[en] Eh was found to have a large effect on the dissolution of UO² in water at pH 4. As was estimated from thermodynamic data, the solubility was found to decrease as the oxygen fugacity, and therefore the Eh of the water, was decreased. Some of the rare earths and other actinides such as europium, cerium, americium, and plutonium released during the leaching of a spent fuel element behaved differently. These elements were not affected to any large extent by the variation in Eh of these experiments. It has been postulated that these elements reached their solubility limits and precipitated as the spent fuel was leached. 2 figures, 2 tables

[en] Low-geometry x-ray spectra from ₂₃₉Pu and ₂₃₇Np, incorporated into simulated high-level radioactive waste-glass, were transformed into depth distributions for these elements. Changes in the depth profiles were observed for a series of static leachings in 75₀C water. Radiochemical assay of the leach solutions revealed that little neptunium or plutonium was leached, and that the amount leached was independent of leaching time. The depth profiles of the leached specimens showed that there was selective leaching of nonradioactive components of the glass, concentrating the remaining neptunium and plutonium in a broad zone near (but not at) the glass surface. Eventual redeposition of nonradioactive material onto the glass surface inhibited further leaching. Since the depth profiles exhibit uniform concentrations of plutonium and neptunium in undisturbed simulated waste-glass, the depth of attack by the solution can be inferred by departures from the uniform concentrations. A different mechanism was observed in leaching the simulated waste-glass by room temperature 0.03 M NaHCO³. There was no evidence for selective leaching or invasion of the glass by the leaching solution. There was, however, a steady buildup of plutonium (but not neptunium) upon the surface of the glass. 4 figures, 3 tables

[en] A review of crevice and pitting corrosion mechanism and testing techniques is given to understand the fundamental problems involved in determining corrosion rates of alloys to be used as nuclear waste containers. The mechanisms are broken down into two sections, initiation and propagation of the crevice or pit. Theories of initiation include the absorbed ion displacement model, ion migration or penetration model and the breakdown-repair model. Experimental results concerning the initiation period include a discussion of the kinetics of initiation, the critical potential for breakdowns, and experimental factors affecting the initiation. The theories of pit and crevice corrosion propagation are mentioned as well as factors affecting propagation. Several experimental techniques are discussed that are useful in determining the probability of pitting and/or crevice corrosion of alloys. In assessing the lifetime of the metallic container accelerated tests are required. With this in mind the testing techniques concentrate on electrochemical techniques and various modifications of these basic techniques. Although susceptibility of alloys to pitting and crevice corrosion can be determined readily, initiation time and propagation rates are not as straightforward particularly over 1000 years. Nevertheless several testing techniques exist that may be used to determine these important values

[en] Before analytical compositional analysis can be applied to the evaluation of the nuclear waste leachates, the credibility of the sample must be demonstrated. During the leaching in neutral to basic aqueous conditions the adsorption of metal ions onto the walls of the Teflon leach vessels was demonstrated. The adsorption was found to be variable for both zinc and neodymium. The resolubilization of zinc and neodymium using dilute nitric acid proved effective prior to analysis. Certain elemental groups represented by cesium, barium, and strontium demonstrated resistance to container adsorption during leaching under these conditions. The maintenance of a dissolved analytical sample was followed and demonstrated for at least seven weeks following the postleaching procedure

[en] Characteristics of risk assessments for the disposal of wastes containing transuranic elements are discussed. A risk assessment requires the characterization of the waste as well as the facility and site in which the waste is disposed. Risk analyses involve the characterization of environmental pathways through which the contamination can travel to reach the accessible environment. The on-site intrusion pathways are generally limiting because significantly less dispersion and dilution of the waste can occur. The key parameters required for an analysis generally have large uncertainties associated with them, so that the primary usefulness of a risk analysis is to determine the sensitive parameters and obtain relative risks among various designs or alternatives. Results of risk analysis are used to obtain representative concentration limits for transuranic radionuclides

[en] The methods of arc melting and inclusion in concrete were used to consolidate slagging-pyrolysis-incinerator (SPI) slags. The SPI is a wet-bottom, modified vertical-shaft furnace being considered for processing low-level radioactive wastes into granulated slag and metal. In the first method tested, three simulated SPI slag compositions (made from copper reverberatory slag and natural minerals) were melted in a 1 (metric) ton electric arc furnace and poured into 210 liter (55 gal), 1.5 mm thick (16 gage), low-carbon steel drums. Energy and electrode consumption were determined, and slag-refractory brick interaction was measured. The cooling rates for filled drums of slag were determined by an array of thermocouples spotwelded to the drums. Analyses of all raw materials and products, including baghouse dusts, are presented. Melting of granulated slag provided a 50% increase in bulk density. In the second method tested, granulated copper reverberatory slag (which has properties similar to those of granulated SPI slag) and Portland cement were mixed with water and poured into a similar low-carbon drum. The concrete was stronger than ordinary concrete, but the filled drum contained only about 70% as much slag as those filled with molten slag

[en] Probabilistic risk assessments are developed for high level waste, low level waste, and radon emission problems from nuclear power, for chemical carcinogens and radon emissions from the ashes produced in coal burning, from chemical carcinogens released in the eventual disposal of photovoltaic cells, and from the effects of coal burning and in producing and deploying photovoltaic arrays. The cancer risk per quantity ingested or inhaled is taken from standard sources, and the transfer rates from the ground to human ingestion are derived from geochemical analogs. Results are given in deaths per GWe-yr obtained from integrating effects over many millions of years, and only over the next 500 years. It is found that wastes from nuclear power cause thousands of times fewer deaths than do the wastes from coal burning and photovoltaic electricity generation. Mining uranium for nuclear fuel reduces future radon exposures, thereby saving hundreds of times as many lives as will be lost from nuclear wastes

[en] An alternative acid digestion system, H₂SO₄-H₂O₂, to a conventional reaction system, H₂SO₄-HNO₃, has been proposed to reduce the volume of spent ion exchange resins generated at nuclear power plants. A comparative study on both reaction systems has been carried out to obtain the relationship between the reaction conditions and the conversion of the resins, the coprecipitation behavior of ⁶⁰Co and ¹³⁷Cs, and the release rate of the radionuclides from a reaction vessel, and to elucidate the feasibility of the system proposed. The mole ratio of an oxidant and carbon contained in the resins, required to digest the resins, was comparable between both systems, while the H₂SO₄-H₂O₂ system gave higher conversion than the H₂SO₄-HNO₃ system. Some degradation products of an anion exchange resin were relatively stable against the oxidation, and required excessive amounts of the oxidant and a higher reaction temperature to digest, especially in the H₂SO₄-HNO₃ system. The coprecipitation behavior of the radionuclides in both systems was almost identical; ⁶⁰Co coprecipitated effectively with Fe(III) sulfate while ¹³⁷Cs did not coprecipitate. The release rates of the radionuclides were on the order of 10^-3%/hr

[en] Sorption coefficients have been determined using site-specific sediments and trench waters, collected from the Maxey Flats, Kentucky, and West Valley, New York, low-level radioactive waste disposal sites. Experimental apparatus and procedures are described to preserve the anoxic character of the liquid phases during experiments. Experiments using anoxic and oxidized trench waters were performed as functions of solution pH, soil/solution ratio, water and soil composition. The lowest sorption was observed with the combination of anoxic waters and untreated soil - the combination most closely resembling the immediate trench environment. For best results in predictive applications, sorption data should be determined under conditions which simulate those in the field as closely as possible. The total radionuclide retention capacity of reducing geochemical environments is the sum of sorption processes on solid phases, as well as precicipation, and coprecipitation reactions involving iron mineral phases

[en] The purpose of the sodium treatment studies reported here was to evaluate the use of the Inert Carrier Process (ICP) for converting sodium metal to a stable disposal form. The ICP demonstration plant consists of a closed loop of silicone oil that is circulated through a reservoir called a disperser. Solid sodium particles were fed to the disperser and kept suspended in the silicone oil carrier by turbulence. The sodium did not react with the silicone oil carrier. The dispersion of sodium in silicone oil was fed to an in-line mixer (''jet'' mixer) where it was mixed with a reactant. Water was used as the reactant in most of the tests, generating sodium hydroxide and hydrogen as the initial products. Analysis of the final solid product from the reaction indicated that the sodium hydroxide initial product interacted with the silicone oil. Complete reaction of the sodium in the demonstration plant required at least a 6/1 molar ratio of water to sodium. Good separation of the product solution was difficult because of the small difference in density between the aqueous product phase and the organic carrier phase. Emulsification of the silicone oil-aqueous solution was minimized by applying heat to the separator. Foaming of the silicone oil in the separator occurred, aggravated by the evolution of hydrogen from the sodium conversion reaction. Bench-scale tests were conducted to analyze and resolve several problems encountered in the plant experiments, such as incomplete reaction in the jet mixer, poor separation of the product from the silicone oil, formation of an oil aqueous solution emulsion in the separator, and oil foaming in the separator. Solidification tests were carried out to immobilize the sodium conversion product by mixing it with various binders. The most satisfactory binder was EPON 828, an epoxy resin