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Recent Developments in the Purex Process for Nuclear Fuel Reprocessing: Complexant Based Stripping for Uranium/Plutonium Separation

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Recent Developments in the Purex Process for Nuclear Fuel Reprocessing: Complexant Based Stripping for Uranium/Plutonium Separation 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 Nuclear Fuel Reprocessing: Complexant Based Stripping for Uranium/Plutonium 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 spent nuclear fuel 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 actinides 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 actinide 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 minor actinide 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 hydrazine 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) Sellafield 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] CHEMISTRY 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 Hanford site 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 lanthanides in solu- contacting equipment, e.g.
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