In pure culture biofilms, 80% of the biofilm weight is occupied by exopoly­saccharides (EPS) (Nelson et al., 1996). In continuous-flow systems, shear forces at the liquid/biofilm surface cause cell detachment and loss of EPS. Viable cells in the biofilm matrix must continuously replace the displaced biofilm materials thereby drawing on the cell’s energy resources. Additionally, cell attachment plays an important role in the cell division cycle inside the biofilm. For example, Meadows (1971) observed that Pseudomonas fluore- scens and Aeromonas liquifaciens cells undergo cell division only during their most stable attachment phase (when lying longitudinal to the solid surface). In more dramatic cases such as the prosthecate bacterium Caulobacter crescentus, cell division occurs only during the attachment phase of the cells life cycle (Neidhardt et al., 1990). In Caulobacter, the prosthecate (stalked) form undergoes cell division giving rise to a swarm cell equipped with a flagellum while the other daughter cell remains attached to the surface with a single prostheca (stalk). The life cycle is completed by differentiation of swarm cells to prosthecates, followed by attachment and cell division.

Scientists and mathematicians have found biofilms too complex to analyze and model especially when working with mixed-culture communities. Biofilm systems in laboratory studies are often oversimplified by using systems with defined chemical and microbial species composition. Attempts to use data from natural systems to develop the models have been unsuc­cessful because of the variability and complexity of the natural aquatic environments. Recently, biofilm studies using microsensors have unravelled the existence of closed-cycles, such as the cycle of sulfate oxidation coupled with sulfate reduction, within biofilms or sediments (Kuhl and Jprgensen, 1992). Models based only on bulk liquid and effluent characteristics often underestimate the overall performance of biofilm systems. The internal metabolic cycles in the biofilms, though hidden from the external observer, determine the final structure of the microbial community thereby affecting the long-term performance of the biofilm (Santegoeds et al., 1998).

Furthermore, bacterial attachment and conglomeration play an impor­tant role in the survival of cells under hostile conditions. For example, attached cells may feed on adsorbed substrates on the surfaces of the immo­bile phase under starvation concentration levels (Zobell, 1943). In toxic environments, biofilm communities may be exposed to lower levels of toxic­ity either due to the masking effect of less susceptible species or due to heterogeneities within the biofilm microenvironment (Chen and Stewart, 1996; Nichols, 1989). Several researchers have proposed mass transport resistance as the main mechanism limiting penetration of toxic substances into biofilms (Xu et al., 1998; Chen and Stewart, 1996; Hodges and Gordon, 1991; Hoyle et al., 1992).

In engineering attached growth bioreactors, the biofilm is allowed to grow thick. This results in a complex dynamic at the surface with a high rate of cell shading and sloughing of the biomass. Inside a thick biofilm, tunnel­ling and honeycombing occurs which makes the prediction of the actual surface area for interfacial diffusion impossible. For this and the above reasons, biofilm models are regarded as mere approximations of the pos­sible operating conditions.

In solutions, all 15 Ln-elements from La to Lu have a common 3+ oxidation state in which they behave chemically in a very similar manner, making their separation very difficult. Only Ce has a strongly oxidizing (but kineti — cally stable) tetravalent oxidation state. Uncommon divalent species of lanthanides (Morss, 1976, Hitchcock, 2008), can be prepared in dilute solu­tion (or solid state) either by gamma irradiation, metallic reduction with alkaline earth metal or electrolysis; only Eu(II) demonstrates appreciable stability.

It is evident that the oxidation states of actinides in solutions are far more variable than those of the lanthanides, particularly in the first half of the series. Multiple oxidation states for the actinide ions are guaranteed by a close proximity of the energy levels of the 7s, 6d, and 5f electrons (Edelstein 2006). The rich chemistry of lighter actinides, from Pa to Am, due to their multiple oxidation states, hydrolytic behavior of their cations and strong coordination of organic ligands, is the most complex and intricate among all elements in the periodic table.

A summary of the oxidation states of actinides is shown in Table 2.2. The bold text indicates the most stable species. The most unstable oxidation states can be produced as transient species in solution by pulse radiolysis (Sullivan et al., 1976a, b, 1982; Gordon et al., 1978). They can also be stabi­lized by structural restrictions imposed in some of their solid coordination compounds (Albrecht-Schmitt, 2008, Edelman, 2006). All actinide divalent species (except for No) are of only transient stability, having only been observed in pulse radiolysis studies; Am(II), Cm(II), and Cf(II) have half­lives of the order of 5-20 ms. While trivalent species are typical for the transplutonium actinides, the lighter actinides are less stable in trivalent oxidation state — in acidic solutions, U(III) is oxidized by water, Np(III) is oxidized by dissolved oxygen in water, Pu(III) is stable, but easily oxidized to Pu(IV) by a variety of mild oxidants. Thorium and protactinium do not

Table 2.2 The oxidation states of the actinide elements. The most stable species are in bold, unstable are in italic, claimed but not important are in () (Edelstein, 2006)

even exhibit the trivalent state in solutions. Stable Th(III) has been reported only in organometalic compounds (Blake et al., 2001).

The tetravalent species also can be considered transient, since a stable 4+ state is observed only for elements from thorium through plutonium and for berkelium. Tetravalent Am in aqueous media can be stabilized by very strong complexing agents like carbonate, phosphate or fluoride. If Am can be stabilized (even as a transient) in its various upper oxidation states, unique options are available for potential Am/Ln group separations (Nash, 2008; 2009). The quantitative oxidation of trivalent americium to the tetra — valent state can be accomplished by strong oxidants using complexants such as lacunary heteropolyanions tungstophosphate and tungstosilicate which are specific for tetravalent species (Donnet et al., 1998). Since tungstosili- cate is a stronger complexing agent towards Am than tungstophosphate, quantitative generation of Am(VI) is possible with both polyanions. Nevertheless, it is easier to obtain Am(IV) with the tungstosilicate and easier to reach Am(VI) with tungstophosphate.

The 5+ oxidation state is well established for the elements protactinium through americium, and the 6+ state in the elements uranium through americium. The 4+ state in curium is confined to a few solid compounds, particularly CmO2 and CmF4, and appears to be present in a stable complex ion that exists in concentrated cesium fluoride solution. The Cf(IV) state is limited to the solid compounds CfO2, CfF4, a complex oxide BaCfO3, and in tungstophosphate solutions; the oxidation of Cf(III) to Cf(IV) in strong carbonate solutions is a disputed topic (Frenkel et al., 1986).

The 2+ oxidation state first appears at americium in a few solid compounds and then at californium in the second half of the series. The II oxidation state becomes increasingly stable in proceeding to nobelium. Md(II) and No(II) have been observed in aqueous solution and this appears to be the most stable oxidation state for nobelium. Am(II) has not only been encoun­tered in solid compounds, but also in electrochemical and pulse radiolysis experiments in acetonitrile solution. The formation of Bk(IV) is associated with enhanced stability of the half-filled 5f configuration (5f7), and the No(II) state reflects the stability of the full 5f shell (5f14). The increase in the stability of the lower oxidation states of the heavier actinide elements relative to the lanthanides may be the result of stronger binding of the 5f (and 6d) electrons in the elements near the heavy end of the actinide series.

Transportable radioactive waste processing facilities can be an attractive alternative to the building of large facilities with fixed canyons or PSCs. Transportable facilities are particularly attractive when processing is expected to be complete within short timescales, when it is advantageous to be able to move the processing equipment within a facility or amongst a series of geographically disperse waste treatment sites, or when the exten­sive decommissioning and clean-up of a large fixed facility is considered undesirable. Phillips (2008) has produced comprehensive review of trans­portable processing systems.

Transportable processing equipment is typically constructed within stan­dard steel ISO-containers, or similar sized spaceframes, and includes por­table ventilation and off-gas equipment added as required. The ISO-containers are shielded and confined as necessary by the addition of steel shielding and confinements to the outside of the containers. Alternatively the ISO-containers are placed within separately constructed concrete enclosures that do not come into contact with any radioactive material and thus do not themselves require any radioactive decommission­ing at mission end.

Within the shielded confinements maximum use is made of systems that greatly minimize the use of moving parts, substituting these with non­moving process equipment and instruments based on fluidics and com­pressed air. Additionally, use is made of “through-wall” drives where moving parts are located outside the shielded ISO-container, with only a shaft drive penetrating the shielding and connecting with the in-container processing equipment. These design provisions enable the placement of most mechani­cal items and instruments requiring maintenance outside the shielding and thus accessible, and ensure that there is no need for personnel to enter the shielded parts of the containers during the life of the system. The Mobile Solidification System (MOSS, Fig. 3.18) is a typical system developed for

use at the Hanford, WA nuclear reservation for the transfer, conditioning and solidification of radioactive sludges from the reservation’s K-Basins, previously used for the underwater storage of UNF.

It is a well-known fact that the organic solvent (both TBP and albeit, to a lesser extent, the diluent) degrades due to the effects of radiolysis, and hydrolysis. These degradation products must be removed prior to solvent recycle, lest they build-up in the organic phase with a concomitant decrease in process performance. Solvent cleanup and treatment, both physical and chemical are an important aspect of PUREX processing. These treatments can include scrubbing or washing the solvent with basic (carbonate and/or hydroxide) solutions to remove acidic degradation products. Carbonate solutions have the ability to form soluble carbonate complexes with residual metal cations in the organic phase, thus precluding any risk of metal cation precipitation during solvent wash. The organic is typically subjected to a subsequent acid rinse operation with dilute HNO3 to remove traces of the basic solution from the organic and slightly re-acidify the organic prior to recycle to the extraction operations. Finally, complementary solvent treat­ment operations involving evaporation and subsequent rectification opera­tions allows recovery of purified TBP solution and diluent. Evaporation makes it possible to remove the heavier, stubborn degradation products such as polymers, while the rectification operation serves to remove the lighter degradation products. These operations are performed on a continu­ous basis with a slip stream on a fraction of the total solvent inventory at the La Hague and Rokkasho plants. The evaporation and rectification cycles are not always performed in continuous mode, but can also be batch type operations preformed on a small fraction of the total solvent inventory. The solvent cleanup operations are carried out as separate unit operations from the various extraction, scrub, and strip cycles.

Liquid-liquid extraction (solvent extraction) is intimately coupled to the PUREX process and has been the workhorse of the nuclear industry for 50+ years. Liquid-liquid extraction plays on the unequal distribution of the components between two immiscible liquid phases, and is therefore highly dependent on the chemistry of those species as well as the extractant molecule(s). Much of this information has been described in the previous discussions for the PUREX process. However, it is important to realize the extension of chemical phenomena is promulgated in the equipment design and operational logistics of a large plant at the industrial scale. While limited mass transfer can be completed in a single, batch equilibrium contact of the two phases, one of the primary advantages of liquid-liquid extraction processes is the ability to operate in a continuous, multistage, countercur­rent flow mode. This allows for very high degree of separation while operat­ing at high processing rates. The aqueous and organic steams flow countercurrently from stage to stage, and the final products are the solvent loaded with the solute(s), and the aqueous raffinate, depleted in solute(s). In this manner, the concentration gradient between the phases remains quite high across all of the stages in the system. This concentration gradient is the motive force for mass transfer and provides the basic phenomena upon which countercurrent solvent extraction is based.

While countercurrent processes could be performed in laboratory glass­ware, their primary advantage is to enable continuous processing at high throughputs. In order to achieve continuous processing, specific equipment is needed that can efficiently mix and separate the two phases in a continu­ous operating mode. In the nuclear industry, specific constraints, such as remote operation and maintenance must be considered, since the solutions processed are highly radioactive. There are three basic types of equipment used in industrial-scale nuclear solvent extraction processes: mixer-settlers, columns, and centrifugal contactors. The basic design and operation of this equipment is well described in the literature (Benedict 1981, Long 1967). It is only noted here that the selection of the type of equipment to be used in large-scale reprocessing hinges on a number of different process param­eters and design considerations, including (but not limited to):

• process foot print and building size/height

• operational flexibility (long-term continuous operation or frequent start/stop operation)

• solvent inventory and in-process volume holdup

• degradation of solvents due to radiolysis/hydrolysis (residence time)

• time required to reach steady-state operation

• potential to operate linked, complex multi-cycle processes

• tolerance to cross-phase entrainment

• tolerance to solids in process solutions

• tolerance to process upsets

• mass transfer kinetics

• process chemistry (e. g. kinetics of valance adjustment)

• remote maintenance capabilities

• criticality constraints.

The equipment type chosen for a particular process application should be based on several factors as indicated in the list above. In-depth reviews and comparisons of pulse columns, mixer-settlers, and centrifugal contactors have culminated with a recent rating of the different equipment designs relative to the criteria aforementioned. Results of a recent review per­formed as part of the United States Department of Energy’s Plutonium Technical Exchange Committee is indicated in Table 6.3 (Todd 1998).

For comparison, the equipment currently used in the world’s major reprocessing facilities is summarized in Table 6.4. Note that the use of pulsed columns and mixer-settlers is far more common than for centrifugal contactors. That trend may change in future applications as additional capacity is brought on-line and the next generation reprocessing plants are designed and built.

In the standard process using air (about 480°C and 4 h processing time), about half of the 14C is volatilized, and minor fractions of fission products are volatilized, including about 5% of krypton and xenon; 1% of iodine and bromine; and ~0.2% of the ruthenium, antimony, and cesium. With higher temperatures and longer reaction times, larger fractions of the noble gases are released — up to 60% of the krypton at 750°C in 8-10 h.10 Using an oxygen-enriched atmosphere enhances the release of 14C to nearly 100%. Higher temperatures, oxidation to UO3, repeated cycles of oxidation — reduction, and mechanical agitation will enhance the release.

T. KOYAMA, Central Research Institute of Electric Power Industry, Japan

Abstract: In this chapter, state of the art in nuclear technology development as well as basic principles of pyrochemical treatment, in other words, dry reprocessing, are described. Although this technology has not been commercialized yet, engineering-scale tests with real spent fuels and/or surrogates are underway in many countries, especially in the US, Russia and Asia. As the technology has an intrinsic nuclear proliferation resistance, due to its inherent difficulty in separation of pure plutonium, it is regarded as one of the most promising nuclear fuel cycle technologies of the next generation.

Key words: pyroprocess, dry reprocessing, electrorefining, molten salt, liquid metal, injection casting.

10.1 Introduction

What do we expect for pyrochemical treatment? As indicated by the prefix, ‘pyro’ (‘fire’ in Greek), we need high temperature, and we must use an unfamiliar liquid medium, a molten salt, in an inert atmosphere. However, pyrochemistry is expected to be the most promising method to solve the difficulties related to the treatment of spent fuel. For example, radiation damage of the solvent is not a concern at all because the molten salt is a fully dissociated ionic liquid. Hence, the treatment of relatively short-cooled spent fuel as well as a decrease in secondary waste due to radiation damage of the solvent are expected. Because water, which acts as a neutron modera­tor, is not present in pyrochemistry, a larger quantity of fissile materials can be handled compared with aqueous processing. In addition, simple but incomplete purification techniques such as electrorefining can be applied because the recycled nuclear fuel will later be used in fast neutron reactors. Hence, pyrochemistry is potentially more compact than aqueous technolo­gies, and a reduction of the fuel cycle cost is expected. The incomplete purification implies the group recovery of actinides in any step of pyro — chemistry, which results in an inherent difficulty in separating weapon- usable plutonium. Although purification is not complete, the recovery of minor actinides (MA) such as neptunium, americium and curium, separating them from fission products (FP), is sufficient to supply the prod­ucts required for fast reactors.

For these reasons, various types of pyrochemical processes using different molten salts have been proposed by many nuclear research institutes. Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) have proposed a pyrochemical treatment, referred to as a ‘pyropro — cess’, suitable for metal fuel processing. Because the process uses chloride molten salts with relatively low melting points, for example, 352 °C for LiCl-KCl, without the evolution of a corrosive gas, engineering difficulties related to the materials are reduced. This might be the reason why this pyrochemical process is the most developed technology and has already been used to treat approximately 4 tons of heavy metals (HM) of spent fuels from the Experimental Breeder Reactor (EBR)-II, and has been applied in many institutes such as the Central Research Institute of Electric Power Industry (CRIEPI) in Japan and the Indira Gandhi Centre for Atomic Research (IGCAR) in India for spent fuel reprocessing, and at the Korea Atomic Energy Research Institute (KAERI) in Korea for spent fuel treatment.

Another pyrochemical process is electrorefining of oxide fuels in NaCl — CsCl molten salt bath, proposed by the Research Institute of Atomic Reactors (RIAR) in Russia. The recovery of plutonium and uranium as oxide forms was demonstrated using irradiated fast breeder reactor (FBR) MOX fuels. As the details of experimental results have not been reported, issues related to nuclear engineering at a higher temperature (650 °C) in the presence of chlorine gas are not apparent. The recovery of MA is another matter to be clarified. The fluoride processes being developed at the Commissariat a l’Energie Atomique (CEA) in France, and other insti­tutes are still at the stage of laboratory-scale experiments and require progress in engineering. The Nuclear Research Institute Rez plc of Czech Republic is developing a fluoride volatility process for the reprocessing of oxide fuels. Fluorination furnaces as well as UF6 condensers have operated well on an engineering scale; however, experiments with irradiated fuel are required to demonstrate the process. Thus, in this chapter, state-of-the-art nuclear engineering for pyrochemistry will be described, mainly focusing on the pyroprocess being developed at ANL, INL, CRIEPI, KAERI and IGCAR.

For newcomers to the study of pyrochemistry, the difficulties lie in the lack of specific technological knowledge on the handling of high tempera­ture molten salts, e. g. purification and control of the atmosphere, the selec­tion of compatible materials, the fabrication of stable reference electrodes, and so on. Although it is not described explicitly, the nuclear engineering discussed in this chapter has been developed on the basis of the technologi­cal knowledge accumulated in each laboratory. Unfortunately, know-how
is rarely described in published papers; thus, literature containing technical details are cited in Section 10.8 for further reference.

It sometimes happens in organic chemistry that already known compounds are re-examined by researchers to meet challenging new demands, or as interest rises in potential applications. This is the case for the calixarenes: cyclic oligomers obtained at the end of the 19th century by condensing formaldehyde on para-substituted (p-tert-butyl, p-octyl, …) phenolic acids to produce Bakelite.

Non functionalized (‘parent’) calix[n]arenes are macrocyclic platforms consisting of p-tert-butyl-phenol units bridged through their ortho positions by methylene spacers (the degree of condensation, [n], usually ranges from 4 to 8). In the solid state, the basket form of the calix[4]arene molecules, resembling a chalice (hence their name: calixarene, in which the arene suffix stems from the aryl rings they are made of), presents two cavities (Fig. 11.2):

11.2 Parent calixinjarenes of chalice shape.

1. A hydrophilic cavity, delimited by the phenol functions: the ‘narrow rim’;

2. A lipophilic cavity, delimited by the p-tert-butyl groups: the ‘wide rim’.

In the late 1980s, Gutsche’s team investigated the condensation reactions of calixarenes and pointed out the importance of the experimental condi­tions (heat, reaction time, nature and concentration of the basic catalyst) on the product yield and composition (Gutsche, 1989, 1998). At the same time, Izatt et al. (1983, 1985) were the first to test ‘parent’ calix[n]arenes for carrying alkali cations through supported liquid membranes. However, since phenolate base is very strong, caesium transportation only occurred from very basic feeds (pH > 12) and not from nitrate feeds. Furthermore, although the caesium flow increases with [n], the selectivity toward caesium surprisingly increases as [n] decreases. Inclusion of a caesium cation in the lipophilic ‘wide rim’ cavity of the ‘parent’ calix[4]arene (probably because of the small cavity size of its ‘narrow rim’) was later demonstrated by X-ray diffraction studies assuming the existence of specific electrostatic interac­tions between the caesium cation and the negative charge delocalized on the aryl rings of the deprotonated ‘parent’ calix[4]arene (Harrowfield et al., 1991).

Unfortunately, ‘parent’ calix[n]arenes appeared totally inappropriate for developing a process for caesium separation from acidic PUREX raffinates or other acidic nuclear waste streams. Nevertheless, this family of macrocy­clic compounds appeared doubly interesting in the European P&T strategy, because:

• the scalable synthesis of ‘parent’ calix[n]arenes was demonstrated (high production yields could be expected);

• the functionalization of calix[n]arenes (either through the hydroxyl groups at the ‘narrow rim’ or through the p-tert-butyl groups at the ‘wide rim’) could lead to a wide range of chemically substituted mole­cules (Vicens and Bohmer, 1991, Ungaro and Pochini, 1991, Creaven et al., 2008).

1,2 — alternate 1,3 -alternate

11.3 The four main conformations of calix[4]arenes.

If the shape of the ‘parent’ calix[4]arene is actually fixed in the solid state, the rotation freedom in solution of its phenol units around its methylene bridges lets the calix[4]arene molecule adopt various conformations (Cornforth et al., 1955). The four main ones were named by Gutsche as: cone, partial cone, 1,2-alternate, and 1,3-alternate, depending on the number of phenol units that have flipped through the mid-plane of the molecule (Fig. 11.3).

This pre-organization feature of calix[4]arenes has definitely been the driving force for the design of caesium selective ligands, because caesium complexes of functionalized calix[4]arenes were expected to be increasingly stable as the changes (induced by the complexation reaction) in the organi­zation of the substrate (calix[4]arene), receptor (caesium), and solvent, were small (Lein and Cram, 1985). It appeared particularly interesting to block specific conformations of the calix[4]arene platforms in order to ori­entate suitable chemical functions (that is to say functions able to bind to caesium cations) in desired directions to fit the alkali cation inner coordina­tion shell.

Ionic liquids (ILs) are salts with low melting points composed of an organic cation and an anion of various forms (Binnemans 2007). The properties of ILs with respect to miscibility with water, solubility of metal salts, polarity, viscosity, etc., can be changed by choice of the anion and the cation. One type of room temperature ILs widely used for current studies is based on the 1-alkyl-3-methylimidazolium cation (bmin) with different forms of anion. With a fluorinated anion, the bmim-based ILs can be water immiscible (hydrophobic). The bmin-based ionic liquid with bis(trifluoromethylsulfonyl)imide anion (Tf2N-) is of particular interest for extraction of metal ions due to their water stability, relative low viscosity, high conductivity, good electrochemical and thermal stability (Mekki et al., 2006). The chemical structure of the ionic liquid [bmin][Tf2N-] is given in Fig. 14.9. Aqueous metal ions are usually not soluble in this type of IL but with the aid of hydrophobic ligands or chelating agents, metal ions may become soluble in the IL phase. Extraction of uranyl ions, trivalent lan­thanides and actinides from nitric acid solutions into the ILs has been investigated using a variety of ligands including TBP, octyl(phenyl)-N, N — diisobutylcarbamoylmethyl phosphine oxide (CMPO), P-diketone, and acidic dialkylorganophosphorous (Binnemans 2007). In general, cation exchange processes have been attributed to the observed metal extraction mechanism for IL systems, in which transfer of positively charged metal ions into the IL phase is accompanied by a simultaneous loss of cations to the aqueous phase (Visser et al. 2003, Jensen et al. 2003). After extraction of metal ions into the ILs, recovering the dissolved metal can be accom­plished by back-extraction with an organic solvent. Using electrochemical methods for recovering metal species in ILs has also been investigated (Rao et al. 2008).

It is known that sc-CO2 dissolves effectively in ILs whereas the solubility of the latter in the former is negligible. Therefore, sc-CO2 provides an effec­tive medium for removing solutes from ILs. Mekki et al. have recently demonstrated that Cu2+ ions and trivalent lanthanides (La3+ and Eu3+) can be extracted from aqueous solutions into an imidazolium-based ionic liquid (1-butyl-3-me thy limidazolium, or bmin) with Tf2N — counter anions using P-diketones as extractants (Mekki et al. 2005, 2006). The metal-P-diketonates in the ionic liquid phase can be effectively transferred to sc-CO2 at 50 °C and 150 atm. The efficiency of extracting lanthanide-P-diketonates from the IL to sc-CO2 is typically greater than 98% under the specified conditions. The reports by Mekki et al. (2005, 2006) suggest that sc-CO2 may be an effective medium for stripping metal species dissolved in an ionic liquid phase without carrying over the ionic liquid to the CO2 phase. Therefore, a two-step extraction process involving three phases (water, ionic liquid, and sc-CO2) may provide an alternative for removing radioactive materials from aqueous solutions to the CO2 phase (Fig. 14.9). The advantages of this new IL-sc-CO2 coupled extraction technique include: (1) radionuclides from the aqueous wastes can be transferred to and concentrated in an ionic liquid under ambient temperature and pressure, and (2) back extraction of the radionuclides from the IL phase to the sc-CO2 phase may be selective because the solvation strength of sc-CO2 is tunable, and (3) no loss of the IL occurs in the back-extraction process and no organic solvent is intro­duced into the ionic liquid phase. Actually, due to its self cleaning nature, sc-CO2 could help removing undesirable organic residues from the IL phase. After the back-extraction step, the IL can be reused and the CO2 recycled after precipitation of the solutes by pressure reduction.

Transport of uranyl ions (UO2)2+ from aqueous nitric acid solutions to [bmin][Tf2N-] using TBP as a complexing agent followed by sc-CO2 strip­ping of the uranyl complex from the ionic liquid phase to the supercritical

[bmin][Tf2N] =

14.9 Extraction of uranium from nitric acid solution using ionic liquid and supercritical CO2 in conjunction — a two-loop extraction system involving three phases.

fluid phase has been reported recently (Wang et al. 2009). The uranyl species involved in this three-phase system were studied using spectroscopic tech­niques. The efficiency of extracting (UO2)2+ in 3 M nitric acid to [bmin] [Tf2N-] with 1.1 M TBP is greater than 95% at room temperature under ambient pressure at an acid to IL phase ratio of 1:1 by volume. The uranyl species extracted into the IL phase showed a UV-Vis absorption spectrum similar to that of UO2(NO3)2(TBP)2 but with some noticeable differences. The uranyl species in the IL phase can be effectively extracted into sc-CO2 phase (>98%) at 40 °C and 150 atm. The rate of supercritical fluid extraction of UO2(NO3)2(TBP)2 from the IL phase appears to follow a zero-order rate equation. The absorption spectrum of the uranyl species extracted into the supercritical fluid phase is similar to that of UO2(NO3)2(TBP)2 known in the literature. The uranyl compound in sc-CO2 was trapped in a hexane solution after depressurizing the system and the absorption spectrum is identical to that of UO2(NO3)2(TBP)2 observed from a standard solution prepared from synthesized UO2(NO3)2(TBP)2 crystals. These studies have provided a basis for developing a new process of extraction and separation of lanthanides and actinides from acidic solutions using a hyphenated IL-sc-CO2 method. Separation of lanthanides and actinides after the sc-CO2 extraction step could be achieved using the counter-current method described in the Areva NP project in the previous section. Separation of uranium from transuranic elements may also be possible using the counter­current method although no experimental results have been reported in the literature yet.

The hyphenated IL-sc-CO2 extraction/separation technique may find applications in reprocessing spent fuel. In a recent report, Billard et al. demonstrated that uranium dioxide can be dissolved directly in [bmin] [Tf2N-] containing nitric acid (Billard et al. 2007). The IL can dissolve about 12,000 ppm of water without forming a separate aqueous phase (Billard and Gaillard, 2009). Our recent study also shows that uranium dioxide and lanthanide oxides can be dissolved directly in [bmin][Tf2N-] containing the TBP(HNO3)18(H2O)06 without forming a second aqueous phase. These new developments suggest the possibility of dissolving spent fuel directly in the IL without an aqueous nitric acid dissolution step. Therefore no aqueous waste would be produced and the dissolution can be done at room tem­perature under ambient pressure. Recovery of uranium from the IL may be achieved using the sc-CO2 back-extraction method or by other tech­niques. For example, using electrochemical techniques for recovering uranium from the IL phase is another option.

Fundamental chemistry of actinides has been reviewed a few times during the past decade. The benchmark book in recent years is The Chemistry of the Actinide and Transactinide Elements, a five-volume work edited by Norman M. Edelstein, Jean Fuger and Lester R. Morss, and published by Springer, 2006. The last, sixth volume was released in 2010. It is a contem­porary and definitive compilation of chemical properties of all of the actinide elements, especially of the technologically important elements uranium and plutonium, as well as the transactinide elements. In addition to the comprehensive treatment of the chemical properties of each element, this book has specialized and definitive chapters on electronic theory, optical and laser fluorescence spectroscopy, X-ray absorption spectroscopy, organoactinide chemistry, thermodynamics, magnetic properties, the metals, coordination chemistry, separations, and trace analysis. Each chapter was written by a team of authors who are recognized experts in their specialty. All chapters represent the current state of research in the chemistry of these elements and related fields. It expounds on topics in actinide science that are undergoing rapid scientific developments and that are germane to the safe development of nuclear energy in the 21st century, from nuclear fuels to the environmental science and management of waste, and this text refers to this book several times, especially to Chapters 15, 23 and 24 (properties, solution chemistry and separations, respectively).

Organometallic and Coordination Chemistry of the Actinides (Springer, 2008), edited by T. E. Albrecht-Schmitt, pages presents critical reviews of the present position and future trends in modern chemical research con­cerned with chemical structure and bonding. It contains short and concise reports, each written by the world’s renowned experts.

The book Recent Advances in Actinide Science edited by R. Alvarez, N. D. Bryan, and I. May (Royal Society of Chemistry, 2006), in six sections (Analysis, the Environment and Biotransformations; Coordination and Organometallic Chemistry; Heavy Elements; Nuclear Fuels, Materials and Waste Forms; Separation and Solution Chemistry; and Spectroscopy and Magnetism) covers more than 200 presentations from leading scientists, presented at the international conference “Actinides 2005” Conference, held at the University of Manchester, UK in July 2005.

Structural Chemistry of Inorganic Actinide Compounds (Elsevier, 2007, edited by Krivorichev, S. V., Burns, P. C. Tananaev, I. G.), is a collection of 13 reviews on structural and coordination chemistry of actinide compounds. Within the last decade, these compounds have attracted considerable atten­tion because of their importance for radioactive waste management, cataly­sis, ion-exchange and absorption applications, etc. Synthetic and natural actinide compounds are also of great environmental concern as they form as a result of alteration of spent nuclear fuel and radioactive waste under Earth surface conditions, during burn-up of nuclear fuel in reactors, repre­sent oxidation products of uranium mines and mine tailings, etc. The actinide compounds are also of considerable interest to material scientists due to the unique electronic properties of actinides that give rise to interesting physical properties controlled by the structural architecture of respective compounds.

Advances in Plutonium Chemistry, 1967-2000 (American Nuclear Society, 2002), is a multi-authored review of advances in plutonium chemistry from 1967 to 2000, documenting the advances in understanding of plutonium chemistry over the more than thirty years since the publication of J. M. Cleveland’s The Chemistry of Plutonium. The book is written by interna­tionally recognized experts in plutonium science, it addresses both the theo­retical interpretations and fundamental properties of plutonium. The detailed technical content is framed together by an impressive and thought­ful summary by the senior editor, D. Hoffman.

The ACS Symposium Series contains high-quality, peer-reviewed books developed from the ACS technical divisions’ symposia. Each volume is a collection of chapters carefully edited by an internationally recognized leader, and chapters are written by experts in the field as invited contribu­tions, usually presented to their peers at the symposia of the ACS annual meetings. The series covers a broad range of chemistry topics. One of the recent volumes, Separations for the Nuclear Fuel Cycle in the 21st Century (Ed. Gregg J. Lumetta, Kenneth L. Nash, Sue B. Clark, and Judah I. Friese), released in 2006 as 933rd volume, is the proceedings from a symposium titled “Separations for the Nuclear Fuel Cycle” in the 21st century which was held in March 2004 and focused on assessing the current state-of-the art in nuclear separations science and technology, and on identifying R&D directions required to enable nuclear separations to meet 21st century demands for waste minimization, environment protection, safety, and security. It was the fifth symposium series devoted to nuclear and radiochemistry. It covers the past 20 years between the previous proceedings, Radioanalytical Methods in Interdisciplinary Research: Fundamentals in Cutting-Edge Applications, volume 868 (released in 2003 and edited by C. A. Laue and K. L. Nash) and Plutonium Chemistry (Vol. 216; 1983; edited by W. T. Carnall and G. R. Choppin), Transplutonium Elements — Production and Recovery (Vol. 161; 1981; edited by J. D. Navratil and W. W. Schulz) and Actinide Separations (vol. 117, 1980, edited by J. D. Navratil and W. W. Schulz).

A series of feed solutions sequentially increasing in nitric acid and then increasing in UO2(NO3)2 concentration were introduced into the “hot” centrifugal contactor system to test the online performance of the spectro­scopic equipment as part of the centrifugal contactor system. At this stage, no contact with TBP/n-dodecane was done. Raman spectra were collected concurrent with the additions of HNO3 and UO2(NO3)2 into the feed. Figure 4.16 shows the accumulated Raman spectra taken over the time frame of the nitric acid and uranyl nitrate additions. Several spectral features are apparent: the water region at 3000-4000 cm-1; the nitrate band at 1050 cm-1; and the UO22+ at 871 cm-1. Collected spectra were subjected to a previously developed chemometric model, and a successful translation of the model based on static UO2(NO3)2/HNO3 measurements to the flow on-line moni­toring was achieved. Figure 4.17 contains the expected and predicted con­centrations of the Raman on-line measurements and shows excellent agreement between values. It is worth noting that the model is capable of not only predicting the UO22+ and nitrate concentrations but is also capable of differentiating between total nitrate and nitric acid. The distinction between nitrate and nitric acid is possible due to the inclusion of all the spectral data within the Raman spectrum, including the water region (3000­4000 cm-1) and multiple nitrate bands (of which 1050 cm-1 is the largest),

nitrate band at 1050 cm-1

0

Time (min)

4.16
On-line Raman monitoring of the composition of the raffinate solution in the centrifugal contactor extraction experiment. Fuel feed simulant solutions: 0-6 M HNO3 and 0-0.6 M UO2(NO3)2. Organic phase: 30% TBP-dodecane solution.

4.17 On-line PLS predictions of nitric acid, total nitrate, and uranyl concentrations based on on-line Raman measurements as a function of experimental time. Measured and predicted Raman on-line measurements showing excellent agreement between values. The light gray lines are the instantaneous concentrations of added analyte in solution (assuming zero mixing time); the dark gray lines are the predicted concentration of HNO3, total nitrate, and UO2(NO3)2 respectively.

which show subtle but reproducible changes based on acid content and the ionic strength of the solution.