USOO8475747B1
(12) United States Patent (10) Patent No.: US 8.475,747 B1 Johnson et al. (45) Date of Patent: Jul. 2, 2013
(54) PROCESSING FISSILE MATERAL (56) References Cited MIXTURES CONTAINING ZIRCONUM AND/OR CARBON U.S. PATENT DOCUMENTS 3,012,849 A 12, 1961 Horn ...... 423.4 Inventors: Michael Ernest Johnson, Richland, WA 3,965,237 A * 6/1976 Paige ...... 423f4 (75) 2003. O156675 A1* 8, 2003 Venneri et al. . ... 376/189 (US); Martin David Maloney, 2003/0234223 A1* 12/2003 Kuraoka et al...... 210,660 Evergreen, CO (US) 2007,0290.178 A1* 12/2007 Baron et al...... 252/643 2008/022410.6 A1* 9, 2008 Johnson et al. ... 252/625 (73) Assignee: U.S. Department of Energy, 2010/0314592 A1* 12/2010 Bourg et al...... 252,636 Washington, DC (US) OTHER PUBLICATIONS General Atomics & US DOE, “Development Plan for Advanced High (*) Notice: Subject to any disclaimer, the term of this Temperature Coated-Particle Fuels'. http://nuclearinl.gov/ patent is extended or adjusted under 35 deliverables/docs/pc-000513 0 relpdf, 2003.* U.S.C. 154(b) by 784 days. Pereira, Candido. “UREX-- Process Overview” www.ne.doe.gov/ pdfFiles/DOENRCUREXSeminar.pdfSimilar, Mar. 26, 2008.* Del Cul et al. “TRISO Coated Fuel Processing to Support High (21) Appl. No.: 12/484,561 Temperature Gas-Cooled Reactors.” http://nuclear.gov/peis/refer ences/RM923 DelCuletal 2002.pdf, p. 1-62, Mar. 2002.* (22) Filed: Jun. 15, 2009 Minato et al. “Retention of fission product caesium in ZrC-coated fuel particles for high-temperature gas-cooled reactors.” J. Nuclear Materials, 279, pp. 181-188, 2000.* Related U.S. Application Data * cited by examiner (60) Provisional application No. 61/061,563, filed on Jun. Primary Examiner — Tri V Nguyen 13, 2008. (74) Attorney, Agent, or Firm — Michael J. Badagliacca: John T. Lucas (51) Int. C. G2IC 9/00 (2006.01) (57) ABSTRACT COIG 56/00 (2006.01) A method of processing spent TRIZO-coated nuclear fuel (52) U.S. C. may include adding fluoride to complex zirconium present in USPC ...... 423/7; 252/625; 252/636; 252/637; a dissolved TRIZO-coated fuel. Complexing the Zirconium 423/8; 423/9 with fluoride may reduce or eliminate the potential for Zirco (58) Field of Classification Search nium to interfere with the extraction of uranium and/or tran USPC ...... 252/500-645; 423/7-10; 376/189 Suranics from fission materials in the spent nuclear fuel. See application file for complete search history. 10 Claims, 7 Drawing Sheets
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US 8,475,747 B1 1. 2 PROCESSING FISSILE MATERAL kernel, oxidizing organic acids, extracting Zirconium from MIXTURES CONTAINING ZIRCONUM the dissolved fuel kernel, and processing the dissolved fuel AND/OR CARBON kernel by at least one or more of the following processes: CCD-PEG, FPEX, TRUEX, and TALSPEAK. CCD-PEG is CROSS REFERENCE TO RELATED an acronym for chlorinated cobalt dicarbollide polyethylene APPLICATION glycol. FPEX is an acronym for Fission Product Extraction. TRUEX is an acronym for Transuranic Extraction. TAL This application claims priority from U.S. Provisional SPEAK is an acronym for Trivalent Actinide Lanthanide Application Ser. No. 61/061,563 filed Jun. 13, 2008. Separations by Phosphorous-reagent Extraction from Aque 10 ouS K complexes. The fuel compact may include at least the STATEMENT REGARDING FEDERALLY carbon fines, the silicon carbide, and the fuel kernel. SPONSORED RESEARCH ORDEVELOPMENT While multiple embodiments are disclosed, still other embodiments of the present invention will become apparent The United States Government has rights in this invention, to those skilled in the art from the following Detailed Descrip as represented by the U.S. Department of Energy, pursuant to 15 tion, which shows and describes illustrative embodiments of agreement DE-FC01-07NE24502. the invention. As will be realized, the invention is capable of modifications in various aspects, all without departing from FIELD OF THE INVENTION the spirit and scope of the present invention. Accordingly, the drawings and detailed description are to be regarded as illus The present disclosure relates to processing nuclear mate trative in nature and not restrictive. rial. More specifically, the present disclosure relates to the processing of fissile material mixtures containing Zirconium BRIEF DESCRIPTION OF THE DRAWINGS and/or carbon and methods and systems for Such processing. FIG. 1 depicts a coated particle fuel kernel. BACKGROUND OF THE INVENTION 25 FIG. 2 depicts a flow chart of one embodiment of the steps for processing TRIZO-coated fuels. Nuclear power plants generate spent nuclear fuels (SNF). FIG. 3 depicts a functional flow diagram for the UREX SNF typically contains uranium, and other radioactive process. actinide elements such as neptunium, plutonium, americium FIG. 4 depicts a functional flow diagram for the chlorinated and curium, radioactive rare earth elements, the radioactive 30 cobalt dicarbollide polyethylene glycol (CCD-PEG) process. transition metal technetium, as well as radioactive cesium and FIG. 5 depicts a functional flow diagram for the TRUEX strontium. feed adjustment and TRUEX process. The spent nuclear fuel may be in the form of a fuel kernel FIG. 6 depicts a functional flow diagram for the TAL or particle and may include a coating, Such as a TRISO-coated SPEAK process. particle fuel or a TRIZO-coated particle fuel. TRISO is an 35 FIG. 7 depicts a functional flow diagram for the ZrTcPX acronym for TRI-structural, ISOtropic, with the coatings process. being a buffer, Such as low-density pyrolytic carbon, high density pyrolytic carbon (IPyC and OPyC), and silicon car DETAILED DESCRIPTION bide (SiC). TRIZO-coated particle fuel is similar to the TRISO-coated particle fuel and includes a thin layer of Zir 40 The disclosure is directed to methods of processing fissile conium (ZrO) deposited on the kernel and/or in the buffer material mixtures, such as TRIZO-coated particle fuel. In one layer. The coating serves as a miniature pressure vessel that aspect, the processing method includes the addition of fluo provides containment of radionuclides and gases. ride anions to complex zirconium present in a dissolved The spent nuclear fuel may be processed to separate reus TRIZO-coated fuel. The complexing of zirconium with fluo able transuranics and uranium from the fission products. The 45 ride may prevent, or otherwise reduce the potential of inter TRIZO-coated fuels contain Zirconium (Zr) in the fuel kernel ference of the UREX (uranium extraction) and TRUEX coatings. Zirconium may contaminate the uranium and tech (TRU, or transuranic extraction) solvent extraction processes netium or the transuranics separated from the TRIZO-coated by Zirconium. The processing method may also include a step fuels during processing. for oxidation of organic acids. 50 In another aspect, the processing method includes a solvent BRIEF SUMMARY OF THE INVENTION extraction processing step called ZrTcEX (Zirconium techne tium extraction). The ZrTcPX process may be applied to Described herein are methods for processing TRIZO TRIZO-coated particle fuels that are made from oxides of coated spent nuclear fuels. One method may include remov transuranics (TRU) that do not contain uranium or where ing a fuel compact from a graphite spent TRIZO-coated fuel 55 uranium recovery is undesirable. The process co-extracts Zir block, separating carbon fines and silicon carbide from a fuel conium and technetium from the dissolved TRIZO TRU kernel, dissolving the fuel kernel, separating Solids from the oxide fuels. If uranium is also present in the TRIZO-coated dissolved fuel kernel, oxidizing organic acids, complexing fuel, uranium is co-extracted with Zirconium and technetium. zirconium with fluoride, and processing the dissolved fuel A step for oxidizing the organic acids may also be included in kernel using a solvent extraction process to separate fission 60 the processing method. products from uranium and transuranics. The fuel compact With reference to FIG. 1, a coated fuel kernel may include may include at least the carbon fines, the silicon carbide, and an inner fuel kernel that provides fission energy and controls the fuel kernel. oxygen potential. TRIZO-coated (and TRISO-coated) fuel Yet another method may include removing a fuel compact kernels can include uranium oxide, mixed uranium and plu from a graphite spent TRIZO-coated fuel block, separating 65 tonium oxide, plutonium oxide, TRU oxides, or other fissile carbon fines and silicon carbide from a fuel kernel, dissolving materials (e.g. thorium). For the TRU oxide fuel, the Pu-239 the fuel kernel, separating solids from the dissolved fuel and Pu-241 content of the TRU functions as the fissile mate US 8,475,747 B1 3 4 rial and the remaining nuclides function as fertile material Nuclear Company, Inc., Idaho Falls, Id., which is hereby and/or contribute to reactivity control during burn-up. The incorporated herein by reference in its entirety. Also, a similar inner fuel kernel may be enclosed by several layers of various method was also used to over-bore fuel channels in the graph materials that collectively make up the coating. The layers ite blocks in the 105-C reactor at the Hanford Site, Richland may include a buffer layer (porous carbon layer), an inner Wash. This method is described in more detail in HW-71711, pyrocarbon layer (IPyC), a silicon carbide (SiC) layer and an 1962, Final Report Production Test Number IP-360-AK outer pyrocarbon (OPyC) layer. The buffer layer may provide Over-Boring Demonstration Cast Iron and Graphite Bore a Void Volume for gaseous fission products and CO (carbon Enlargement 105-C, General Electric Company, Richland monoxide), accommodate Swelling of the kernel and attenu Wash., which is hereby incorporated by reference herein in its ate fission recoils. The inner pyrocarbon layer may protect the 10 kernel from chlorine (Cl) during deposition of the silicon entirety. The Hanford Site system used a shell end mill tool carbide layer, reduce tensile stress in the SiC layer and may for cutting the graphite and collected the graphite fines using retain gaseous fission products. The SiC layer may serve as a vacuum system. the primary loadbearing member and may retain gaseous and After drilling, the fuel compacts are pushed out of the fuel metallic fission products. The outer pyrocarbon layer may 15 channels. It has been demonstrated that compacts removal reduce tensile stress in the SiC layer and retain gaseous fission from Fort St. Vrain graphite blocks may be achieved by products, and may protect the SiC layer from a chemical pushing with an average force of 10.7 lb,(ranged from 1.5 to attack by coolant impurities. The buffer, inner pyrolytic car 33 lb). The graphite blocks are removed from the disassem bon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon bly cell for dispositioning. The compacts are transferred to (OPyC) layers are referred to collectively as a TRISO coating. another cell for size reduction and to minimize contamination Some fuel kernels may include a thin layer of zirconium in the graphite block disassembly cell. The fuel compacts are carbide (ZrO) coating around the kernel which may prevent further processed for size reduction. carbon monoxide (CO) formation. The ZrC coating may pre A first method for size reduction of fuel compacts uses vent the release of fission products from the fuel kernel. ZrC grinding and milling. Use of ajaw-type crusher to breakup the can be substituted in the buffer and SiC layers to allow higher 25 fuel compacts has been proposed. Carbonaceous materials fuel burn-up. Excess oxygen may oxidize the ZrC to form are separated from the fuel kernels and disposed. The fuel zirconium dioxide (ZrO). Fuel particles that include the ZrC kernels are processed by jet-milling to produce Small size layer (or multiple layers) around the kernel are referred to as (<50 um) particles. a TRIZO-coated particle fuel. A second method for size reduction of fuel compacts uses TRIZO-Coated Fuels Processing 30 pulsed current to breakup the fuel compacts and kernels. The As illustrated in FIG. 2, the steps for processing TRIZO coated fuels may include mechanical and aqueous processing fuel compacts are placed in water and are exposed to high steps. The mechanical processes may include separating the voltage/high intensity pulses of 200 to 500 kV, and discharge TRIZO-coated fuel from the prismatic blocks and preparing currents from 10 to 20 kA. The duration of pulses being short, the fuel for aqueous processing. Mechanical processing steps 35 the implemented energy is a few kJ. In these conditions, in the may include separating the fuel compacts from the prismatic presence of water, an electric arc is created near the Solid and blocks, crushing/milling the fuel compacts to separate carbon propagates preferentially through it. Energy with density fines and silicon carbideas well as crush/expose the fuel. The ranging from 10 to 100J/cm is deposited infew microseconds disposition of the Solid wastes separated during mechanical in holes of discharge of less than 10 microns in diameter that processing are discussed in more detail below. 40 induced a temperature increase which can reach up to 10,000 The aqueous processing steps for the TRIZO-coated fuel K together with a pressures rise of 10" Pa. These extreme may include leaching the crushed fuel kernels, separating conditions split up the solid to reduce it to fine fragments. If Solids from the dissolved fuel, oxidizing organic acids, com the electric arc does not cross the solid material, the electrical plexing Zirconium, and UREX--1a solvent extraction pro energy is transformed into mechanical energy that propagates cessing to separate fission products from uranium and tran 45 in the water in the form of a shock wave provoking a similar suranics. The UREX--1a solvent extraction process is effect of dismantling. The pulsed current method for size comprised of the UREX, CCD-PEG (or FPEX), TRUEX, and reduction of fuel compacts using nuclear grade graphite TALSPEAK solvent extraction processes. The aqueous pro samples has been demonstrated. The graphite samples (typi cessing steps are described in more detail below. The dispo cally 27x27x40 mm) were fractured into pieces with particle sition of gaseous waste generated from the mechanical head 50 size distribution less than 100 um using the pulsed current end and fuel dissolvers and the treatment of the dissolver technique. The pulsed current technique may be suitable for sludges and fission products separated during the UREX--1a either separating the carbonaceous material from the fuel Solvent extraction process are discussed in more detail below. kernels to enable separate disposal or for size reduction of the Head-End Mechanical Preparation of TRIZO-Coated Fuels fuel kernels. The graphite prismatic blocks are transported from the fuel 55 Aqueous Processing of TRIZO-Coated Fuels storage location to a disassembly cell. The compacts are taken TRIZO-coated fuels kernels may include uranium oxide, out of the graphite block using a drilling method to expose the mixed uranium and plutonium oxide, plutonium oxide, TRU fuel channels and pressure to push out the compacts. The oxides, or other fissile materials (e.g., thorium). The aqueous graphite fines generated from cutting into the graphite block processing steps for the uranium oxide and the mixed ura to expose the fuel channel are collected by a vacuum system 60 nium and plutonium oxide fuels are generally similar and are for disposal. A similar system was developed for extracting discussed in more detail below. The plutonium oxide and irradiated compacts from the Fort St. Vrain HTR reactor TRU oxides TRIZO-coated fuels contain very little uranium. prismatic blocks, which used a gunmill cutting tool and a While these fuels may be processed using the same aqueous vacuum system to collect graphite powder. This method is processing steps as the uranium containing TRIZO-coated described in more detail in WINCO-1159, 1993, Fort St. 65 fuels, a method, known as ZrTcEX, described in more detail Vrain Graphite Fuel Mechanical Separation Conceptual below takes advantage of the near absence of uranium in these Selection Final Report, S. M. Berry, Westinghouse Idaho fuels. US 8,475,747 B1 5 6 Dissolution of Uranium Containing TRIZO-Coated Fuels acid at 20 to 120° C. Crushed uranium carbide samples were The size reduced fuel kernels are leached using a suitable dissolved with nitric acid and uranium extraction was con acid such as nitric acid or nitric acid mixed with hydrofluoric ducted by equilibrating the solution with 20% TBP in 80% acid to dissolve the fuel. Zirconium carbon present in the fuel Amsco solvent, an odorless kerosene-like diluent. Little or no coatings will dissolve in the acid solution along with the fuel. extraction of the organic species was observed. Organic spe Additionally, Zirconium oxide formed in the TRIZO-coated cies as Solids were recovered by evaporating the raffinate fuel is also known to dissolve in nitric acid based on dissolu from the solvent extraction tests to dryness. Of the original tion tests of calcined zirconium oxide (INEL-95/0225, 1995, carbide carbon present, 50 to 70% was converted to carbon Actinide Partitioning from Actual ICPP Dissolved Zirconium dioxide and 21 to 44% converted to oxalic acid dihydrate, Oxide Calcine using the TRUEXSolvent, K. N. Brewer et al., 10 mellitic acid, and other soluble organic acids when uranium Idaho National Engineering Laboratory, Lockheed Idaho carbide was dissolved in 4 to 16M nitric acid. The effect or Technologies Company, Idaho Falls Id.). The pyrolytic car lack thereof of the organic acids on uranium extraction by bon and silicon carbide coatings, along with carbon fines and TBP was not reported. (Reactions of the Uranium Carbides noble metals (e.g. rhodium, palladium, etc.) that do not dis with Nitric Acid, 1965, J. Am. Chem. Soc., Vol. 87, pp 1710 solve in the acid are separated from the dissolver solution by 15 1714, L. M. Ferris and M. J. Bradley, Oak Ridge National filtration. The clarified dissolver solution is then precondi Laboratory, Oak Ridge Tenn.) tioned for UREX--1a solvent extraction processing. During PUREX (an acronym for Plutonium Uranium Preconditioning of the clarified dissolver solution may Extraction) process tests with uranium carbide dissolved in include decomposing organic acids and complexing Zirco nitric acid, and in the presence of Zirconium, an unknown nium. In one embodiment, the dissolution step includes the Zirconium organic complex formed that could not be use of a composition including hydrogen fluoride (HF) and scrubbed from the 30% TBP in normal paraffin hydrocarbon nitric acid (HNO). If such a composition is used, addition of Solvent. Uranium extraction was not reported as being fluoride, as described herein, may not be necessary. As dis effected by the presence of organic acids. Treatment of the cussed in the section entitled Organic Acid Oxidation, dissolved uranium carbide Solution with an oxidizing agent organic acid decomposition may reduce interference with 25 destroyed the Zirconium organic species and allowed effec UREX--1a solvent extraction steps. Following decomposi tive zirconium scrubbing from the PUREX solvent. The tion of the organic acids, fluoride may be added to the TRIZO organic acids were shown to be effectively oxidized with coated fuel dissolver Solution to complex zirconium, reduce dichromate and cerium (IV). The process conditions tested or prevent interference during uranium solvent extraction were oxidation at 50° C. for 24-hours using 0.002M dichro (UREX), and enable Zirconium stripping during the transu 30 mate, 0.02M dichromate, or 0.005M cerium (IV) added to ranics and lanthanide (TRUEX) solvent extraction process, as dissolver solution containing 1.8M uranium and 0.5M nitric discussed in section entitled Zirconium Complexing. The acid. Hydrogen peroxide was reported to be effective in oxi preconditioned dissolver feeds are reacted with a reductant/ dizing the organic acids, though to a lesser extent. However, complexant (acetohydroxamic acid) to reduce the Valence no specific process conditions were reported in the reference state and hinder the extractability of plutonium and nep 35 for the hydrogen peroxide tests. (BNWL-1573, 1971, tunium. The solution is then fed to the UREX--1a solvent Improved Zirconium Decontamination in PUREX Process, extraction process. The UREX+1a process consists of four page 16, J. L. Swanson, Battelle Pacific Northwest Labora solvent extraction processes: UREX, CCD-PED (or FPEX), tories, Richland Wash.) TRUEX, and TALSPEAK, which are described further in the Processing the High Temperature Gas-Cooled Reactor section entitled UREX-1A Solvent Extraction Process for 40 (HTR) fuels in either the THOREX or PUREX solvent Uranium Containing Solutions. extraction processes has also been investigated. Fuel ele Acetohydroxamic acid (AHA) is added in the UREX pro ments containing graphite of the HTR’s may require a head cess to Suppress plutonium extraction and prevent the reten end procedure which is more laborious than for metal clad tion of certain problematic fission products (e.g., Mo, Zr, and fuel. Generally, the head-end procedure may not affect the Ru) by the tributyl phosphate (TBP) solvent. The degree to 45 aqueous chemical process because following dissolution, the which AHA forms complexes with zirconium under condi history of the fuel has almost completely disappeared. How tions experienced during processing TRIZO-coated fuels is ever, if too much carbon remains in the fuel after the graphite not known and the effectiveness of preventing Zirconium is burned off (viz. more than 0.1% carbon in the ash), prob extraction by TBP is uncertain. If AHA proves suitable to lems may be encountered in the solvent extraction. Oxalic complex zirconium and prevent extraction by TBP, then fluo 50 acid may be produced by the reaction of HNO with graphite. ride would not be added to the dissolver solution. Instead, Because oxalic acid may form complexes with plutonium, a fluoride would be added after the AHA had been decomposed satisfactory plutonium recovery from the extraction step may in the TRUEX feed preparation step to complex zirconium not occur. (Chemical Processing of HTR Fuels Applying and prevent co-extraction with the actinides and lanthanides. Either THOREX or PUREX Flow Sheets, 1983 International Organic Acid Oxidation 55 Atomic Energy Agency, International Working Group on The presence of carbon during dissolution of uranium can Gas-Cooled Reactors, Vienna (Austria); State Committee on result in the formation of oxalic acid and mellitic acid, as well the Utilization of Atomic Energy of the USSR, Moscow (Rus as potentially other organic acids. The organic acids can inter sian Federation) IWGGCR-8, pp: 333-344, E. Zimmer and E. fere with the extraction of uranium (or other elements) in Merz, Kernforschungsanlage Juelich GmbH, Institut fuer Solvent extraction systems. The effect of organic acids on 60 Chemische Technologie der Nuklearen Entsorgung, Juelich, solvent extraction systems similar to the UREX process are Germany.) discussed in the next section. Oxidation of these organic acids The potential formation of mellitic acid during dissolution may avoid unwanted effects during solvent extraction pro of crushed TRISO-coated fuel surrogates that contained car cessing and is discussed in more detail below. bon has been reported (Processing of Spent TRISO-Coated Organic Acids Effects on Solvent Extraction Systems 65 GEN IV Reactor Fuels, 2004, Actinide and Fission Product Experiments have been conducted to determine the reac Partitioning & Transmutation Eighth Information Exchange tions of uranium mono- and di-carbides with 2 to 16M nitric Meeting Las Vegas, Nev., USA 9-11 Nov. 2004, B. B. Spen US 8,475,747 B1 7 8 cer, C. H. Mattus, G. D. Del CuI, D. Hunt and E. D. Collins, of hydrogen peroxide to oxidize organic acids appears to have Oak Ridge National Laboratory, Oak Ridge Tenn.). Batch the least effect on the UREX--1a solvent extraction process mixing tests to measure uranium extraction with TBP were and, in one embodiment, is the oxidizing reagent used in the conducted. While some of the crushed TRISO-coated fuel process. Surrogates contained Zirconium, measured or calculated Zir conium distribution coefficients with the TBP were not The oxidization of organic acids present in nitric acid dis reported. Such a result may be because an organic acid was solved TRIZO-coated fuels with hydrogen peroxide is now extracted into the organic (TBP) phase and enhanceduranium discussed. For development of a preliminary flowsheet, it is extraction based on tests conducted with and without carbon conservatively assumed that 44% of the original carbide car being present during dissolution of the crushed TRISO 10 bon present is converted to organic acids and one mole of coated fuel Surrogates. No physical, mechanical, foam, or hydrogen peroxide reacts with one mole of organic carbon. emulsion issues were encountered during the batch mixing Furthermore, one mole of hydrogen peroxide is assumed to tests. The organic acid species could effect the extraction react with one mole of plutonium as a result of the aforemen behavior of the solvent if it accumulates over time. tioned plutonium oxidation and reduction reactions. The oxi Oxidation of Organic Acids 15 dation of organic acids with hydrogen peroxide is assumed to The preceding section describes some potential negative be conducted at 50° C. for 24-hours. impact organic acids may have on Solvent extraction pro Zirconium Complexing cesses for separating uranium or transuranics including inter ference with extraction, accumulation of organic acids in Although other aqueous processes for separating uranium Solvent, and Zirconium complexing. Oxidation of the organic and transuranic elements from spent nuclear fuel (SNF) are acids appeared to be effective in destroying the organic spe available, the following discussion refers to the UREX+1a cies that were complexing Zirconium and preventing effective process. However, other processes may be utilized. scrubbing from uranium extracted with 30% TBP solvent. The UREX--1a process is a series of four solvent-extrac The organic acids were shown to be effectively oxidized with tion flow sheets that perform the following operations: (1) dichromate and cerium (IV). The process conditions tested 25 recovery of Tc and U (UREX), (2) recovery of '7Cs and 'Sr were oxidation at 50° C. for 24-hours using 0.002M dichro fission products (CCD-PEG or FPEX), (3) transuranics and mate, 0.02M dichromate, or 0.005M cerium (IV) added to dissolver solution containing 1.8M uranium and 0.5M nitric rare earths separation from non-rare-earth (TRUEX), and (4) acid. The use of the higher dichromate concentration did not separation of transuranic elements from the rare earths (TAL appear to significantly enhance the Zirconium scrubbing from 30 SPEAK). the 30% TBP solvent. Hydrogen peroxide was reported to be The TRIZO-coated fuels may contain a significant inven effective in oxidizing the organic acids, though to a lesser tory of Zr present in the fuel kernel coatings. Zirconium may extent. However, no specific process conditions were contaminate the uranium and technetium or the transuranics reported for the hydrogen peroxide tests. separated from the TRIZO-coated fuels during UREX--1a While the concentration of the dichromate and cerium (IV) 35 processing. The following Subsections discuss the impact of used to oxidize the organic acids was relatively low, it may not zirconium on the UREX, CCD-PED (or FPEX), TRUEX, and be desirable to introduce additional metal cations in the TALSPEAK solvent extraction processes, as well as a method UREX--1a solvent extraction process. In particular, the use of for complexing Zirconium. dichromate and cerium (IV) to oxidize organic acids may Zirconium Behavior in UREX Process increase the mass of high-level waste generated during pro 40 cessing of TRIZO-coated fuels. Dichromate is known to oxi The zirconium distribution coefficient is less than 0.01 for dize plutonium (IV) to plutonium (VI) in nitric acid, which process conditions experienced in the UREX uranium and was used in the bismuth phosphate precipitation process to technetium separation process when the tributyl phosphate recover plutonium and in the REDOX (an acronym for (TBP) solvent (30% TBP in n-dodecane) is highly saturated Reduction Oxidation) solvent extraction process to co-extract 45 with uranium as experienced when processing light water plutonium (VI) and uranium (VI), but not in the PUREX reactor (LWR) fuels. Furthermore, the UREX process adds Solvent extraction process Co-extraction of plutonium along acetohydroxamic acid (AHA) to the feed to the extraction with uranium in the UREX solvent extraction process may be section and in the scrub section to Suppress plutonium extrac undesirable. Cerium is difficult to separate from the transu tion and prevent the retention of certain problematic fission ranics in the TALSPEAK solvent extraction process and will 50 products (e.g., Mo, Zr, and Ru) by the TBP solvent that would similarly oxidize plutonium (IV) to plutonium (VI) in nitric otherwise contaminate the uranium and technetium product acid. The UREX process does include addition of a reductant/ streams. The TRU oxide TRIZO-coated fuel does not contain complexant (0.47 Macetohydroxamic acid and 0.3 MHNO) a significant mass of uranium, therefore the TBP solvent in to reduce the valence state and limit the extractability of the UREX process will not be saturated with uranium when plutonium and neptunium, and additional reductant would be 55 processing these fuels. Again, AHA is added to the feed to the required if either dichromate or cerium (IV) were added to extraction and scrub sections to Suppress plutonium extrac oxidize organics in the processing of TRIZO-coated fuels. tion and hinder Zirconium extraction. Hydrogen peroxide is known to react with various organic Zirconium extraction tests conducted with 30% TBP in acids in nitric acid, including oxalate and citrate. Hydrogen normal paraffin hydrocarbon (similar to n-dodecane) for the peroxide is also known to reduce plutonium(VI) to plutonium 60 PUREX process have demonstrated that Zirconium can be (IV) in nitric acid. Plutonium (IV) is further reduced to plu complexed with fluoride and effectively scrubbed from the tonium (III) in the presence of hydrogen peroxide and nitric solvent. Additional testing (as described more below in the acid, but oxidizes back to plutonium (IV) once the hydrogen section entitled Zirconium Behavior in TRUEX Process) has peroxide is depleted. The reactions of hydrogen peroxide shown that Zirconium complexed with fluoride results in the with organic acids (or plutonium) present in the dissolved 65 uranium product being free from Zirconium in the UREX TRIZO-coated fuel results in the formation of water, carbon process. Therefore, Zirconium complexed with fluoride will dioxide, and Smaller chain organics. Therefore, the addition be minimally extracted by the UREXTBP solvent and can US 8,475,747 B1 9 10 readily be scrubbed. Zirconium and fluoride will report to the scrub step uses moderately concentrated nitric acid to scrub raffinate stream from the UREX process. oxalic acid from the solvent. The third scrub step uses rela As previously noted, AHA is added in the UREX process to tively dilute nitric acid to lower the nitric acid concentration Suppress plutonium extraction and hinder the retention of in the solvent to allow effective stripping. The third scrub step certain problematic fission products (e.g., Mo, Zr, and Ru) by 5 uses a weak complexant salt (lactic acid with diethylenetri the TBP solvent. The degree to which AHA forms complexes aminepentaacetic acid) to strip the transuranics and rare earth with Zirconium under conditions experienced during process elements from the solvent. ing TRIZO-coated fuels is not known and the effectiveness of The TRUEX solvent may extract zirconium in sufficient preventing zirconium extraction by TBP is uncertain. IfAHA quantities that can hinder the desired extraction of transuran 10 ics and lanthanides. Testing of the TRUEX process has been proves suitable to complex Zirconium and prevent extraction conducted in order to separate transuranics from: (1) actual by TBP, then fluoride would not be added to the dissolver Zirconium oxide calcine that was dissolved in nitric acid; (2) solution. Instead, fluoride would be added after the AHA had simulated dissolved Zirconium oxide calcine spiked with been decomposed in the TRUEX feed preparation step to Am' and Zr; and (3) simulated sodium bearing waste, a complex zirconium and prevent co-extraction with the 15 nitric acid solution containing Zirconium fission products, actinides and lanthanides. uranium and transuranics. Table 1 shows the composition of The presence of carbon during dissolution of uranium can the dissolved solutions used in TRUEX process tests. The result in the formation of oxalic acid and mellitic acid, as well TRUEX process tests with the sodium bearing waste demon as potentially other organic acids. PUREX process tests with strated about 60% of the Zirconium in the feed would remain uranium carbide dissolved in nitric acid resulted in a forma with the transuranics. Similar TRUEX process tests con tion of Some Zirconium organic species that could not be ducted with dissolved zirconium oxide calcine demonstrated scrubbed from the 30% TBP in normal paraffin hydrocarbon that 99% of the zirconium in the feed would remain with the solvent. However, treatment of the dissolved uranium carbide Solution with an oxidizing agent destroyed the Zirconium transuranics. organic species and allowed effective Zirconium scrubbing The TRUEX process has also been tested for extracting from the PUREX solvent. An organic acid oxidation step may 25 transuranics from a Zirconium cladding waste (referred to as be included in the UREX+1a process when processing NCRW) sludge that was dissolved in nitric acid. The NCRW TRIZO-coated fuels to destroy the zirconium organic species sludge contains Sodium, Zirconium, uranium, actinides, lan and allow effective zirconium scrubbing from the UREX thanides, aluminum, silicon, iron, chromium, as well as fis solvent. sion products (e.g. 'Ce, ''Cs, '7Cs, ''Eu, Sb, Sr. 30 Tc, 'Co). The NCRW sludge was generated as a result of Zirconium Behavior in CCD-PEG (or FPEX) Process dissolving the Zircaloy-cladding on metallic uranium fuel in The raffinate stream from the UREX process will contain a PUREX plant. The dissolution process dissolved the Zir the spent fuel components except for uranium and techne caloy-cladding as well as a small fraction of the metallic tium. The UREX process raffinate is fed to the extraction uranium fuel. Tests were conducted to dissolve the NCRW section of the CCD-PEG (or alternatively the FPEX) process. 35 sludge using a mixture of nitric acid and hydrofluoric acid and The CCD-PEG solvent is a mixture of 0.011 M chlorinated process the dissolved waste solution using the TRUEX pro cobalt dicarbollide (CCD) and 0.027 Mpolyethylene glycol cess. Table 1 also provides the composition of dissolved (PEG-400) in phenyl-trifluoromethyl sulfone (FS-13). The NCRW sludge used in these TRUEX process tests. Uranium alternative FPEX process uses a solvent composed of 0.0075 was first separated from the dissolved NCRW sludge using M 4.4".(5)-Di(t-butyldicyclo-hexano-)-18-crown-6 40 30% TBP in n-dodecane as process conditions similar to those used in the UREX process. The uranium product from (DtBuCH18C6), 0.007 M Calix4arene-bis-(tertoctylbenzo the 30% TBP in n-dodecane solvent did not appear to contain crown-6) (BOBCalixC6), 0.75 M 1-(2,2,3,3-tetrafluoropro zirconium. Transuranics were separated using the TRUEX poxy)-3-(4-sec-butylphenoxy)-2-propanol) (Cs-7SB), and process. 0.003 M trioctylamine (TOA) in Isopar R. L. Alternative sol 45 vent mixtures may be employed to accomplish the extraction TABLE 1 of cesium and strontium. The behavior of zirconium in the CCD-PEG (or FPEX) process is not known and requires Composition of Wastes used for TRUEX Process Tests further study. However, Zirconium will likely not remain in Actual Simulated Simulated the CCD-PEG (or FPEX) solvent following scrubbing. The 50 Com- Dissolved Dissolved Sodium Dissolved ponent Zirconium Zirconium Bearing NCRW CCD-PEG (and FPEX) solvent is a highly specific extractant (M) Calcine Calcine Waste Sludge for cesium, strontium, and few of the other group IA (e.g. H 2.58 3.4 1.31 2.2 rubidium) and IIA (e.g. barium) cations. Therefore, the scrub NO3 4.33 6.4 4.46 2.2 section of the CCD-PEG (or FPEX) process should contain 55 Al 0.72 O49 O.S63 O.O34 Ca 0.57 O.78 3.39E-02 transuranics, lanthanides, Zirconium, and other fission prod Ce 3.63E-04 ucts, which are fed to the TRUEX process. C 3.S2E-O2 Zirconium Behavior in TRUEX Process Cr 2.6E-O3 54E-O3 5.63E-03 The transuranics and the rare earth elements are extracted Cs 6. S2E-OS F O.64 1.2 9.66E-O2 O.96 by the TRUEX solvent, which is 0.2 Moctyl (phenyl)-N,N- 60 Fe O.O1 O.O15 2.45E-O2 disobutylcarboylmethyl-phosphine oxide (CMPO) and 1.4 Hg 1.17E-03 MTBP diluted by n-dodecane. Lesser amounts of other fis K O.O2 O.141 Mn 1.42E-O2 sion products (e.g., Zr and 'Y) are also extracted and may Na O.20 O.O12 1.17 O.19 be scrubbed from the solvent so as not to contaminate the N 1.63E-O3 transuranics. The TRUEX process employs three scrub steps. 65 Nd 3.71E-03 In the first scrub step, the impurities are removed from the Pb 9.8E-04 Solvent using oxalic acid mixed with nitric acid. The second US 8,475,747 B1 11 12 TABLE 1-continued mixed fission products from uranium and technetium present in the dissolved SNF. Technetium is subsequently separated Composition of Wastes used for TRUEX Process Tests from uranium to enable recycle ofuranium and disposal of the Actual Simulated Simulated technetium. The transuranics and mixed fission products are Com- Dissolved Dissolved Sodium Dissolved 5 then fed to the CCD-PEG (or FPEX) process for further ponent Zirconium Zirconium Bearing NCRW treatment. (M) Calcine Calcine Waste Sludge Similar to the solvent used in the PUREX process, the Sir 5.28E-04 solvent for the UREX process is tributyl phosphate (TBP) SO4 3.86E-O2 dissolved in n-dodecane. A reductant/complexant (which Zr O.066 O.23 7.7E-05 O.22 10 may be 0.47 MAHA and 0.3 MHNO) is added to the process U 2.69E-03 O.O14M (gmL) through the scrub to limit the extractability of plutonium and CiL neptunium. The feed and the scrub contain low concentra tions of nitric acid to enhance the complexation of Pu and Np Cs-137 O.23 and to increase the extractability of pertechnetate ion. AHA Am-241 1.25E-04 2.8E-04 15 Pu-238 1.3E-O3 also hinders the retention of certain problematic fission prod Pu-239 SE-OS 2.8E-04 ucts (e.g., Mo, Zr, and Ru) by the TBP solvent that may otherwise contaminate the uranium and technetium product Note: StreamS. Blanks in table indicate no data available for these elements Turning now to the UREX process, the uranium and tech The TRUEX process tests with dissolved NCRW sludge netium in the feed are extracted into the solvent in the extrac tion section while other extractible species are scrubbed from demonstrated that Zirconium could form a separate Solid the solvent in the scrub section. The solvent, now loaded with phase in the TRUEX process under specific conditions. How uranium and technetium, is stripped of technetium in the ever, the formation of zirconium solids could be eliminated Tc-Strip section using 6 M nitric acid. The Tc product stream by adjusting the F/(Al-Zr) ratio to 3.8 in the dissolved NCRW 25 is scrubbed of uranium in the U-Re-extraction section. The Solution. Aluminum is present to complex fluoride and act as combined solvent enters the U-Strip section, where a dilute a corrosion inhibitor for stainless steel equipment. Addition (0.01 M) nitric acid feed removes uranium from the solvent. ally, the fluoride ion formed a stable complex with zirconium The solvent is then washed with 0.25 M sodium carbonate and reduced the zirconium distribution coefficient in the Solution to remove degradation products and contacted with TRUEX solvent, thus enabling scrubbing of zirconium from 30 0.1 M nitric acid to acidify the solvent before recycling the the transuranics. Transuranics and lanthanides were then solvent to the front end of the process. stripped from the TRUEX solvent. For the NCRW TRUEX The uranium recovered in the UREX process can be process, the zirconium fluoride complex was scrubbed from returned to the fuel manufacturing process for recycle or the TRUEX solvent using 0.15M oxalic acid scrub. Extracted discarded as waste. Technetium separated in the UREX pro oxalic acid was scrubbed from the TRUEX solvent using 5M 35 cess can be made into a waste form for disposal as high-level nitric acid, and excess nitric acid was scrubbed from the waste or used in other manufacturing processes. solvent using 0.01M nitric acid. These scrub solutions are the CCD-PEG: CS/Sr Extraction same scrub solutions used in the UREX+1a TRUEX process. FIG. 4 depicts a functional flow diagram for the chlorinated Zirconium Behavior in TALSPEAK Process cobalt dicarbollide polyethylene glycol (CCD-PEG) process. The discussion above demonstrates that Zirconium, when 40 The raffinate of the UREX process is fed to the extraction not complexed, interferes with the extraction and stripping of section of the CCD-PEG process. The CCD-PEG process is transuranics and lanthanides in the TRUEX process. How selective for cesium (Cs) and strontium (Sr) extraction for ever, adding fluoride to the dissolved TRIZO-coated fuel will feeds containing less than 1 M nitric acid. Alternatively, the form a stable zirconium fluoride complex which then can be Fission Product Extraction (FPEX) process, when suffi scrubbed from the TRUEX solvent prior to stripping of tran- 45 ciently developed, could be used instead of the CCD-PEG Suranics and lanthanides. The transuranics and lanthanides process. The FPEX process is discussed in more detail below. can then be separated in the TALSPEAK process without Cesium and strontium (as well as barium and rubidium) minimal to none Zirconium contamination present in the tran may be extracted by the CCD-PEG solvent, which is a mix Suranics. ture of 0.011 M chlorinated cobalt dicarbollide (CCD) and UREX--1a Solvent Extraction Process for Uranium Contain- 50 0.027 M polyethylene glycol (PEG-400) in phenyl-trifluo ing Solutions romethyl sulfone (FS-13). Barium and rubidium are also The preconditioned dissolver feeds are reacted with AHA quantitatively extracted by the solvent. In the scrub section, to complex/reduce plutonium and neptunium, thereby limit transuranics and lanthanides are removed from the Solvent ing their extraction by tributyl phosphate in the UREX pro using 3M nitric acid. The strip section uses 100 g/L guanidine cess. The solution is then fed to the UREX-1a solvent extrac- 55 (or methylamine carbonate) and 20 g/L DTPA to strip the tion process. The UREX--1a process is a series of four cesium and strontium from the solvent. The guanidine forms solvent-extraction flow sheets that perform the following a cation that replaces cesium (and rubidium) that was operations: (1) recovery of Tc and U (UREX), (2) recovery of exchanged with the CCD. DTPA forms a complex with stron '7Cs and Sr fission products (CCD-PEG or FPEX), (3) tium (as well as barium and other +2 or +3 cations) and transuranics and rare earths separation from non-rare-earth 60 prevents the strontium from precipitating in the alkaline Strip (TRUEX), and (4) separation of transuranic elements from Solution. A regenerable strip reagent, methylamine carbon the rare earths (TALSPEAK). The processing of TRIZO ate, has also been demonstrated, but further testing is needed coated fuels using these four solvent extraction processes is to optimize the process. The solvent is then washed with 0.1 described in the following subsections. M nitric acid to remove impurities and acidify the solvent UREX: Uranium and Technetium Separation 65 before recycling the solvent to the front end of the process. FIG. 3 depicts a functional flow diagram for the UREX The cesium and strontium separated in the CCD-PEG process process. The UREX process separates the transuranics and may be treated for disposal as waste. US 8,475,747 B1 13 14 Alternative FPEX Process for Cs/Sr Separation 1 to 3 M nitric acid solution may be used for the extraction An alternative process, FPEX, for separating Cs and Sr of the transuranics (as neutral nitrate complexes) by the from the UREX raffinate is being developed. If successfully TRUEX solvent. The nitric acid concentration of the CCD developed, the FPEX process could be substituted for the PEG raffinate is nominally 0.74 M. Therefore, addition of CCD-PEG process. concentrated nitric acid to the CCD-PEG raffinate may be The FPEX process uses a solvent composed of 0.0075 M used for transuranics extraction in the TRUEX process. 4.4".(5)-Di(t-butyldicyclo-hexano)-18-crown-6 TRUEX Process: Transuranics--Lanthanide Separation (DtBuCH18C6), 0.007 M Calix4arene-bis-(tertoctylbenzo After adjustment, the raffinate of the CCD-PEG (or FPEX) crown-6) (BOBCalixC6), 0.75 M 1-(2,2,3,3-tetrafluoropro process is fed to the extraction section of the TRUEX process. poxy)-3-(4-sec-butylphenoxy)-2-propanol) (Cs-7SB), and 10 0.003 M trioctylamine (TOA) in Isopar R. L. The strontium A functional flow diagram for the TRUEX feed adjustment (Sr) and cesium (Cs) extractants are DtBuCH18C6 and BOB and TRUEX process is depicted in FIG. 5. CalixC6, respectively. The modifier Cs-7SB enhances the The transuranics and the rare earth elements are extracted extraction of Sr from the solvent mixture. Due to viscosity by the TRUEX solvent, which may be 0.2 MCMPO and 1.4 issues associated with the modifier, concentrations must be 15 MTBP (phase modifier to prevent third phase formation) less than 0.8 M. TOA enhances the stripping of Cs from the diluted by n-dodecane. Lesser amounts of other fission prod solvent. Replacing the BOBCalixC6 and TOA with calix4 ucts (e.g., Zr and Y”) are also extracted and are scrubbed arene-bis(2-ethylhexylbenzo-18-crown-6) (BEHBCalixC6) from the solvent. The TRUEX flow sheet may have three and a commercial guanidine reagent (LIX(R79), respectively, scrub sections. In the first Scrub section, the impurities are to improve solvent performance has been evaluated. removed from the solvent using oxalic acid (0.1M) mixed The FPEX solvent mixture provides distribution coeffi with nitric acid (0.1M). The second scrub section uses mod cients useful for Cs and Sr extraction over a range of 0.5 erately concentrated (4M) nitric acid to scrub oxalic acid M