Argonne National Laboratory
Reduced Enrichment for Research and Test Reactors
Nuclear Science and Engineering Division at Argonne
U.S. Department of Energy

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Abstracts and Available Papers Presented at the
2005 International RERTR Meeting


D.W. Vinson, R.L. Sindelar, and N.C. Iyer
Savannah River National Laboratory - Materials Science & Technology
Aiken, SC 29808 – USA


Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are “failed” or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, LR, for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems.

The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A2, are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate.

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Dr. Jordi Roglans-Ribas
Technical Director, RERTR Department
Nuclear Engineering Division – 362
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2016 RERTR Meeting

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Last modified on July 29, 2008 11:32 +0200