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.

Copy to clipboard


                    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.

Copy to clipboard


                    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.

Copy to clipboard


                    
@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}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (DOE)

Publisher's Version of Record

DOI: 10.13182/NT13-154

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 4 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Works referenced in this record:

Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE
journal, May 2011

Rearden, B. T.; Williams, M. L.; Jessee, M. A.
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT174-236

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

Resonance Self-Shielding Methodologies in SCALE 6
journal, May 2011

Williams, Mark L.
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT09-104

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE
journal, May 2011

Gauld, Ian C.; Radulescu, Georgeta; Ilas, Germina
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT11-3

Similar Records in DOE PAGES and OSTI.GOV collections:

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

Conference Radulescu, Georgeta ; Gauld, Ian C ; Ilas, Germina ; ...

The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address themore »« less
An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

Conference Scaglione, John M ; Mueller, Don ; Wagner, John C

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only ismore »« less
An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (k_eff) Predictions

Journal Article Scaglione, John M. ; Mueller, Don E. ; Wagner, John C. - Nuclear Technology

One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (k_eff) evaluations based on best-available data and methodsmore »« less
Cited by 2
DOI: 10.13182/NT13-151

Full Text Available
Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

Conference Mueller, Don ; Bowen, Douglas G ; Marshall, William BJ J

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k_eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept themore »« less
Full Text Available
Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

Technical Report Radulescu, Georgeta

Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNFmore »« less
DOI: 10.2172/982739

Full Text Available

Similar Records

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

Abstract

Citation Formats

Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE journal, May 2011

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

Resonance Self-Shielding Methodologies in SCALE 6 journal, May 2011

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE journal, May 2011

Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE
journal, May 2011

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

Resonance Self-Shielding Methodologies in SCALE 6
journal, May 2011

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE
journal, May 2011