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Title: Integral Nuclear Data Validation Using Experimental Spent Nuclear Fuel Compositions

 [1];  [1];  [2];  [1]
  1. ORNL
  2. OECD Nuclear Energy Agency
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
Work for Others (WFO)
OSTI Identifier:
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: M&C International Conference on Mathematics & Computational Methods Applied to Nuclear Science and Engineering, Jeju Island, South Korea, 20170416, 20170420
Country of Publication:
United States

Citation Formats

Gauld, Ian C, Williams, Mark L, Michel-Sendis, Franco, and Martinez-Gonzalez, Jesus S. Integral Nuclear Data Validation Using Experimental Spent Nuclear Fuel Compositions. United States: N. p., 2017. Web.
Gauld, Ian C, Williams, Mark L, Michel-Sendis, Franco, & Martinez-Gonzalez, Jesus S. Integral Nuclear Data Validation Using Experimental Spent Nuclear Fuel Compositions. United States.
Gauld, Ian C, Williams, Mark L, Michel-Sendis, Franco, and Martinez-Gonzalez, Jesus S. Sun . "Integral Nuclear Data Validation Using Experimental Spent Nuclear Fuel Compositions". United States. doi:.
title = {Integral Nuclear Data Validation Using Experimental Spent Nuclear Fuel Compositions},
author = {Gauld, Ian C and Williams, Mark L and Michel-Sendis, Franco and Martinez-Gonzalez, Jesus S},
abstractNote = {},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sun Jan 01 00:00:00 EST 2017},
month = {Sun Jan 01 00:00:00 EST 2017}

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  • Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less
  • The availability of measured isotopic assay data to validate computer code predictions of spent fuel compositions applied in burnup-credit criticality calculations is an essential component for bias and uncertainty determination in safety and licensing analyses. In recent years, as many countries move closer to implementing or expanding the use of burnup credit in criticality safety for licensing, there has been growing interest in acquiring additional high-quality assay data. The well-known open sources of assay data are viewed as potentially limiting for validating depletion calculations for burnup credit due to the relatively small number of isotopes measured (primarily actinides with relativelymore » few fission products), sometimes large measurement uncertainties, incomplete documentation, and the limited burnup and enrichment range of the fuel samples. Oak Ridge National Laboratory (ORNL) recently initiated an extensive isotopic validation study that includes most of the public data archived in the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) electronic database, SFCOMPO, and new datasets obtained through participation in commercial experimental programs. To date, ORNL has analyzed approximately 120 different spent fuel samples from pressurized-water reactors that span a wide enrichment and burnup range and represent a broad class of assembly designs. The validation studies, completed using SCALE 5.1, are being used to support a technical basis for expanded implementation of burnup credit for spent fuel storage facilities, and other spent fuel analyses including radiation source term, dose assessment, decay heat, and waste repository safety analyses. This paper summarizes the isotopic assay data selected for this study, presents validation results obtained with SCALE 5.1, and discusses some of the challenges and experience associated with evaluating the results. Preliminary results obtained using SCALE 6 and ENDF/B-VII cross sections libraries are also briefly summarized. Oak Ridge National Laboratory (ORNL) has been performing spent-fuel isotopic validation studies using the depletion analysis methods in the SCALE [1] code system for the past 20 years. These studies involve comparisons of calculated inventories against measured isotopic composition data obtained from destructive radiochemical analysis of commercial spent nuclear fuel samples. The results of these benchmark studies are used to quantify the bias and uncertainties associated with isotopic calculations and ultimately determine appropriate margins for uncertainty that can be applied in safety-related analyses such as burnup credit in criticality calculations, decay heat analysis, and source terms. Previous studies using several versions of SCALE and nuclear data libraries have been published in multiple validation reports [2-6] that evaluate selected experimental data obtained largely from public sources. A study was recently initiated at ORNL with the objectives of updating and expanding the validation calculations using a comprehensive database of experimental isotopic assay data that includes isotopic composition data obtained from both publicly available sources and international commercial programs. As part of the study, an extensive isotopic database of nearly 120 measured spent fuel samples with an expanded range of initial enrichments and burnup values compared to previously analyzed data was reviewed and analyzed. The calculations were performed using two-dimensional (2-D) assembly models and a consistent set of modeling assumptions using the SCALE 5.1 code system and ENDF/B-V 44-group cross section library. As part of the current study, detailed benchmark modeling information and measurement data are being documented in a format that is readily usable for validating depletion and decay codes. The work is being extended to include analysis results using SCALE 6 and the ENDF/B-VII 238-group cross section library. This paper describes the isotopic composition data evaluated in this study and highlights the preliminary findings to date.« less
  • The Generic TRUEX Model (GTM) was recently validated with data from three demonstrations of the TRUEX process. These demonstrations were run by researchers from the Power Reactor and Nuclear Fuel Development Corp. of Japan (PNC) at the Tokai Works. The feed for the PNC runs was the highly active raffinate from reprocessing of spent fuel from fast breeder reactors. The GTM is designed to calculate TRUEX solvent extraction flowsheets based on input of a specific feed and a specific set of process goals and constraints. This model was used to predict concentration profiles for the three flowsheet demonstrations of themore » TRUEX process run by PNC. A 19-stage mixer settler was used for the extraction and scrub sections, and a 16- to 19-stage unit for stripping. Stagewise data were collected on the behavior of nitric acid and several fission products and actinides during these runs. Results of the GTM calculations showed good agreement with the actual data published by PNC researchers. Differences between model predictions and experimental data are discussed in terms of the process chemistry and demonstration conditions.« less
  • New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.