[en] This report was prepared as a summary of a fourfold effort: (1) to examine schemes for defining and categorizing the field of separation science and technology; (2) to review several of the major categories of separation techniques in order to determine the most recent developments and future research needs; (3) to consider selected problems and programs that require advances in separation science and technology as a part of their solution; and (4) to propose suggestions for new directions in separation research at Oak Ridge National Laboratory (ORNL)

[en] Solvent extraction and ion exchange have been the most widely used separation techniques in nuclear and radiochemistry since their development in the 1940s. Many successful separations processes based on these techniques have been used for decades in research laboratories, analytical laboratories, and industrial plants. Thus, it is easy to conclude that most of the fundamental and applied research that is needed in these areas has been done, and that further work in these ''mature'' fields is unlikely to be fruitful. A more careful review, however, reveals that significant problems remain to be solved, and that there is a demand for the development of new reagents, methods, and systems to solve the increasingly complex separations problems in the nuclear field. Specifically, new separation techniques based on developments in membrane technology and biotechnology that have occurred over the last 20 years should find extensive applications in radiochemical separations. Considerable research is needed in such areas as interfacial chemistry, the design and control of highly selective separation agents, critically evaluated data bases and mathematical models, and the fundamental chemistry of dilute solutions if these problems are to be solved and new techniques developed in a systematic way. Nonaqueous separation methods, such as pyrochemical and fluoride volatility processes, have traditionally played a more limited role in nuclear and radiochemistry, but recent developments in the chemistry and engineering of these processes promises to open up new areas of research and application in the future

[en] The solvent extraction of heptavalent technetium from aqueous nitric or hydrochloric acid by tributyl phosphate in n-dodecane (TBP-NDD) has been studied over a wide range of TBP and acid concentrations at 25, 50, and 60⁰C. The extraction was found to proceed according to the reaction 3 TBP + H⁺ + TcO₄^- → (HTCl₄ . 3TBP). A discussion of possible reaction mechanisms is presented, along with values for ΔG, ΔH, ΔS and the equilibrium constant for the extraction reaction. Finally, evidence for the coextraction of technetium by uranyl ions is presented

[en] The literature pertaining to the solvent extraction of heptavalent technetium and rhenium from aqueous solution by tributyl phosphate (TBP) has been compiled, critically evaluated, and supplemented with new data in some areas. The effects of adding mineral acids, alkali metal nitrates, alkali metal chlorides, uranyl nitrate, thorium nitrate, and plutonium(IV) nitrate to these systems were also examined. Discussions of the possible nature of the organic-phase complexes are presented, along with values of ΔG, ΔH, ΔS, and the equilibrium constant for the extraction reaction in several systems. Mathematical models correlating the distribution behavior over a wide range of conditions were also developed. Equations are given for calculating the distribution coefficients for the extraction of Re(VII) or Tc(VII) from 0.2 to 4 M HCl by 0.339 to 2.90 M TBP at 298 to 333⁰K. Equations are given for calculating the distribution coefficients under the same conditions, but with HNO₃ (instead of HCl) present in the aqueous phase. 28 references, 27 figures, 3 tables

[en] This task is concerned primarily with the fundamental chemistry of the actinide and fission product elements. Special efforts are made to develop research programs in collaboration with researchers at universities and in industry who have need of national laboratory facilities. Specific areas currently under investigation include: (1) spectroscopy and photochemistry of actinides in low-temperature matrices; (2) small-angle scattering studies of hydrous actinide and fission product polymers in aqueous and nonaqueous solvents; (3) kinetic and thermodynamic studies of complexation reactions in aqueous and nonaqueous solutions; and (4) the development of inorganic ion exchange materials for actinide and lanthanide separations. Recent results from work in these areas are summarized here

[en] Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc⁺⁷ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase

[en] Several methods for removing fluoride from aqueous nitric acid were investigated and compared with the frequently used aluminum nitrate-calcium nitrate (Ca²⁺-Al³⁺) chemical trap-distillation system. Zirconium oxynitrate solutions were found to be superior in preventing volatilization of fluoride during distillation of the nitric acid, producing decontamination factors (DFs) on the order of 2 x 10³ (vs approx. 500 for the Ca²⁺-Al³⁺ system). Several other metal nitrate systems were tested, but they were less effective. Alumina and zirconia columns proved highly effective in removing HF from HF-HNO₃ vapors distilled through the columns; fluoride DFs on the order of 10⁶ and 10⁴, respectively, were obtained. A silica gel column was very effective in adsorbing HF from HF-HNO₃ solutions, producing a fluoride DF of approx. 10⁴

[en] The high selectivity of tributylphosphate (TBP) for U(VI) and Pu(IV) makes it possible to extract and purify these elements from the complex mixtures that result when irradiated nuclear fuel is dissolved in nitric acid. This selectivity, along with relatively high resistance of TBP to degradation by hydrolysis and radiolysis, combines to make the plutonium-uranium extraction (PUREX) solvent extraction process a remarkably successful method for the remote processing of large amounts of irradiated reactor fuel under harsh chemical conditions and in the presence of high radiation fields. However, while the selectivity of TBP for uranium and plutonium can be very high, it is not absolute; TBP extracts some fission products and transplutonium elements even under the most favorable conditions. Fortunately, the list of fission products that extract enough to be of concern in the PUREX process is relatively short. Ruthenium and zirconium (and zirconium's daughter, niobium) are by far the most troublesome. Technetium is not well-extracted by pure TBP, but coextracts with uranium, plutonium, and other metals to an appreciable extent. Iodine can react with the hydrocarbon diluents used in the process to form organic iodides that degrade process performance. Most of the discussion in this chapter focuses on these elements. The remaining fission products cause little trouble in the solvent extraction portion of the PUREX process, though some cause difficulties in off-gas treatment, waste management, or other parts of the fuel cycle; these latter elements will be only briefly discussed

No abstract available

[en] The solvent extraction of heptavalent technetium from aqueous nitric or hydrochloric acid by tributyl phosphate in n-dodecane (TBP-NDD) has been studied over a wide range of TBP and acid concentrations at 25, 50, and 60⁰C. The extraction was found to proceed according to the reaction 3TBP + H⁺ + TcO₄^- → (HTcO₄ 3TBP). A discussion of possible reaction mechanisms is presented, along with values for ΔG, ΔH, ΔS, and the equilibrium constant for the extraction reaction. Finally, evidence for the coextraction of technetium by uranyl ions is discussed. 10 figures, 5 tables