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

Journal Article · · Nuclear Technology
DOI:https://doi.org/10.13182/NT13-154· OSTI ID:1163574
 [1];  [1];  [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
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This paper describes a depletion code validation approach for 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 a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in 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 SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

Research Organization:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Organization:
USDOE
Grant/Contract Number:
AC05-00OR22725
OSTI ID:
1163574
Journal Information:
Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 2 Vol. 188; ISSN 0029-5450
Publisher:
Taylor & Francis - formerly American Nuclear Society (ANS)Copyright Statement
Country of Publication:
United States
Language:
English

References (13)

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Cited By (1)

Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters journal February 2019

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Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ISG-8
burnup credit
criticality
depletion
validation