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

doi:10.13182/NT13-154

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

Abstract

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.

Authors:

Radulescu, Georgeta ^[1]; Gauld, Ian C. ^[1]; Ilas, Germina ^[1]; Wagner, John C. ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Publication Date:: 2014-11-01

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE

OSTI Identifier:: 1163574

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Nuclear Technology

Additional Journal Information:: Journal Volume: 188; Journal Issue: 2; Journal ID: ISSN 0029-5450

Publisher:: Taylor & Francis - formerly American Nuclear Society (ANS)

Country of Publication:: United States

Language:: English

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

Citation Formats


                    Radulescu, Georgeta, Gauld, Ian C., Ilas, Germina, and Wagner, John C. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses.  United States: N. p., 2014. 
        Web.  doi:10.13182/NT13-154.

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                    Radulescu, Georgeta, Gauld, Ian C., Ilas, Germina, & Wagner, John C. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses.  United States.  doi:10.13182/NT13-154.

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                    Radulescu, Georgeta, Gauld, Ian C., Ilas, Germina, and Wagner, John C. Sat .  
        "Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses".  United States.  doi:10.13182/NT13-154.  https://www.osti.gov/servlets/purl/1163574.

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@article{osti_1163574,

  title        = {Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses},

  author       = {Radulescu, Georgeta and Gauld, Ian C. and Ilas, Germina and Wagner, John C.},

  abstractNote = {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.},

  doi          = {10.13182/NT13-154},

  journal      = {Nuclear Technology},

  number       = 2,

  volume       = 188,

  place        = {United States},

  year         = {2014},

  month        = {11}

}

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DOI: 10.13182/NT13-154

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Works referenced in this record:

An analysis of variance test for normality (complete samples)
journal, December 1965

Shapiro, S. S.; Wilk, M. B.
Biometrika, Vol. 52, Issue 3-4
DOI: 10.1093/biomet/52.3-4.591

SCALE 6: Comprehensive Nuclear Safety Analysis Code System
journal, May 2011

Bowman, S. M.
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT10-163

Reactor Physics Methods and Analysis Capabilities in SCALE
journal, May 2011

DeHart, Mark D.; Bowman, Stephen M.
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT174-196

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

Abstract

Citation Formats

An analysis of variance test for normality (complete samples) journal, December 1965

SCALE 6: Comprehensive Nuclear Safety Analysis Code System journal, May 2011

Reactor Physics Methods and Analysis Capabilities in SCALE journal, May 2011

An analysis of variance test for normality (complete samples)
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SCALE 6: Comprehensive Nuclear Safety Analysis Code System
journal, May 2011

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