CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 898 CHIMIA 2005, 59, No. 12

Chimia 59 (2005) 898–904 © Schweizerische Chemische Gesellschaft ISSN 0009–4293

Recent Developments in the Purex Process for Reprocessing: Complexant Based Stripping for / Separation

J. Eddie Birkett, Michael J. Carrott, O. Danny Fox, Chris J. Jones, Chris J. Maher, Cécile V. Roube, Robin J. Taylor*, and David A. Woodhead

Abstract: In order to recycle potentially valuable uranium and plutonium, the Purex process has been successfully used to reprocess for several decades now at industrial scales. The process has developed over this period to treat higher burnup fuels, oxide as well as metal fuels within fewer solvent extraction cycles with reduced waste arisings. Within the context of advanced fuel cycle scenarios, there has been renewed international interest recently in separation technologies for recovering from spent fuel. Aqueous fuel processing re- search and development has included further enhancement of the Purex process as well as the development of minor partitioning technologies that use new extractants. The use of single cycle Purex solvent extraction flowsheets and centrifugal contactors are key objectives in the development of such advanced Purex processes in future closed fuel cycles. These advances lead to intensified processes, reducing the costs of plants and the volumes of wastes arising. By adopting other flowsheet changes, such as reduced fission product decontamination factors, U/Pu co-processing and Pu/Np co-stripping, further improvements can be made addressing issues such as proliferation resistance and burning, without adverse effects on the products. One interesting development is the demonstration that simple hydroxamic acid complexants can very effectively separate U from Np and Pu in such advanced Purex flowsheets. Keywords: Acetohydroxamic acid · Actinides · Centrifugal contactors · Nuclear fuel reprocessing · Purex process

4+ – 1. Introduction then vitrified ready for disposal. By far the Pu + 4NO3 + 2TBP ↔ most successful reprocessing technology Pu(NO3)4.2TBP (1) to date has been the Purex process, which Nuclear fuel reprocessing [1] is the sepa- uses solvent extraction between aqueous UO 2+ + 2NO – + 2TBP ↔ ration and purification of reusable uranium 2 3 nitric acid solutions and organic solutions UO2(NO3)2.2TBP (2) and plutonium products from irradiated of tri-n-butyl phosphate (TBP) diluted in a nuclear fuel. The recovered U and Pu can 4+ 4+ paraffinic diluent, such as Exxon D-80. Re- 2Pu + U + 2H2O ↔ then be converted in to new uranium oxide 3+ 2+ + processing plants in the United Kingdom, 2Pu + UO2 + 4H (3) (UOx) or mixed oxide (MOx) fuels for re- France, Japan and Russia all use versions of cycle to reactors. The highly active (HA) the Purex process and it is likely that Purex A complication is that Pu(III) is quite fission product wastes are evaporated and will remain the basis of any future aque- easily re-oxidised to Pu(IV) by nitric acid ous reprocessing technology developed for in a series of reactions that is autocata- advanced fuel cycles such as ‘Generation lysed by nitrous acid (Eqn. (4) and (5)). IV’ reactors. To interrupt this process and stabilise Pu The separation of U and Pu from fis- as Pu(III), an aqueous phase nitrous acid sion products is achieved by extraction of scavenger such as is added to hexavalent and tetravalent metal nitrate the U/Pu separation contactor (Eqn. (6) complexes (Eqn. (1) and (2)) followed and (7)) [2]. by reductive stripping of Pu, whereby the *Correspondence: Dr. R. Taylor extractable tetravalent Pu is reduced to in- Pu3+ + HNO + H+ → Pu4+ + NO + H O Nexia Solutions Ltd. 2 2 BNFL Technology Centre extractable trivalent Pu and stripped to an (4) aqueous phase, leaving U(VI) in the organic CA20 1PG phase. Typically, a reductant such as ura- 2NO + HNO + H O → 3HNO (5) United Kingdom 3 2 2 IV Tel.: 44 1946779266 nous nitrate {U( )} is used for this purpose Fax: 44 1946779007 (Eqn. (3)) [2]. N2H4 + HNO2 → HN3 + 2H2O (6) E-Mail: [email protected] AND MATERIALS IN NUCLEAR POWER PRODUCTION 899 CHIMIA 2005, 59, No. 12

NH3 + HNO2 → N2 + N2O + H2O (7) However, although it is nominally con- sidered to be ‘inextractable’, trace quanti- ties of Pu(III) are extracted into the solvent phase where, in the absence of a scavenger, re-oxidation to Pu(IV) does occur. In two- phase mixed systems this process is en- hanced as the system tries to equilibrate; the ultimate effect being a requirement to add a substantial excess of reductant to maintain Pu in the trivalent state [2].

2. Purex Process Development and International Experience

The Purex process was developed in the late 1940s in the United States. Previously large-scale plutonium and fuel processing used a bismuth phosphate precipitation method; this was then superseded by sol- vent extraction processes that could be run Fig. 1. Schematic representation of a typical configuration for an oxide fuel reprocessing plant using continuously. The Redox process using me- the Purex process, indicating (a) Head End, (b) Chemical Separation, (c) Product Conversion and (d) thyl isobutyl ketone or hexone as the solvent Waste and Effluent Management was used at the in the USA and the Butex process using dibutyl carbi- tol was used in the UK at Windscale. How- The Thermal Oxide Reprocessing Plant Hence, in recent years there has been ever, the Purex process was quickly shown (Thorp) at the British Nuclear Group Sell- renewed interest in the development of to possess a number of process chemistry afield site, UK, exemplifies this approach. enhanced separation processes for the and chemical engineering advantages over It uses three solvent extraction cycles with management of spent nuclear fuel within these other processes and has remained the an early split flowsheet compared to four possible future closed fuel cycles [3][5]. dominant technology for processing nucle- cycles and a late split used by the earlier Both aqueous partitioning and molten salt ar fuel since the 1950s/60s [3]. Magnox reprocessing plant on the Sellafield pyro-processing are under development, A Purex reprocessing plant comprises a site. Ferrous sulphamate and sodium nitrite with a particular focus on developing a number of facilities: (a) a head end plant to reagents used by Magnox are replaced by capability to recycle and burnup (or trans- receive and store spent fuel and to convert salt-free reagents, U(IV) and nitrogen oxide mute) the transuranic elements. This is in the fuel to a solution in HNO3 ready for (b) gases (NOx). Pulsed columns are used for order to reduce by factors in the order of chemical separation using solvent extrac- the Pu-bearing streams and the separations 103 the time taken to reduce nuclear wastes tion to produce separate aqueous nitrate plant is integrated with the other supporting to the activity of the original natural U com- products that can be (c) converted to solid operations. This is achieved despite Thorp ponent, which would make the long-term oxide products. A substantial supporting processing higher burnup, enriched oxide management of nuclear waste much easier infrastructure (d) is necessary to treat solid fuel from Advanced Gas Cooled Reactors [6]. There are a number of variations on ad- wastes and liquid and gaseous effluents (AGRs) and Light Water Reactors (LWRs) vanced fuel cycles, but most scenarios share arising from reprocessing operations [1] as opposed to lower burnup, short cooled, a common aim of recycling TRU actinides [4]. A schematic representation is given in natural enrichment U metal Magnox fuel for transmutation. This generally requires a Fig. 1 for a typical oxide fuel reprocessing [1][2][4]. move from current thermal reactors to ei- facility. ther fast reactors (FR) or accelerator driven Large-scale Purex reprocessing plants systems (ADS) [6]. Notably, four different (~800 te/a) are in operation in the UK 3. Advanced Fuel Cycles (AFCs) FR systems are being developed within the (Sellafield), France (La Hague), and Japan ‘Generation IV International Forum’, which (commissioning at Rokkasho) but a number In the last ten years or so there has been is now probably the most prominent inter- of other countries have past experience of a growing international interest in future or national programme developing advanced Purex reprocessing at various scales, e.g. advanced nuclear fuel cycles and many dif- reactor systems [7]. Consequently, whether USA, Germany, Russia, Belgium, India ferent scenarios have been proposed around such fuel cycles are based on a measured [1][2][4]. the basic alternatives of open, partially transition, the so-called ‘double strata’ ap- Chemical separation flowsheets are all closed or fully closed cycles (with respect proach or a more aggressive transition to broadly similar, using a number of solvent to the actinide inventories) [5]. Partially fast reactors, actinide recovery and recycle extraction cycles to separate and purify U closed cycles in which Pu is recycled as is a common requirement [6]. However, and Pu from all other elements. Process MOx fuel and fully closed cycles, which re- recovery of the trans-Pu (TRPu) elements development has been in the direction of cycle all the transuranium (TRU) elements (Am, Cm) is not achievable within the reducing the number of solvent extraction (Np, Pu, Am, Cm), obviously require fuel Purex process and is chemically very dif- cycles, adopting an early separation of Pu processing and separations technologies. ficult due to the very similar properties of from U, using salt-free reagents to mini- Whilst Np and Pu can be recovered using these trivalent actinides {An(III)} with the mise waste arisings and intensifying the Purex technology, the recovery of Am and trivalent fission product in solu- contacting equipment, e.g. pulsed columns Cm is more problematic and requires alter- tion. Much effort has therefore been dedi- rather than mixer-settlers. native separation methods. cated to the development of new partition- CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 900 CHIMIA 2005, 59, No. 12 ing technologies for the trivalent actinides ing concerns on future energy and fuel cy- kinetics can be slow. Most recent efforts (chemical separations and fuel fabrication). cle needs. have been directed at finding variants with As well as TRU actinide recycle, further The French Commissariat à l’Énergie increased stability [24–26]. The CEA have requirements for advanced separation tech- Atomique (CEA) have made very signifi- also proposed a calixarene-based process nologies include: increased proliferation cant progress in developing solvent extrac- for Cs separation and an electrochemical resistance; reduced costs and lower envi- tion processes that can partition trivalent process for Am/Cm separation [10][27]. ronmental impact and a reduced impact on actinides from Purex high-level wastes. European development of partitioning geologic repositories such as Yucca Moun- Their reference process again uses a modi- routes, including DIAMEX and SANEX tain in the USA [6][8–10]. fied Purex flowsheet to control U, Pu and processes, has been coordinated through the These drivers have led to substantial Np followed by new solvent extraction European Commission EURATOM Frame- research worldwide to either improved processes termed DIAMEX (Diamide Ex- work programmes. Projects such as NEW- aqueous based processes or alternative traction) and SANEX (Selective Actinide PART, PARTNEW, CALIXPART and most technology, most notably pyroprocessing Extraction) [10]. The DIAMEX process recently EUROPART have investigated a in molten halide salt media [11]. Some of uses a malonamide extractant, such as DM- wide range of complexing agents such as the consequent objectives of development DOHEMA (N,N’-dimethyl N,N’-dioctyl BTP and hemi-BTP ligands and calixarene- programmes for advanced aqueous-based hexyloxyethyl malonamide) to co-extract based ligands for the separation of trivalent process flowsheets and separation technol- An(III) and Ln(III) ions [20]. Various di- actinide and caesium ions from high level ogies are given in Table 1. amides have been investigated extensively wastes [28–31]. In the development of aqueous-based and ‘hot’ testing on real high-level raffi- The American Advanced Fuel Cycle separation processes, the Purex process is nates and even high-level concentrates fol- Initiative (AFCI), is aimed at developing the well-established basis for recovering lowing evaporation has been successfully separations and recycle technology for fu- U and Pu, and with slight modifications, carried out. A centrifugal contactor trial on ture US spent fuel in order to substantially Np, I and Tc products (129I and 99Tc are real high-level wastes (HLW) in 2000, for extend the lifetime of the Yucca Mountain long-lived fission products with half lives instance, achieved ca. 99.9% recovery of repository [8]. The proposed aqueous proc- of 1.6E7 and 2.14E5 years respectively). Am and Cm [21][22], and a hot test using ess is termed UREX+, which aims to pro- Various forms of ‘advanced’ Purex proc- highly active concentrate (HAC) recovered duce a bulk U product suitable for disposal esses have been proposed to enable Purex 99.7% Am and 99.9% Cm [23]. Following as low-level waste or recycle to LWRs, and technology meet the requirements for the co-extraction of the trivalent actinides and Pu-Np and Am products for recycle. Cs and recovery of these species in advanced fuel lanthanides, the SANEX process is used to Sr separations are also planned to reduce cycles, e.g. [10][15][19]. Recovery of Am, separate Am and Cm from the lanthanides the short-term heat loading in the reposi- and other species including Cm and Cs, Sr [10]. A variety of extractants have been tory, which is one of the major limiting fac- (heat generating fission products) requires proposed within the SANEX framework, tors on the allowed repository inventory. A the development of alternative processes but all with the objective of exploiting the number of separation processes are being and extractants. small differences in covalency between 5f studied. An advanced Purex process us- and 4f electrons to achieve separation. Neu- ing the complexant acetohydroxamic acid tral multi-N-donor ligands have perhaps (AHA) for Np and Pu control (see Section 7 4. Current and Recent International proved the most successful with BTP (bis- below) seems to be accepted as the basis of Programs triazinylpyridine) ligands showing quite re- the Purex flowsheet. Cs/Sr separation using markable selectivity for Am(III) over Ln(III) a cobalt dicarbollide extractant or possibly Some, or all, of these requirements (Ta- (Fig. 2 illustrates an example of a BTP lig- calixarenes and TRPu element separations ble 1) are drivers for a number of major and). The BTP ligands however are rather using the French DIAMEX-SANEX proc- international programmes that have been unstable in process solutions with respect esses appear to be leading options to follow established in recent years to address grow- to radiolysis and hydrolysis and extraction the Purex section, although other options have been tested [12][32][33]. Table 1. Requirements for advanced aqueous reprocessing and partitioning technologies for AFCs In Japan, the Japan Nuclear Fuel Cy- Full TRU actinide recycle for A requirement for >99% separation, recovery and recycle of TRU cle Development Company (JNC) are transmutation elements (Np, Pu, Am and possibly Cm, Cf) [6][12] leading the development of technology to demonstrate the feasibility of fast reactor Increased proliferation resis- Flowsheet changes such as co-processing, avoiding separated (FR) fuel cycles; this includes the back tance pure Pu, increased fission product contamination of actinde end fuel processing technologies [13][34]. products (lower decontamination factors [DFs]), co-routing of They are investigating both aqueous and Np with Pu in flowsheets pyro-processing routes. On the aqueous Reduced waste arisings and Development of single cycle solvent extraction flowsheets with side, they propose an interesting innova- lower environmental impact lower solvent inventories [3], and/or the use of a crystallisation tion utilising crystallisation to remove the process [13] bulk of the U from solution followed by a much simplified Purex type process to co- Reduced costs Smaller intensified plant; the use of centrifugal contactors is wi- extract U, Np and Pu away from the fission dely accepted as the basis for next generation solvent extraction product wastes [13]. The actinides are then technology [14]. co-stripped to a U-Np-Pu product ready for Greater flexibility of proces- An ability to process within the flowsheet much higher burn up product finishing. Am and Cm are recovered ses fuels, including mixed oxide (MOx), Pu, and FR fuels by a modified TRUEX type process termed SETFICS [34]. (The TRUEX [Transura- Reduced impact on geologic Consideration of the partitioning of heat generating species such nium Extraction] process was developed by repositories as Cs and Sr from high level wastes for decay storage rather than the Argonne National Laboratory (ANL) in disposal. Also the capability of meeting specifications for the U product that allow the reclassification of U to low level waste the USA in the 1980s to extract TRU ele- (LLW) [12]. ments from wastes. It uses octyl-(phenyl)- N,N-diisobutylcarbamoylmethylphosphine CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 901 CHIMIA 2005, 59, No. 12

velopment of advanced or modified Purex processes must be a major component of Fig. 2. Examples any programme considering aqueous-based of some leading technology for fuel processing. The focus of extractants proposed for minor actinide such programmes is generally to meet the cri- partitioning. (a) iPr- teria listed in Table 1. Hence, the actinides U, BTP (iso-propyl Np and Pu need to be fully controlled within bistriazinylpyridine) a single Purex solvent extraction cycle that is proposed by CEA flexible to a wide range of feeds. The proc- for the separation ess must probably meet high product speci- of actinide(III) from fications on the U stream, with Pu streams (III) that offer increased proliferation resistance. ions in the SANEX This must be achieved within short residence process, (b) CMPO time annular centrifugal contactors, using (octyl-(phenyl)-N,N- diisobutylcarbamoyl- kinetically fast ‘CHON’ [42] reagents only methylphosphine oxide) [27], producing minimised volumes of easily proposed by ANL for treated product and effluent streams. TRUEX separations Notable early progress in the develop- of trivalent actinides, ment of such advanced reprocessing flow- (c) the malonamide sheets was the German IMPUREX proc- DMDOHEMA (N,N’- ess. However, the aims of this work were to dimethyl N,N’-dioctyl achieve high purity products (high fission hexyloxyethyl product decontamination factors [DFs]) malonamide) rather than TRU actinide recovery and other proposed by CEA for co-extraction of requirements for AFCs, as given in Table trivalent actinides and 1. They proposed using a high degree of lanthanides from high feed clarification coupled with a very high level wastes (DIAMEX saturation flowsheet to achieve very high process). fission product DFs in the first contactor [43][44]. A potentially useful innovation is com- oxide [CMPO] as the extractant (Fig. 2) velopment of different types of thermal and plexant-based rather than reductive strip- [11]. fast reactors using UOx, MOx, Pu, highly ping of Np and Pu using simple hydroxamic There are numerous other aqueous-based enriched uranium, carbide, metal fuels etc. acids [45–49]. processes that have been proposed for the dif- In particular, the development of simplified ficult task of separating An(III) from fission Purex-based flowsheets capable of process- products, particularly the trivalent lantha- ing a broad envelope of feeds, including 7. Complexant-Based Stripping nides (see e.g. [11]). Most of these processes very high Pu loadings, and converting these use counter-current solvent extraction tech- heterogeneous materials to standard oxide 7.1. Introduction nology but the proposed extracting solvents forms, is highly relevant to strategies for the As discussed previously, current Purex vary. Hybrid processes combining aqueous management of such Pu residues and ex- flowsheets use reductive stripping to sepa- and volatile fluoride routes, or aqueous and perimental fuels. Potential applications of rate Pu from U by exploiting the change pyro-routes have also been suggested [35]. simplified flowsheets, initially developed in Pu extractability as the oxidation state The Nuclear Energy Agency (NEA) of the for AFCs, to process UK legacy materials changes from IV to III. Fe(II), U(IV) and Organisation for Economic Co-operation have been reported [39]. Similar work sup- HAN are commonly used reducing agents. and Development (OECD) have established porting the development of the chemical The reductant must be used in combination various working parties to link international flowsheet for processing ex-weapons ma- with a nitrite scavenger such as sulphamic developments in partitioning and transmuta- terials in the US and Russian disposition acid or hydrazine [2]. tion (P&T) technologies and a series of sta- programmes have been reported by Bros- An alternative option is to replace the tus reports reviewing the progress made by sard and co-workers (in the context of joint reductive stripping of Pu by complexation: various member states are being published French – Russian – US programmes). They Pu(IV) is selectively complexed by a hy- [36–38]. report the development of aqueous proc- drophilic ligand and stripped in to the aque- esses that include pre-treatment to remove ous phase, leaving U in the solvent. The impurities such as chloride and to prepare advantages of complexation include fast 5. Legacy and Waste Processing the feed, a Ag(II) catalysed dissolution, fol- kinetics, relative temperature insensitivity lowed by a Purex Pu purification flowsheet. compared to redox reactions and no reoxi- Besides the development of flowsheets The solvent extraction flowsheet is based dation of Pu(III), hence no need for a stabi- for AFCs, modified aqueous processes can on Purex technology, with a hydroxylamine liser and likely improved nuclear criticality also be applied to deal with processing leg- nitrate (HAN) reductant stabilised by hy- control. It is also more likely that compl- acy fuels, actinide wastes, orphan materi- drazine [40][41]. exant-based stripping can be extended to als and similar problematic materials that fuels that have high Pu contents (e.g. up to many countries and laboratories have left 40 wt. %), where reductive stripping would over from development 6. Advanced Purex Process be difficult to implement and control due programmes. For instance, the UK has a Development to high reductant concentrations needed to wide range of legacy fuels and residues maintain Pu in the trivalent state. reflecting sixty years’ plus pioneering nu- Considering the proven utility and es- The most studied ligand for complexant clear industry development, including de- tablished operational experience, the de- stripping of Pu is the sulphate anion. Indeed, CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 902 CHIMIA 2005, 59, No. 12 a flowsheet based on sulphate stripping was designed and operated at the UKAEA Dounreay site in the UK for reprocessing fast reactor fuel [2]. This method suffers from the salt-loading of the waste streams and the corrosive nature of the sulphate anion and is not considered suitable for ad- vanced fuel cycles. Alternatives based on organic anions de- rived from common weak acids have been proposed; notably, formate, acetate and lac- tate but these generally are not favoured due to the weak nature of the complexes formed in nitric acid [2][49]. More promising studies on the use of or- ganic complexants led to the proposition of the use of hydroxamic acids for Pu stripping [19]. Interest in the use of hydroxamic acids arose because of the noted differences in the stability constants of Pu(IV) and U(VI) for a series of organic soluble hydroxamates, such as benzohydroxamate [50][51]. Hydroxam- Fig. 3. Effect of acetohydroxamic acid (AHA) concentration on the distribution of Pu at variable Pu ates are also known to be very good com- and variable AHA concentrations (batch distribution data measured at 25 ˚C) plexing agents for metals with high charge densities (e.g. Fe(III), Pu(IV)) through studies Consequentially, we have been studying the Pu stream still meets product speci- of siderophores and their analogues – natu- the chemistry of simple hydroxamic acids fications for current Thorp Pu products. rally occurring biomolecules used by fungi and their applications in Purex reprocessing The Pu DF on the U stream was fairly low for sequestration of Fe from the environment since the late 1980s. in the first trial (1000). It has been noted [52][53]. For consideration in reprocessing, earlier that the U product should meet cri- the simplest hydroxamates, formo- and aceto- 7.2. Plutonium Stripping teria for either disposal as LLW, prolonged hydroxamic acids, were chosen because they The applications of AHA in advanced storage as UO3, or recycle, either in LWRs have the least impact on the process, are most single cycle Purex flowsheets for separat- or as FR blanket fuel. The flowsheet there- easily managed in waste treatment and are the ing high levels of Pu from U have been fore needs to be capable of producing high most hydrophilic forms [46]. tested recently in a series of centrifugal specification U products, relatively free The stability constants for formo- and contactor trials [47]. The centrifugal con- of TRU α-activity. Modifications made acetohydroxamic acids with a range of met- tactor cascade used is shown in Fig. 4. in the later trials increased the Pu DF on al ions have been determined. The stability Three flowsheets with increasing Pu feed the U stream to 1.45E6, despite increasing constants in HClO4 (logβ1–logβ4) for the concentrations of 7, 20, and 40 wt.% were the Pu content in the HA feed to 100 g/l. Pu(IV) AHA complexes were found to be tested, representing an increasing challenge Mass balances across the flowsheet were 13.9, 24.1, 32.7 and 38.8 compared to 7.94 on the performance of the flowsheet and generally very good and percent recover- and 14.11 (logβ1, logβ2) for the U(VI) com- the ability of hydroxamic acids to separate ies excellent. These trials have given us plexes and a pKa for AHA of 9.02. These U and Pu. The levels of Pu in these flow- great confidence that hydroxamic acids demonstrate the strong complexation with sheets were selected to exemplify process- are very effective and robust reagents for tetravalent actinides even in very acidic so- ing MOx fuels, FR fuels and, at 40 wt.% Np+Pu recovery across a broad range of lutions and their specificity for Pu(IV) over Pu, some U.K. legacy materials as well as AFC scenarios. U(VI) [46][48]. to provide a strenuous test of the flowsheet. The ability of these hydroxamic acids The flowsheet for the 7% trial ([Pu] in the to strip tetravalent actinides from TBP, in Highly Active Feed [HAF] = 18 g/l) is re- 8. Conclusions the absence and presence of U(VI) was con- produced in Fig. 5, and the other flowsheets firmed by simple solvent extraction batch were simply slight modifications on this, The Purex process has been the basis distribution experiments, in which aqueous with small changes to parameters such as of nuclear fuel reprocessing schemes since and organic phases in a 1:1 ratio, contain- acidities, flowrates, reagent concentrations the 1950s and there is now substantial suc- ing Np(IV), Pu(IV) and (as required) U(VI) in and the number of stages in the U/Pu split, cessful operational experience at industrial the organic phase and the hydroxamic acid in order to optimise performance. scales coupled with a large historic knowl- phase are equilibrated [54][55]. Analysis of A detailed description of the results from edge base. However, further enhancements the metal ions in each phase after equili- these trials is beyond the scope of this cur- to the process are still of significant value, bration gives a distribution ratio (the ratio rent paper. However, the trials were very particularly in the context of developing of the concentration of the metal ion in the successful, demonstrating that excellent technologies for TRU actinide recycle in solvent to its concentration in the aqueous separation of U and Pu is achievable using AFCs and Generation IV reactors. phase). Lower distribution coefficients (D), AHA, even at very high Pu feeds; this is In particular, the separation of trivalent indicate better stripping of the metal ion illustrated in Table 2, which shows some actinides from the lanthanides has been a into the aqueous phase. To generalise, re- key data from the trials. The first trial ob- major challenge to aqueous processing, in- duced acidity and increased AHA (AHA:Pu tained a very high U DF on the Pu stream volving the design of new extractants ca- ratio) decrease DPu(IV) whereas increased (1.3E5), far in excess of what would be pable of exploiting the small differences Pu concentration increases DPu(IV). For an required for AFCs. This was therefore in covalency between 4f and 5f elements. example, Fig. 3 illustrates the relationship reduced in subsequent trials, although it Much progress has been made internation- between DPu and the AHA:Pu ratio. should be noted that even in the third trial, ally in demonstrating the technical feasibil- CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 903 CHIMIA 2005, 59, No. 12

ity of such new partitioning flowsheets. The development of advanced Purex processes that achieve U, Np, and Pu recov- ery within a single solvent extraction cycle, using centrifugal contactors for phase mix- ing and separation and that meet the criteria for AFCs set out in Table 1 is also a signifi- cant challenge. Simple hydroxamic acids have been shown to be extremely useful reagents in this context. Fig. 4. Image of the 28-stage miniature Received: August 18, 2005 centrifugal contactor rig, contained within a [1] P.D. Wilson, ‘The Nuclear Fuel Cycle’, Pu glovebox at Nexia Oxford University Press, Oxford, 1996. Solutions laboratories, [2] J.H. Miles, in ‘Science and Technology BNFL, Sellafield of ’ Volume III, Part II, Ed. W.W. Schulz, J.D. Navratil, CRC Press, Florida, 1990, p. 11. [3] G.R. Choppin, in ‘Chemical Separati- ons in Nuclear Waste Management’, Ed. G.R. Choppin, M.K. Khankhasayev, H.S. Plendl, Battelle Press, Columbus, 2002, p. 3. [4] H.A.C. McKay, J.H. Miles, J.L. Swanson, in ‘Science and Technology of Tributyl Phosphate’ Volume III, Part II, Ed. W.W. Schulz, J.D. Navratil, CRC Press, Florida, 1990, p. 1. [5] Nuclear Energy Agency, ‘Trends in the Nuclear Fuel Cycle’, OECD, Paris, 2001. [6] H. Takano, T. Ikegami, in ‘Actinide and Fission Product Partitioning and Trans- mutation, 7th Information Exchange Meeting, Jeju, Republic of Korea, 14–16 October, 2002’, Nuclear Energy Agency, OECD, Paris, 2003, p. 23. [7] http://gen-iv.ne.doe.gov/. [8] B. Savage, in ‘Proc. Int. Conf. Atalante 2004, Advances for Future Nuclear Fuels Cycles’, CEA, Nimes, 2004, Inv-06. [9] J.W. Herczeg, in ‘Actinide and Fission Product Partioning and Transmutation, Fig. 5. Example chemical separation flowsheet designed to separate U from Pu using acetohydroxamic 7th Information Exchange Meeting, Jeju, acid (7 wt.% Pu feed). [The nomenclature adopted is from conventional Purex solvent extraction Republic of Korea, 14–16 October, 2002’, cycles: HAF = Highly Active Feed; HA/HS = Highly Active extract and Scrub; AP/SP = Aqueous/ Nuclear Energy Agency, OECD, Paris, Solvent Product; A/S = Aqueous/ Solvent feed; URX = Uranium Re-eXtract; PuS = Plutonium Strip]. 2003, p. 37. [10] D. Warin, in ‘Actinide and Fission Product Partioning and Transmutation, 7th Infor- Table 2. U/Pu separation flowsheets: Key data from flowsheet trials (7, 20, and 40 wt. % Pu feeds, mation Exchange Meeting, Jeju, Republic respectively). of Korea, 14–16 October, 2002’, Nuclear Energy Agency, OECD, Paris, 2003, p. Trial 1 Trial 2 Trial 3 53. [11] G.R. Choppin, in ‘Chemical Separations [Pu] in HAF [g/l] 18 55.2 99.6 in Nuclear Waste Management’, Ed. G.R. [Pu] in HAR (g/l) 0.0041 0.0018 0.0017 Choppin, M.K. Khankhasayev, H.S. Plendl, Battelle Press, Columbus, 2002, p. 49. Pu % recovery >99.97 >99.995 >99.99 [12] J.J. Laidler, J.C. Bresee, in ‘Proc. Int. Conf. Atalante 2004, Advances for Fu- U % recovery >99.999 >99.999 >99.999 ture Nuclear Fuels Cycles’, CEA, Nimes, 2004, O11-01. U stream Pu DF 1000 60000 1.45E+06 [13] K. Nomura, A. Shibata, A. Aoshima, in ‘Proceedings of the International Sympo- Pu stream U DF 1.30E+05 5500 360 sium NUCEF 2001 – Scientific basis for criticality safety, separation process and Mass balance Pu [%] 107.7 97.8 97.1 waste disposal’, Ed. T. Banba, Y. Tsuba- Mass balance U [%] 98.4 101.4 99 ta, JAERI-Conf 2002-004, Tokai-Mura, 2002, p. 213. CHEMISTRY AND MATERIALS IN NUCLEAR POWER PRODUCTION 904 CHIMIA 2005, 59, No. 12

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