rather than an extensive variable (inventory) of the system. A similar insensitivity on repository performance has been identified in analyses of partitioning and transmutation of SNF; reduction in Np-237 inventory of >99.9 percent is needed to cause a reduction in the peak release rate of this solubility-limited nuclide.3 Furthermore, extended containment would not alleviate concerns regarding the chemical reactivity of metallic-uranium SNF in an underground repository.

If a HIC option were pursued, the first preparation step would be to vacuum-dry the SNF to remove free water. Next, the SNF would be packaged in a standard-size canister fabricated from C-22 or possibly some other highly corrosion-resistant material such as titanium (Ti) Grade-12. C-22 (UNSN06022) is a Ni-Cr-Mo alloy, which is highly corrosion resistant due to its high (22%) Cr and high (13%) Mo content. Recent laboratory studies in solutions simulating environments expected in potential underground, high-level nuclear waste repositories have suggested that C-22 and Ti Grade-12 are immune to localized corrosion in these environments.4 However, it is extremely difficult to make reliable long-term predictions concerning the susceptibility to localized corrosion based on the electrochemical polarization technique used in this study which evaluates the probability that pitting will occur. Nevertheless, the design of this system should provide exceptional corrosion resistance since it combines the corrosion allowance of the outer container and its eventual role as sacrificial anode with the exceptional corrosion resistance of the inner container.

As an option, neutron poison inserts could be added within the canister to help mitigate potential nuclear criticality concerns presented by untreated, highly enriched SNF, including EBR-II.

The high-integrity canister would then be placed in dry storage at an appropriate location. Prior to emplacement in a geological repository, the SNF-containing HIC could be placed into an overpack designed to promote containment under repository conditions. Variations of this direct disposal concept include co-disposal of SNF canisters with canisters of reprocessed high-level waste (HLW) borosilicate glass within a single overpack.

After waste emplacement, the encapsulated SNF (e.g., EBR-II) would be isolated from direct contact by water or steam for the lifetime of the overpack and the HIC. This containment period will depend on a variety of environmental factors, such as water infiltration rate, temperature, and future climatological conditions. Certainly a minimum containment time of 1,000 years might be envisioned. Containment times several times longer than this 1,000-year period might be postulated, if substantiated by future “service condition” tests on overpack and HIC materials. Reliable predictions of HIC lifetimes to tens of thousands of years or greater are, however, unsupported by any data known to the committee.


The glass material oxidation and dissolution system (GMODS) was described by the presenter as a general-purpose process for treating miscellaneous fissile materials including EBR-II SNF.6 The basic concept of GMODS is to add unprocessed SNF and a sacrificial oxide to a glass melter at a temperature of 800-1,000 ºC. The solvent is a lead-borate glass (2PbO•B2O3). The oxides in the SNF dissolve in the molten glass. The PbO serves as the oxidant for the SNF metal components, e.g., uranium, plutonium, zirconium, etc., converting them in situ to metal oxides, which dissolve in the glass, and metallic lead. Metals more noble than Cu are not oxidized and remain dissolved in the molten lead. The lead metal


Nuclear Wastes: Technologies for Separation and Transmutation, National Research Council, National Academy Press, Washington, D.C., 1996.


A.K. Roy, D.L. Fleming, and B.Y. Lum, “Localized Corrosion Behavior of Candidate Nuclear Waste Package Container Materials,” p. 54 in Materials Performance, National Association of Corrosion Engineers, Houston, TX, March 1998.


Based in part on information presented to the committee by Charles W. Forsberg (ORNL), March 16, 1998, Washington, D.C.


For a representative reference to the GMODS process, see C. W. Forsberg, E.C. Beahm, G.W. Parker, and K.R. Elam, “Conversion of Radioactive and Hazardous Chemical Wastes into Borosilicate Glass Using the Glass Material Oxidation and Dissolution System,” Waste Management, Vol. 16, 1996, p. 615.

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