Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

Johnson, A.B.; Dobbins, J.C.; Zaloudek, F.R.

doi:10.2172/60183

Cited by 3 publications

(2 citation statements)

References 10 publications

(14 reference statements)

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“…The mechanism appears to be thermally-induced oxidation, perhaps assisted by the g»mma radiation field. This is consistent with the observation that hematite was apparently formed from the crud [52]. For the fuel assembly we used, the total release amounted to about 1.8 mCi.…”

Section: Airsupporting

confidence: 89%

Carbon-14 in waste packages for spent fuel in a tuff repository

Konynenburg

Smith

Culham

et al. 1986

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Carbon-14 is produced naturally by cosmic ray neutrons in the upper atmosphere. It is also produced in nuclear reactors, in amounts much smaller than the global inventory. About one-third of this is released directly to the atmosphere, and the other two-thirds remains in the spent fuel. Both the Environmental Protection Agency and the Nuclear Regulatory Commission have established limits on release of the \ in spent fuel. This is of particular concern for the proposed repository in tuff, because of the unsaturated conditions and the consequent possibility of gaseous transport of l *C as COj. existing measurements and calculations of the C inventory in spent.fuel are reviewed. The physical distribution and chemical forms of the C are discussed. Available data on the release of C from spent fuel in aqueous solutions and in gaseous environments of air, nitrogen, and helium are reviewed. Projected '*C behavior in a tuff repository is described. It is concluded that **C release measurements from spent fuel into moist air at temperatures both above and below the in situ boiling point of water as well as detailed transport calculations for the tuff geological environment will be needed to determine whether the 10CFR60 and 40CFR191 requirements can be met.

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Section: Airsupporting

confidence: 89%

Carbon-14 in waste packages for spent fuel in a tuff repository

Konynenburg

Smith

Culham

et al. 1986

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“…The impacts of cladding defects in dry storage are discussed in detail in PNL-4835 ( 25 ) and in PNL-6207. ( 30 ) In general, the impact of a cladding breach that develops during dry storage is anticipated to be minimal: Fission gas would be released to the sealed cask or canister but there would be essentially no release of fuel particles. (25) Evidence suggests that there is no mechanism that will result in growth of cladding defects during storage in inert cover gases.…”

Section: Effect Of Failed or Damaged Fuel Assemblies On Storage Handmentioning

confidence: 99%

Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

Bailey¹

1987

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The results of a study that was jointly sponsored by the U.S. Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to 1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, 2) categorize those fuel assemblies, and 3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies • i i i ACKNOWLEDGMENTS This report summarizes a jointly sponsored study conducted by Pacific Northwest Laboratory (PNL) for the Commercial Spent Fuel Management Program, which is sponsored by the U.S. Department of Energy (DOE) under Contract DE-AC06-76RLO 1830, and for the Electric Power Research Institute (EPRI) under Agreement RP 2062-11. The principal investigator was W. J. Bailey of the Reactor Systems, Fuels, and Materials Department. The author acknowledges the contributions of R. C. Walling (Waste Systems Department) on input for Section 3.5 and Appendix A, D. E. Blahnik (Reactor Systems, Fuels, and Materials Department) on Appendix A, and C. E. Beyer (Reactor Systems, Fuels, and Materials Department) as a peer reviewer. Grateful appreciation is expressed to DOE for guidance and support on the project and to R. W. Lambert of EPRI for continuing guidance on the study and for supporting the wet storage technology program • xi

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