Hostname: page-component-6bf8c574d5-r4mrb Total loading time: 0 Render date: 2025-03-09T15:35:04.493Z Has data issue: false hasContentIssue false
  • English
  • Français

Comparison of Uranium Release from Spent Fuel and Unirradiated U02in Salt Brine

Published online by Cambridge University Press:  28 February 2011

W. J. Gray
W. J. Gray*
Affiliation:
Pacific Northwest Laboratory, P.O. Box 999, Richland, Washington 99352
Get access

Abstract

Spent fuel and unirradiated U02were previously reported [1] to have exhibited a 100-fold difference in the total amount of uranium released during static leach tests in salt brine. Two additional studies designed to investigate that difference have now been completed. The first study involved leach tests with U02pellets that had been doped with plutonium to simulate the alpha activity in spent fuel. The second study consisted of leach tests on spent fuel specimens that originated in the same reactor assembly as those used in previous tests but handled differently following discharge from the reactor. Results from the recently completed tests show that at least 90% of the reported difference between the dissolution behavior of spent fuel and U02can be attributed to an apparent surface oxidation of the spent fuel specimens after their discharge from the reactor. The effect of alpha radiolysis on the leaching of spent fuel and U02in salt brine in the presence of an air atmosphere was found to be relatively small in these tests.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 84: Symposium L – Scientific Basis for Nuclear Waste Management X , 1986 , 141
Copyright
Copyright © Materials Research Society 1987

Access options

Get access to the full version of this content by using one of the access options below. (Log in options will check for institutional or personal access. Content may require purchase if you do not have access.)

References

1. Barner, J.O., Gray, W.J., McVay, G.L. and Shade, J.W., Interactive Leach Tests of UO2 and Spent Fuel with Waste Package Components in Salt Brine, PNL-4898-SRP (1986), Pacific Northwest Laboratory, Richland, Washington.CrossRefGoogle Scholar
2. U.S. Nuclear Regulatory Commission. Disposal of High-Level Radioactive Wastes in Geologic Repositories--Technical Criteria. 10 CFR Part 60. (Washington, D.C., 1983).Google Scholar
3. Thomas, G.F. and Till, G., Nucl. Chem. Waste Manage. 5, 141 (1984).CrossRefGoogle Scholar
4. Bailey, M.G., Johnson, L.H. and Shoesmith, D.W., Corros. Sci. 25, 233 (1985).CrossRefGoogle Scholar
5. Gray, W.J. and McVay, G.L., FY-1984 Annual Report: Spent Fuel and UO2 Source Term Evaluation Results, PNL-5650 (1986), Pacific Northwest Laboratory, Richland, Washington.Google Scholar
6. Johnson, L.H., Garisto, N.C. and Stroes-Gascoyne, S., in Waste Manage. 85, Vol. 1, High-Level Waste, edited by Post, R.G. (University of Arizona, Tucson, Arizona, 1985) pp. 479–482.Google Scholar
7. Barner, J.O., Characterization of LWR Spent Fuel MCC-Approved Testing Material -- ATM-101, PNL-5109 Rev.1 (1985), Pacific Northwest Laboratory, Richland, Washington.CrossRefGoogle Scholar
8. Steacy, H. R. and Kaiman, S., in Short Course Handbook; Vol. 3: Uranium Deposits, Their Mineralogy and Origin, edited by Kimberley, M.M. (University of Toronto Press, Canada, 1978) pp.107–140.Google Scholar