[en] This document was prepared by Pacific Northwest Laboratory (PNL) and is an attempt to analyze and estimate the effects of feed composition variables and reducing agent variables on the expected chemistry of reactions occurring in the cold cap and in the glass melt in the nuclear waste glass Slurry-fed, joule-heated melters as they might affect foaming during the glass-making process. Numerous redox reactions of waste glass components and potential feed additives, and the effects of other feed variables on these reactions are reviewed with regard to their potential effect on glass foaming. A major emphasis of this report is to examine the potential positive or negative aspects of adjusting feed with formic acid as opposed to other feed modification techniques including but not limited to use of other reducing agents. Feed modification techniques other than the use of reductants that should influence foaming behavior include control of glass melter feed pH through use of nitric acid. They also include partial replacement of sodium salts by lithium salts. This latter action (b) apparently lowers glass viscosity and raises surface tension. This replacement should decrease foaming by decreasing foam stability

[en] Amberlite, Duolite, Ionac A-581 and A-580, and Dowex resins were screened for use in plutonium and neptunium recovery and purification processes. The sorption and elution kinetics and column loading and elution characteristics were investigated

[en] Solubilities of key solid compounds that are either present in the waste form or can readily precipitate from waste package leachates under repository conditions can be used to set maximum limits on radionuclide concentrations expected in ground water. This is because the solubility limited concentrations are independent of the release scenarios, hydrologic transport characteristics, and absorption-desorption reactions. Some of the important factors that control solubilities are pH, pe, type of solid phase, and nature of complexing ligands in the ground waters. Most of the above factors are affected by radiolysis due to the inherent radiation field of the waste form. Experimental results pertaining to the solubilities of selected Am, U, Np, and Pu compounds and the effects of radiolysis are discussed. These results show that: (1) at expected repository pH and reducing conditions, solubility controlled concentrations of several actinides are low and near acceptable limits, (2) the redox conditions at the waste form-water interface may be very oxidizing due to radiolytic effects, despite the fact that normal repository conditions are assumed to be reducing, (3) additional data on solubility limits and key thermodynamic parameters are needed. 26 references, 4 figures

[en] The present invention is a method of obtaining a radionuclide product selected from the group consisting of ²²³Ra and ²²⁵Ac, from a radionuclide ''cow'' of ²²⁷Ac or ²²⁹Th respectively. The method comprises the steps of (a) permitting ingrowth of at least one radionuclide daughter from said radionuclide ''cow'' forming an ingrown mixture; (b) insuring that the ingrown mixture is a nitric acid ingrown mixture; (c) passing the nitric acid ingrown mixture through a first nitrate form ion exchange column which permits separating the ''cow'' from at least one radionuclide daughter; (d) insuring that the at least one radionuclide daughter contains the radionuclide product; (e) passing the at least one radionuclide daughter through a second ion exchange column and separating the at least one radionuclide daughter from the radionuclide product and (f) recycling the at least one radionuclide daughter by adding it to the ''cow''. In one embodiment the radionuclide ''cow'' is the ²²⁷Ac, the at least one daughter radionuclide is a ²²⁷Th and the product radionuclide is the ²²³Ra and the first nitrate form ion exchange column passes the ²²⁷Ac and retains the ²²⁷Th. In another embodiment the radionuclide ''cow'' is the ²²⁹Th, the at least one daughter radionuclide is a ²²⁵Ra and said product radionuclide is the ²²⁵Ac and the ²²⁵Ac and nitrate form ion exchange column retains the ²²⁹Th and passes the ²²⁵Ra/Ac. 8 figs

No abstract available

[en] The present invention is a method of removing an impurity of plutonium, lead or a combination thereof from a mixture of radionuclides that contains the impurity and at least one parent radionuclide. The method has the steps of (a) insuring that the mixture is a hydrochloric acid mixture; (b) oxidizing the acidic mixture and specifically oxidizing the impurity to its highest oxidation state; and (c) passing the oxidized mixture through a chloride form anion exchange column whereupon the oxidized impurity absorbs to the chloride form anion exchange column and the ²²⁹Th or ²²⁷Ac ''cow'' radionuclide passes through the chloride form anion exchange column. The plutonium is removed for the purpose of obtaining other alpha emitting radionuclides in a highly purified form suitable for medical therapy. In addition to plutonium, lead, iron, cobalt, copper, uranium, and other metallic cations that form chloride anionic complexes that may be present in the mixture are removed from the mixture on the chloride form anion exchange column. 8 figs

No abstract available

No abstract available

No abstract available

[en] The oxides of four transuranium isotopes (²³⁸Pu, ²³⁹Pu, ²⁴¹Am, and ²⁴⁴Cm), prepared by identical methods of calcining the oxalate at 750⁰C for two hours, had different physico-chemical properties. For all four oxides the density ranges from 9.8 to 11.4 g cm^-3 and initial ultrafilterability (suspended fraction of activity less than 24 A) varied from 0.002 percent for ²³⁹PuO₂ to 2.24 percent for ²³⁸PuO₂. Dogs were exposed by nose-only techniques to aerosols generated by nebulizing water suspensions of the oxides. The dogs were sacrificed at intervals from one week to about a year postexposure. The rate of translocation of material from lung to other tissues increased from ²³⁹Pu to ²³⁸Pu to ²⁴¹Am to ²⁴⁴Cm, possibly reflecting the decrease in mean particle size from an MMD of 0.7 μm to ²³⁹PuO₂ to 0.6 μm for ²³⁸PuO₂ to 0.4 μm for ²⁴¹AmO₂ to 0.1 μm for ²⁴⁴CmO/sub x/. Accumulation of the isotopes in the liver and skeleton as a percentage of final body burden was 1 percent ²³⁹Pu and 7 to 23 percent for ²³⁸Pu at about a year postexposure, while at 270 days postexposure, values were 40 percent for ²⁴¹Am and 40 to 30 percent for ²⁴⁴Cm