A voloxidizer and electroreduction cell are pieces of equipment specific to spent oxide fuels treatment. Voloxidation is a well-known technique; however, most experiments have been carried out at the laboratory scale. KAERI is developing engineering-scale voloxidizers for decladding and pulverization of oxide fuel. A voloxidizer of 20 kg HM/batch was designed to oxidize fuels on a vibrating mesh where powdered products are sepa­rated from unreacted chunks of fuel. The effectiveness of the vibration on the oxidation rate was reported for the tests with 20 kg unirradiated UO2 (Kim, 2005). Figure 10.24 shows a new model of the voloxidizer, equipped with a rotating mesh basket in the furnace. The chopped spent fuels are crushed by impact from a ceramic ball, and small pieces of pellet are pulver­ized by oxidation. The fuel powder is collected in the collector located under the mesh basket. In this voloxidizer, simulated fuels (90% W and 10% SiO2) equivalent to 20 kg spent fuel were tested. Oxidation was completed in 13 hrs, and the obtained recovery ratios for hull and powders were 100% and 97.6%, respectively (Kim, 2009).

T. J. TRANTER, Idaho National Laboratory, USA

Abstract: Solid-phase extraction has become a subject of increased interest for separation applications specific to radioisotopes of the nuclear fuel cycle. The objective of this chapter is to discuss recent advances in the technology with a focus on the separations of minor actinides and lanthanides from streams associated with nuclear fuel reprocessing. The discussion covers various techniques for making solid-phase extraction resins, recent applications, separation flow sheets, column modeling and the potential advantages and disadvantages of the technology.

Key words: solid-phase extraction, extraction chromatography, solvent — impregnated resins, actinide separation, lanthanide separation, nuclear fuel reprocessing, radioactive waste treatment.

13.1 Introduction

The term solid-phase extraction, as used in the context of this chapter, generally refers to macroporous polymers that hold an organic complexing compound or extractant within the pore structure of the polymer. Various literature sources may also refer to this material as extraction chromatog­raphy resins, solvent-loaded resins or solvent-impregnated resins. The mac­roporous polymer supports are roughly spherical beads, appearing much like the resins typical to ion exchange, and may be used in a similar fashion in a column or packed bed configuration. It is often said that solid-phase extraction resins combine the metal selectivity of liquid-liquid solvent extraction with the operational benefits of packed bed ion exchange (Warshawsky, 1981, Cortina et al., 1997). Some of the seminal work in syn­thesizing the first variants of this material was performed by Small (1961), Fritz et al., (1971) and Spevackova et al., (1970) wherein organic polymers were used as solvent supports for analytical separations. Further pioneering work was done by Warshawsky (1974, 1978) and Grinstead et al., (1974) who investigated solvent-impregnated resins for use in hydrometallurgical and effluent treatment applications.

Solid-phase extraction has become an accepted separations technology in pharmaceutical and analytical organic chemistry applications. Over the last twenty years it has also gained wide acceptance in radioanalytical

methods for effecting very clean separations of various radionuclides, espe­cially lanthanides and actinides. The technology of solid-phase extraction was, however, in its infancy when much of the industrial-scale separation schemes were developed for reprocessing nuclear fuel. Thus, it has not been implemented at large scales for this purpose to any significant degree. Nonetheless, liquid solvent extraction techniques for partitioning the trans­actinide (An) and lanthanide (Ln) elements are fairly well understood and it follows that solid supported extractants may offer beneficial improve­ments for select portions of fuel reprocessing flow sheets. The technology has been steadily improved and has reached a level of maturity for analyti­cal and smaller-scale applications. It is therefore the objective of this chapter to discuss the applications and potential merits of solid-phase extraction technology for separating the common actinide and lanthanide isotopes specific to nuclear fuel reprocessing.

Biosorption of Sr2+ was observed under equilibrium conditions in 2 L bench scale anaerobic bioreactors (2 L). The initial Sr2+ concentration in the

experiments was varied between 75 and 1000 mg/L, while the sulfate reduc­ing bacteria (SRB) cell density was kept constant at 1 mg/L. The suspen­sions were agitated for 3 hours and then samples were withdrawn for residual Sr2+ concentration analysis.

Sr2+ concentration was measured in the medium, and a desorption process was conducted to determine the amount sorbed on cells. Based on the amounts in the medium and the amount recovered from cells, it was observed that adsorption on the cells followed the Langmuir model. This suggests that Sr2+ removal occurred until equilibrium was reached, as opposed to precipitation reactions where the data cannot be fitted within a simple Langmuir model.

Straightforward precipitation follows a multi-layer model including surface precipitates. The SRB cells were then demonstrated conclusively to adsorb Sr2+ sorbent. The sorption capacity of the cells (qmax) was relatively high (444 mg/g), a value much higher than values obtained from other cell types (Shaukat et al., 2005; Dabbagh et al., 2007, Chegrouche et al., 2009) and purified cultures of sulfate reducing bacteria (Vijayaraghavan and Yun, 2008).

The solution to the classical Langmuir model is shown in Equation 15.6 and the Freundlich isotherm in Equation 15.7 with the optimum values obtained for the sulfate reducing consortium shown in Table 15.4.

where: q = sorption uptake (MM-1), qmax = maximum sorbate uptake (MM-1), b = coefficient related to the affinity between the sorbent and sorbate, Ceq = equilibrium concentration of the sorbate remaining in the solution (ML-3), k = constant corresponding to the binding capacity and n = coefficient related to the affinity between the sorbent and sorbate.

In advanced fuel cycles, it is likely that fast spectrum, high temperature reactors will be utilized to transmute actinide isotopes that are inefficiently transmuted in light water moderated reactors. Many such reactors will utilize metallic fuels and electrometallurgical processing might become a practical alternative to aqueous processing methods like solvent extraction and ion exchange. The solution chemistry of actinides and fission products is significantly different in the molten salt/molten metal media used for such processing.

Dry processing (pyrometallurgy) has historically found use in electrolytic production of some metallic products, notably for refining very electroposi­tive elements and strong reducing agents like alkali metals. Potassium and sodium metals were first prepared in 1807 by using melt electrolysis of respectively potash and soda. [11,12] Today melt electrolysis (alkali chlo­ride) remains the only process for production of metallic sodium or lithium. Molten salt refining is also appropriate for aluminum, which is obtained by

electrolytic decomposition of aluminum oxide in molten cryolite (Hall — Heroult process). [13]

Pyrochemical processing involves dry chemical reactions at high tem­perature where reactions occur in solid, liquid and gas phases. Oxidation — reduction, volatilization (of halide or metal), slagging (melt refining, molten salt extraction, carbide slagging), liquid metal (melt refining, liquid metal extraction, liquation, precipitation), and electrolytic processes are the most common types. Implicit in this chemistry is the requirement of conducting separations operations when the metals or metal salts are in a fluid condi­tion, which typically occurs only at moderate to high temperatures. A sub­category of such media are the so-called room temperature ionic liquids (RTILs) which by convention form molten salts below 100°C. This new category of materials would appear to offer the greatest promise for elec­trometallurgical partitioning, but is still quite new, thus most important details of metal ion coordination chemistry in these media are unknown.

Pyrometallurgy operates in a quite different manner from solvent extrac­tion. Such dry processing offers some advantages, but also suffers limita­tions. The first significant attempts at actinide metal production in molten salts started with the Manhattan Project in the 1940s. [14] Significantly, Kolodney confirmed that uranium and plutonium could be electrodeposited from molten chlorides. [15] The literature on U and Pu electrorefining in molten salts has been reviewed by Willit et al. [16] Basic chemistry and technologies developed in Russia have been described by Bychkov and Skiba. [17] Molten alkali and alkaline-earth chlorides have been most extensively studied for plutonium conversion and separation. Three pro­cesses are in use thorough the world at significant scale-up: (i) DOR process (direct oxide reduction) which consists in PuO2 reduction by calcium in Ca-based chloride salt, (ii) MSE process (molten salt extraction) for 241Am removal from weapon-grade plutonium by MgCl2 in alkali chloride salt, (iii) ER process (electrorefining) for high plutonium purification using molten alkali and/or alkaline earth chloride electrolyte. [18]

3.4.1 Materials selection

The selection of a material for any chemical process plant is a complicated decision involving many parameters such as corrosion resistance, mechani­cal properties, availability and cost. What makes nuclear plant more demand­ing is the presence of radiation, particularly gamma, and of radioactive contamination of surfaces, usually alpha radiation emitting particles, which need to be decontaminated either on a routine basis or at the end of their life. Not only must the materials perform their function for extended periods of time in non-maintainable areas but this performance has to be demon­strated. The application of the most common materials for nuclear process plant is described below.

Thermodynamically, U(IV) is a stronger reducing agent for Pu(IV) than either HAN or Fe(II) and has the additional advantage that it does not introduce any additional metal into the system. However, it is imperative to understand that Pu(III) is an unstable species at the conditions of Pu partitioning, and rapidly reoxidizes to Pu(IV) in the presence of nitrous acid. The oxidation reaction is autocatalytic in that there is no net consump­tion of HNO2 in accord with the proposed reactions (Schultz 1984):

HNO3 + HNO2 ^ N2O4 + H2O 6.6

N2O4 + Pu3+ + H+ ^ Pu4+ HNO2 + NO2 6.7

This is also the case for U(IV), which is oxidized by a similar mechanism. The point is that either of the aforementioned reducing agents, U(IV) and Fe(II), must be supported by a nitrite scavenger to prevent this autocata­lytic reoxidation of Pu(III) to Pu(IV) and thereby inhibit Pu partitioning.

Hydrazine nitrate (N2H5NO3) is commonly added as a nitrite scavenger, preventing Pu(III) oxidation and helps stabilize U(IV). Hydrazine destroys NO2- in accordance with the following reactions:

N2H5+ + NO2- ^ HN3 + 2H2O 6.8

HN3 + H+ + NO2- ^ N2O + N2 + H2O 6.9

Since excess hydrazine is always added in the process, reaction (6.9) can be neglected. Note that HAN is also a nitrite scavenger, albeit less effective than hydrazine, (vide infra, equation 6.11). Consequently, HAN is always used in conjunction with hydrazine in industrial flowsheets. At this point, it is also relevant to mention another nuance of Tc chemistry that becomes prominent in connection with the Pu partitioning operation under discus­sion. Coextracted Tc remaining in the loaded organic and entering the Pu partitioning process complicates the Pu partitioning of U and Pu when N2H4 is used as a nitrite scavenger. Technetium is a catalyst for both hydrazine destruction and the U(IV) oxidation, which results in the concomitant increase in the consumption of N2H4. In modern PUREX operations, it is now well recognized that to avoid the excessive consumption of N2H4 it is necessary to remove coextracted Tc in the solvent prior to U-Pu partition­ing. This is the predominant consideration for including the Tc scrub opera­tion in the aforementioned codecontamination step of the first extraction cycle.

Since U(IV) is extractable by TBP, albeit to a lesser extent than U(VI) (refer to Table 6.1), at least two hydrazine stabilized U(IV) aqueous streams are used in the Pu partitioning contactor. Referring again to Fig. 6.2, a stream containing U(IV) at a moderate ~1 to 2 M HNO3 concentration is fed to the Pu partitioning contactor at a location near the organic feed entry (Sood 1996). The acidity in this stream is useful in salting the U(IV) into the organic phase, thereby facilitating the reduction of Pu(IV). A second U(IV) containing aqueous stream with a lower 0.2 M HNO3 concentration is fed near the organic effluent (Fig. 6.2, Pu scrub contactor), effectively scrubbing Pu from the U laden organic product; the lower acidity also facilitates reduction of Pu (IV) to Pu(III) by U(IV). A fresh organic scrub stream of 20-30% TBP is fed at the opposite end (Fig. 6.2, bottom of the U scrub contactor) to remove residual uranium from the aqueous pluto­nium product stream. This organic phase, containing some uranium and plutonium, is recombined with the organic feed in the stripping operation. Normally, the amount of U(IV) used in the partitioning process is four to six times the stoichiometric requirement (Sood 1996).

Other design criteria built into the Pu partitioning step are worth men­tioning. In addition to purity of the Pu stream, another important consid­eration in the plutonium partitioning cycle is that it be operated in a manner conducive to concentration of Pu in the final aqueous product from this operation. This is one of the benefits of the reductive stripping operation, since Pu concentration in the partitioning step makes it possible to elimi­nate intercycle evaporative concentration operations between the codecon­tamination and the second Pu purification cycles. The partitioning of uranium and plutonium early in the overall process flowsheet has been universally adapted to segregate issues related to treatment capacity from those relating to criticality risk management. For example, the equipment used for the codecontamination and uranium purification cycles requires a large throughput of several tonnes/day, with very high liquid flowrates. For the plutonium purification cycles, the equipment design is more complex to preclude the associated criticality risks; conversely, the fact that smaller quantities of material streams are being handled calls for equipment of greatly reduced size.

The UREX+ separations strategy provides the advantage of a focused optimization approach for the recovery of products that meet pre­determined key objectives for a fuel cycle or a waste disposal strategy. The concept has been successfully tested for the recycle of LWR SNF for a number of different product and waste form configurations. The complexity

of this separations approach increases as the desired number of products or waste forms increases. As this complexity increases so does the cost of the separations facility. The use of various separations modules requires that the integrated process be tested in order to determine any incompat­ibilities between process modules.

To regenerate the spent UNEX-extractant, a technique for the distillation of the diluent (phenyltrifluoromethylsulfone — FS-13) water steam in the presence of sodium carbonate was elaborated. In this way it is possible to distill more than 90% of practically pure diluent (with purification coeffi­cients from the radionuclides — above 1000). During this process, over 40% CCD transfers into the sodium carbonate aqueous solution, from where it can be precipitated in the form of cesium salt suitable for re-use.

Specific experiments have shown that the properties of UNEX-extractant prepared from regenerated FS-13 and CCD do not differ from those of a fresh extractant. By using this regeneration technique, the radionuclides contained in the spent extractant remain almost completely in the still residue along with CMPO, PEG and a portion of CCD. This residue is a viscous liquid which can be removed and then treated with other high-level waste.

The transformation of uranium or mixed uranium/plutonium (U/Pu) oxides through fission and/or neutron capture reactions in pressurized (PWR) or boiling (BWR) water nuclear reactors generates more than 400 radionu­clides (representing 40 elements of the periodic table) with differing physi­cal and chemical properties and making spent nuclear fuels extremely radioactive. The fissile and/or fertile properties of the major actinides, U and Pu, heighten the interest of recycling them (they are moreover the major contributors to the long-term radiotoxicity of spent nuclear fuel). The PUREX process not only separates valuable U and Pu from the waste by taking advantage of the extracting properties of tributyl-phosphate (TBP), it also allows nuclear waste to be conditioned in a safe, inert glass matrix while generating a minimum of secondary technological waste. Spent nuclear fuel reprocessing thus significantly reduces the mass, volume, toxic­ity, and thermal activity of the ultimate nuclear waste to be disposed of in deep geological repositories (Schulz et al., 1990). Furthermore, recycling purified Pu in MOX fuel, which in turn can be reprocessed after fission, makes the option of closing the fuel cycle ever more possible.

Further possible innovative routes have been identified beyond this first but nevertheless essential step towards developing sustainable nuclear energy, preserving natural fossil resources and minimizing the impact on the human environment. Future nuclear fuel cycles will undeniably require recycling of not only U and Pu but of all the valuable and hazardous radio­nuclides, such as long-lived fission products and minor actinides (MA: nep­tunium, Np, americium, Am, and curium, Cm). Although they account for only a thousandth of the mass of the major actinides, the MA remain, after several centuries, the major contributors to the radiotoxicity of spent nuclear fuel (provided Pu has been eliminated). The current P&T research policy is not driven strictly by the problem of waste repository safety since it has been demonstrated that, under certain reductive geological condi­tions, the glass matrix alteration would be very slow and the release of LLRN from the ultimate waste negligible (although variable according to the geological repository conditions). The issue is rather that of limiting the potential radiotoxicity of nuclear waste.

Nevertheless, the potential noxiousness of Am is different from those of Cm and Np, since Am is the major long-term radiotoxic contributor before Cm and Np. After a century of decay of the thermal fission products and Cm isotopes, Am-241 is the largest contributor to the heat release of nuclear waste. On the other hand, the problems encountered when recycling LLRN are not equivalent: recycling Am appears less detrimental than for Cm because of the presence of Cm-244, a radioactive, thermal, and neutron — emitting isotope that worsens the difficulties of fabricating Cm-based fuels or targets.

A few hundred years after nuclear waste disposal, the fission products would contribute to only a negligible fraction of its radiotoxicity. Although computational studies carried out on the chemical behaviour of deep geo­logical repositories revealed that some fission products, such as iodine-129, caesium-135, and technetium-99, could be released to the biosphere after several hundred thousand years, because of their high solubility in water (OECD/NEA, 1999), recent studies of nuclear waste storage scenarios have confirmed the mobility of iodine, but invalidated those of caesium and technetium, especially in clay repositories. Nonetheless, the strategy adopted in France — where two waste management acts were enacted in 1991 and 2006 to organize French research programs on P&T and help prepare the construction of both an industrial workshop for manufacturing MA-based fuels and an actinide burner reactor by 2020 (Warin, 2007) — has focused on the development of separation processes for iodine, caesium, technetium, neptunium, americium, and curium. This policy was further followed in Europe, where EURATOM has funded many collaborative projects since the 1990s (Dozol et al., 1997, Madic and Hudson, 1998, Dozol et al., 2000a, Madic et al., 2000, Dozol et al., 2004, Madic et al., 2004). It still consists in:

• Taking advantage of the potentialities of the PUREX process to separate most of the iodine and technetium and at least 99% of the neptunium by optimizing the operating conditions: increasing the feed acidity actually forces the oxidation of Np(V) into its +VI valence state, which is better extracted by TBP. The remaining Np is expected to be quantitatively recovered in the following Am and Cm recovery step.

• Devising and developing complementary separation steps, making use of highly selective molecules (since TBP is only suitable for extract­ing elements at valence states above +III) to recover caesium and more than 99.9% of americium and curium downstream from the PUREX process.

Calculations actually proved that the time required for the radiotoxicity of a spent nuclear fuel to decrease below that of natural U (extracted from the mine to manufacture the fuel) is estimated to exceed 100 000 years. This period shortens to 10 000 years if Pu is recycled, and shrinks to only three centuries if MA are also recycled (leaving less than 0.1% of the latter radio­nuclides in the nuclear glass waste). However, for physical reasons of trans­mutation efficiency in reactors, the decontamination factors of Am and Cm with regard to the lanthanides (Ln, known as neutron-absorbing elements unfavourable to MA transmutation) were set so as to limit to 5 wt.% the fraction of Ln in the final ‘Am + Cm’ product.

C. M. WAI, University of Idaho, USA

Abstract: Minimizing liquid waste generation in the nuclear fuel cycle is of great importance to the future of nuclear energy. Separation techniques utilizing green solvents, supercritical fluid carbon dioxide and ionic liquids, for dissolution and extraction of uranium dioxide and fission products relevant to nuclear waste management are described in this chapter. An industrial demonstration of the supercritical fluid technology for recovering enriched uranium from the incinerator ash produced by the light water reactor fuel fabrication process by Areva NP in Richland, Washington is a good example of the new trend for treating nuclear wastes. Prospects and advantages of these emerging green techniques for nuclear fuel reprocessing and radioactive waste treatment are discussed.

Key words: green extraction techniques, supercritical fluid, ionic liquid, nuclear waste, spent fuel.

14.1 Introduction

With the threat of global warming, developing environmentally sustainable energy sources to replace traditional fossil fuels is of ultimate importance for the survival of our civilization. Nuclear power is free of carbon emission. Currently, nuclear energy contributes to about 20% of the electricity gener­ated in the USA compared to 80% of that in France. One public concern regarding expanding use of nuclear energy for power generation in many countries including the USA is the economic and environmental issues associated with managing the wastes produced by nuclear power genera­tion. Traditional methods of treating nuclear wastes and reprocessing spent fuel require aqueous acids and organic solvents for dissolution and separa­tion of radioactive elements with the unavoidable consequence of generat­ing large volumes of liquid wastes. Minimizing waste generation in the nuclear fuel cycle is obviously of great importance to the future of nuclear energy. Searching new technologies for managing nuclear wastes in an environmentally sustainable way has become an active research area in recent years. This chapter presents two emerging techniques utilizing envi­ronmentally sustainable solvents, namely supercritical fluid carbon dioxide and room temperature ionic liquids, for separation of actinides, lanthanides and fission products relevant to nuclear waste management and spent fuel reprocessing.

Supercritical fluid carbon dioxide (sc-CO2) and ionic liquids (ILs) are considered green solvents for chemical reactions and separations (Phelps et al. 1996, Welton 1999). Research in sc-CO2 dissolution and extraction of metal species started in the early 1990s (Laintz et al. 1991, 1992). Even in the early stage of the technology development, it was realized that one potential application of this new extraction technique would be in the area of nuclear waste treatment, because supercritical fluid extraction does not require conventional liquid solvents. Two decades later, supercritical fluid extraction technology has emerged as an acceptable green technique for treating nuclear wastes. One example is a recent announcement by AREVA NP to construct a supercritical fluid extraction plant for recovering enriched uranium from the incinerator ash waste produced by the light water nuclear reactor fuel fabrication process (Smith and Thomas 2008). The AREVA NP project currently in progress in Richland, Washington represents the first industrial effort of adopting a non-traditional green technology for nuclear waste management. Some information regarding the AREVA’s supercritical fluid extraction technology will be described later in Section 14.4.

Ionic liquids have unique properties including non-flammable nature, near zero vapor pressure and high solubilities for a variety of compounds (Welton 1999, Dietz and Dzielawa 2001). These properties make ILs attrac­tive for replacing volatile organic solvents traditionally used in various liquid-liquid extractions. In addition, ILs like sc-CO2 show good radiation stability (Visser and Rogers 2003, Dietz and Dzielawa 2001, Mekki et al. 2006) an attractive property for their utilization as media for processing radioactive materials. Both sc-CO2 and IL-based separation techniques are emerging as potential alternatives for replacing the conventional aqueous acids and organic solvent-based techniques for managing nuclear wastes.