Six alternative TRU waste forms and seven waste immobilization processes are comparatively assessed on the basis of both product properties and process costs and risks. The waste forms are characterized for their leachability, mechanical strength, and thermal and radiation stability. The processes are evaluated in terms of costs (for processing, transportation, and repository disposal) and in terms of occupational exposure, industrial hazard, and quality assurance. Cast cement is recommended for immobilization of defense TRU wastes. A glass system, either borosilicate or aluminosilicate, is recommended for immobilization of commercial TRU wastes.

The safety of low-level radioactive waste disposal sites is analyzed using the computer code LOTRAN. Radiation doses to off-site individuals and to future land reclaimers are computed for a generic disposal site located in a humid Eastern environment. The predicted doses are compared to illustrative performance criteria developed under this project. Results provide insight into those characteristics of low-level waste disposal systems most important to public safety and will allow analyses of tradeoffs among natural and engineered features.

This paper reports on studies of the leachability of Co–60 and Cs–137 from simulated LWR radwastes solidified with thermosetting resin and evaluates the effects of chemical fixation on leachability. It is concluded that insolubilization by a nickel-ferrocyanide compound offers an effective chemical fixation of these radionuclides and is a recommended pretreating method for radwastes that are to be solidified.

The long-term integrity of low-level waste shallow land burial sites is dependent on the interaction of physical, chemical, and biological factors that modify the waste containment system. Past research on low-level waste shallow land burial methods has emphasized physical (i.e., water infiltration, soil erosion) and chemical (radionuclide leaching) processes that can cause waste site failure and subsequent radionuclide transport.

The purpose of this paper is to demonstrate the need to consider biological processes as being potentially important in reducing the integrity of waste burial site cover treatments. Plants and animals not only can transport radionuclides to the ground surface via root systems and soil excavated from the cover profile by animal burrowing activities, but they modify physical and chemical processes within the cover profile by changing the water infiltration rates, soil erosion rates and chemical composition of the soil.

One approach to limiting biological intrusion through the waste cover is to apply a barrier within the profile to limit root and animal penetration with depth. Experiments in the Los Alamos Experimental Engineered Test Facility were initiated to develop and evaluate biological barriers that are effective in minimizing intrusion into waste trenches. The experiments that are described employ four different candidate barrier materials of geologic origin. Experimental variables that will be evaluated, in addition to barrier type, are barrier depth and soil overburden depth. The rate of biological intrusion through the various barrier materials is being evaluated through the use of activatable stable tracers.

Krypton and other gases are readily adsorbed in zeolite 5A molecular sieve or porous Vycor “thirsty glass” and may be trapped there by sintering 2–4 hours at 700° or 900°C, respectively. Encapsulation in these materials is the basis of a method to “immobilize” krypton-85 for storage. Amounts of gas trapped are approximately proportional to gas density, and are about 50 and 40 cc(standard conditions = STP)/cc of solid at 1000 atm, for zeolite 5A and glass, respectively. Based on tests of a few months duration a leakage of 0.3% of the krypton is predicted for 10 years at 400°C for either material.

Demonstration tests were performed on iron-enriched basalt (IEB), a dissolution and immobilization medium for TMI radioactive wastes. Zeolite of the type used for cesium and strontium decontamination of TMI containment water was mixed with 20 wt% additives and melted at 1500°C to form IEB. Cesium volatility from the IEB melts was low. Leaching tests in 90°C deionized water showed leach rates of 6 μg/cm2 d for both cesium and strontium. IEB melts were used to dissolve Type 304 SS pellets and UO2 pellets clad with zircaloy in order to simulate immobilization of TMI core debris. Bubbling air through the melts greatly enhanced the dissolution of these components.

A possible method of treatment for Magnox cladding waste is by dissolution in nitric acid and precipitation of barium sulphate-based floc with which radioactive ions are co-precipitated. The floc could then be immobilised in a matrix material such as cement or bitumen to give the waste form, or alternatively can be converted directly into a waste form by hot pressing.

This paper describes the direct conversion of barium sulphate floc, containing simulated radwaste, into a synthetic, ceramic version of the natural mineral barite by a hot-pressing route. By variation of the parameters pressure, temperature and time, optimum conditions for consolidation of the floc to > 90% theoretical density on a laboratory scale are found to be 22.5 MPa, 900°C for 10 minutes. Using a pressure of 15 MPa, at 900°C for 30 min., hot-pressed billets of BaSO4 have been made on a 5 kg scale. In going from the Magnox waste to the hot-pressed barium sulphate a volume reduction factor ∼ 18 is achieved. The principal phases in the product are found to be BaSO4 , MgO and Fe3O4, and the degree of consolidation achieved depends on the MgO content.

The leaching behaviour of the hot-pressed materials in 100°C, 3 day Soxhlet tests also depends on the MgO content, and on the consequent level of open porosity. If there is porosity accessible to the leach water, MgO at the internal surfaces is converted to Mg(OH)2, which deposits within the pores, and a weight gain is registered in the Soxhlet test. If, however, there is no open porosity, a weight loss occurs, and leach rates ∼ 4 × 10−7 kg/m2/sec are found. In contrast, pure BaSO4, hot-pressed to similar densities, shows no variation in leaching behaviour over a wide range of o en porosities, and gives Soxhlet leach rates ∼ 8 × 10−8 kg/m2/sec.

In this study of subseabed disposal, the two dimensional (axial and radial) migration of radionuclide chains released from a canister located in a sedimentary layer bounded at the top by the ocean and at the bottom by an impermeable basalt zone is analyzed to determine the escape rate of radlonuclides into the seawater. Analytical solutions have been derived to represent the transient concentration profiles within the sediment, flux and discharge rates to the water column of each member present in a decay chain. Using the properties of chain members present in actinide decay systems, the effects of half-life, adsorption equilibrium and other relevant parameters are elucidated.

The hydrofracturing process is a waste disposal process in use at the Oak Ridge National Laboratory for the permanent disposal of locally generated waste solutions. This process is now being modified for use in the disposal of sludge that results from the sodium hydroxide neutralization of acid waste solutions. In this process, the sludges will be slurried in a bentonite clay suspension and mixed with a solids blend of cement and other additives. The amount of dry solids required for each liter of waste slurry will be determined from a rheogram that relates the viscosity of the slurry with the grams per liter recommended for grouts with desirable flow properties.

A description of the process and the development of rheograms are included. Data are presented on the use of chemical additives to control the flow properties of grouts.

Two basic crystalline phase assemblages have been developed for incorporating the full range of Savannah River Plant waste compositions into polyphase ceramic forms. Both phase assemblages provide crystalline host phases, with stable mineral analogues, for all radionuclides in the waste. The first, an alumina based assemblage, immobilizes the radioactive elements in solid solutions of magnetoplumbite and uraninite with the bulk non-radioactive waste elements being present in spinel and nepheline. The second assemblage uses the titanate based “zirconolite” type fluorite structure and the alumina/iron based magnetoplumbite phases to host the radioactive nuclei with spinel and nepheline, again providing crystalline hosts for the non-radioactive elements. Both phase assemblages can be consolidated to a fine grain ceramic by hot isostatic pressing at 1040°C pressures from 20,000 to 30,000 psi. Redox control during processing, just sufficient to reduce uranium to the tetravalent state, is used.

The studies were performed using Ru–103 as a tracer. The volatilization of Ru from ruthenium containing liquids, like HAW-solutions, is about 15%. A chemical denitration of these solutions with HCHO or HCOOH reduces the volatilization to about 2%. Additional catalytic reduction with Fe-powder lowers the volatilization to about 4 ‰. The carry over of Ru seems to depend on the aerosol production rate and not the vapor pressure. Only RuO2 could be detected in the off-gas.

Concretes formed under elevated temperatures and pressures (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.

The melting process of noncombustible solid wastes seems to be one of the most promising processes for volume reduction and immobilization. This melting process needs a high temperature of around 1500° C, but the behavior of radionuclides in such a process and at such a temperature has not been fully examined. We melted small contaminated samples to examine the behavior of radionuclides. Melted products were subjected to leaching tests.

Many assorted samples were melted, and we found that the vapour pressure of the radionuclide determined the distribution between melted products and dusts. The standard free energy of oxide formation and the quantity of stable nuclide of the radionuclide determined the distribution between metal and slag products. The leaching rate of melted products was sufficiently low.

Vitrification is the reference process for the immobilization of radioactive waste from the production of defense materials at the Savannah River Plant (SRP). Since 1979, a small vitrification facility (1 Ib/hr) has been operated at the Savannah River Laboratory using actual SRP waste. In previous studies. dried waste was fed to this smaller melter. This report discusses direct feeding of actual liquid-waste slurries to the small melter. These liquidfeeding tests demonstrated that addition of premelted glass frit to the waste slurry reduces the amount of material volatilized. Results of these tests are in accord with results of large-scale tests with actual waste.

For glass systems related to two potential nuclear-wasteglasses, ABS–39 and ABS–41, the 600°C isotherm of the metastable immiscibility field has been located experimentally. Both glasses are in the one phase glass field down to temperatures below 600°C. ABS–39 and ABS–41 are Na2O-B2O3-SiO2 glasses containing 1.66 wt% UO2 and 9 wt% of simulated fission products. Additional components in ABS–39 are Al2O3 and Fe2O3. In ABS–41 these two oxides are partly substituted with ZnO and Li2O. This substitution causes an extension of the metastable immiscibility field relative to the ABS–39 system.

A reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control.

Several simulated interim waste forms have been investigated in the laboratory to study their suitability for application in handling and transportation of high-level radioactive wastes to terminal processing sites. In the fused-salt/sludge option, the neutralized supernatant liquid and the precipitated sludge are treated simultaneously to form fused-salt cakes. Silicate-based options, in which sodium silicate or sodium silicate and Ca(OH)2 act as binders for the sludge, require prior separation of the sludge and the soluble radioactive constituents from the supernatant before the waste form can be prepared. The results from tests on simulated fused-salt waste forms indicated that the process simplicity of this option is partially offset by the high water solubility and hygroscopicity of the product, which would necessitate special precautions during transportation and storage. The most promising silicate-based option is the ambienttemperature silicate sludge process, in which the sludge is mixed with sodium silicate [and sometimes with Ca(OH)2] and subsequently exposed to a contrelled-humidity environment at room temperature to form a chemical bond. Solid material containing 75 wt % synthetic calcined sludge, prepared by this process, has sufficient physical, chemical, and mechanical stability for use as an interim waste form.

Novel organic resins that achieve high selectivities for both cesium and strontium have been synthesized. They are condensation polymers of resorcinol and formaldehyde withattached chelating groups. Their column performance compares favorably with that of commercially available resins for either cesium or strontium removal. By combining Cs+ and Sr2+ removal in the same bed, these resins can significantly reduce the size and complexity of proposed facilities for processing defense high-level waste.

Nuclear reaction analysis, Rutherford backscattering and ion implantation techniques are applied to the study of the reaction between glass and aqueous solutions. The use of these techniques is described as are some of the results obtained. Solutions of NaCl are seen to etch some glasses several orders of magnitude faster than distilled water. Radiation damage created by high energy heavy ion bombardment is seen to have great effects on the penetration of hydrogen (water) into glass.

The ion implantation-based simulation of α-recoil aging in radwaste glasses has been applied to several simulated HLW glasses. The results are qualitatively described by a new model of leaching for the implanted glasses although several specific features of irradiated surfaces are not clearly understood. This model suggests that for the most likely case of diffusion-controlled corrosion, where the reaction scheme involves several steps (hydration, dissolution of hydrated silica gel, etc…), the kinetics-controlling one would not necessarily be radiation sensitive. Indeed for these more representative glasses, no evidence for a drastic increase in the dissolution rate of the silica gel, induced by ion implantation has been so far detected, whereas the hydration rate seems in some cases more significantly enhanced. However the implications of these observations for the durability of the glasses remain to be assessed.