[en] A method is described for immobilizing waste chloride salts containing radionuclides and hazardous nuclear material for permanent disposal starting with a substantially dry zeolite and sufficient glass to form leach resistant sodalite with occluded radionuclides and hazardous nuclear material. The zeolite and glass are heated to a temperature up to about 1000 K to convert the zeolite to sodalite and thereafter maintained at a pressure and temperature sufficient to form a sodalite product near theoretical density. Pressure is used on the formed sodalite to produce the required density

[en] Preliminary short-term experiments to study the effect of gamma radiation on the waste package materials interactions in the environment of a nuclear waste repository in basalt were completed. Experimental parameters were similar to those expected during the saturated post-closure period of the repository, i.e., a temperature of 200⁰C and a hydrostatic pressure of ≥ 70 bar. The test components included various combinations of synthetic groundwater (GR4), low carbon steel coupons, CH₄, basalt, and basalt/bentonite mixtures (packing). The tests were run in low carbon steel pressure vessels. It was found that gamma radiation at a dose rate of 10⁴ rad/hr for a duration of one and two months increased the H₂ yield in all tests. Increases in the organic carbon yield, the sulfate ion concentration, and the corrosion rate of the coupon were observed in some of the tests. These latter results varied with the combination of waste package components included in the tests. Evidence for the quench effect was observed in tests which included basalt. 12 references, 6 figures, 2 tables

[en] Short-term experiments which studied the effect of gamma radiation on the waste package components of a basaltic repository were completed. Experimental parameters were similar to those expected during the saturated post-closure period of the repository, i.e., a temperature of 200⁰C and a hydrostatic pressure ≥ 70 bar. The waste package components included low carbon steel pressure vessels, synthetic ground water formulation (GR4), CH₄, basalt or packing, and low carbon steel coupons. It was found that gamma radiation at a dose rate of 10⁴ rad/h for a duration of one and two months increased the H₂ yield in all tests. Increases in the organic carbon yield, the sulfate ion concentration, and the corrosion rate of the coupon were observed in some of the tests. These latter results varied with the combination of waste package components included in the tests. Evidence for the quench effect was observed in tests which included basalt

[en] Cementitious waste forms are being considered for immobilizing nuclear waste before disposal. In earlier work, it was found that irradiation of a mortar formulation consisting of slag, portland cement, fly ash, water, and up to 10 wt% KCl endash LiCl salt resulted in the generation of hydrogen. Yields were relatively high and the rates of generation were constant for the irradiation period investigated. The addition of small amounts of oxygen-rich electron scavengers to the mortar was investigated as a means for reducing hydrogen yields. The addition of NaNO₃ reduced the hydrogen yield; changed the radiolytic products from hydrogen to a mixture of hydrogen, nitrogen, and N₂O; and reduced the pressurization rate after exposure to 400 Mrads. The addition of NaIO₄ and KMnO₄ reduced hydrogen yields slightly while the addition of Ag₂O increased the yield. Moreover, the addition of FeS to a non-slag mortar changed the radiolysis mechanism but the addition of FeO did not. The results of these experiments provided an insight into the nature of the radiolytic reactions occurring in the mortar formulations and indicated that the radiolytic generation of gases might be controlled with the proper choice of additive. 14 refs., 3 figs., 2 tabs

[en] A mortar formulation capable of immobilizing chloride salts with high levels of radioactivity is being developed. As part of the developmental effort, radiation effects are being investigated. The radiolytic generation of gas(es) from irradiated mortar formulations was determined for several formulations with variable salt loadings at several test temperatures. The irradiation of a mortar formulation consisting of cement, slag, fly ash, water and 0 to 10 wt % salt led to the generation of hydrogen. The rate of generation was approximately constant, steady state pressures were not attained and final pressures were comparatively high. Higher salt concentrations were correlated with higher hydrogen generation rates for experiments at ambient temperature while lower rates were observed at 120/degree/C. The irradiation of a mortar consisting of cement, fly ash, water and salt led to the radiolytic generation of both oxygen and hydrogen. The addition of 2 wt % FeS or CaS inhibited oxygen generation and changed the hydrogen production rate. 10 refs., 4 figs., 3 tabs

[en] The use of pulse radiolysis to learn about processes which occur before the beginning of chemical times is discussed. Two examples, the distance distribution of positive and negative ions in hydrocarbons, and the state of the dry electron are discussed in detail

[en] Argonne is developing a method to treat spent nuclear fuel in a molten salt electrorefiner. Wastes from this treatment will be converted into metal and mineral forms for geologic disposal. A glass-bonded zeolite is being developed to serve as the mineral waste form that will contain the fission products that accumulate in the electrorefiner salt. Fission products are ion exchanged from the salt into the zeolite A structure. The crystal structure of the zeolite after ion exchange is filled with salt ions. The salt-loaded zeolite A is mixed with glass frit and hot pressed. During hot pressing, the zeolite A may be converted to sodalite which also retains the waste salt. The glass-bonded zeolite is leach resistant. MCC-1 testing has shown that it has a release rate below 1 g/(m²day) for all elements

[en] A ceramic waste form is being developed for waste generated during electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCl-KCl eutectic. The waste form is a composite fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Normalized release rate is less than 1 g/m²d for all elements in MCC-1 leach test run for 28 days in deionized water at 90 C. This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in cationic form of zeolite and in glass composition. Composites were made with 3 forms of zeolite A and 6 glasses. We used 3-day ASTM C1220-92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. Loss of Cs is small (0.1-0.5 wt%), while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses rich in silica and poor in alkaline earth oxides. XRD show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, absence of salt phase corresponds to improved leach resistance. Interactions between zeolite and glass depend on composition of both

[en] The pyrometallurgical processing of spent fuel from the Integral Fast Reactor (IFR) results in a waste of LiCl-KCl-NaCl salt containing approximately 10 wt% fission products, primarily CsCl and SrCl₂. For disposal, this waste must be immobilized in a form that it is leach resistant. A salt-occluded zeolite has been identified as a potential waste form for the salt. Its leach resistance properties were investigated using powdered samples. The results were that strontium was not released and cesium had a low release, 0.056 g/m² for the 56 day leach test. The initial release (within 7 days) of alkali metal cations was rapid and subsequent releases were much smaller. The releases of aluminum and silicon were 0.036 and 0.028 g/m², respectively, and were constant. Neither alkali metal cation hydrolysis nor exchange between cations in the leachate and those in the zeolite was significant. Only sodium release followed t^0.5 kinetics. Selected dissolution of the occluded salt was the primary release process. These results confirm that salt-occluded zeolite has promise as the waste form for IFR pyroprocess salt

[en] Pyrochemical processing of spent fuel from the Integral Fast Reactor (IFR) yields a salt waste of LiCl-KCl that contains approximately 6 wt% fission products, primarily as CsCl and SrCl₂. Past work has shown that zeolite A will preferentially sorb cesium and strontium and will encapsulate the salt waste in a leach-resistant, radiation-resistant aluminosilicate matrix. However, a method is sill needed to convert the salt-occluded zeolite powders into a form suitable for geologic disposal. We are thus investigating a method that forms bonded zeolite by hot pressing a mixture of glass frit and salt-occluded zeolite powders at 990 K (717 degree C) and 28 MPa. The leach resistance of the bonded zeolite was measured in static leach tests run for 28 days in 363 K (90 degree C) deionized water. Normalized release rates of all elements in the bonded zeolite were low, <1 g/m² d. Thus, the bonded zeolite may be a suitable waste form for IFR salt waste