Leslie R. Dole

Cementitious Radioactive Waste Hosts Formed Under Elevated Temperatures and Pressures (Fuetap Concretes)

Concretes f ormed u nder e levated t emperatures a nd p ressures (FUETAP concretes) are effective hosts for transuranic (TRU) and high-level defense and commercial wastes. Tailored cement formulations developed at Oak Ridge National Laboratory (ORNL) use Portland cement, fly ash, sand, and clay additives. These FUETAP concretes are cured under mild autoclave conditions, then the unbound water is removed. This paper summarizes the FUETAP development program. These continuous studies address the major questions concerning the performance of radioactive waste forms.

Maintaining Chemically Reducing Waste-Package Conditions

Most fission products and actinides in spent nuclear fuel (SNF) are trapped in the uranium dioxide (UO2) crystal structure and cannot escape until the UO2 is oxidized. Oxidation destroys the crystal structure, exposes the radionuclides to groundwater, and creates the potential for dissolution. For a repository in an oxidizing geochemical environment, the potential exists to delay the release of radionuclides for hundreds of thousands to millions of years by using a combination of excess depleted uranium dioxide and iron in the waste package (WP) to locally maintain chemically reducing conditions. This approach limits the rate of oxygen transport to the SNF after WP failure. Alternative methods of oxygen transport to the SNF were investigated, including transport by groundwater, diffusion through air, and diffusion through water.

Phosphate Ions: Does Exposure Lead to Degradation of Cementitious Materials?

An assessment of the potential effects of phosphate ions on cementitious materials was made through a review of the literature, contacts with concrete research personnel, and conduct of a “bench-scale” laboratory investigation [1]. The objectives of this limited study were to: (1) review the potential for degradation of cementitious materials due to exposure to high concentrations of phosphate ions; (2) provide an improved understanding of any significant factors that may lead to a requirement to establish exposure limits for concrete structures exposed to soils or ground waters containing high levels of phosphate ions; (3) recommend, as appropriate, whether a limitation on phosphate ion concentration in soils or ground water is required to avoid degradation of concrete structures; and (4) provide a “primer” on factors that can affect the durability of concrete materials and structures in nuclear power plants. Results of a literature review, contacts with industry personnel, and a laboratory investigation indicate that no harmful interactions occur between phosphate ions and cememtitious materials unless phosphates are present in form of phosphoric acid. Relative to the “primer,” a separate NUREG report has been prepared that provides a review of pertinent factors that can affect the durability of nuclear power plant reinforced concrete structures.

Cermet Spent Nuclear Fuel Casks and Waste Packages

Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

Depleted Uranium Dioxide Waste Package for Spent Nuclear Fuel

Depleted uranium dioxide (DUO2) waste packages (WPs) for disposal of spent nuclear fuel (SNF) are being investigated to (1) reduce radionuclide releases from WPs, (2) decrease the potential for repository nuclear criticality events, (3) provide radiation shielding, and (4) provide a means to beneficially use excess depleted uranium (DU). The DUO2 is incorporated into the WP as (1) a particulate fill for void spaces within the package and (2) a component of a DUO2steel cermet (DUO2 embedded in steel) that replaces the steel components of the WP. Depending upon the design, there is 3 to 8 times as much DUO2 as SNF UO2 in the WP. Most radionuclides in the SNF cannot be released until the UO2 crystal structure is destroyed. The DUO2 surrounding the SNF slows the degradation of the SNF UO2 in the interior. This behavior is similar to the mechanisms that slow the degradation of natural uranium ore bodies containing UO2. The results of initial investigations and the expected thermodynamic WP behavior are described.

Formulation and durability of tailored cementitious hosts applied to TRU waste generated at the Rocky Flats Plant

This paper discusses the development of cementitious grout formulations at the Oak Ridge National Laboratory which immobilize up to 50% of Rocky Flats Building 374 startup TRU waste. Trial grout mixes were tested in order to minimize the water content and maximize the waste loading. The effects of water reducers and set regulators on these trial mixes were also examined. Durability and physical-property measurements were performed on formulations which are compatible with standard processing equipment. The densities, porosities, and compressive strengths of these solids are reported. Compressive strengths of the samples were found to increase by as much as 69% after the samples had been exposed to leachants for 28 days at 90/sup 0/C.