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A review on UREX processes for nuclear spent fuel reprocessing
Nuclear Engineering and Design 358 (2020) 110410
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A review on UREX processes for nuclear spent fuel reprocessing ⁎
T
Indu Kumari , B.V.R. Kumar, Ashok Khanna Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur, Uttar Pradesh 208016, India
A R T I C LE I N FO
A B S T R A C T
Keywords: UREX process Reprocessing Nuclear fuel
This paper presents a review on the UREX and variants of UREX + processes used for reprocessing of the spent nuclear fuels. The UREX + process consists of nine versions from which seven product streams are obtained. Uranium and technetium are recovered as separate products with the yield > 99% in the first stage of the UREX process. Cesium and strontium are recovered from the UREX waste using CCD-PEG or FPEX process. Plutonium and neptunium are separated using NPEX process with high level of impurity. The NPEX raffinate is treated in the TRUEX process from which minor actinides along with rare-earth elements are separated. Purification of minor actinides from the rare-earth elements is carried out in the Cyanex 301 process. The technologies used for each reprocessing process are described in detail. The flowsheets and process chemistry of all the versions of UREX processes are also reported and discussed in detail. The key issues related to UREX processes for reprocessing the spent nuclear fuels have been summarized. The strategy of the non-proliferation resistance using the advanced reprocessing process is described in the present work.
1. Introduction The spent fuel discharged from the nuclear reactors contains approximately 95% uranium, 4% fission products, 0.9% plutonium and 0.1% minor actinides. Uranium is a major contributor in the volume and mass in the spent nuclear fuel. Uranium being an energy resource can be used as nuclear fuel in the nuclear reactor after reprocessing and recycling it. Plutonium is also a resource of energy but at the same time it is a major contributor in heat loads and long term radiotoxicity. Plutonium cannot be reprocessed and recycled in pure form due to proliferation issues. Fission products cesium and strontium contribute to heat load and short term radiotoxicity to the spent nuclear fuel while technetium and iodine are contributors to the long term radiotoxicity. Minor actinides (americium, curium and neptunium) are also major contributor to the long term radiotoxicity. Uranium product obtained from UREX process can be used to manufacture MOX fuel for the light water reactor.” the above statement explains that MOX fuel is manufactured from extracted plutonium and uranium. Uranium can be depleted uranium or low-enriched uranium or it can also be recovered uranium. (Basak et al., 2010; Ghosh et al., 2009). Several methods are there to treat the spent nuclear fuel. Partitioning and transmutation of the spent nuclear fuel is one of these which can be used for the reduction of both volume as well as the radiotoxicity of the spent fuel. Nuclear fuel reprocessing is a chemical separation of spent nuclear fuel into its major components. The purpose of reprocessing the spent
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fuel is to reduce volume and radiotoxicity by recovering the useful elements and hence reduce the need of supplementary repositories (Law et al., 2007). It is used to recover actinides from spent nuclear fuel reducing the amount of nuclear waste. UREX is uranium and technetium extraction process developed under the US Advanced Fuel Cycle Initiative (AFCI) program (Srimok, 2011; Thompson et al., 2002). UREX is the advanced version of PUREX process in which plutonium remains in the raffinate along with other minor actinides and fission products. To account for proliferation issue acetohydroxyl amine is used as a complexing agent to prevent extraction of plutonium in the product stream. UREX is also used to recover fission products: iodine by volatilization and technetium by electrolysis. Main goal of the UREX process is to recover > 99.9% Uranium, > 95% Technetium in separate product streams (Srimok, 2011; Pruett, 1981). The UREX process has been demonstrated at the laboratory scale at the Savannah River National Laboratory using irradiated fuel from the Dresden BWR (Meikrantz et al., 2007). According to the demonstration, the U losses to the Tc and raffinate streams were less than 0.02%, and the Tc losses to the U stream were < 1.2% and losses to the raffinate were low. This demonstrated that > 95% of the Tc could be recovered. Loss of Pu and other actinides to the Tc and U product streams was < 0.02% in all tests, with > 99.98% going to the raffinate. Uranium product obtained from UREX process can be used to manufacture MOX fuel for the light water reactor. Technetium as separate product can either be used in transmutation or go for direct
Corresponding author. E-mail address: [email protected] (I. Kumari).
https://doi.org/10.1016/j.nucengdes.2019.110410 Received 26 June 2019; Received in revised form 24 October 2019; Accepted 25 October 2019 0029-5493/ © 2019 Elsevier B.V. All rights reserved.
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Nomenclature
solvent in U-strip section using dilute nitric acid (Thompson et al., 2002; Vandegrift et al., 2004).
TRU FP Ln
3. Versions of UREX process
Transuranic elements Fission products Lanthanide
The versions of UREX + processes provide many options for nonproliferation strategy. Plutonium product is either mixed with other products or present in non-extractable form to prevent its extraction. All the versions of the UREX + processes are listed in Table 1 along with the products stream and the combination of plutonium with other products (Np, U). The major advantages of the UREX + processes are following: (1) the efficiencies of U and Pu recovery are > 99.9%, (2) volume of the spent fuel reduced greatly by separating uranium from the spent nuclear fuel, (3) radiotoxicity and heat generation of the spent fuel decreases by the recovery of Tc and Cs/Sr respectively (Ghosh et al., 2009; Nash et al., 2011).
disposal. UREX raffinate can further be treated to separate the desired products like Cesium (Cs) and Strontium (Sr) in order to minimize the heat production from the waste (Srimok, 2011). Aim of this review paper is to describe UREX reprocessing process and its variants along with the process chemistry involved. This paper also presents advantages and limits of the processes. 2. Urex process The UREX process flowsheet as shown in Fig. 1 is divided into three parts: extraction, scrub and strip. Feed containing spent fuel dissolved in nitric acid is fed into the extraction section; contacted with fresh solvent 30% tri-butyl phosphate (TBP) dissolved in n-dodecane. Uranium and technetium are separated from the dissolved spent fuel in the extraction section. Complexant/reductant is added in the scrub section to form complex with the plutonium (Pu) and neptunium (Np) and hence prevent their extraction. Uranium and technetium is stripped off from the loaded solvent by using dilute nitric acid in the strip section. The raffinate containing transuranic (TRU) and all fission products (FP) except technetium becomes the feed for CCD-PEG process. The CCD-PEG stands for chlorinated cobalt dicarbollide (CCD) and polyethylene glycol (PEG) used for the cesium and strontium extraction will be discussed in detail CCD-PEG process section (Pereira et al., 2007). Low concentration of nitric acid is maintained in the feed and scrub sections to increase the complex formation of plutonium and neptunium and also to increase the extraction of pertechnetate ion (Vandegrift et al., 2004). Fig. 2 consists of extraction, scrub, U Re-extraction and U, Tc-strip sections describing separation of uranium and technetium from loaded solvent as well as from each other. Generally high concentration of nitric acid is used in the strip section to strip off uranium and technetium from loaded solvent. The loaded solvent obtained from the scrub section is stripped off of technetium in Tc strip section using high concentration of nitric acid. The Tc product is separated from uranium in the U-Re-Extraction section. Excess nitric acid is scrubbed off from loaded solvent before U-strip section. Uranium is recovered from the
3.1. UREx + 1 In UREX + 1 process uranium and technetium are extracted as separate streams. Cesium and strontium are extracted as combined stream. Transuranics and lanthanides are also separated together leaving fission product as separated product. 3.2. UREX + 1a Fig. 3 represents schematic of UREX + 1a process. The aim of the UREX + 1a is > 99.9% recovery of the transuranics (TRUs), > 95% recovery of uranium, > 95% recovery of technetium, 99% recovery of cesium and strontium and only 0.1% of plutonium and 1% of cesium and strontium left in the raffinate (Pereira et al., 2007). Main characteristics of UREX + 1a process are that plutonium and curium are kept together. Curium emits more neutrons than plutonium which prevents measurement of plutonium. Plutonium product obtained from UREX + 1a is impure (Feener, 2010). UREX + 1a process consists of four solvent extraction processes:(a) extraction of uranium and technetium (UREX), (b) separation of uranium from technetium (ion exchange), (c) extraction of cesium and strontium (CCD-PEG), (d) separation of fission products from lanthanides and TRU (TRUEX) and (e) separation of lanthanide fission products (TALSPEAK) and (f) separation of americium, curium, neptunium and plutonium (TALSPEAK) (Pereira et al., 2007). Spent fuel from the reactor is stored in the spent fuel storage. After breaking in small pieces, voloxidation (heating of the
Fig. 1. UREX process flowsheet (Pereira et al., 2007). 2
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Fig. 2. UREX process flowsheet (Vandegrift, et al., 2004; Nash et al., 2011). Table 1 Versions of UREX processes. Process
Product 1
Product 2
Product 3
Product 4
Product 5
Product 6
Product 7
UREX + 1 UREX + 1a UREX + 1b UREX + 2 UREX + 2a UREX + 3 UREX + 3a UREX + 4 UREX + 4a
U U U U U U U U U
Tc Tc Tc Tc Tc Tc Tc Tc Tc
Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr
TRU/Ln TRU U/TRU Pu/Np U/Pu/Np Pu/Np U/Pu/Np Pu/Np U/Pu/Np
FP FP/Ln FP/Ln Am/Cm/Ln Am/Cm/Ln Am/Cm Am/Cm Am Am
FP FP FP/Ln FP/Ln Cm Cm
FP/Ln FP/Ln
Notes: TRU = Transuranic elements: Am, Cm, Np, Pu. FP = Fission products. Ln = Lanthanide fission products.
Fig. 3. UREX + 1a process flowsheet (Feener, 2010). 3
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industries. Tributyl phosphate is used as extractant. In the UREX + 3 process uranium is removed and plutonium and neptunium are separated together, (ii) minor actinides, americium and curium are separated using di-2-ethyl-hexyl phosphoric acid and (iii) partition of cesium and strontium from fission product waste (Collins, 2005). Plutonium/neptunium product recovered in the first and americium/ curium in the second separation process. Americium and curium are separated from the lanthanides leaving fission products and lanthanides together in the waste. Fig. 7 represents flowsheet for urex + 3 process in detail (Ghosh et al., 2009).
fuel in oxygen rich environment) is done to decrease the size of the fuel particles and separate it from the cladding. The fuel is diluted in nitric acid in the dissolution step to separate from the hulls and cladding. The remaining cladding and the hulls are stored in the waste storage. The dissolved fuel enters the first step of solvent extraction, where it contacts the organic extractant. Uranium and technetium are extracted in the first step of the solvent extraction leaving everything else in the raffinate. Cesium and strontium are extracted in the Chlorinated Cobalt Dicarbollide and Poly-ethylene Glycol (CCD-PEG) step leaving behind the transuranics (TRUs) and the fission products. The waste obtained from CCD-PEG is treated in Transuranic Extraction (TRUEX) step. Transuranics and lanthanides are extracted in the TRUEX step and the fission product goes into the raffinate (Pereira, et al., 2007; Feener, 2010). Lanthanides are separated from transuranics in the Trivalent Actinide Lanthanide Separation by Phosphorus Extractants and Aqueous Komplexes (TALSPEAK) step. TRU goes into the raffinate stream of the TALSPEAK step. TRUs and plutonium are solidified and packaged and then sent for storage. Partitioning of uranium and technetium is done after the UREX step. Uranium is converted into uranium oxide followed by evaporation and then goes to the storage. The waste products (Cs, Tc, Sr, FP and lanthanides) obtained from all the process steps are packaged in the solid form for further shipment (Feener, 2010).
3.6. UREX + 3a Fig. 8 represents flowsheet of the UREX + 3a consisting of five extraction processes (UREX, FPEX, NPEX, TRUEX and TALSPEAK) and one ion exchange process. Technetium and uranium are extracted from dissolved spent fuel in the UREX section. Uranium is separated from technetium using ion exchange process. The raffinate from UREX is fed to FPEX (Fission Product Extraction) segment leading cesium and strontium extraction along with Rb and Ba. The NPEX (Neptunium Plutonium Extraction) segment makes the UREX + 3a process different from the UREX + 3. Uranium, neptunium and plutonium are extracted all together in NPEX process. The raffinate from the NPEX becomes feed for the TRUEX (Transuranic Extraction) (Goddard, 2010; Goddard et al., 2010).
3.3. UREX + 1b Pyroprocessing is grouped with UREX + 1b since both produce plutonium that is not separated from uranium, americium, neptunium, and curium in the final product (Bari et al., 2009). Recovery of traces of uranium along with the transuranic elements differentiates UREX + 1b from UREX + 1a.
3.7. UREx + 4 Americium is separated from curium using UREX + 4 processes. The main characteristics of UREX + 4 processes are that americium is burnt in the specially designed target assemblies. The target fabrication does not require shielding after separation of americium from curium (Nash et al., 2011).
3.4. UREx + 2 The process flowsheet of the UREX + 2 as shown in Fig. 4 consists of three solvent extraction and one ion exchange processes: (a) extraction of plutonium and neptunium in the co-extraction section, (b) extraction of technetium and uranium in co-extraction, (c) extraction of cesium and strontium using CCD-PEG and (d) recovery of uranium from technetium in ion exchange process. The UREX + 2 process flowsheet was developed using AMUSE (Argonne Model for Universal Solvent Extraction) code at Argonne National Laboratories and Idaho National Engineering and Environmental Laboratories (Pereira et al., 2005). The dissolved fuel is introduced in the co-extraction section consisting of extraction, scrub and strip sections. The co-extraction process flowsheet has been shown in Fig. 5. Uranium, plutonium, neptunium and technetium are recovered from the dissolved fuel using 30% TBP in n-dodecane. The extract (U, Tc, Np, and Pu) obtained from extraction/Scrub section is fed to Np/Pu-strip section where re-extraction of uranium and technetium is done using the solvent 30% TBP in n-dodecane. Complexant/reductant is used in extraction section to prevent extraction of neptunium and plutonium by forming complexes or reducing to nonextractable form. Uranium and technetium are stripped off using dilute nitric acid in U/Tc strip section. The raffinate of co-extraction section is fed to CCD-PEG section after reducing nitric acid concentration. Cesium and strontium are extracted in CCD-PEG section using the combined solvent Chlorinated Cobalt Dicarbollide and Poly-ethylene Glycol. The raffinate of CCD-PEG section contains TRU and FP which are separated from each other in the An/Ln separation section. Ion exchange method as described previously is used for separation of uranium from technetium (Nash et al., 2011; Pereira et al., 2005).
4. Process chemistry Uranium and technetium are the major products of UREX process. Complexant/reductant are used to form complexes with neptunium and plutonium preventing their extraction into the product stream (Collins, 2005). The complex formation of neptunium and plutonium and extraction of technetium can be increased by using low concentration of nitric acid both in the feed as well as in the scrub. However, highly concentrated nitric acid is used for stripping off technetium from the solvent. Uranium is separated from the solvent using dilute nitric acid (Pereira et al., 2007; Vandegrift et al., 2004). The actinides with +6 and +4 oxidation states form complexes with TBP and hence can easily be extracted into organic phase. TBP has lower affinity with +5 and almost nil for +3 and lower oxidation states (Am (III), Cm (III), Pu (III), Cs(I) and Sr (II)) (Nash et al., 2011). 4.1. Uranium and technetium separation Technetium is a long-lived (half-life 214,000 y) radioisotope present
3.5. UREx + 3 UREX + 3 process flowsheet shown in Fig. 6 consists of three parts: (i) first part is modified version of solvent extraction process used in the
Fig. 4. Schematic diagram of UREX + 2 process. 4
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Fig. 5. Co-extraction process segment flowsheet (Nash et al., 2011).
4.2. Recovery of cesium and strontium
in significant quantities in the spent nuclear fuel. Technetium as pertechnetate ion is highly mobile and soluble in the environment. This is a matter of concern for disposal of technetium in repository (Campbell, 1963; Poineau et al., 2008). Separation of technetium from the spent fuel can resolve these issues. Uranium and technetium are co-extracted from the dissolved spent nuclear fuel using TBP in n-dodecane in UREX process. Pu(IV), Np(IV) and Np(VI) present in the solution can readily be extracted by TBP. However Np(V) is not extractable by TBP. Acetohydroxamic acid (AHA) is used as reducing/complexijng agent to prevent extraction of Pu(IV), Np(IV) and Np(VI). AHA reduces Np(VI) to non-extractable Np(V). AHA prevents extraction of Np(IV) and Pu (IV) in the organic phase by forming complexes with them. Technetium is separated from the U/Tc product by using ReillexTM HPQ resin in ion exchange process. Tc is recovered as pertechnetate ion, TcO4-. Pertechnetate ion is extracted by forming complex with the TBP (Chung et al., 2007). Thus uranium and technetium are extracted by TBP/ndodecane leaving fission products and transuranics in the raffinate.
Recovery of cesium and strontium from the spent nuclear fuel minimizes short term heat load from the spent fuel. Cesium and strontium can be extracted simultaneously using either CCD-PEG or FPEX solvent extraction process. CCD-PEG is more effective for the feed containing < 1 M nitric acid. The solvent Chlorinated Cobalt Dicarbollide (CCD) is used for cesium separation and Poly Ethylene Glycol (PEG) for strontium recovery. The solvent used is the mixture of CCD and PEG diluted in phenyltrifluoromethyl sulfone (FS-13). Strip solution containing guanidine carbonate/diethylenetriamine pentaacetic acid (DTPA) is used to strip off cesium and strontium effectively from the solvent (Law et al., 2005). The second process FPEX solvent extraction is used for combined extraction of cesium and strontium. The solvent used in FPEX is the combination of calix[4]arene-bis-(tert-octylbenzo-crown-6)(BOBCalixC6),4′,4′ (S')-Di-(t-butyldicyclo-hexano)18-crown-6 (DtBuCH18C6) and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4sec-butylphenoxy)-2-propanol (Cs-7SB) diluted in Isopar-L. BOBCalixC6
Fig. 6. UREX + 3 process flowsheet (Collins, 2005). 5
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Fig. 7. UREX + 3 process flowsheet (Ghosh et al., 2009).
Fig. 8. High-level process flow diagram for UREX + 3a (Goddard, 2009, Goddard et al., 2010).
The process flowsheet of FPEX (Fission Product Extraction) is shown in Fig. 10 for the combined extraction of Cs and Sr. FPEX process recovers > 99.9% Cs and Sr by leaving < 100 nCi/g TRU in the strip solution (Law et al., 2007). Both CCD-PEG and FPEX processes can be used directly on UREX raffinate for recovery of cesium and strontium. Simultaneous recovery of cesium and strontium results in single product facilitating easy storage and reduction of overall process complexicity (Nash et al., 2011; Law et al., 2005). The CCD-PEG is more mature process but the solvent used in the FPEX process is more compatible with the other processes which are based on n-dodecane. The raffinate from the TRUEX process can also be treated using either CCDPEG or FPEX process rather than the UREX raffinate (Nash et al., 2011).
and DtBuCH18C6 are extractants; Cs-7SB is used as a phase modifier. The extractant BOBCalixC6 is responsible for cesium extraction and DtBuCH18C6 is for strontium. The above mixed solvent can have composition of is 0.007 M BOBcalixC6, 0.15 M DtBuCH18C6, 0.75 M Cs-7SB in Isopar-L (Nash et al., 2011; Law et al., 2005, 2006; Meikrantz, et al., 2005; Todd, 2005; Mincher et al., 2006). The process flowsheet of CCD-PEG is shown in Fig. 9. The raffinate obtained from UREX process contains fission products and transuranics. This is fed to CCD-PEG process separate cesium and strontium. CCD-PEG process shown in Fig. 9 consists of four sections: extraction, scrub, strip and wash. Cesium and strontium along with Rb and Ba are extracted in the extraction section. Scrubbing of transuranics and rare earths from the solvent are done in the scrub section using moderate concentration of nitric acid. Ammonium carbonate salt together with a complexing agent is used to strip off the alkaline-earth cations and alkali in the strip section. Since the solvent supply is limited solvent wash is done to recycle the solvent (Pereira et al., 2007).
4.3. Recovery of plutonium and neptunium Plutonium and neptunium are extracted from the spent fuel using NPEX process as shown in Fig. 11. NPEX process is utilized either on 6
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Fig. 9. CCD-PEG process flowsheet (Pereira et al., 2007).
Fig. 10. FPEX process flowsheet (Nash et al., 2011; Law et al., 2007).
Raffinate obtained from solvent extraction process (CCD-PEG raffinate) is fed into NPEX extraction section where it contacts with the organic solvent TBP in n-dodecane. Nitric acid scrub is used to remove the impurities (Americium/ Curium, Fission Products and Lanthanides) in the scrub section. This impurity goes in NPEX raffinate stream. Same
CCD-PEG raffinate or FPEX raffinate. Before following NPEX process, feed adjustment should be done to make plutonium and neptunium in the extractable form, concentration of nitric acid should be increased and to eliminate complexant and reductant present in the feed (Nash et al., 2011).
Fig. 11. NPEX process flowsheet (Hodges, 2006). 7
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diisobutylcarboylmethyl-phosphine oxide) and TBP in n-dodecane for the extraction of americium, curium, the rare earths and residual plutonium and neptunium (Ghosh et al., 2009; Vandegrift et al., 2004; Bond et al., 1987; Arai et al., 1997; Herbst et al., 2000). The procedure of scrub section is the same as discussed the above section. The rare earths and the actinides are stripped off from the solvent using weak complexant salt. The weak solvent is used to keep the pH values in the range of 3–4 for recovery of actinides using Cyanex 301 solvent (Aneheim et al., 1985, 2011, 2012). If the TRUEX process treats NPEX raffinate, the resulting product would be Am-Cm-Ln. However if the NPEX raffinate is not treated in the TRUEX process, the product obtained would be Pu-Np-TRUs (Nash et al., 2011). Cyanex 301 flowsheet is shown in Fig. 13. The product Cyanex 301 (chemical name: bis (2, 4, 4-trimethylpentyl) dithiophosphinic acid) is supplied by Cytec Industries, Canada in the impure form. The purification of Cyanex 301 is done by the suggested method of Zhu et al. The product obtained from the TRUEX process is fed directly in the extraction section of the Cyanex process. Solvent containing purified 301 diluted in TBP/n-dodecane is used for separation of actinide/lanthanide. The impurities are separated using weak complexant in the acid. Stripping of americium and curium from the solvent is done using ammonium salt (Vandegrift et al., 2004). The drawback of the use of Cyanex 301 is that it decomposes both in solution as well as in solid. Therefore, Cyanex 301 is not recommended for separation of actinides and lanthanides in the industries and that leads to the use of TALSPEAK process (Nash et al., 2011). Actinide and lanthanide are recovered using TALSPEAK process. HDEHP (bis(2-ethylhexyl)phosphoric acid) in n-dodecane is used as solvent in the TALSPEAK. However the aqueous phase contains buffered solution of lactate and DTPA (Nash et al., 2011). The solvent containing lactate buffer and a complexant is the feed for the TALSPEAK process. The TALSPEAK process flowsheet as shown in Fig. 14 contains three sections: extraction, scrub and strip. The actinides are separated by forming complex with aminopolyacetic acid. The trivalent lanthanides are extracted by bis (2-ethylhexyl) phosphoric acid (HDEHP). Weaker complex of lanthanides are extracted. However, stronger complexes of actinides remain in the aqueous phase. Fractions of Am and Cm are removed by the scrub which are extracted by HDEHP. The lanthanides are removed from the solvent in the strip section using concentrated nitric acid (Pereira et al., 2007; Nash et al., 2011; Croff et al., 2008).
complexant/reductant that is used in UREX process is used to strip off plutonium and neptunium from the organic stream in the Pu/Np section. The solvent used in the process is washed using sodium bicarbonate and rinsed using nitric acid in the solvent wash section (Hodges, 2006; Mitchell, 2008). U-Pu-Np product is obtained by adding uranium to the recovered plutonium and neptunium in the strip section (Nash et al., 2011; Hodges, 2006; Cheng et al., 2006).
4.4. Extraction of transuranics Pu(IV), Am(III), Cm(III) and Np(VI) are extracted by 6,6′ Bis (5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-1,2,4-triazin-3-yl)-2,2′[bipyridine] (CyMe4-BTBP) and TBP dissolved in cyclohexanone (Thompson et al., 2002; Pruett, 1981; Aneheim et al., 1985, 2011, 2012). Trivalent and pentavalent actinides are separated from trivalent lanthanides by using CyMe4-BTBP. However, TBP can be used to extract tetravalent and hexavalent actinides. CyMe4-BTBP does not react with TBP. Therefore, they can be combined into one solvent. Thus, CyMe4-BTBP and TBP together is used for the recovery of trivalent, tetravalent, pentavalent and hexavalent actinides (Aneheim et al., 1985, 2011, 2012; Thompson et al., 2002; Vandegrift et al., 2004; Horwitz et al., 1985; Bond et al., 1987). TRUEX (TRansUranium EXtraction) process is used to separate transuranics and the rare earth elements from the CCD-PEG raffinate (Pereira et al., 2007). Radio-toxicity of the waste can be minimized significantly by the extraction of transuranics (Ghosh et al., 2009). TRUEX process flowsheet shown in Fig. 12 consists of five sections: one extraction, three scrubs and one strip section. The raffinate obtained from the CCD-PEG process becomes feed for the extraction section of the TRUEX process. Concentrated nitric acid is used to adjust composition of the feed. Reducing agent is added to the feed to reduce Np(V) to extractable Np(IV). Impurities of the solvent are removed in the first scrub section using oxalic acid. Scrubbing of oxalic acid is done in the second scrub section using the moderate concentration of nitric acid. For effective stripping, acid concentration of the solvent is lowered by the addition of dilute nitric acid in the third scrub section. Lactate buffer along with a complexant are used for stripping off the rare earths and the actinides from the solvent (Pereira et al., 2007). However, raffinate from NPEX process can be fed directly to TRUEX process without doing feed adjustment. The process flowsheet of TRUEX process using the feed from NPEX raffinate is shown in Fig. 13. The solvent used in this process is CMPO (octyl(phenyl)-N,N-
Fig. 12. TRUEX process flowsheet (Pereira et al., 2007). 8
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Fig. 13. TRUEX and Cyanex process flowsheet (Pereira et al., 2007).
5. Discussion
exchange process. Plutonium and neptunium can be extracted from the products stream of third stage using NPEX process. These processes have some advantages and disadvantages also which are being discussed in the subsequent sections.
The desired components are extracted efficiently using the all the versions of the UREX + process. The entire UREX + processes can be generalized as following. All versions of the UREX + processes can be performed in three major stages. In the first stage uranium and technetium can be recovered using the solvent TBP in n-dodecane and reducing/complexing agent acetohydroxamic acid (AHA) leaving everything else in the raffinate. Since Pu(IV), Np(IV) and Np(VI) present in the spent nuclear fuel can readily be extracted by TBP. Therefore, AHA is used here as a reducing/complexing agent which prevents extraction of Pu(IV), Np(IV) and Np(VI) by reduction of Np(VI) to non-extractable Np(V) and forming complexes with Np(IV) and Pu(IV). Cesium and strontium are recovered from the raffinate of first stage using the combined solvent of calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6), 4′,4′ (S')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (Cs-7SB) diluted in Isopar-L and rest of the materials go in the raffinate. Pu(IV), Am(III), Cm(III) and Np(VI) are extracted together in the third stage by 6,6′ Bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-1,2,4-triazin-3-yl)-2,2′-[bipyridine] (CyMe4-BTBP) and TBP dissolved in cyclohexanone leaving trace amount of the fission products and lanthanides in the raffinate. The combined products stream obtained from the above three major stages can be treated to separate from each other using suitable processes. Technetium is separated from the U/Tc product by using ReillexTM HPQ resin in ion
5.1. Advantages of UREX + processes The advantage of using UREX + process is that it eliminates the risk related to the spent nuclear fuel; waste to be disposed of may be less radioactive, less HLW repositories required, risks due to leakage of mobile elements, decay heat and radiation hazards from the spent nuclear fuel will be less, fissile and fertile elements (U, Pu) can be better used as resources, mining of uranium and the associated risks may be reduced (Cordfunke, 1995). The typical UREX process is used to separate uranium and technetium leading to volume reduction. Application of UREX + 3 flowsheet minimizes the heat-generating product in the waste hence maximizes the lifetime of high level waste repository. Mass of the spent nuclear fuel is minimized by recycling. Proliferation resistance increases (Collins, 2005). Recovery of Uranium reduces the volume and dose rate of the spent fuel. Technetium is a long-lived FP; extraction of this leads to reduction of long-term dose of the spent fuel. Cesium and strontium having short half-lives of approximately 30 years are highly radioactive and heat generating elements which can reduce the mechanical strength of the long-term storage (Wu et al., 2016). Cesium and strontium are highly soluble in water. Due to this they easily spread in the environment and can be very hazardous to plants
Fig. 14. TALSPEAK process flowsheet (Pereira et al., 2007; Nash et al., 2011). 9
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unit equipment cost has been multiplied by number of equipment and adding all the cost gives total cost of UREX process. It is clear from Table 2 that Equipment cost for UREX process is minimum and for it is maximum for UREX + 3a process because product streams obtained in UREX process lesser than UREX + 3a process. Separation of more products increases the cost of the process.
and animals. Ingestion of these by the human can be major health hazards. Cesium and strontium are contributor to short-term radiotoxicity and heat load to the spent nuclear fuel. Recovery of cesium and strontium from the spent nuclear fuel minimizes short-term heat load and radiotoxicity of the spent fuel. Transuranic elements (Pu, Np, Am and Cm) having very long half-lives are alpha-emitting isotopes. Decay time of these elements is much longer. Therefore, problems of radioactive hazards due to these isotopes are long-term problem. Recovery of these elements reduces long-term dose-rate and toxicity from the spent fuel (Nash et al., 2011). Plutonium having long half-life is a major contributor to long-term heat-load and radio-toxicity to spent nuclear fuel. Plutonium can be used as a source of energy but due to proliferation issues it cannot be separated as pure Pu. Therefore Pu is recovered with minor actinides. Minor actinides neptunium, americium, and curium are major contributors to long-term heat-load and radiotoxicity of the spent nuclear fuel. Plutonium and neptunium recovered in NPEX process are highly pure which can be used in MOX fuel (Abdallah et al., 2019; Vandegrift, et al., 2004).
6. Conclusions UREX processes are more versatile than the PUREX due to proliferation issues. In UREX + 1 plutonium after forming complex with acetohydroxamic acid remains in the raffinate along with the other components. Reduction and complexation make recovery of plutonium impossible. Cesium and strontium are extracted profitably > 99.85 using CCD-PEG and FPEX techniques. Solvent used in CCD-PEG is mixture of Chlorinated Cobalt Dicarbollide and Poly Ethylene Glycol. Stripping of Cs and Sr from the solvent is done with guanidine carbonate/diethylenetriamine pentaacetic acid (DTPA). Cesium and strontium are extracted simultaneously by FPEX process using the combined solvent of DtBuCH18C6, BOBCalixC6 and Cs-7SB dissolved Isopar-L. The technique used in NPEX is feasible for the extraction of both Pu-Np and U-Pu-Np product using TBP in n-dodecane. Transuranics (Pu(IV), Am(III), Cm(III) and Np(VI)) are recovered by CyMe4-BTBP and TBP dissolved in cyclohexanone. Uranium (> 99.9%) and technetium (> 95%) are recovered as separate product in all versions of UREX + processes. Recovery of uranium from the spent fuel reduces its volume for the disposal. Recovery of technetium reduces long-term dose. Cesium and strontium separation minimizes repository heat load. TRUEX and TALSPEAK processes are used for extraction of actinides from lanthanides at high purities. Recovery of transuranics from the spent nuclear fuel results in reduction of long-term radiotoxicity. Recovery of plutonium, americium and lanthanides meet the goal of > 99.99%, > 99.97% and < 0.03% respectively. Equipment cost for UREX + processes depends upon the number of product streams.
5.2. Key issues of UREX + processes The complexity of the UREX + processes increases when the separation of desired or waste products increases. Due to increase in the processes complexicity cost of the processes also increases. The overall cost of the energy production using nuclear reactor will increase. Additional costs (cost of partitioning and fuel fabrication, equipment cost, piping cost, labour and land cost) and risks of operation (leakage, handling and transport of hazardous materials) of reactors used in recovery processes will increase (Cordfunke, 1995). Table 2 represents estimated cost (in Rupees) of equipment used in the UREX + process flow sheet for pilot plant. All the data for estimating the equipment cost have been taken from online suppliers (https://www.indiamart.com). Cost of the piping, materials and operation have not been included. Cost estimation can be understood by taking an example. UREX flow sheet consists of two extraction columns, two scrubbers, two strippers, one dissolver and four storage tanks. Per Table 2 Equipment cost of UREX + processes. Process
Storagea
Cost (Rupees) Feed Dissolverb
Extractorc
Scrubberd
Stripperd
UREX
75,000*4
85,000
1.5 Lakh*2
1.0 Lakh*2
1.0 Lakh*2
UREX + 1a CCD-PEG TRUEX TALSPEAK UREX + 2 CCD-PEG
75,000*6 75,000*4 75,000*3 75,000*3 75,000*6 75,000*4
85,000
1.5 1.5 1.5 1.5 1.5 1.5
Lakh*1 Lakh*1 Lakh*1 Lakh*1 Lakh*2 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*1 Lakh*3 Lakh*1 Lakh*3 Lakh*1
1.0 1.0 1.0 1.0 1.0 1.0
Lakh*1 Lakh*1 Lakh*1 Lakh*1 Lakh*2 Lakh*1
UREX + 3 UREX TRUEX FPEX TALSPEAK
75,000*4 75,000*3 75,000*3 75,000*3
85,000
1.5 1.5 1.5 1.5
Lakh*2 Lakh*1 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0
Lakh*2 Lakh*3 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*1
UREX + 3a UREX FPEX NPEX TRUEX TALSPEAK
75,000*4 75,000*3 75,000*4 75,000*3 75,000*3
85,000
1.5 1.5 1.5 1.5 1.5
Lakh*2 Lakh*1 Lakh*1 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*3 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*1 Lakh*1
a b c d e f g
85,000
Evaporatore
Washera
Ion-Exchangerf
Separatorg
Total (Rs.) 1,085,000
1,000
2,936,000 75,000
5,000
50,000
2,115,000
75,000 3,245,000
75,000 5,000
Karadani Engineering Pvt Ltd, Ahmedabad, Gujarat, India. Krish Pvt Ltd. Trident Labortek, Mumbai, Thane, India. Yansons Private Limited, India. KEP Engineering Services Private Limited. Thermochem Corporation Private Limited, Bengaluru, Karnataka. Marine Craft Exports. 10
75,000 75,000
3,890,000
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I. Kumari, et al.
References
monitoring during nuclear fuel reprocessing using the UREX+3a method. Nucl. Eng. Des. 240, 3904. Herbst, R.S., et al., 2000. Integrated AMP-PAN, TRUEX, and SREX flowsheet test to remove cesium, surrogate actinide elements, and strontium from INEEL tank waste using sorbent columns and centrifugal contactors. Idaho National Engineering and Environment Laboratory INEEL/EXT-2000-00001. Hodges, M., 2006. Development of an integrated process simulation system model for spent fuel treatment facility (SFTF) design, Master of Science. Department of Mechanical Engineering, Howard R. Hughes College of Engineering, University of Nevada, Las Vegas. Horwitz, E.P., et al., 1985. The TRUEX process - a process for the extraction of the transuranic elements from nitric acid in wastes utilizing modified PUREX solvent. Solvent Extr. Ion Exch. 3, 75. Law, J.D., et al., 2005. AIChE Spring National Meeting. Idaho National Laboratory, U.S. NL/CON-05-00087. Law, J.D., et al., 2006. WM'06 Conference. Tucson, AZ. Law, J.D., et al., 2007. 11th International conference on environmental remediation and radioactive waste management, ICEM2007, Bruges, Belgium. Meikrantz, D.H., et al., 2007. US7524469 B2, US 11/859,453. Meikrantz, H., et al., 2005. WO2006016892 A3, 28 Jan 2005. Mincher, B.J., et al., 2006. 9th Information Exchange Meeting for Actinide and Fission Product Partitioning and Transmutation And Seminar. University of Manchester. Mitchell, J.A., 2008. Removing acetic acid from the UREX+ Process. The University of Tennessee, Knoxville Master of Science. Nash, K.L., Lumetta, G.J., 2011. Advanced Separation Techniques for Nuclear Fuel Reprocessing and Radioactive Waste Treatment. Woodhead Publishing, Oxford Cambridge Philadelphia. Pereira, C., et al., 2005. WM’05 Conference. Tucson, AZ. Pereira, C., et al., 2007. Lab-scale demonstration of the UREX+1a process using spent fuel, WM’07 Symposium, Tucson, AZ. Poineau, F., et al., 2008. Uranium/technetium separation for the UREX process –synthesis and characterization of solid reprocessing forms. Radiochim. Acta 96, 52. Pruett, D.J., 1981. Symposium on separation science and technology for energy applications, Gatlinburg, Tennessee. Srimok, B., 2011. Investigation of Nuclear Waste Management for Advanced Fuel Cycles. Ph.D. thesis. North Carolina State University, Raleigh, North Carolina. Thompson, M.C., et al., 2002. WSRC-TR-2002-00444, http://www.osti.gov/bridge. Todd, T.A., 2005. WM’05 Conference: Tucson, AZ. Vandegrift, G.F., et al., 2004. Designing and demonstration of the UREX+ process using Spent nuclear fuel, ATATLANTE, Advances for future nuclear fuel cycles International Conference, Nimes, France. Wu, Y., et al., 2016. Simultaneous separation and recovery of Cs(I) and Sr(II) using a hybrid macrocyclic compounds loaded adsorbent. Kinetic, equilibrium and dynamic adsorption studies. J. Nucl. Sci. Technol. 53, 1968.
Abdallah, et al., Closing the nuclear fuel cycle: reprocessing, Energy and energy policy. https://www.indiamart.com. Aneheim, E., et al., 1985. Solvent Extr. Ion Exch. 28, 437. Aneheim, E., et al., 2011. A TBP/BTBP-based GANEX separation process-part 2: ageing, hydrolytic, and radiolytic stability. Solvent Extr. Ion Exch. 29, 157. Aneheim, E.H.K., 2012. Development of a solvent extraction process for group actinide recovery from used nuclear fuel. Ph.D thesis. Chalmers University of Technology, Gothenburg, Sweden, Department of Chemical and Biological Engineering. Arai, K., et al., 1997. Modified TRUEX process for the treatment of high-level liquid waste. J. Nucl. Sci. Technol. 34, 521. Bari, R., et al., 2009. Institute of Materials Management (INMM) 50th Annual Meeting. Tucson, Arizona, USA. Basak, U., Dyck G.R., 2010. IAEA activities on assessment of partitioning processes for transmutation of actinides, proceedings, first ACSEPT international workshop, Lisbon, Portugal. Bond, W.D., et al., 1987. ORNL review of TRUEX flowsheet proposed for development at the Rockwell Hanford Plutonium Finishing Plant. ORNL/TM-10334. Campbell, M.H., 1963. Separation of technetium from mixed fission products by solvent extraction with tributyl phosphate. Anal. Chem. 35, 2052. Cheng, M., et al., 2006. Proceedings, IMECE2006. ASME International Mechanical Engineering Congress and Exposition, Chicago, Illinois, USA. Chung, D.Y., et al., 2007. Transactions of the Korean Nuclear Society Autumn Meeting Pyeong Chang, Korea. Collins, E.D., 2005. Development of the UREX+3 flowsheet-an advanced separations process for spent fuel processing. American Nuclear Society Winter Meeting, Oak Ridge National Laboratory. Cordfunke, E.H.P., 1995.Transmutation of nuclear waste, status report. RAS programme 1994: recycling and transmutation of actinides and fission products, ECN-R-95-025. Croff, A.G., et al., 2008. Background, status, and issues related to the regulation of advanced spent nuclear fuel recycle facilities. Advisory Committee on Nuclear Waste and Materials White Paper, NUREG-1909. Feener, J.S., 2010. Safeguards for the uranium extraction (Urex) +1a process, Master of Science, Nuclear Engineering, Texas A&M University, TAMU-2010-05-270. Ghosh, Tushar K., Prelas, Mark A., 2009. Energy Resources and Systems. Springer Netherlands, Dordrecht. Goddard, B., 2009. Development of a real-time detection strategy for material accountancyand process monitoring during nuclear fuel reprocessing using the urex +3a method. Master of Science. Texas A&M University. Goddard, B., 2010. Development of a Real-Time Detection Strategy for Material Accountancy and Process Monitoring during Nuclear Fuel Reprocessing Using the UREX+3A Method. Master of Science. A&M University, Texas. Goddard, B., et al., 2010. Development of a real-time detection strategy for process
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Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
A review on UREX processes for nuclear spent fuel reprocessing ⁎
T
Indu Kumari , B.V.R. Kumar, Ashok Khanna Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur, Uttar Pradesh 208016, India
A R T I C LE I N FO
A B S T R A C T
Keywords: UREX process Reprocessing Nuclear fuel
This paper presents a review on the UREX and variants of UREX + processes used for reprocessing of the spent nuclear fuels. The UREX + process consists of nine versions from which seven product streams are obtained. Uranium and technetium are recovered as separate products with the yield > 99% in the first stage of the UREX process. Cesium and strontium are recovered from the UREX waste using CCD-PEG or FPEX process. Plutonium and neptunium are separated using NPEX process with high level of impurity. The NPEX raffinate is treated in the TRUEX process from which minor actinides along with rare-earth elements are separated. Purification of minor actinides from the rare-earth elements is carried out in the Cyanex 301 process. The technologies used for each reprocessing process are described in detail. The flowsheets and process chemistry of all the versions of UREX processes are also reported and discussed in detail. The key issues related to UREX processes for reprocessing the spent nuclear fuels have been summarized. The strategy of the non-proliferation resistance using the advanced reprocessing process is described in the present work.
1. Introduction The spent fuel discharged from the nuclear reactors contains approximately 95% uranium, 4% fission products, 0.9% plutonium and 0.1% minor actinides. Uranium is a major contributor in the volume and mass in the spent nuclear fuel. Uranium being an energy resource can be used as nuclear fuel in the nuclear reactor after reprocessing and recycling it. Plutonium is also a resource of energy but at the same time it is a major contributor in heat loads and long term radiotoxicity. Plutonium cannot be reprocessed and recycled in pure form due to proliferation issues. Fission products cesium and strontium contribute to heat load and short term radiotoxicity to the spent nuclear fuel while technetium and iodine are contributors to the long term radiotoxicity. Minor actinides (americium, curium and neptunium) are also major contributor to the long term radiotoxicity. Uranium product obtained from UREX process can be used to manufacture MOX fuel for the light water reactor.” the above statement explains that MOX fuel is manufactured from extracted plutonium and uranium. Uranium can be depleted uranium or low-enriched uranium or it can also be recovered uranium. (Basak et al., 2010; Ghosh et al., 2009). Several methods are there to treat the spent nuclear fuel. Partitioning and transmutation of the spent nuclear fuel is one of these which can be used for the reduction of both volume as well as the radiotoxicity of the spent fuel. Nuclear fuel reprocessing is a chemical separation of spent nuclear fuel into its major components. The purpose of reprocessing the spent
⁎
fuel is to reduce volume and radiotoxicity by recovering the useful elements and hence reduce the need of supplementary repositories (Law et al., 2007). It is used to recover actinides from spent nuclear fuel reducing the amount of nuclear waste. UREX is uranium and technetium extraction process developed under the US Advanced Fuel Cycle Initiative (AFCI) program (Srimok, 2011; Thompson et al., 2002). UREX is the advanced version of PUREX process in which plutonium remains in the raffinate along with other minor actinides and fission products. To account for proliferation issue acetohydroxyl amine is used as a complexing agent to prevent extraction of plutonium in the product stream. UREX is also used to recover fission products: iodine by volatilization and technetium by electrolysis. Main goal of the UREX process is to recover > 99.9% Uranium, > 95% Technetium in separate product streams (Srimok, 2011; Pruett, 1981). The UREX process has been demonstrated at the laboratory scale at the Savannah River National Laboratory using irradiated fuel from the Dresden BWR (Meikrantz et al., 2007). According to the demonstration, the U losses to the Tc and raffinate streams were less than 0.02%, and the Tc losses to the U stream were < 1.2% and losses to the raffinate were low. This demonstrated that > 95% of the Tc could be recovered. Loss of Pu and other actinides to the Tc and U product streams was < 0.02% in all tests, with > 99.98% going to the raffinate. Uranium product obtained from UREX process can be used to manufacture MOX fuel for the light water reactor. Technetium as separate product can either be used in transmutation or go for direct
Corresponding author. E-mail address: [email protected] (I. Kumari).
https://doi.org/10.1016/j.nucengdes.2019.110410 Received 26 June 2019; Received in revised form 24 October 2019; Accepted 25 October 2019 0029-5493/ © 2019 Elsevier B.V. All rights reserved.
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Nomenclature
solvent in U-strip section using dilute nitric acid (Thompson et al., 2002; Vandegrift et al., 2004).
TRU FP Ln
3. Versions of UREX process
Transuranic elements Fission products Lanthanide
The versions of UREX + processes provide many options for nonproliferation strategy. Plutonium product is either mixed with other products or present in non-extractable form to prevent its extraction. All the versions of the UREX + processes are listed in Table 1 along with the products stream and the combination of plutonium with other products (Np, U). The major advantages of the UREX + processes are following: (1) the efficiencies of U and Pu recovery are > 99.9%, (2) volume of the spent fuel reduced greatly by separating uranium from the spent nuclear fuel, (3) radiotoxicity and heat generation of the spent fuel decreases by the recovery of Tc and Cs/Sr respectively (Ghosh et al., 2009; Nash et al., 2011).
disposal. UREX raffinate can further be treated to separate the desired products like Cesium (Cs) and Strontium (Sr) in order to minimize the heat production from the waste (Srimok, 2011). Aim of this review paper is to describe UREX reprocessing process and its variants along with the process chemistry involved. This paper also presents advantages and limits of the processes. 2. Urex process The UREX process flowsheet as shown in Fig. 1 is divided into three parts: extraction, scrub and strip. Feed containing spent fuel dissolved in nitric acid is fed into the extraction section; contacted with fresh solvent 30% tri-butyl phosphate (TBP) dissolved in n-dodecane. Uranium and technetium are separated from the dissolved spent fuel in the extraction section. Complexant/reductant is added in the scrub section to form complex with the plutonium (Pu) and neptunium (Np) and hence prevent their extraction. Uranium and technetium is stripped off from the loaded solvent by using dilute nitric acid in the strip section. The raffinate containing transuranic (TRU) and all fission products (FP) except technetium becomes the feed for CCD-PEG process. The CCD-PEG stands for chlorinated cobalt dicarbollide (CCD) and polyethylene glycol (PEG) used for the cesium and strontium extraction will be discussed in detail CCD-PEG process section (Pereira et al., 2007). Low concentration of nitric acid is maintained in the feed and scrub sections to increase the complex formation of plutonium and neptunium and also to increase the extraction of pertechnetate ion (Vandegrift et al., 2004). Fig. 2 consists of extraction, scrub, U Re-extraction and U, Tc-strip sections describing separation of uranium and technetium from loaded solvent as well as from each other. Generally high concentration of nitric acid is used in the strip section to strip off uranium and technetium from loaded solvent. The loaded solvent obtained from the scrub section is stripped off of technetium in Tc strip section using high concentration of nitric acid. The Tc product is separated from uranium in the U-Re-Extraction section. Excess nitric acid is scrubbed off from loaded solvent before U-strip section. Uranium is recovered from the
3.1. UREx + 1 In UREX + 1 process uranium and technetium are extracted as separate streams. Cesium and strontium are extracted as combined stream. Transuranics and lanthanides are also separated together leaving fission product as separated product. 3.2. UREX + 1a Fig. 3 represents schematic of UREX + 1a process. The aim of the UREX + 1a is > 99.9% recovery of the transuranics (TRUs), > 95% recovery of uranium, > 95% recovery of technetium, 99% recovery of cesium and strontium and only 0.1% of plutonium and 1% of cesium and strontium left in the raffinate (Pereira et al., 2007). Main characteristics of UREX + 1a process are that plutonium and curium are kept together. Curium emits more neutrons than plutonium which prevents measurement of plutonium. Plutonium product obtained from UREX + 1a is impure (Feener, 2010). UREX + 1a process consists of four solvent extraction processes:(a) extraction of uranium and technetium (UREX), (b) separation of uranium from technetium (ion exchange), (c) extraction of cesium and strontium (CCD-PEG), (d) separation of fission products from lanthanides and TRU (TRUEX) and (e) separation of lanthanide fission products (TALSPEAK) and (f) separation of americium, curium, neptunium and plutonium (TALSPEAK) (Pereira et al., 2007). Spent fuel from the reactor is stored in the spent fuel storage. After breaking in small pieces, voloxidation (heating of the
Fig. 1. UREX process flowsheet (Pereira et al., 2007). 2
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Fig. 2. UREX process flowsheet (Vandegrift, et al., 2004; Nash et al., 2011). Table 1 Versions of UREX processes. Process
Product 1
Product 2
Product 3
Product 4
Product 5
Product 6
Product 7
UREX + 1 UREX + 1a UREX + 1b UREX + 2 UREX + 2a UREX + 3 UREX + 3a UREX + 4 UREX + 4a
U U U U U U U U U
Tc Tc Tc Tc Tc Tc Tc Tc Tc
Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr Cs/Sr
TRU/Ln TRU U/TRU Pu/Np U/Pu/Np Pu/Np U/Pu/Np Pu/Np U/Pu/Np
FP FP/Ln FP/Ln Am/Cm/Ln Am/Cm/Ln Am/Cm Am/Cm Am Am
FP FP FP/Ln FP/Ln Cm Cm
FP/Ln FP/Ln
Notes: TRU = Transuranic elements: Am, Cm, Np, Pu. FP = Fission products. Ln = Lanthanide fission products.
Fig. 3. UREX + 1a process flowsheet (Feener, 2010). 3
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industries. Tributyl phosphate is used as extractant. In the UREX + 3 process uranium is removed and plutonium and neptunium are separated together, (ii) minor actinides, americium and curium are separated using di-2-ethyl-hexyl phosphoric acid and (iii) partition of cesium and strontium from fission product waste (Collins, 2005). Plutonium/neptunium product recovered in the first and americium/ curium in the second separation process. Americium and curium are separated from the lanthanides leaving fission products and lanthanides together in the waste. Fig. 7 represents flowsheet for urex + 3 process in detail (Ghosh et al., 2009).
fuel in oxygen rich environment) is done to decrease the size of the fuel particles and separate it from the cladding. The fuel is diluted in nitric acid in the dissolution step to separate from the hulls and cladding. The remaining cladding and the hulls are stored in the waste storage. The dissolved fuel enters the first step of solvent extraction, where it contacts the organic extractant. Uranium and technetium are extracted in the first step of the solvent extraction leaving everything else in the raffinate. Cesium and strontium are extracted in the Chlorinated Cobalt Dicarbollide and Poly-ethylene Glycol (CCD-PEG) step leaving behind the transuranics (TRUs) and the fission products. The waste obtained from CCD-PEG is treated in Transuranic Extraction (TRUEX) step. Transuranics and lanthanides are extracted in the TRUEX step and the fission product goes into the raffinate (Pereira, et al., 2007; Feener, 2010). Lanthanides are separated from transuranics in the Trivalent Actinide Lanthanide Separation by Phosphorus Extractants and Aqueous Komplexes (TALSPEAK) step. TRU goes into the raffinate stream of the TALSPEAK step. TRUs and plutonium are solidified and packaged and then sent for storage. Partitioning of uranium and technetium is done after the UREX step. Uranium is converted into uranium oxide followed by evaporation and then goes to the storage. The waste products (Cs, Tc, Sr, FP and lanthanides) obtained from all the process steps are packaged in the solid form for further shipment (Feener, 2010).
3.6. UREX + 3a Fig. 8 represents flowsheet of the UREX + 3a consisting of five extraction processes (UREX, FPEX, NPEX, TRUEX and TALSPEAK) and one ion exchange process. Technetium and uranium are extracted from dissolved spent fuel in the UREX section. Uranium is separated from technetium using ion exchange process. The raffinate from UREX is fed to FPEX (Fission Product Extraction) segment leading cesium and strontium extraction along with Rb and Ba. The NPEX (Neptunium Plutonium Extraction) segment makes the UREX + 3a process different from the UREX + 3. Uranium, neptunium and plutonium are extracted all together in NPEX process. The raffinate from the NPEX becomes feed for the TRUEX (Transuranic Extraction) (Goddard, 2010; Goddard et al., 2010).
3.3. UREX + 1b Pyroprocessing is grouped with UREX + 1b since both produce plutonium that is not separated from uranium, americium, neptunium, and curium in the final product (Bari et al., 2009). Recovery of traces of uranium along with the transuranic elements differentiates UREX + 1b from UREX + 1a.
3.7. UREx + 4 Americium is separated from curium using UREX + 4 processes. The main characteristics of UREX + 4 processes are that americium is burnt in the specially designed target assemblies. The target fabrication does not require shielding after separation of americium from curium (Nash et al., 2011).
3.4. UREx + 2 The process flowsheet of the UREX + 2 as shown in Fig. 4 consists of three solvent extraction and one ion exchange processes: (a) extraction of plutonium and neptunium in the co-extraction section, (b) extraction of technetium and uranium in co-extraction, (c) extraction of cesium and strontium using CCD-PEG and (d) recovery of uranium from technetium in ion exchange process. The UREX + 2 process flowsheet was developed using AMUSE (Argonne Model for Universal Solvent Extraction) code at Argonne National Laboratories and Idaho National Engineering and Environmental Laboratories (Pereira et al., 2005). The dissolved fuel is introduced in the co-extraction section consisting of extraction, scrub and strip sections. The co-extraction process flowsheet has been shown in Fig. 5. Uranium, plutonium, neptunium and technetium are recovered from the dissolved fuel using 30% TBP in n-dodecane. The extract (U, Tc, Np, and Pu) obtained from extraction/Scrub section is fed to Np/Pu-strip section where re-extraction of uranium and technetium is done using the solvent 30% TBP in n-dodecane. Complexant/reductant is used in extraction section to prevent extraction of neptunium and plutonium by forming complexes or reducing to nonextractable form. Uranium and technetium are stripped off using dilute nitric acid in U/Tc strip section. The raffinate of co-extraction section is fed to CCD-PEG section after reducing nitric acid concentration. Cesium and strontium are extracted in CCD-PEG section using the combined solvent Chlorinated Cobalt Dicarbollide and Poly-ethylene Glycol. The raffinate of CCD-PEG section contains TRU and FP which are separated from each other in the An/Ln separation section. Ion exchange method as described previously is used for separation of uranium from technetium (Nash et al., 2011; Pereira et al., 2005).
4. Process chemistry Uranium and technetium are the major products of UREX process. Complexant/reductant are used to form complexes with neptunium and plutonium preventing their extraction into the product stream (Collins, 2005). The complex formation of neptunium and plutonium and extraction of technetium can be increased by using low concentration of nitric acid both in the feed as well as in the scrub. However, highly concentrated nitric acid is used for stripping off technetium from the solvent. Uranium is separated from the solvent using dilute nitric acid (Pereira et al., 2007; Vandegrift et al., 2004). The actinides with +6 and +4 oxidation states form complexes with TBP and hence can easily be extracted into organic phase. TBP has lower affinity with +5 and almost nil for +3 and lower oxidation states (Am (III), Cm (III), Pu (III), Cs(I) and Sr (II)) (Nash et al., 2011). 4.1. Uranium and technetium separation Technetium is a long-lived (half-life 214,000 y) radioisotope present
3.5. UREx + 3 UREX + 3 process flowsheet shown in Fig. 6 consists of three parts: (i) first part is modified version of solvent extraction process used in the
Fig. 4. Schematic diagram of UREX + 2 process. 4
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Fig. 5. Co-extraction process segment flowsheet (Nash et al., 2011).
4.2. Recovery of cesium and strontium
in significant quantities in the spent nuclear fuel. Technetium as pertechnetate ion is highly mobile and soluble in the environment. This is a matter of concern for disposal of technetium in repository (Campbell, 1963; Poineau et al., 2008). Separation of technetium from the spent fuel can resolve these issues. Uranium and technetium are co-extracted from the dissolved spent nuclear fuel using TBP in n-dodecane in UREX process. Pu(IV), Np(IV) and Np(VI) present in the solution can readily be extracted by TBP. However Np(V) is not extractable by TBP. Acetohydroxamic acid (AHA) is used as reducing/complexijng agent to prevent extraction of Pu(IV), Np(IV) and Np(VI). AHA reduces Np(VI) to non-extractable Np(V). AHA prevents extraction of Np(IV) and Pu (IV) in the organic phase by forming complexes with them. Technetium is separated from the U/Tc product by using ReillexTM HPQ resin in ion exchange process. Tc is recovered as pertechnetate ion, TcO4-. Pertechnetate ion is extracted by forming complex with the TBP (Chung et al., 2007). Thus uranium and technetium are extracted by TBP/ndodecane leaving fission products and transuranics in the raffinate.
Recovery of cesium and strontium from the spent nuclear fuel minimizes short term heat load from the spent fuel. Cesium and strontium can be extracted simultaneously using either CCD-PEG or FPEX solvent extraction process. CCD-PEG is more effective for the feed containing < 1 M nitric acid. The solvent Chlorinated Cobalt Dicarbollide (CCD) is used for cesium separation and Poly Ethylene Glycol (PEG) for strontium recovery. The solvent used is the mixture of CCD and PEG diluted in phenyltrifluoromethyl sulfone (FS-13). Strip solution containing guanidine carbonate/diethylenetriamine pentaacetic acid (DTPA) is used to strip off cesium and strontium effectively from the solvent (Law et al., 2005). The second process FPEX solvent extraction is used for combined extraction of cesium and strontium. The solvent used in FPEX is the combination of calix[4]arene-bis-(tert-octylbenzo-crown-6)(BOBCalixC6),4′,4′ (S')-Di-(t-butyldicyclo-hexano)18-crown-6 (DtBuCH18C6) and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4sec-butylphenoxy)-2-propanol (Cs-7SB) diluted in Isopar-L. BOBCalixC6
Fig. 6. UREX + 3 process flowsheet (Collins, 2005). 5
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Fig. 7. UREX + 3 process flowsheet (Ghosh et al., 2009).
Fig. 8. High-level process flow diagram for UREX + 3a (Goddard, 2009, Goddard et al., 2010).
The process flowsheet of FPEX (Fission Product Extraction) is shown in Fig. 10 for the combined extraction of Cs and Sr. FPEX process recovers > 99.9% Cs and Sr by leaving < 100 nCi/g TRU in the strip solution (Law et al., 2007). Both CCD-PEG and FPEX processes can be used directly on UREX raffinate for recovery of cesium and strontium. Simultaneous recovery of cesium and strontium results in single product facilitating easy storage and reduction of overall process complexicity (Nash et al., 2011; Law et al., 2005). The CCD-PEG is more mature process but the solvent used in the FPEX process is more compatible with the other processes which are based on n-dodecane. The raffinate from the TRUEX process can also be treated using either CCDPEG or FPEX process rather than the UREX raffinate (Nash et al., 2011).
and DtBuCH18C6 are extractants; Cs-7SB is used as a phase modifier. The extractant BOBCalixC6 is responsible for cesium extraction and DtBuCH18C6 is for strontium. The above mixed solvent can have composition of is 0.007 M BOBcalixC6, 0.15 M DtBuCH18C6, 0.75 M Cs-7SB in Isopar-L (Nash et al., 2011; Law et al., 2005, 2006; Meikrantz, et al., 2005; Todd, 2005; Mincher et al., 2006). The process flowsheet of CCD-PEG is shown in Fig. 9. The raffinate obtained from UREX process contains fission products and transuranics. This is fed to CCD-PEG process separate cesium and strontium. CCD-PEG process shown in Fig. 9 consists of four sections: extraction, scrub, strip and wash. Cesium and strontium along with Rb and Ba are extracted in the extraction section. Scrubbing of transuranics and rare earths from the solvent are done in the scrub section using moderate concentration of nitric acid. Ammonium carbonate salt together with a complexing agent is used to strip off the alkaline-earth cations and alkali in the strip section. Since the solvent supply is limited solvent wash is done to recycle the solvent (Pereira et al., 2007).
4.3. Recovery of plutonium and neptunium Plutonium and neptunium are extracted from the spent fuel using NPEX process as shown in Fig. 11. NPEX process is utilized either on 6
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Fig. 9. CCD-PEG process flowsheet (Pereira et al., 2007).
Fig. 10. FPEX process flowsheet (Nash et al., 2011; Law et al., 2007).
Raffinate obtained from solvent extraction process (CCD-PEG raffinate) is fed into NPEX extraction section where it contacts with the organic solvent TBP in n-dodecane. Nitric acid scrub is used to remove the impurities (Americium/ Curium, Fission Products and Lanthanides) in the scrub section. This impurity goes in NPEX raffinate stream. Same
CCD-PEG raffinate or FPEX raffinate. Before following NPEX process, feed adjustment should be done to make plutonium and neptunium in the extractable form, concentration of nitric acid should be increased and to eliminate complexant and reductant present in the feed (Nash et al., 2011).
Fig. 11. NPEX process flowsheet (Hodges, 2006). 7
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diisobutylcarboylmethyl-phosphine oxide) and TBP in n-dodecane for the extraction of americium, curium, the rare earths and residual plutonium and neptunium (Ghosh et al., 2009; Vandegrift et al., 2004; Bond et al., 1987; Arai et al., 1997; Herbst et al., 2000). The procedure of scrub section is the same as discussed the above section. The rare earths and the actinides are stripped off from the solvent using weak complexant salt. The weak solvent is used to keep the pH values in the range of 3–4 for recovery of actinides using Cyanex 301 solvent (Aneheim et al., 1985, 2011, 2012). If the TRUEX process treats NPEX raffinate, the resulting product would be Am-Cm-Ln. However if the NPEX raffinate is not treated in the TRUEX process, the product obtained would be Pu-Np-TRUs (Nash et al., 2011). Cyanex 301 flowsheet is shown in Fig. 13. The product Cyanex 301 (chemical name: bis (2, 4, 4-trimethylpentyl) dithiophosphinic acid) is supplied by Cytec Industries, Canada in the impure form. The purification of Cyanex 301 is done by the suggested method of Zhu et al. The product obtained from the TRUEX process is fed directly in the extraction section of the Cyanex process. Solvent containing purified 301 diluted in TBP/n-dodecane is used for separation of actinide/lanthanide. The impurities are separated using weak complexant in the acid. Stripping of americium and curium from the solvent is done using ammonium salt (Vandegrift et al., 2004). The drawback of the use of Cyanex 301 is that it decomposes both in solution as well as in solid. Therefore, Cyanex 301 is not recommended for separation of actinides and lanthanides in the industries and that leads to the use of TALSPEAK process (Nash et al., 2011). Actinide and lanthanide are recovered using TALSPEAK process. HDEHP (bis(2-ethylhexyl)phosphoric acid) in n-dodecane is used as solvent in the TALSPEAK. However the aqueous phase contains buffered solution of lactate and DTPA (Nash et al., 2011). The solvent containing lactate buffer and a complexant is the feed for the TALSPEAK process. The TALSPEAK process flowsheet as shown in Fig. 14 contains three sections: extraction, scrub and strip. The actinides are separated by forming complex with aminopolyacetic acid. The trivalent lanthanides are extracted by bis (2-ethylhexyl) phosphoric acid (HDEHP). Weaker complex of lanthanides are extracted. However, stronger complexes of actinides remain in the aqueous phase. Fractions of Am and Cm are removed by the scrub which are extracted by HDEHP. The lanthanides are removed from the solvent in the strip section using concentrated nitric acid (Pereira et al., 2007; Nash et al., 2011; Croff et al., 2008).
complexant/reductant that is used in UREX process is used to strip off plutonium and neptunium from the organic stream in the Pu/Np section. The solvent used in the process is washed using sodium bicarbonate and rinsed using nitric acid in the solvent wash section (Hodges, 2006; Mitchell, 2008). U-Pu-Np product is obtained by adding uranium to the recovered plutonium and neptunium in the strip section (Nash et al., 2011; Hodges, 2006; Cheng et al., 2006).
4.4. Extraction of transuranics Pu(IV), Am(III), Cm(III) and Np(VI) are extracted by 6,6′ Bis (5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-1,2,4-triazin-3-yl)-2,2′[bipyridine] (CyMe4-BTBP) and TBP dissolved in cyclohexanone (Thompson et al., 2002; Pruett, 1981; Aneheim et al., 1985, 2011, 2012). Trivalent and pentavalent actinides are separated from trivalent lanthanides by using CyMe4-BTBP. However, TBP can be used to extract tetravalent and hexavalent actinides. CyMe4-BTBP does not react with TBP. Therefore, they can be combined into one solvent. Thus, CyMe4-BTBP and TBP together is used for the recovery of trivalent, tetravalent, pentavalent and hexavalent actinides (Aneheim et al., 1985, 2011, 2012; Thompson et al., 2002; Vandegrift et al., 2004; Horwitz et al., 1985; Bond et al., 1987). TRUEX (TRansUranium EXtraction) process is used to separate transuranics and the rare earth elements from the CCD-PEG raffinate (Pereira et al., 2007). Radio-toxicity of the waste can be minimized significantly by the extraction of transuranics (Ghosh et al., 2009). TRUEX process flowsheet shown in Fig. 12 consists of five sections: one extraction, three scrubs and one strip section. The raffinate obtained from the CCD-PEG process becomes feed for the extraction section of the TRUEX process. Concentrated nitric acid is used to adjust composition of the feed. Reducing agent is added to the feed to reduce Np(V) to extractable Np(IV). Impurities of the solvent are removed in the first scrub section using oxalic acid. Scrubbing of oxalic acid is done in the second scrub section using the moderate concentration of nitric acid. For effective stripping, acid concentration of the solvent is lowered by the addition of dilute nitric acid in the third scrub section. Lactate buffer along with a complexant are used for stripping off the rare earths and the actinides from the solvent (Pereira et al., 2007). However, raffinate from NPEX process can be fed directly to TRUEX process without doing feed adjustment. The process flowsheet of TRUEX process using the feed from NPEX raffinate is shown in Fig. 13. The solvent used in this process is CMPO (octyl(phenyl)-N,N-
Fig. 12. TRUEX process flowsheet (Pereira et al., 2007). 8
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Fig. 13. TRUEX and Cyanex process flowsheet (Pereira et al., 2007).
5. Discussion
exchange process. Plutonium and neptunium can be extracted from the products stream of third stage using NPEX process. These processes have some advantages and disadvantages also which are being discussed in the subsequent sections.
The desired components are extracted efficiently using the all the versions of the UREX + process. The entire UREX + processes can be generalized as following. All versions of the UREX + processes can be performed in three major stages. In the first stage uranium and technetium can be recovered using the solvent TBP in n-dodecane and reducing/complexing agent acetohydroxamic acid (AHA) leaving everything else in the raffinate. Since Pu(IV), Np(IV) and Np(VI) present in the spent nuclear fuel can readily be extracted by TBP. Therefore, AHA is used here as a reducing/complexing agent which prevents extraction of Pu(IV), Np(IV) and Np(VI) by reduction of Np(VI) to non-extractable Np(V) and forming complexes with Np(IV) and Pu(IV). Cesium and strontium are recovered from the raffinate of first stage using the combined solvent of calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6), 4′,4′ (S')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (Cs-7SB) diluted in Isopar-L and rest of the materials go in the raffinate. Pu(IV), Am(III), Cm(III) and Np(VI) are extracted together in the third stage by 6,6′ Bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-1,2,4-triazin-3-yl)-2,2′-[bipyridine] (CyMe4-BTBP) and TBP dissolved in cyclohexanone leaving trace amount of the fission products and lanthanides in the raffinate. The combined products stream obtained from the above three major stages can be treated to separate from each other using suitable processes. Technetium is separated from the U/Tc product by using ReillexTM HPQ resin in ion
5.1. Advantages of UREX + processes The advantage of using UREX + process is that it eliminates the risk related to the spent nuclear fuel; waste to be disposed of may be less radioactive, less HLW repositories required, risks due to leakage of mobile elements, decay heat and radiation hazards from the spent nuclear fuel will be less, fissile and fertile elements (U, Pu) can be better used as resources, mining of uranium and the associated risks may be reduced (Cordfunke, 1995). The typical UREX process is used to separate uranium and technetium leading to volume reduction. Application of UREX + 3 flowsheet minimizes the heat-generating product in the waste hence maximizes the lifetime of high level waste repository. Mass of the spent nuclear fuel is minimized by recycling. Proliferation resistance increases (Collins, 2005). Recovery of Uranium reduces the volume and dose rate of the spent fuel. Technetium is a long-lived FP; extraction of this leads to reduction of long-term dose of the spent fuel. Cesium and strontium having short half-lives of approximately 30 years are highly radioactive and heat generating elements which can reduce the mechanical strength of the long-term storage (Wu et al., 2016). Cesium and strontium are highly soluble in water. Due to this they easily spread in the environment and can be very hazardous to plants
Fig. 14. TALSPEAK process flowsheet (Pereira et al., 2007; Nash et al., 2011). 9
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unit equipment cost has been multiplied by number of equipment and adding all the cost gives total cost of UREX process. It is clear from Table 2 that Equipment cost for UREX process is minimum and for it is maximum for UREX + 3a process because product streams obtained in UREX process lesser than UREX + 3a process. Separation of more products increases the cost of the process.
and animals. Ingestion of these by the human can be major health hazards. Cesium and strontium are contributor to short-term radiotoxicity and heat load to the spent nuclear fuel. Recovery of cesium and strontium from the spent nuclear fuel minimizes short-term heat load and radiotoxicity of the spent fuel. Transuranic elements (Pu, Np, Am and Cm) having very long half-lives are alpha-emitting isotopes. Decay time of these elements is much longer. Therefore, problems of radioactive hazards due to these isotopes are long-term problem. Recovery of these elements reduces long-term dose-rate and toxicity from the spent fuel (Nash et al., 2011). Plutonium having long half-life is a major contributor to long-term heat-load and radio-toxicity to spent nuclear fuel. Plutonium can be used as a source of energy but due to proliferation issues it cannot be separated as pure Pu. Therefore Pu is recovered with minor actinides. Minor actinides neptunium, americium, and curium are major contributors to long-term heat-load and radiotoxicity of the spent nuclear fuel. Plutonium and neptunium recovered in NPEX process are highly pure which can be used in MOX fuel (Abdallah et al., 2019; Vandegrift, et al., 2004).
6. Conclusions UREX processes are more versatile than the PUREX due to proliferation issues. In UREX + 1 plutonium after forming complex with acetohydroxamic acid remains in the raffinate along with the other components. Reduction and complexation make recovery of plutonium impossible. Cesium and strontium are extracted profitably > 99.85 using CCD-PEG and FPEX techniques. Solvent used in CCD-PEG is mixture of Chlorinated Cobalt Dicarbollide and Poly Ethylene Glycol. Stripping of Cs and Sr from the solvent is done with guanidine carbonate/diethylenetriamine pentaacetic acid (DTPA). Cesium and strontium are extracted simultaneously by FPEX process using the combined solvent of DtBuCH18C6, BOBCalixC6 and Cs-7SB dissolved Isopar-L. The technique used in NPEX is feasible for the extraction of both Pu-Np and U-Pu-Np product using TBP in n-dodecane. Transuranics (Pu(IV), Am(III), Cm(III) and Np(VI)) are recovered by CyMe4-BTBP and TBP dissolved in cyclohexanone. Uranium (> 99.9%) and technetium (> 95%) are recovered as separate product in all versions of UREX + processes. Recovery of uranium from the spent fuel reduces its volume for the disposal. Recovery of technetium reduces long-term dose. Cesium and strontium separation minimizes repository heat load. TRUEX and TALSPEAK processes are used for extraction of actinides from lanthanides at high purities. Recovery of transuranics from the spent nuclear fuel results in reduction of long-term radiotoxicity. Recovery of plutonium, americium and lanthanides meet the goal of > 99.99%, > 99.97% and < 0.03% respectively. Equipment cost for UREX + processes depends upon the number of product streams.
5.2. Key issues of UREX + processes The complexity of the UREX + processes increases when the separation of desired or waste products increases. Due to increase in the processes complexicity cost of the processes also increases. The overall cost of the energy production using nuclear reactor will increase. Additional costs (cost of partitioning and fuel fabrication, equipment cost, piping cost, labour and land cost) and risks of operation (leakage, handling and transport of hazardous materials) of reactors used in recovery processes will increase (Cordfunke, 1995). Table 2 represents estimated cost (in Rupees) of equipment used in the UREX + process flow sheet for pilot plant. All the data for estimating the equipment cost have been taken from online suppliers (https://www.indiamart.com). Cost of the piping, materials and operation have not been included. Cost estimation can be understood by taking an example. UREX flow sheet consists of two extraction columns, two scrubbers, two strippers, one dissolver and four storage tanks. Per Table 2 Equipment cost of UREX + processes. Process
Storagea
Cost (Rupees) Feed Dissolverb
Extractorc
Scrubberd
Stripperd
UREX
75,000*4
85,000
1.5 Lakh*2
1.0 Lakh*2
1.0 Lakh*2
UREX + 1a CCD-PEG TRUEX TALSPEAK UREX + 2 CCD-PEG
75,000*6 75,000*4 75,000*3 75,000*3 75,000*6 75,000*4
85,000
1.5 1.5 1.5 1.5 1.5 1.5
Lakh*1 Lakh*1 Lakh*1 Lakh*1 Lakh*2 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*1 Lakh*3 Lakh*1 Lakh*3 Lakh*1
1.0 1.0 1.0 1.0 1.0 1.0
Lakh*1 Lakh*1 Lakh*1 Lakh*1 Lakh*2 Lakh*1
UREX + 3 UREX TRUEX FPEX TALSPEAK
75,000*4 75,000*3 75,000*3 75,000*3
85,000
1.5 1.5 1.5 1.5
Lakh*2 Lakh*1 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0
Lakh*2 Lakh*3 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*1
UREX + 3a UREX FPEX NPEX TRUEX TALSPEAK
75,000*4 75,000*3 75,000*4 75,000*3 75,000*3
85,000
1.5 1.5 1.5 1.5 1.5
Lakh*2 Lakh*1 Lakh*1 Lakh*1 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*3 Lakh*1
1.0 1.0 1.0 1.0 1.0
Lakh*2 Lakh*1 Lakh*1 Lakh*1 Lakh*1
a b c d e f g
85,000
Evaporatore
Washera
Ion-Exchangerf
Separatorg
Total (Rs.) 1,085,000
1,000
2,936,000 75,000
5,000
50,000
2,115,000
75,000 3,245,000
75,000 5,000
Karadani Engineering Pvt Ltd, Ahmedabad, Gujarat, India. Krish Pvt Ltd. Trident Labortek, Mumbai, Thane, India. Yansons Private Limited, India. KEP Engineering Services Private Limited. Thermochem Corporation Private Limited, Bengaluru, Karnataka. Marine Craft Exports. 10
75,000 75,000
3,890,000
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I. Kumari, et al.
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