Polymers are used for valve diaphragms, gaskets and seals, insulation and sheathing for electric cables, lubricants, greases, pneumatic pipes and adhe­sives. Certain processes are carried out in sealed gloveboxes — the transpar­ent windows are usually polycarbonate, acrylic or PVC, sometimes leaded for extra shielding; seals for the windows can be fluoropolymers or specially formulated nitrile or EPDM seals. The “gloves” themselves are made from chlorosulphonated polyethylene.

Much work has been carried out to characterize polymers for use in radiation areas. Of the common materials polythene, polyamides, poly — imides, fluoroelastomers and, to a lesser extent, PEEK (polyetheretherke — tone) have been used in seals and bearings when radiation levels are low enough. For non-active or low active pipework, the commonest form of gasket is based on PTFE, which has now been used for many years with no difficulties. Unfortunately, PTFE is one of the least radiation-resistant poly­mers and cannot be used in high radiation areas. Indeed, there are no polymers available which will survive high radiation levels for protracted periods and they are usually avoided by the use of metal seals or inorganic materials such as graphite and, to a lesser extent, asbestos.

The purpose of the plutonium purification cycle is to complete Pu decon­tamination and to concentrate the plutonium stream prior to the conversion step to the solid oxide. The plutonium entering this step is in the Pu(III) oxidation state, and as a prerequisite to extraction, the plutonium must undergo reoxidation to Pu(IV). This oxidation is typically accomplished by sparging the solution with nitrous vapors (in essence, the chemical reaction in Eq. 6.7), and subsequently with air to remove excess nitrous acid. Subsequent to this reoxidation step, the plutonium extraction and back — extraction operations both result in increased concentration. It should be noted that such concentration of the plutonium stream may be enhanced by providing recycle of part of the production stream to the head end of the cycle.

Flowsheets have been developed using uranous cation, U(IV), as the reductant for Pu partitioning in the purification cycle. However, reductive stripping in the Pu purification cycle is sometimes performed using hydrox — ylamine nitrate (HAN) as the reductant. Hydroxylamine nitrate, although not as effective as U(IV) for Pu reduction, is considered better suited than U(IV) for plutonium cycles involving higher plutonium concentrations. Furthermore, the re-addition of uranium back into the Pu purification cycle seems counter-intuitive since that negates the desired separation obtained in the initial partitioning cycle. Hydroxylamine nitrate (HAN) can readily reduce Pu(IV) to Pu(III) as follows:

4Pu4+ + 2NH3OH+ ^ 4Pu3+ + N2O + H2O + 6H+ 6.10

Furthermore, HAN will also scavenge HNO2 at low acidities, up to ~1 to 1.5 M HNO3:

NH3OHNO3 + HNO2 ^ N2O + HNO3 + 2H2O 6.11

The additional nitrate available from HAN also provides a salting effect at low acidity which partly compensates to help keep uranium in the organic phase. HAN is not required in large excess and it can be readily destroyed by heating the solution to >60 °C before the subsequent plutonium finishing steps.

Even with the use of HAN as the primary Pu reductant, purification flowsheets can include a “plutonium barrier”, using hydrazine stabilized U(IV) to complete plutonium stripping from the solvent, prior to the solvent regeneration operation. In the La Hague plants, HAN is used as the reductant in the Pu purification cycles: UP-3 uses the HAN flowsheet in pulsed columns; UP-2 uses HAN in centrifugal contactors. In the UP-2 plant, the Pu product from the first (codecontamination) cycle requires no additional decontamination from U; additionally, the introduction of a single reductant (HAN) is much simpler than adding several as would be the case for uranous nitrate. The Pu product from UP-2 is reportedly quite pure with <5000 ppm total impurities, <0.1 pCi P-у activity, and <100 ppm U (Baron 2008).

Uranium purification cycle

A fraction of the neptunium coextracted with uranium, and plutonium follows the uranium stream through the process. The uranium stripping operation in the partitioning step is operated such that the uranium product stream is at a relatively low concentration in dilute nitric acid. This stream is concentrated in evaporators prior to treatment in the uranium purifica­tion cycle. This concentration increase is necessary if the extraction opera­tion is to be carried out with favorable organic to aqueous flow ratios. During the concentration step, neptunium is oxidized to Np(V) and Np(VI) and the acid concentration is increased in the aqueous solution fed to the U purification cycle. In order to complete uranium purification from Np, hydrazine nitrate is introduced into the extraction step to reduce extract­able Np(VI) to the inextractable Np(V) species. The uranium stripping operation is then similar to that carried out in the first cycle.

As originally developed, the voloxidation process is carried out in an air atmosphere at 480°C to 600°C, primarily to remove tritium from the fuel. During oxidation of the UO2-based fuel, tritium, which may be present in the fuel in elemental form, diffuses to the surface of the particles where it reacts with oxygen to form water, which enters the gas stream.6,7 The rate of reaction at 480°C is such that >99.9% of the tritium is released from the fuel in about 4 h.

A significant portion of the tritium in the used fuel assemblies is associ­ated with the cladding as zirconium hydride (ZrTx).5 The fraction of tritium in the cladding has been reported at 40%8 and higher.6 The hydride should readily oxidize, releasing the tritium, if it were accessible by the oxygen. However, experimental data indicate that oxidation has little effect on the tritium held in the cladding when processed for 6 h at a temperature of 480°C.9

Of special interest from a practical point of view is the modification of the extraction mixture, i. e. the use of dipicolinic acid diamides (DPA) instead of carbamoylphosphineoxide. The mixture of CCD with DPDA shows the synergistic effect for Am extraction [37]. The proposed mixture of CCD + DPDA + polyethylene glycols manifests the extraction of Cs, Sr, REE and TPE, similar to the classical UNEX-extractant [38]. The formulae of some investigated DPDAs are presented in Figs 9.11 and 9.12.

The data on extraction of radionuclides are given in Table 9.18 and Fig. 9.13, from where it can be seen that diamides effectively extract radionu­clides even from 3 M HNO3.

The advantages of DPAs compared to carbamoylphosphineoxides are their rather simple synthesis and the possibility for a wide variation in their

9.12 N, N,-diphenyl-N, N,-dimethyldiamide of dipicolinic acid (PhMDPA).

1

[HNO3], M

9.13 Extraction of europium by mixture of 0.13 M CCD + 0.027 M PEG-400 + TBDPA in FS-13 as function of HNO3 concentration and content of diamide in organic phase.

structure. Further, metal solvates with DPDA are more easily soluble when compared to those with CMPO, and thus one can use higher concentrations of DPDA in the extraction mixture which leads to an increase in the extrac­tion mixture capacity for trivalent metals. This is of importance for the treatment of solutions with a high content of REE. Furthermore, DPAs extract TPE to a greater extent than REE.

As in other systems based on CCD, the extraction ability of CCD — DPDA and CCD — DPDA — PEG systems varies depending on the diluent in the row of F-3 > FS-13 > F-8 (see Fig. 9.14).

The above data indicate that diamides of dipicolinic acid are of interest as synergistic additions to CCD. The CCD — DPDA mixtures effectively extract REE and An from acidic solutions of HNO3. Diamides are easier to synthesize and thus they are significantly cheaper than carbamoylphos — phineoxides. At the same time, the extraction properties of CCD — DPDA and CCD — CMPO mixtures are practically the same. It should be also emphasized that, by applying DPDA instead of CMPO one can use higher concentrations of DPDA and thus increase the limiting concentration of REE and An in the extractant. This property is of prime importance for treatment of waste with a high REE content. The modified UNEX-extractant CCD-TBDPA — Slovafol-909 in F-3 was tested on simulated waste with high REE content. Cesium, strontium and minor actinides were recovered rather efficiently. At extraction from the simulated solution bearing more than 4 g/L REE, the high separation factors of americium remained even after three successive contacts with fresh portions of aqueous solutions. Dynamic testing of this extractant was performed at Mayak PA in collaboration with RI. The extraction process was stable and americium was extracted more effectively than europium. Hydrodynamic and extraction properties were unaffected during the tests.

Selective stripping is one possible direction for UNEX process develop­ment. The data on selective stripping of metals from both organic solvents are presented in Tables 9.19 and 9.20. It can be seen that, in the case of a saturated UNEX solvent (0.08 M CCD + 0.015 M PEG-400 + 0.013 M CMPO in FS-13), actinides and lanthanides can first be stripped with a solution (A) of ammonium carbonate and acetohydroxamic acid (AHA), and then Sr can be stripped separately from Cs with a solution of (NH4)2CO3 with AHA and DTPA (B). Cesium is stripped in the last stage using 2 M methylamine carbonate solution [39].

The same stripping solutions were examined for the recovery of metals of interest from the modified UNEX-solvent (0.1 M CCD + 0.05 M TBDPA + 0.025 M PEG in FS-13). Unlike in the previous case, the use of 1 M (NH4)2CO3 + 0.013 M AHA does not provide stripping of Ln and An ele­ments; therefore, different complexants were examined, such as citric acid, nitrilotriacetic acid (NTA), hydroxyethylene-diphosphonic acid (HEDPA), diethylene triamine pentaacetic acid (DTPA), mixed with ammonium car­bonate. The most promising data were obtained with aminoacetic acid (glycine) used as a buffer compound (solutions H and I). The principal scheme for selective stripping is indicated in Fig. 9.15. As a first stage, the selective stripping of Sr with a solution H (1 M (NH4)2CO3 + 0.02 M glycine + 0.013 M DTPA) is applied. At the second stage, a solution G (1 M (NH4)2CO3 + 0.039 M DTPA) provides the selective separation of An and Ln. Cesium can be stripped in the last stage using 2 M methylamine carbon­ate solution.

The data presented confirm that it is possible to achieve the separation of nuclide groups by selective stripping.

9.4

Conclusions

As a result of collaboration between RI and INL, the universal extraction system based on CMPO or DPDA, CCD and PEG in polar diluents have been developed, which permit the recovery of Cs, Sr, An and Ln.

The optimal composition of the UNEX-solvent was determined and the conditions for extraction (combined and fraction) of long-lived radionu­clides were established. The high chemical and radiation resistance of the UNEX system, as well as its corrosion, explosion and fire safety, were dem­onstrated under operating conditions.

Variants of work flows based on the UNEX process were developed and tested at pilot plants belonging to KRI, NIKIMT and MCC in Russia, and at INL in the USA, with the use of real and simulated HLW of different compositions. Optimization of the UNEX process resulted in creation of a simple-to-realize work flow involving the three following operations: com­bined extraction of Cs, Sr, An and REE, scrubbing of the extract and com­bined stripping of all radionuclides under study; the recovery rates of radionuclides attained in the course of tests allowed the main HLW bulk to be transferred into the category of low-level waste.

The possibility for realizing the UNEX process on a commercial scale was verified by its testing with the use of the commercial EZR125 centrifugal contactors and with simulated HLW. As to the secondary (end) products of the UNEX process (raffinate and strip product), techniques for their solidi­fication were proven; the traditional technique of cementing was checked for low-level raffinate; the vitrification process for high-level strip products dem­onstrated the possibility of producing glass blocks with a volume of 5 L for every 1 m3 of HLW being treated. The method for regeneration of spent extractant of the UNEX process was elaborated and tested, which made it possible to return more than 90% FS-13 diluent and 40% CCD for re-use.

Thus, the technologies developed and tested for HLW treatment have shown the possibility of deep recovery of radionuclides which allows to transfer the main bulk of wastes into the category of low-level wastes (LLW) suitable for inexpensive near-surface storage. The results of feasibil­ity study, conducted by Idaho National Laboratory have confirmed that the use of UNEX process should reduce the amount of solidified HLW by a factor of 23.

Challenges of the selective extraction of caesium

After five years of cooling, one tonne of a spent PWR uranium oxide fuel (with a burnup of 60 GWd/t) still contains 4.6 kg of caesium (1.89 kg of stable isotope Cs-133, 50 g of two-year half-life Cs-134, 769 g of 2.3 million- year half-life Cs-135, and 1.93 kg of 30-year half-life Cs-137), which together with strontium is the main short-term heat source in the vitrified waste. In addition to the need for selective caesium extraction from dissolver liquors to decrease this thermal source, selective extraction of caesium from acidic effluents containing large quantities of sodium nitrate arising from the neutralization of nuclear plant technological waste streams by concentrated soda, as well as from basic sludge contaminated with alpha emitters inherited from decades of military research programs, is of primary impor­tance in nuclear waste management.

In aqueous nitric acid solutions, the caesium metallic cation is monova­lent and considered as a ‘hard acid’ in Pearson’s theory. It therefore inter­acts preferentially with ‘hard bases’, such as anions or molecules containing fluoride or oxygen donor atoms and inducing electrostatic interactions. Studies reported in the literature point out the difficulties of extracting caesium selectively from acidic solutions, especially in the presence of other alkali elements (Moyer and Su, 1997, Herbst et al., 2002a).

The first extracting agents envisaged in the 1980s were the crown ethers: macrocyclic polyethers discovered by Pedersen (1967). Parametric studies carried out on the extraction of alkali cations by crown ethers (Danesi et al., 1975, Sadakane et al., 1975, Gerow et al., 1981, Schulz and Bray, 1987, Wood et al., 1995, Dozol et al., 1995, Dietz et al., 1996, Kumar et al., 1998, Kikuchi and Sakamoto, 2000) have shown that the quantity of metallic cation extracted in the organic phase depends on the following factors: the polarity of the organic diluent used to dissolve the crown ether, the nature of the co-extracted anion, the size of the crown ether cavity relative to the alkali cation diameter, the chemical composition of the aqueous solution (acidity, ionic strength), and the type of the chemical functions grafted onto the crown ether. The most evolved design of a crown ether for caesium extraction is that of di-(tert-butyl-benzo)-21-crown-7 (Fig. 11.1) because the cavity size of its crown, containing seven oxygen atoms, matches the coor­dination shell of the caesium cation, and because the tert-butyl-benzo groups grafted onto its skeleton enhance its hydrophobicity.

Although di-(tert-butyl-benzo)-21-crown-7 can extract caesium from nitrate feeds of low acidity, it definitely requires the addition of a synergistic cation exchanger, such as dinonyl-naphthalene sulfonic acid (Kozlowski et al., 2002, 2007), or even better hexabrominated bis(dicarbollide) anion (Gruner et al., 2002), a very strong cation exchanger used in the UNEX process (Herbst et al., 2002a, b, 2003), to extract caesium from PUREX acidic raffinates. Nevertheless, the drawback of synergistic mixtures com­posed of solvating agents (such as crown ethers, which extract metallic cations with increasing efficiency as the ionic strength increases) and cation exchangers (such as cobalt dicarbollides, which extract metallic cations more efficiently at low acidity) is the difficulty encountered when stripping the metallic cations, because of competition between the two extractants. The same problem occurs for crown ethers that have been functionalized with a cobalta bis(dicarbollide) cage (Fig. 11.1) to extract caesium from acidic feeds (Gruner et al., 2002). Their caesium extraction efficiency is comparable with that of the corresponding synergistic mixtures composed of hexabrominated bis(dicarbollide) anion and crown ethers. Their selectiv-

ity with respect to Na+ cation is sometimes even higher (separation factor: SFcs/Na[10] > 100 for trace elements) than that of the corresponding synergistic mixtures, yet too low in the presence of high contents of sodium ions to develop a process for decontaminating technological nuclear waste streams.

When dissolved in organofluorine diluents such as 1,1,7-trihydrodo — decafluoroheptanol, di-(tert-butyl-benzo)-21-crown-7 and other crown ethers appear to extract caesium from acidic feeds ([HNO3] > 3 mol. L-1, Yakshin et al., 2008), but the complexity of their preparation and conse­quently their high cost, as well as their poor Cs+/Na+ selectivity, still hinder their use on an industrial scale.

In the fuel fabrication process for light water nuclear power reactors, enriched uranium hexafluoride (UF6) is converted to UO2 as illustrated in Fig. 14. 7. Disposable solid waste generated by the process is reduced by a factor of 1/25 through a carefully controlled incineration process. The incin­erator ash contains approximately 10% by weight of enriched uranium (about 3.5% 235U). Some gadolinium (Gd) which is added to the fuel as a neutron absorber is also present in the waste ash. In 2003, AREVA NP started looking into the possibility of using sc-CO2 as a medium for recover­ing enriched uranium from the incinerator ash waste generated by the fuel fabrication process in collaboration with the University of Idaho. With the aid of the TBP-HNO3 complex described in the previous section, enriched uranium and gadolinium in the incinerator ash can be effectively extracted

Incoming UF6 cylinders

14.7 Illustration of a typical light water reactor fuel fabrication facility (Source: US Nuclear Regulatory Commission. "Light Water Reactor Low-Enriched Uranium Fuel," Fuel Fabrication. ONLINE. 2009. Nuclear Regulatory Commission. Available: http://www. nrc. gov/materials/fuel- cycle-fac/fuel-fab. html [5 Aug. 2009].

able commodity (enriched uranium) from a material that previously was considered waste.

Another industrial demonstration of the sc-CO2 extraction technology was conducted in Japan by Mitsubishi Heavy Industries, Japan Nuclear Cycle Corp. and Nagoya University testing the feasibility of reprocessing spent nuclear fuel using this green solvent (Shimada et al. 2002). The process is called “Super-DIREX” process which stands for supercritical fluid for direction extraction. The main purpose of this project is to evaluate the fea­sibility of direct dissolution of uranium and plutonium in irradiated uranium oxides and in spent fuel by sc-CO2 containing the TBP(HNO3)18(TBP)06 complex. After dissolution, uranium and plutonium are stripped from the sc-CO2 phase with water. Recovery of UO2 from the strip solution follows the conventional PUREX process. The details of the Super-DIREX project are not fully known. The nitric acid concentration in the acid droplets sur­rounded by TBP (like reverse micelles) is estimated to be greater than 12 mol/L (Sawada et al. 2005). A recent report by Sawada et al. (2009) described the distribution coefficients of U(VI) and some simulated fission products measured between aqueous phase of high nitrate concentrations (4-15 mol/L) and dodecane with the purpose of understanding the extrac­tion behavior of uranium and fission products in the Super-DIREX process. Reprocessing spent nuclear fuel in sc-CO2 is obviously a complicated process and requires many efforts in chemical and engineering studies to evaluate its feasibility. Nevertheless, the Super-DIREX project should provide some valuable information regarding the chemistry of uranium, plutonium and fission products in the sc-CO2 dissolution process and safe handling of highly radioactive materials in a pressurized system.

PUREX process is the most mature solvent extraction separation process in use today. Its principles are described in a great detail in Chapter 6 of this book. PUREX was first implemented in the 1950s, thus its radiation chemistry is the best studied. In general, the PUREX solvent consists of 30 vol % n-tributylphosphate (TBP) in an alkane diluent. In this process, as in its advanced version, UREX (Uranium Extraction), a multistage usage of the solvent is expected. With each use it is exposed to the high radiation field arising from dissolved used nuclear fuel. Considering either the group separation design or the multistep UREX+ process, the PUREX/UREX is the starting process step and hence, TBP obtains the highest radiation dose possible during the processing. The aqueous phase initially contains all fission products, which emit strong beta and gamma radiation, and most of which are much shorter lived than actinides. It has long been recognized that the major products of TBP radiolysis are hydrogen, methane, and dibutylphosphoric acid (HDBP), with monobutylphosphoric acid (H2MBP) and phosphoric acid produced in lesser amounts, but increasing with pro­longed exposure without cleanup. Dibutylphosphoric acid is a major radio­lytic product with ion-exchange extraction properties (Egorov, 1986; Zilberman et al., 2002). The radiation chemistry of TBP was recently reviewed (Mincher et al., 2008; 2009).

Many of the extractants applied in the post-PUREX separation stages of advanced nuclear fuel cycle concepts are functionalized aromatic compounds. Their irradiation in the form of a nitric acid saturated organic solvent and in the presence of aqueous nitric acid may result in aromatic nitration reactions. This may lead to degradation or formation of different bonds and, actually, new coordination chemistry between the metal of inter­est and the ligands. This may have adverse effects on solvent extraction. Dependent on the dose and LET of irradiation, different effects are pos­sible. The degradation of ligand or extracted complexes doesn’t have to be progressive. For extraction of Am3+ with diamidic dipicolinates, it was observed that extraction yields first increased with low radiation doses, then with increased dose significantly dropped (Lapka et al., 2010). Stronger binding also may lead to difficulties in stripping, and in such a way, to lower separation yields than desired.

Initial non-radioactive tests to evaluate performance of the spectroscopic equipment under the flow conditions employed 30% TBP/n-docecane and five aqueous solutions containing variable nitric acid and neodymium nitrate concentrations were introduced into the extractor bank in a counter­current manner. The spectroscopic probes were positioned to measure the effluent from the extractor banks of the raffinate (aqueous) stream. Starting with a water wash and then proceeding with solutions 1-5 (compositions are given in Fig. 4.13) in sequential order, approximately 200 mL of each aqueous phase solution was delivered through the contactors, while the organic phase delivery was constant during the experiment. The Raman spectra collected during the experiment are shown in Fig. 4.13.

The nitrate and neodymium concentrations of the solutions were deter­mined using PLS analysis of the spectroscopic data. Figure 4.14 shows the plots of both the nitrate and neodymium concentrations as a function of experimental time from the Raman and vis-NIR data, respectively. The initial feed concentration is outlined in light grey, and the measured con­centration in the raffinate is indicated as dark grey symbols. As expected, the nitrate concentration in the raffinate (Fig. 4.14 a) nearly mirrors the

4.13 Raman spectra of aqueous raffinate solutions collected on-line during counter current flow extraction experiment showing variable nitrate response at 1050 cm-1.

concentration in the feed, attributed to the insignificant extraction of total nitrate into the organic phase. In contrast, the neodymium shows appre­ciable extraction when the total nitrate concentration exceeds approxi­mately 1 M (Fig. 4.14 b). This figure also depicts the noticeable mixing regions reflecting the consequent switching of the feed solutions; the mea­sured concentration of the analyte lags in time because of the time needed for the new solution to be integrated into the counter-current contactor bank.

The results of a batch contact distribution experiment for Nd(NO3)3 in variable nitrate with 30% TBP/n-dodecane is shown in Fig. 4.15. The measured concentrations for Nd3+ in the feed and raffinate are plotted in the figure at left, and the resulting distribution values for Nd (DNd) as a function of total nitrate concentration are shown in the figure on the right. Distribution values for Nd3+ depend linearly on the nitrate concentration

4.15

Batch contact distribution experiment for Nd(NO3)3 in variable nitrate (composition of solutions 1-5 are shown in Fig. 4.13) with TBP/ dodecane.

in the aqueous phase with the corresponding slope of nearly 3 (3.22) con­firming the extraction of Nd(NO3)3 by TBP. This verifies that the decrease of Nd optical signal observed in the raffinate stream (for solutions 4 and 5) is due to the enhanced Nd extraction leading to the depletion of the Np in aqueous phase and not due to the instrumental failure.

M. C. REGALBUTO, Argonne National Laboratory, USA

Abstract: The separations strategy for the UREX+ type processes is based on an optimization approach for which key objectives for a closed or partially closed fuel cycle are pre-determined. The overall process is composed of a sequence of separation steps or modules linked to generate a desired set of outputs, whether products or intermediates. Processing options are shown for the case of LWR recycle, where incentives exist to separate and recycle the actinides, and separate and manage fission products. The UREX+ approach was successfully tested for the recycle of LWR SNF for a number of different product and waste forms configurations and the results from these tests are given.

Key words: LWR recycle, UREX+, GNEP, AMUSE, FPEX, NPEX, TALSPEAK, TRUEX, CCD-PEG.

7.1 Introduction

The original goal of actinide separations research in the US, which dates back to the 1940s, was the recovery of plutonium for defense purposes. The ultimate result was the development of large-scale chemical processes for the separation of plutonium from irradiated uranium. A number of chemi­cal processes were tested (US Nuclear Regulatory Commission, 2008, p. 13) at the Hanford and Savannah River sites. By far the most successful was the PUREX process, which has now been developed commercially to treat spent power reactor fuel. By the mid 1950s, advancements in nuclear power technology, the then-perceived scarcity of uranium-bearing ore, and the high energy requirements of gaseous diffusion for uranium enrichment, resulted in the belief that there would be a shortage of uranium fuel sup­plies if nuclear power expanded significantly. In response, efforts were begun to develop reactors that could breed plutonium to serve as the fissile component in fuel, along with reprocessing facilities to recover the bred plutonium.

Although the plutonium was intended for use in commercial power reac­tors (rather than defense applications), the goal remained the isolation of plutonium from the other components of spent fuel, as it was the product of interest. Uranium was also collected as a product in PUREX but required re-enrichment to serve as a fuel for light water reactors. Because of this early development work, PUREX is the only current industrial process for recovery of plutonium (and uranium) from commercial spent fuel. In the early 2000s, the reprocessing of spent fuel re-emerged as an area of interest driven by the forecasted growth in nuclear power (MIT, 2003) and the need to develop sustainable, environmentally acceptable and economic closed fuel cycles (Williamson et al., 2004). The separations goal this time had changed from recovery of plutonium to management of all minor actinides and fission products not only for reuse of material as fuel but also to help address the waste management challenges still facing nuclear power gen­eration (Wigeland et al., 2006).

Today, in many countries possessing highly developed nuclear power, the concept of a closed nuclear fuel cycle (NFC) with reprocessing of spent nuclear fuel (SNF) prevails. One advantage of the closed NFC is the pos­sibility of a radical solution to the problem concerning long-term safe management of long-lived radionuclides, as SNF reprocessing allows these radionuclides to be recovered and handled individually.

Transmutation is a reliable method for the management of long-lived radionuclides. Another promising method is based on the creation of espe­cially strong matrices to be disposed of in geological formations. In both cases, the long-lived radionuclides contained in SNF must be selectively recovered. The bulk of long-lived radionuclides are contained in liquid high-level wastes (HLW) arising from SNF reprocessing. Therefore, the problem of individual or group separation of long-lived radionuclides from HLW calls for the development of effective partitioning methods. The com­position of the fractions depends on the chosen method for further manage­ment of ecologically hazardous radionuclides, i. e. on their transmutation or production of stable matrices for their prolonged storage or final disposal. The most important fractions generated by HLW reprocessing are:

• cesium and strontium (in combination or separately);

• actinides and lanthanides (in combination or separately).

Among the different ways of HLW partitioning (precipitation, sorption, chromatography, etc.), extraction processes are of special interest. The fol­lowing extraction systems should be considered as most promising:

• neutral organophosphorus compounds (alkylphosphine oxides [1,2,3], carbamoylphosphine oxides [4,5], phosphorilated calixarenes [6]);

• acidic organophosphorus compounds (diisodecylphosphoric acid [7], zirconium salts of dialkylphosphoric acids [8]);

• macrocyclic compounds (crown-ethers [9], calix-crowns [10]);

• diamides (DIAMEX-process [11,12] or ARTIST-process [13]);

• hydrophobic anions in polar diluents (chlorinated cobalt dicarbolide — CCD [14]);

• synergistic mixtures of different extractants [15-18].

So far, the only method which has found commercial application is extrac­tion of cesium and strontium by CCD [19]. CCD was first proposed for Cs and Sr extraction by Czech scientists [20] and the technological bases of this process were then elaborated by Czech and Russian scientists; thereafter, specialists at the Khlopin Radium Institute (KRI) and Production Association Mayak (Mayak PA) developed the process up to its introduction at a radiochemical plant [21].

A problem encountered in the development of technology for HLW management concerned the recovery of actinides (uranium, neptunium, plutonium, americium, curium) and rare-earth elements (REE), along with cesium and strontium. The problem could be solved by using the above- listed extraction systems, for example the UREX (Uranium Extraction) process [22]. However, the extraction of several extractants in several extraction cycles is more expensive when compared to extracting every­thing in one extraction cycle, so extractants for the simultaneous recovery of different fractions from HLW are of great interest.