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Title: Containment Analysis Capability of UNF-ST&DARDS

Abstract

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodologymore » for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.« less

Authors:
ORCiD logo [1];  [1];  [1];  [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1394125
DOE Contract Number:  
AC05-00OR22725
Resource Type:
Journal Article
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 199; Journal Issue: 3; Journal ID: ISSN 0029-5450
Publisher:
Taylor & Francis - formerly American Nuclear Society (ANS)
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; UNF-ST&DARDS; containment; spent nuclear fuel

Citation Formats

Radulescu, Georgeta, Banerjee, Kaushik, Lefebvre, Robert A., Miller, L. Paul, and Scaglione, John M. Containment Analysis Capability of UNF-ST&DARDS. United States: N. p., 2017. Web. doi:10.1080/00295450.2017.1348800.
Radulescu, Georgeta, Banerjee, Kaushik, Lefebvre, Robert A., Miller, L. Paul, & Scaglione, John M. Containment Analysis Capability of UNF-ST&DARDS. United States. doi:10.1080/00295450.2017.1348800.
Radulescu, Georgeta, Banerjee, Kaushik, Lefebvre, Robert A., Miller, L. Paul, and Scaglione, John M. Fri . "Containment Analysis Capability of UNF-ST&DARDS". United States. doi:10.1080/00295450.2017.1348800.
@article{osti_1394125,
title = {Containment Analysis Capability of UNF-ST&DARDS},
author = {Radulescu, Georgeta and Banerjee, Kaushik and Lefebvre, Robert A. and Miller, L. Paul and Scaglione, John M.},
abstractNote = {The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodology for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.},
doi = {10.1080/00295450.2017.1348800},
journal = {Nuclear Technology},
issn = {0029-5450},
number = 3,
volume = 199,
place = {United States},
year = {2017},
month = {9}
}