Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

Georgeta Radulescu, Ian C. Gauld, Germina Ilas, John C. Wagner

Research output: Contribution to journalArticlepeer-review

12 Scopus citations

Abstract

This paper describes a depletion code validation approach/or criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of criticality safety analysis models by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in the effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1 and the Evaluated Nuclear Data FilelB (ENDF/B) Version VII nuclear data. The validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance (ISG)-8.

Original languageEnglish
Pages (from-to)154-171
Number of pages18
JournalNuclear Technology
Volume188
Issue number2
DOIs
StatePublished - Nov 1 2014

Funding

FundersFunder number
Oak Ridge National Laboratory
U.S. Department of Energy

    Keywords

    • Criticality
    • Depletion
    • Validation

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