Title: Disposal characteristics of plutonium and americium in a high salt acid waste

We have studied the solubilities of neptunium, plutonium, and americium in J-13 groundwater from Yucca Mountain (Nevada) at three temperatures and hydrogen ion concentrations. They are 25{degree}, 60{degree}C, and 90{degree}C and pH 5.9, 7.0, and 8.5. The results for 25{degree}C are from a study which we did during FY 1984. We included these previous results in the tables to give more information on the solubility temperature dependence; they were, however, done at only one pH (7.0). The solubilities were studied from oversaturation. The nuclides were added at the beginning of each experiment as NpO{sub 2}{sup +}, Pu{sup 4+}, and Am{supmore » 3+}. The neptunium solubility decreased with increasing temperature and with increasing pH. The soluble neptunium did not change oxidation state at steady state. The pentavalent neptunium was increasingly complexed by carbonate with increasing pH. All solids were crystalline and contained carbonate, except the solid formed at 90{degree}C and pH 5.9. We identified this solid as crystalline Np{sub 2}P{sub 5}. The 25{degree}C, pH 7 solid was Na{sub 3}NpO{sub 2}(CO{sub 3}){sub 2} {center_dot} nH{sub 2}O. Plutonium concentrations decreased with increasing temperature and showed no trend with pH. Pu(V) and Pu(VI) were the dominant oxidation states in the supernatant solution; as the amount of Pu(V) increased with pH, Pu(VI) decreed. The steady-state solids were mostly amorphous, although some contained a crystalline component. They contained Pu(IV) polymer and unknown carbonates.« less

This report summarizes the work done to date on the application of the TRUEX solvent extraction process for removing and separately recovering plutonium and americium from a nitric acid waste solution containing these elements, uranium, and a complement of inert metal ions. This simulated waste stream is typical of a raffinate from a tributyl phosphate (TBP)-based solvent extraction process for removing uranium and plutonium from dissolved plutonium-containing metallurgical scrap. The TRUEX process solvent in these experiments was a solution of TBP and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) dissolved in carbon tetrachloride. A flowsheet was designed on the basis of measured batch distributionmore » ratios to reduce the TRU content of the solidified raffinate to less than or equal to 10 nCi/g and was tested in a countercurrent experiment performed in a 14-stage Argonne-model centrifugal contractor. The process solvent was recycled without cleanup. An unexpectedly high evaporative loss of CCl/sub 4/ resulted in concentration of the active extractant, CMPO, to nearly 0.30M in the solvent. Results are consistent with this higher CMPO concentration. The raffinate contained only 2 nCi/g of TRU, but the higher CMPO concentration resulted in reduced effectiveness in the stripping of americium from the solvent. Conditions can be easily adjusted to give high yields and good separation of americium and plutonium. Experimental studies of the hydrolytic and gamma-radiolytic degradation of the TRUEX-CCl/sub 4/ showed that solvent degradation would be (1) minimal for a year of processing this typical feed, which contained no fission products, and (2) could be explained almost entirely by hydrolytic and radiolytic damage to TBP. Even for gross amounts of solvent damage, scrubbing with aqueous sodium carbonate solution restored the original americium extraction and stripping capability of the solvent. 43 refs., 5 figs., 36 tabs.« less

In the case of repositories in salt, adsorption/desorption reactions with salt are expected to be minimal. If solubility-controlling solids of radionuclides are present in the waste or can form in the engineered barrier system, the upper concentration limits of radionuclides that can be leached from the wastes will be solubility-limited but independent of the release scenarios, hydrologic transport characteristics, and adsorption/desorption reactions. The available thermochemical data show that most of the radioactive elements, such as actinides, that are of concern over long repository storage times form solubility-controlling solids. Therefore, it should be possible to set upper limits on the concentrationsmore » of these radioactive elements that can be leached from wastes disposed of in salt repositories. To set upper limits, data are needed for solid phases that form readily, have low solubilities, and either are present in wastes (spent fuel, waste glasses, etc.) or can form readily in the geologic environment. Factors that would lower the maximum concentrations in leachates include an increase in the crystallinity of amorphous precipitates, the presence of crystalline solids in the wastes, and the formation of solid solutions of the actinides. 11 refs.« less

This report provides conceptual waste package designs for use by the Office of Nuclear Waste Isolation (ONWI) in preparing a repository conceptual design in salt. Included are designs for the current reference waste form configurations of Defense High Level Waste, which consists of Savannah River Laboratory wastes immobilized in borosilicate glass, Commercial High Level Waste, which is a borosilicate glass waste form that results from the immobilization of commercial spent fuel reprocessing wastes, and Spent Fuel - Form 2, which consists of consolidated spent fuel rods from PWR or BWR assemblies. Reference designs are presented which are used as amore » baseline to evaluate design alternatives resulting from variations in the waste form configuration, the design approach, and the design data base. This broad spectrum of conceptual designs for salt has been included to provide ONWI a basis for concept comparison, an indication as to which of several package and repository design approaches are more cost effective, and guidance as to which approaches should be pursued in future design efforts. Based on available data and the analyses performed, all the concepts in this report offer technically viable approaches to the containment of the waste form for at least 1000 years and for adequate isolation of radionuclides thereafter. The reference borehole-type design is one which provides containment through the use of a titanium alloy corrosion-resistant overpack, backed with a carbon steel structural reinforcing member. An alternate borehole design is one which provides containment with an all-steel overpack which is sufficiently thick to withstand expected crushing loads and to accommodate expected corrosion. The self-shielded package approach provides containment through use of a thick section of moderately corrosion resistant ferrous alloy material. Development programs are identified that will be required to support designs during licensing.« less

This report details the results of a demonstration for disposal of the legacy solution containing americium and curium through the High Level Waste system.