Nuclear Fuel, Reprocessing of
The objective of this entry is to give a basic overview of the technology elements behind nuclear fuel reprocessing. It should serve as a starting point for more detailed study with the aid of the Bibliography section to obtain more technical details on this subject. Several more process concepts have been proposed, tested, and demonstrated other than those listed in this entry. For the sake of conciseness, only two fundamentally different technologies have been described here – aqueous and pyrochemical fuel reprocessing. In the case of pyrochemical fuel reprocessing, focus has been placed on the LiCl-KCl electrorefining technology developed originally at Argonne National Laboratory. The overall scope of nuclear fuel reprocessing technology is too broad to cover in this entry.
KeywordsZirconium Lithium Cadmium Sodium Chloride Immobilization
All elements including and beyond actininium (Z > 89) in the periodic table. In spent fuel, the major actinides of interest are uranium, plutonium, neptunium, americium, and curium.
- Cathode processor
A high-temperature vacuum distillation furnace used to separate salt from metallic actinides deposited on an electrorefiner cathode.
- Centrifugal contactors
Liquid–liquid extraction equipment used for aqueous solvent extraction that consists of a spinning rotor to intensely mix the different phases.
- Ceramic waste
The glass-bonded sodalite matrix used to encapsulate waste salt from electrorefiner operation.
French process for coextracting uranium and plutonium using extraction methods similar to PUREX.
An electrochemical system used to separate actinides from spent fuel using a molten salt electrolyte.
- Experimental Breeder Reactor-II
A sodium-cooled, fast test reactor operational at Argonne National Laboratory-West from 1963 to 1994.
- Geologic repository
A permanent nuclear waste disposal site located deep within a geological formation.
- Metal waste
The stainless steel–zirconium matrix used to encapsulate cladding hulls and noble metals left in anode baskets after U electrorefining is completed.
Liquid–liquid extraction equipment used for aqueous solvent extraction requiring a relatively large footprint.
- Minor actinides
Actinide elements other than U and Pu. In spent fuel, the primary minor actinides of concern are Np, Am, and Cm.
- Noble metals
Elements found in spent nuclear fuel that have chloride forms that are thermodynamically less stable than uranium chloride.
- Pulsed columns
Liquid–liquid extraction equipment used for aqueous solvent extraction involving a single column consisting of trays of perforated plates to promote interphase mass transport.
Nuclear fuel treatment technology that uses electrochemical reactors with molten salt electrolytes to separate actinides from fission products.
Nuclear reprocessing technology that separates actinides from the spent fuel via liquid–liquid extraction involving acidic aqueous and organic liquid phases.
- Spent fuel
Nuclear fuel that has gone through at least one irradiation cycle in a nuclear reactor. It contains a mixture of actinides and fission products.
- Solvent extraction
A separations method for extracting species from a liquid phase. In this entry, it refers to a process for removing uranium from spent fuel that utilizes dissolution in acid followed by liquid–liquid extraction between aqueous and organic liquid phases.
A variant of the PUREX process that separates uranium from spent fuel without recovering pure plutonium
A v-shaped vessel that is designed to efficiently blend two or more different kinds of powders with or without applied heat.
- 1.Cochran RG, Tsoulfanidis N (1993) The nuclear fuel cycle: analysis and management, 2nd ed. American Nuclear Society, Washington, p. 214Google Scholar
- 2.OECD, IAEA (2008) Uranium 2007: resources, production, and demand. Nuclear Energy Agency, Washington, June 2008Google Scholar
- 3.Gray LW (1999) From separations to reconstitution – A short history of plutonium in the US and Russia. Lawrence Livermore National Laboratory, UCRL-JC-133802Google Scholar
- 5.(1951) REDOX technical manual, Hanford Works, HW-18700Google Scholar
- 6.Hore-Lacy I (2009) Mixed oxide fuel (MOX). World Nuclear Association (Content Partner); Cutler J Cleveland (Topic Editor). In: Cutler J (ed) Encyclopedia of earth. Environmental Information Coalition, National Council for Science and the Environment, Cleveland/Washington, DCGoogle Scholar
- 7.Denniss IS, Jeapes AP (2001) Reprocessing irradiated fuel. In: Wilson PD (ed) The nuclear fuel cycle: from ore to wastes. Oxford University Press, Oxford, p 120Google Scholar
- 9.Long JT (1967) Engineering for nuclear fuel reprocessing. Gordon Breach Sci Publ, New YorkGoogle Scholar
- 10.Petitjean V, Fillet C, Boen R, Veyer C, Flament T (2002) Development of vitrification process and glass formulation for nuclear waste conditioning, Proceedings of Waste Management 2002. Tucson, AZ USAGoogle Scholar
- 11.Spent Fuel Reprocessing Options (2008) International Atomic Energy Administrations, IAEA-TECDOC-1587, Vienna, Austria 2008Google Scholar
- 12.Boullis B (2008) Future nuclear fuel cycles: prospects and challenges. In: Bruce Moyer (ed) Solvent Extraction: fundamentals to industrial applications, Proceedings of ISEC 2008 International Solvent Extraction Conference, vol 1., pp 29–42Google Scholar
- 13.Nash K (2008) Key features of the TALSPEAK and similar trivalent actinide-lanthanide partitioning processes. In: Bruce Moyer (ed) Solvent extraction: fundamentals to industrial applications, Proceedings of ISEC 2008 International Solvent Extraction Conference, vol 1., pp 511–519Google Scholar
- 14.Laidler J (2008) An overview of spent-fuel processing in the Global Nuclear Energy Partnership. In: Bruce Moyer (ed) Solvent extraction: fundamentals to industrial applications, Proceedings of ISEC 2008 International Solvent Extraction Conference, vol 1., pp 695–701Google Scholar
- 17.Miguirditchian M, Chareyre L, Hérès X, Hill C, Baron P, Masson M (2007) GANEX: adaptation of the DIAMEX-SANEX process for the group actinide separation, Proceedings of GLOBAL 2007 Advanced Nuclear Fuel Cycles and Systems. Bosie, IdahoGoogle Scholar
- 18.Wigeland R, Bauer T, Fanning T, Morris E (2006) Separations and transmutation criteria to improve utilization of a geologic repository. Nuclear Technol 154(1):95–106Google Scholar
- 19.Drain F, Emin JL, Vinoche R, Baron P (2008) COEX process: cross-breeding between innovation and industrial experience. Proceedings from Waste Management 2008, Tucson, AZGoogle Scholar
- 20.Katsuta T, Suzuki T (2009) Japan’s spent fuel and plutonium management challenge. Energy Policy doi: 10.1016/j.enpol.2009.05.075Google Scholar
- 21.Pereira C, Vandegrift G, Regalbuto M, Bakel A, Bowers D, Gelis A, Hebden A, Maggos L (2007) Lab-scale demonstration of the UREX+1a process using spent fuel. Proceedings from Waste Management 2007, Tucson, AZGoogle Scholar
- 22.Nuñez L, Vandegrift G (2000) Evaluation of hydroxamic acid in uranium extraction process: literature review, Argonne National Laboratory, ANL00/35.Google Scholar
- 23.Colven, TJ Jr, (1956) Mixer-Settler development-operating characteristics of a large-scale mixer-seller. Savannah River Laboratory, DP-140Google Scholar
- 24.Davidson JK, Shafer AC, Haas WO (1957) Application of Mixer-Settlers to the PUREX Process. In: The Symposium on the Reprocessing of Irradiated Fuels, Book 1. United States Atomic Energy Commission, TID-7534Google Scholar
- 25.Benedict M, Pigford TH, Levi HW (1981) Nuclear chemical engineering. McGraw-Hill, New York, p 210Google Scholar
- 27.Sege G, Woodfield FW (1954) Chem Eng Progress 50(8)Google Scholar
- 28.Geier RG (1954) Application of the Pulse Column to the PUREX Process. USACC, Report TID-7534Google Scholar
- 29.Richardson GL, Platt AM (1961) Progress in nuclear energy, Series IV, Technology engineering and safety, vol 4. Pergammon Press, New YorkGoogle Scholar
- 31.Jubin RT et al (1988) Developments in centrifugal contactor technology. Oak Ridge National Laboratory, ORNL/TM-10768Google Scholar
- 32.Meikrantz DH, et al (2001) Annular Centrifugal Contactors for Multiple Stage Extraction Processes. Chem Eng Comm 188: 115–127Google Scholar
- 33.Watts C (1977) Solvent Extraction Equipment Evaluation Study – Part 2. Battelle Northwest Laboratory, BNWL-2186 Pt. 2Google Scholar
- 34.Bernstein GL et al (1973) A high-capacity annular centrifugal contactor. Nuclear Technol 20Google Scholar
- 35.Drain F et al (2003) Forty years of experience with liquid-liquid extraction equipment in the nuclear industry. Proceedings from Waste Management Conference 2003, Tucson, AZGoogle Scholar
- 36.Meikrantz DH et al (1996) Rotor sleeve for a centrifugal separator. U.S. Patent # 5,571,070Google Scholar
- 37.Macaluso LL, Meikrantz DH (1999) Self-cleaning rotor for a centrifugal separator. U.S. Patent # 5,908,376Google Scholar
- 38.Garn, TG, Meikrantz DH, Law JD (2008) Remote evaluation of a three-stage 5 cm annular centrifugal contactor remote module at the INL. Idaho National Laboratory, INL/EXT-08-13670Google Scholar
- 39.Meikrantz DH, Garn TG, Law JD, Macaluso LL (2009) Evaluation of a new remote handling design for high throughput annular centrifugal contactors. Idaho National Laboratory INL/EXT-09-16824Google Scholar
- 40.Chang YI (1989) The integral fast reactor. Nuclear Technol 188(2):129–138Google Scholar
- 42.Benedict RW (1997) EBR-II spent fuel treatment demonstration project. Trans Amer Nuclear Soc 77:75–76Google Scholar
- 44.Lee SY et al (2007) A preliminary study on the safeguardability of a Korean Advanced Pyroprocessing Facility (KAPF). Proceedings of Global 2007, Boise, IdahoGoogle Scholar
- 45.Lee HS, Hur JM, Ahn DH, Kim IT, Lee JH (2009) Development of Pyroprocessing Technology at KAERI. Proceedings of Global 2009, Paris, FranceGoogle Scholar
- 47.Goff KM, Benedict RW (2005) Electrorefining Experience for Pyrochemical Reprocessing of Spent EBR-II Fuel. Proceedings of Global 2005, Tsukuba, Ibaraki (Japan)Google Scholar
- 48.Karell EJ, Gourishankar KV, Smith JL, Chow LS, Redey L (2001) Separation of actinides from LWR fuel using molten-salt-based electrochemical processes. Nuclear Technology 136:342–353Google Scholar
- 49.Gourishankar K, Redey L, Williamson M (2002) Electrochemical Reduction of Metal Oxides in Molten Salts. Light Metals 2002, TMSGoogle Scholar
- 50.Westphal BR, Keiser DD, Rigg RH, Laug DV (1994) Production of Metal Waste Forms from Spent Fuel Treatment. Proceedings of the DOE Spent Nuclear Fuel Meeting: Challenges and Initiatives, Salt Lake City, Utah; December 13–16, 1994Google Scholar
- 51.Abraham DP, McDeavitt SM, Park J (1996) Metal waste forms from the electrometallurgical treatment of spent nuclear fuel. Proceedings of the Embedded Topical Meeting on DOE Spent Nuclear Fuel and Fissile Material Management, Reno, Nevada, June 16–20, 1996Google Scholar
- 52.Pereira C, Hash MC, Lewis MA, Richmann MK, Basco J (1999) Incorporation of Radionuclides from the Electrometallurgical Treatment of Spent Fuel into a Ceramic Waste Form. Materials Research Society Symposium Proceedings 556 (1999), 115–120Google Scholar
- 56.Kim EH, Park GI, Cho YZ, Yang HC (2008) A new approach to minimize pyroprocessing waste salts through a series of fission product removal process. Nuclear Technol 162(2):208–218Google Scholar
- 57.Simpson MF, Sachdev P (2008) Development of electrorefiner waste salt disposal process for the EBR-II spent fuel treatment project. Nuclear Eng Technol 40(3), April 2008Google Scholar
- 58.Simpson MF, Goff KM, Johnson SG, Bateman KJ, Battisti TJ, Toews KL, Frank SM, Moschetti TL, O’Holleran TP (2001) A description of the ceramic waste form production process from the demonstration phase of the electrometallurgical treatment of EBR-II spent fuel. Nuclear Technol 134:263–277Google Scholar
- 59.Thomas JL, Mange M, Eyraud C (1971) Molecular Sieve Zeolites-I., R.F. Gould, Ed., Amer Chem Soc (1971), 101:443–449Google Scholar
- 60.Ebert WE (2005) Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain repository. Argonne National Laboratory, ANL-05/43, September 2005Google Scholar