Metal-Ion Separation and Preconcentration - American Chemical Society

Chapter 2

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Future of Separation Science in U.S. Department of Energy Processing and Remediation Gregory R. Choppin Department of Chemistry, Florida State University, Tallahassee, FL32306-4390

The Department of Energy has stored large quantities of nuclear wastes from defense activities in many sites around the U.S., and is beginning to process them for final disposition as well as to clean the associated environmental contamination. The magnitude and diversity of the wastes and its radioactive nature present major challenges to separation technologies, and the waste processing and land remediation is expected to require decades and large costs. A number of actinide separation methods, both in use or in the R & D stage, are discussed with emphasis on metal specific separation agents, extractants, natural and biological agents, and some nonaqueous systems. The principles, advantages, and problems of these separation methods are the focus of the review.

Historical Review Separations of radioactive elements began with the discovery of radioactivity. W. Crookes and H. Becquerel found that carbonate anions kept dissolved uranium in solution as a soluble uranyl carbonate complex, allowing purification from insoluble carbonates. The Curies separated components of pitchblende to isolate the radioactive elements polonium and radium. The methods used by these pioneers depended on separation by precipitation, which remained the dominant separation technique in radiochemistry until the Manhattan Project of World War II. Radiochemical separation technology on the kg scale was developed as part of the wartime effort to separate plutonium from irradiated uranium and its fission and decay products. The first such separation process was based on the use of bismuth phosphate precipitation (1) as a carrier for Pu and Pu \ This precipitation technique was replaced by solvent extraction methods in which both uranium and plutonium were isolated from fission products. However, the underlying principle of reduction and oxidation employed in the bismuth phosphate process was retained in the solvent extraction processes. Among the extraction systems developed, the PUREX process, using tributyl phosphate (TBP) as the organic phase, became the most used and remains today the international choice for nuclear fuel reprocessing (2). +

©1999 American Chemical Society

In Metal-Ion Separation and Preconcentration; Bond, A., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1999.

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14 Several nonaqueous processes have been developed for separation purposes in the nuclear fuel cycle. For example, the uranium hexafluoride production processes, based on the volatility of UF , were used on a large scale to produce the feed material for the enrichment of U by gaseous diffusion. Plutonium also forms a volatile hexafluoride, PuF , which allows separation of uranium and plutonium from bulk impurities or fission products by volatilization. Extractions with molten salts and molten metal purification by electrolysis are examples of other nonaqueous processes that have been used for some separation systems. These methods have the advantage compared to aqueous systems of higher radiation resistance and much smaller secondary waste streams. 6 2 3 5


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Present Technology Drivers Research and development in separation technology has been stimulated in recent years by the need to initiate treatment leading to the ultimate disposal of nuclear wastes in a safe and cost-effective manner. While the greater volume of these radioactive wastes in the U. S. has been generated in the nuclear weapons program (defense wastes), the spent fuel from the operation of nuclear power plants which has accumulated has almost filled the temporary storage facilities and must be treated in the near future for permanent disposal. The defense wastes were generated mainly in the reprocessing of irradiated uranium to produce plutonium. The radioactive components are fission products, transuranic elements and their decay products, and nuclides resulting from the neutron activation of structural materials such as cladding. The U. S. has large volumes of these stored defense radioactive wastes as well as contaminated equipment, soils, etc. at a number of sites. The largest volume at a single site is at the Hanford area in the state of Washington where about 2.6 χ 10 liters of waste, as a mixture of liquid, sludge, and solid, are stored in 177 buried tanks. In the storage process sodium hydroxide was added to neutralize the nitric acid (from the original PUREX processing) to minimize corrosion of the mild steel tanks, resulting in precipitation of salts and sludge and considerably increasing the volume and chemical complexity of the wastes. A number of methods have been proposed for separating the intensely radioactive (e.g., Cs, Sr) and longest lived (e.g., the actinides) components from the bulk of the wastes. Many of these technologies are being studied in laboratories while others are ready to be or are being tested in pilot plant operations. In this paper, some of the methods, with emphasis on those for actinide separations, are reviewed and their present state of development assessed. 8

Important Characteristics of Actinides To understand the science of actinide separation chemistry, it is necessary to be familiar with certain characteristic properties of actinide elements. The more important of these are discussed subsequently while more complete discussions can be found in reviews and books (e.g., Chapter 16 of réf. 1). The actinides of Ζ < 96 can exist in oxidation states of III to VI in solution, some in 2 or 3 such states in equilibrium. The change between III and IV and between V and VI is rapid, involving only electron exchange. However, between III, IV and V, VI, the rate is slower as chemical bonds must be broken or made in addition to electron exchange since the An(V) and An(VI) species exist as linear dioxo cations. In all oxidation states, the actinide cations are hard acids and show little interaction with soft base donors such as S, P, etc. in aqueous media. Also, as expected for hard-hard interactions, the bonding to donors is primarily ionic. In Metal-Ion Separation and Preconcentration; Bond, A., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1999.

15 Αη(ΙΠ) and An(IV) cations have coordination numbers (CN) of 6 to 12 while the dioxo actinyls have CN = 4-6. These ranges in CN reflect the ionic bonding which results in the coordination number and symmetry being dominated by steric and net charge effects. The complexation strength normally follows the pattern: 4+

An > An0

2+ 2

3+

> A n > An0

+ 2

2+

which has been attributed to an effective charge on the An in A n 0 of 3.3 ± 0.1 and in A n 0 of 2.2 ± 0.1 (3). In designing specific ligands, it is helpful to be aware of the following: i. An-N bonds are longer-lived than An-0 bonds. ii. An-Ο bonds promote and may be necessary to the formation of An-N bonds. iii. Five-membered chelateringsare the most stable. iv. Prearranged structures more stable. ν. If too rigid, prearranged structures may be too slow to form. vi. Bulky steric groups can slow dissociation kinetics. vii. Redox, steric effects, high CN, and weak bond covalency (in extractant ligands) can be exploited to improve specificity in separations.


2

+

2

Aqueous Processes Aqueous processes have played important roles in the separation of irradiated uranium and are likely, over the near future, to be the principal separation processes for the treatment of nuclear wastes. The separation of radioactive elements in these processes is based on the differences in such chemical properties of the dissolved species as redox potentials, complexation strength with various ligands, affinity to extractants and/or to ion exchange resins or inorganic ion exchangers, transport behavior through membranes or in an electric field, etc. Variations within these processes are necessary, however, to treat successfully wastes of different compositions, pH, etc. Solvent Extraction. In the PUREX process, uranium and plutonium are coextracted (as tetravalent cations) into the organic phase from an aqueous nitric acid solution (2 to 3 M) by TBP while the fission and other (e.g., Am, Np) products remain in the aqueous phase. The plutonium is subsequently separated from the uranium by selectively reducing it to Pu , which is stripped into the aqueous phase. Depending on the operating conditions, neptunium may remain in the uranium streams and can be separated from uranium by adjusting extraction conditions in subsequent steps. This process could be adapted to treat the defense wastes in order to concentrate the actinides into fractions of relatively small volumes for storage or for destruction by fissioning in reactors. However, depending on the composition of the wastes, additional separations (e.g., solvent extraction or ion exchange) might have to be added to the PUREX process to obtain sufficient separation of the actinides from other radioactive species in the wastes. The TRUEX process uses CMPO (octyl(phenyl)-N, N diisobutylcarbamoyl-methylphosphine oxide) to separate the transuranium elements from acidic high level waste (HLW) solutions (4). This process has been demonstrated in the laboratory. Diphenyl dibutylcarbamoyl methyl phosphine oxide in polar inorganic solvents is also a promising extractant (5). Both extractants can be sorbed on a solid porous support and used in a liquid chromatographic elution system. 3+


16 The Talspeak process (6, 7) is based on separation of lanthanides from trivalent actinides by extraction of the former into di(2-ethylhexyl) phosphoric acid (HDEHP) solution from an aqueous phase of lactic + DTPA (diethylenetriaminepentaacetic acid) acids at pH 2.5 - 3.0. Subsequently, 6 M H N 0 is used to strip the Ln(III). A modification - the "reverse" Talspeak process in which the An(III) + Ln(III) elements are extracted into an organic phase by HDEHP then stripped into an aqueous phase of lactate + DTPA has also been proposed (8). Since the Talspeak process operation may be adversely affected by radiation damage, testing at full radiation levels in pilot plant operations is needed for full evaluation of the process. A number of bidentate extractants have been studied. Diphosphine dioxides (9) and diamides (10) show good extraction and radiation resistant properties but, like CMPO, show poor separation of An(III) and Ln(III). In general, they show extraction properties similar to that of CMPO and warrant further evaluation as alternatives to CMPO. The diamides have the advantage that they do not cause formation of phosphate which can be a problem in the processing of the waste for final disposal. Good separation factors have been reported between trivalent lanthanides and trivalent actinides for solvent extraction systems based on complexants with soft donor groups (e.g., N, S). In general, the complexants are based on amide functional groups or on sulfur-based β-diketones (77). These systems show good promise for separations of the trivalent cations. However, redox-based separations can easily separate the actinides of Ζ < 95 more simply and, usually, more efficiently from the Ln(III) fission products. Consequently, there is little incentive to develop soft donor ligand systems for nontrivalent actinide separations. Bis-dicarbollycobaltate ((^(3)-(l,2-C2B H )2Co") or dicarbollide) has been shown to have very high selectivity and efficiency for extraction of Cs(I) and Sr(II) from 2 - 3 M HN0 (72). Efficient extraction of Cs, Sr, Ba, and Αη(ΙΠ) from deacidified PUREX wastes has been achieved using a solution of dicarbollide with p-nonylphenolnonaethyleneoxide with subsequent separation of Ln(ffl) and An(III) species by stripping into nitric acid. Dicarbollides have desirable radiation resistant properties. Crown ethers, cryptands, podands, and similar macrocyclic ligands with cavities of specific size and shape can show high selectivities for atoms or ions which match the cavity size and shape. Studies on these and other stereospecific extractants could be of great value for isolation of radioactive elements such as cesium, strontium, technetium, and plutonium. A crown ether is the extractant used in the SREX process for strontium isolation (13). Specific chelating ligands for various actinides have been developed and tested in the laboratory (14), but have not been used in the treatment of nuclear waste streams. The efficiency of regeneration for recycling can be a problem if the specificity is due to the rigidity and, hence, the inertness of the encapsulating ligand structure. In treatment of nuclear wastes, the role of selective extractants would be in the later parts of the processing sequence where it is desirable to remove and/or isolate a particular element. The potential value of reversible, highly specific, and recyclable extractants offers great opportunities for use in efficient waste treatment and justifies a major research effort; however, selective specificity is not the only criterion to be used in assessing the value.


3

9

11

3

Ion Exchange Materials. A number of interesting new solid ion exchange materials are in various stages of development. One of the more promising is a substituted diphosphonic acid resin, "Diphonix", which is commercially available and relatively inexpensive. This resin is a very strong complexing agent, removes actinides from 10 M H N O 3 solutions, and is effective for In Metal-Ion Separation and Preconcentration; Bond, A., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1999.

17 removing a wide variety of toxic heavy metals (including Pb, Hg, Cd, Zn, Ni, Co, and Cr) from waste water. It has promise for use in removing certain radionuclides from nuclear wastes. Some high temperature zeolites, such as titanium phosphate, may prove useful for isolation of specific cations such as Cs . Silicon titanates have very high capacity for removing Cs and Sr from radioactive solutions of high salt content such as that in the Hanford tanks. The clay sodium fluorophogopite mica is reported (15) to be superior to the zeolite clinoptilolite for removing Sr from nuclear waste and also can remove Cs from solutions of high sodium content. The high capacity and the chemical radiation stability of these inorganic exchangers also make them interesting candidates for study as materials within which to incorporate the concentrated radionuclides for final geologic storage.


+

Nonaqueous Processes Nonaqueous processes, based on the difference in such properties as volatility of various compounds or redox thermodynamics of elements in molten-salt media, have been used over many years in uranium isotope separation, electrorefining of plutonium metal, and production of metallic fuel for advanced nuclear reactors. There is interest in conducting research of nonaqueous processes as separation technologies for treatment of nuclear wastes. These processes have the advantage of relatively high insensitivity to radiation effects - in contrast to aqueous processes for which radiolysis can be a serious problem, causing degradation of the organic extractants and changing the aqueous-phase chemistry through the radiolysis of water. Volatility Processes. Uranium hexafluoride has been used for 50 years in the gaseous diffusion process for uranium isotopic enrichment (I). Volatility techniques with fluorides have also been used to purify plutonium in isotope separation plants (16) and were studied for use in fuel processing in the moltensalt reactor project at Oak Ridge National Laboratory. The separation of uranium and plutonium from fission products is limited in these processes by the fact that volatile fluorides are formed by several fission products, and, in particular, by iodine and tellurium. However, iodine and tellurium can be separated from uranium and plutonium by distillation after oxidation. The decontamination of technetium remains a difficult task in the fluoride volatility process because its fluoride diffuses with the UF and PuF streams. Other volatility separation processes may have promise for separating particular elements from certain types of wastes, but they are not as well developed as thefluoridevolatility process. For example, the volatility of ZrCl could be used to remove the zirconium cladding on spent fuel elements. The β-diketone complexes of trivalent actinides are volatile but their possible use in separations needs more research (17). In a proposed scheme for transmutation of technetium to the nonradioactive ruthenium, the product Ru is converted to Ru0 by ozonolysis and separated from the remaining Tc 0 , using the higher volatility of Ru0 (18). These volatility processes have not been studied sufficiently to evaluate them for practical use in full-scale separation systems. 6

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Pyrochemical Processes. Molten salt (19) and molten metal (20) systems are among the pyroprocesses investigated as possible technologies to treat spent fuel. These types of processes could be considered for treatment of nuclear wastes but since the latter are usually in wet, oxic conditions, the pretreatment to prepare them for treatment in the anhydrous, anoxic pyrochemical systems In Metal-Ion Separation and Preconcentration; Bond, A., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1999.

18 would be a major disadvantage to use of such methods. An advantage of these nonaqueous systems is the much reduced (relative to aqueous processes) volume of secondary, low level wastes resulting from the treatment of the high level wastes.


Natural Agents The use of specific microbial siderophores is receiving attention since these reagents could result in important separation and concentration applications in waste treatment and/or land remediation. It may be possible to use siderophores (microbially produced chelating agents) to sequester species such as Pu(IV) from the environment (21). In contact with jimson weed radioactive sludges have plutonium removed via binding to cell walls (22). Moreover, the jimson weed cells sorbed the plutonium when dead and when alive. It has been known for some time that such cells accumulate uranium from waste streams by hydrolytic sorption on the walls. An interesting possibility for use of biological material is the use of a derivative of chitin, in the form of porous beads, to remove heavy metals from ground water. The beads would be collected with a magnetic field and the sorbed heavy metal stripped. These processes are only at the level of laboratory study presently. A disadvantage may be the sensitivity of biological materials to destruction in high radiation fields. Summary Many novel and promising separation agents and methods have been proposed. The new initiative by the USDOE of the Environmental Management Science Program can be expected to continue the progress in this area in developing more efficient, more specific, and more economical separation techniques. It is necessary that the development of new specific ligands and extractants in research laboratories be followed by the R & D in radioactive facilities to demonstrate their applicability in high radiation fields, complex waste mixtures, etc. Literature Cited (1) (2) (3) (4) (5) (6)

(7) (8)

Choppin, G. R.; Rydberg, J.; Liljenzin, J. O. Radiochemistry and Nuclear Chemistry; Butterworth-Heinemann Ltd: Oxford, 1995. Culler, F. L. In Progress in Nuclear Energy, Series III, Process Chemistry; Bruce, F. R.; Fletcher, I. M.; Hyman, H. H.; Katz, J. J., Eds.; McGraw-Hill: New York, 1956; Vol. 1, Ch. 5.2. Choppin, G. R.; Rao, L. F. Radiochim. Acta 1984, 37, 143. Schulz, W. W.; Horwitz, E. P. Sep. Sci. Technol. 1988, 23, 1191. Dzekun, E. G., et al. In Proc. Symp. On Waste Management Post, R. G.; Wacks, M . E., Eds.; Arizona Board of Regents: Tucson, AZ, 1992. Weaver, B.; Kappelmann, F. A. Talspeak: A New Method of Separating Americium and Curium from Lanthanides by Extraction from an Aqueous Solution of Aminopolyacetic Acid Complex with a Monoacidic Phosphate or Phosphonate; Report ORNL-3559, Oak Ridge National Laboratory: Oak Ridge, TN, 1964. Kolarik, E.; Koch, G.; Kuesel, H. H.; Fritsch, J. Separation of Am and Cm from Highly Radioactive Waste Solution, KFK-1533, Karlsruhe Nucl. Res. Center: Germany, 1972. Persson, G. E.; Svantesson, S.; Wingefors, S.; Liljenzin, J. O. Solvent Extr. Ion Exch. 1984, 2, 89. In Metal-Ion Separation and Preconcentration; Bond, A., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1999.

19 (9) (10) (11) (12)


(13) (14) (15) (16)

(17) (18)

(19) (20) (21) (22)

Rosen, A. M.; Nikolotova, Ζ. I. Radiokhimiya 1991, 33, 1. Musikas, C.; Hubert, H.SolventExtr.IonExch.1987, 5, 877. Ensor, D. D.; Jarvinen, G. D.; Smith, B. F. Solvent Extr. Ion Exch. 1988, 6, 439. Esimantoviskii, D.; Romanovskii, V.; Shishkin, N., Dzekun, E. G., Proc. Symp. On Waste Management, Tucson, 1992, Arizona Board of Regents. Horwitz, E. P.; Dietz, M. L.; Fisher, D. E. Solvent Extr. Ion Exch. 1980, 8, 557. Kappel, M. J.; Nitsche, H.; Raymond, Κ. N. Inorg. Chem. 1985, 24, 605. Paulus, W. J.; Komarmeni, S.; Roy, R. Nature 1992, 357, 571. Hyman, H. H.; Vogel, R.; Katz, J. J. In Progress in Nuclear Energy, Series III, Process Chemistry; Bruce, F. R.; Fletcher, I. M.; Hyman, H. H.; Katz, J. J., Eds.; McGraw-Hill: New York, 1956; Vol. 1, Ch. 6.1. Steinberg, M.; Powell, J. R.; Takahashi, H. Nucl. Tech. 1982, 58, 437. Dewey, H. J.; Jarvinen, G. D.; Marsh, S. F.; Marsh, N . C.; Schroeder, N. C.; Smith, B. F.; Villareal, R.; Walker, R. B.; Yarbro, S. L.; Yates, M . A. Status of Development of Actinide Blanket Processing Flowsheets for Accelerator Transmutation of Nuclear Waste, Report LA-UR-93-2944, Los Alamos National Laboratory: Los Alamos, NM, 1993. Steunenberg, R. K.; Pierce, R. D.; Johnson, I. Symp. on Reproc. Nucl. Fuels; CONF-690801; USAEC: Washington, D.C.; Vol. 15. Coops, M . S.; Knighton, J. B.; Mullins, L. J. In Plutonium Chemistry; Carnall, W. T.; Choppin, G. R., Eds.; ACS Symp. Ser. 216, Amer. Chem. Society: Washington, D. C., 1983. Wildung, R. E.; Garland, T. R.; Rogers, J. E. In Environmental Research on Actinide Elements; Pinder, J. E., et al., Eds.; OSTI; U. S. Department of Energy, 1987. Nuclear Wastes: Technologies for Separations and Transmutation; National Research Council; National Academy Press: Washington, D. C., 1995; p 177.


Metal-Ion Separation and Preconcentration - American Chemical Society

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