fuel has been designated as the representative fuel for the mixed-oxide (MOX) fuel group which is a mixture of uranium and plutonium oxides.
Intact and Degraded Mode Criticality Calculations for the Codisposal of ATR Spent Nuclear Fuel in a Waste Package
Intact and Degraded Mode Criticality Calculations for the Codisposal of ATR Spent Nuclear Fuel in a Waste Package
SAS2H Generated Isotopic Concentrations for B&W 15xl5 PWR Assembly (SCPB: N/A)
SAS2H Generated Isotopic Concentrations for B&W 15xl5 PWR Assembly (SCPB: N/A)
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1
SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary
SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary
The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit is to be taken for
the reduced reactivity of burned or spent fuel relative to its original $fresh# composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to
OECD/NEA Burnup Credit Criticality Benchmarks Phase IIIB: Burnup Calculations of BWR Fuel Assemblies for Storage and Transport
OECD/NEA Burnup Credit Criticality Benchmarks Phase IIIB: Burnup Calculations of BWR Fuel Assemblies for Storage and Transport
The report describes the final results of the Phase IIIB Benchmark conducted by the
Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy
Agency (NEA) of the Organization for Economic Cooperation and Development (OECD).
The Benchmark was intended to compare the predictability of current computer code and
data library combinations for the atomic number densities of an irradiated BWR fuel
assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM
Evaluation of Measured LWR Spent Fuel Composition Data for Use in Code Validation End-User Manual
Evaluation of Measured LWR Spent Fuel Composition Data for Use in Code Validation End-User Manual
Burnup credit (BUC) is a concept applied in the criticality safety analysis of spent nuclear fuel
in which credit or partial credit is taken for the reduced reactivity worth of the fuel due to both fissile
depletion and the buildup of actinides and fission products that act as net neutron absorbers.
Typically, a two-step process is applied in BUC analysis: first, depletion calculations are performed
to estimate the isotopic content of spent fuel based on its burnup history; second, three-dimensional
Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel
Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel
This report investigates various calculational modeling issues associated with boilingwater-
reactor (BWR) fuel depletion relevant to burnup credit. To date, most of the efforts in
burnup-credit studies in the United States have focused on issues related to pressurized-waterreactor
(PWR) fuel. However, requirements for the permanent disposal of BWR fuel have
necessitated the development of methods for predicting the spent fuel contents for such fuels.
Concomitant with such analyses, validation is also necessary. This report provides a summary of
DISPOSABILITY OF LOADED U.S. DUAL-PURPOSE CANISTERS FROM A CRITICALITY STANDPOINT
DISPOSABILITY OF LOADED U.S. DUAL-PURPOSE CANISTERS FROM A CRITICALITY STANDPOINT
This paper assesses the feasibility of direct disposal of loaded dual-purpose canisters (DPCs) from a criticality standpoint by evaluating attributes that could be credited to justify that the DPCs remain subcritical over a repository performance period. This study investigates the uncredited criticality margin associated with actual fuel loading compared with the regulatory licensing design basis limits and evaluates the percentage of DPCs that remain subcritical solely based on the uncredited criticality margin.