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This study addresses the physical-chemical processes that might naturally or inadvertently occur and that would lead to a separation of the poisoning nonfissionable actinides (232Th, 238U) from the fissionable ones (239Pu, 235U) by selective dissolution and redeposition over a prolonged storage of the waste. Of the various chemistries that were evaluated, carbonate complexation reaction is the most plausible means of achieving the separation of these actinides. Carbonate ions (formed by the dissolution and hydrolysis of atmospheric CO2) can selectively dissolve the actinide oxides through the formation of soluble carbonate complexes, which could result in the separation of poisoning actinides from the fissionable ones. The concentrations of these soluble carbonate species are dependent on the pH, temperature, and other ions; therefore, changes in any of these parameters over time—especially cyclic changes (daily or seasonal)—could cause a selective dissolution and redeposition of the more soluble species ...

Enriched isotopes of sulfur are becoming attractive for use as tracers in a variety of environmental and scientific experiments. Many stages of separation are needed, especially for 33S and 36S, which exist in low natural abundances. Small amounts of highly enriched sulfur isotopes can be provided by means of electromagnetic separation in the calutrons. However, the cost is prohibitively high for enrichment of large amounts needed by many experiments. Thus, a two-stage process is planned to obtain large amounts using gas centrifuges for preenrichment, followed by final enrichment in the calutrons. Sulfur hexafluoride is the most efficient form for processing sulfur in the centrifuges; however, SF6 is unacceptable as feed to the calutrons because it has excellent insulating properties for the high-voltage apparatus used to ionize the feed. An efficient process is needed to convert the preenriched sulfur isotopes from SF6 to a suitable form for calutron feed. Development of a conversion process is ...

The concern that there might be some physical-chemical process which would lead to a separation of the poisoning actinides ({sup 232}Th, {sup 238}U) from the fissionable ones ({sup 239}Pu, {sup 235}U) in waste storage tanks at Oak Ridge National Laboratory has led to a paper study of potential separations processes involving these elements. At the relatively high pH values (>8), the actinides are normally present as precipitated hydroxides. Mechanisms that might then selectively dissolve and reprecipitate the actinides through thermal processes or additions of reagents were addressed. Although redox reactions, pH changes, and complexation reactions were all considered, only the last type was regarded as having any significant probability. Furthermore, only carbonate accumulation, through continual unmonitored air sparging of the tank contents, could credibly account for gross transport and separation of the actinide components. From the large amount of equilibrium data in the literature, concentration differences in Th, U, and Pu due to carbonate complexation as a function of pH have been presented to demonstrate this phenomenon. While the carbonate effect does represent a potential separations process, control of long-term air sparging and solution pH, accompanied by routine determinations of soluble carbonate concentration, should ensure that this separations process does not occur.

The potential hazard posed to future generations by long-lived radionuclides such as the transuranic elements (TRU) is perceived as a major problem associated with the use of nuclear power. TRU wastes have to remain isolated from the environment for ``geological`` periods of time. The costs of building, maintaining, and operating a ``geological TRU repository`` can be very high. Therefore, there are significant economical advantages in segregating the relatively low volume of TRU wastes from other nuclear wastes. The chemical behavior of lanthanides and actinides, 4f and 5f elements respectively, is rather similar. As a consequence, the separation of these two groups is difficult. The ``TALSPEAK`` process (Trivalent Actinide Lanthanide Separations by Phosphorus-reagent Extraction from Aqueous Complexes) is one of the few means available to separate the trivalent actinides from the lanthanides. The method is based on the preferential complexation of the trivalent actinides by an aminopolyacetic acid. Cold experiments showed that by using citric acid the deleterious effects produced by impurities such as zirconium are greatly reduced.

A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

Use of advanced carbide or nitride fuels in place of the reference oxide fuel in LMFBRs may be required in the future to optimize the utilization of this reactor system. Differences in the environmental impact associated with such fuel substitutions are of critical importance in determining the real future viability of various advanced fuels. Calculations indicate that the concentration of /sup 14/C can be significantly different in the various fuels following their residence in the reactor. Most of the /sup 14/C present in the spent fuels is converted to /sup 14/CO/sub 2/ during fuel reprocessing, and /sup 14/C released to the environment will be in this form. Best estimates indicate a release to the environment of no more than 1 percent of the /sup 14/C contained in the spent fuels. Radiation dose commitments due to /sup 14/CO/sub 2/ release from advanced-fuel reprocessing plants were calculated. They range between 1.5 x 10/sup -5/ mrem from nitride fuel containing no /sup 14/N to a maximum of 50 mrem from nitride fuel synthesized with natural nitrogen. Population doses within an 80-km (50-mile) radius of the reprocessing plant were calculated to be 13 man-rems for carbide fuel containing 1000 ppm /sup 14/N and 760 man-rems for nitride fuels synthesized from natural nitrogen.

The experimental studies carried out included (1) the electrolytic production of Ce(IV) from Ce(III), (2) the leaching of refractory PuO/sub 2/ from HEPA filter materials with maintenance of Ce(IV) concentrations by anodic oxidation during leaching, and (3) evaluation of methods for contacting the HEPA solids with the leaching solution and for separating the solid residue from the leaching liquor. Anodic oxidation of Ce(III) was accomplished with an electric current efficiency of about 85% at current densities of 0.04 to 0.4 A/dm/sup 2/ at a platinum anode. Refractory PuO/sub 2/ was dissolved by a 4.0 M HNO/sub 3/ - 0.1 M Ce(IV) solution in 1.5 h at 100/sup 0/C using stirred-contact leaching of the solids or by recirculating the leachant through a packed column of the solids. Cerium (IV) concentrations were maintained continuously by anodic oxidation throughout leaching. Dissolution times up to 10 h were required unless the HEPA media were oxidized initially in air at 300/sup 0/C to destroy carbonaceous species which consumed Ce(IV) more rapidly than it could be regenerated by anodic oxidation. Leaching solids in packed columns avoided the relatively difficult liquid-solids separation by centrifugation which was required after stirred-contact leaching; however, the solids handling difficulties associated withmore » charging and discharging of the packed columns in a remote environment remain a significant design obstacle. A chemical flowsheet is proposed for the recovery of actinides from HEPA filters. A 4 M HNO/sub 3/ - 0.1 M Ce(IV) nitrate solution is used as the leachant and the Ce(III) is recycled to the leaching operation using bidentate solvent extraction.« less