R & D of Pyrometallurgical Reprocessing Technology 
in CRIEPI

Tadafumi Koyama

Central Research Institute of Electric Power Industry (CRIEPI)
2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511, Japan
E-mail: koyama@criepi.denken.or.jp

The recently increasing interest in pyrometallurgy can be attributed to the drastic change of boundary conditions around the nuclear fuel cycle in the world.  Today’s main emphasis is put on a maximal cost reduction of the fuel cycle. Furthermore recovery of long-lived nuclides becomes a new requirement Recovery of long-lived nuclides means to use of various reactor systems for transmutation, resulting in new requirements to reprocess different fuel types e.g. MOX, metal fuel, nitride fuel, high burn-up fuel, etc. These new requirements may result in a different choice for future fuel cycle technology.  Pyrometallurgical processing is one of the most attractive alternatives to meet these requirements.  The requirement for product purity being much less stringent, the recovery of minor actinides (MA: Np, Am, Cm) will take place simultaneously with plutonium due to the thermodynamic properties of molten salt media.  The recovery of MA allows the reduction of TRU wastes, and decrease at the same time the risk of nuclear proliferation.  The molten salt media also have two important advantageous properties as a solvent material in nuclear processing.  The radiation stability of molten salt allows the processing of spent fuels of high radioactivity (e.g. spent fuel with short cooling time) without any increase of solvent waste. Since molten salt is not a neutron moderator such as water is, comparatively large amount of fissile material can be handled in the process equipment, i.e. experimental facilities are compact and economical. 

The metal fuel cycle combined with metal-fuelled fast reactor allows the on-site reprocessing proposed on mid 1980’s by Argonne National Laboratory (ANL) [1], and further investigation has been carried out by Central Research Institute of Electric Power Industry (CRIEPI) since 1987[2].  The pyrometallurgical process consists of fuel chopping, electrorefining, cathode processing, injection casting, and waste treatment.  In addition to these processes, pyrochemical processes such as the conversion of oxide fuel and HLW into metal or chloride form have been developped for applying these materials into pyrometallurgical processing.  As a key technology, electrorefining and reductive extraction to recover actinides has benn examined by lots of experiments using actinide materials.  The electrorefining using liquid Cd cathode indicated to deposit Pu with an effective collection rate by the joint study with Japan Atomic Energy Research Institute[3]. The reductive extraction test using small amount of MA showed more than 99% of TRU nuclides can be recovered[4]  The lithium reduction of oxide fuels is confirmed to produce a metal of uranium, plutonium, neptunium, americium from each oxide and mixed alloy from MOX pellet[5].  In order to demonstrate the process with using not only gram-scale of americium and hundreds gram of plutonium but also real irradiated materials, the pyroprocess facility was installed in the Institute of Transuranium Elements in Germany[6]. As a first experiment, electrorefining of U-Pu-Zr alloy has been carried out.  The new glove box facility which can handle several hundred of plutonium are in construction to examine the material flow through oxide reduction step to the distillation step as a cooperation with Japan Nuclear Fuel Cycle Institute. 

References

[1] Y.I.Chang (1989), “The integral fast reactor”, Nucl. Technol., 88, 129.
[2] T.Inoue and H.Tanaka (1997), “Recycling of actinides produced in LWR and FBR fuel cycles by applying pyrometallurgical process”, Proceedings of GLOBAL1997, pp.646-652, Yokohama, October 1997.
[3] M.Iizuka, K.Uozumi, T.Inoue, T.Iwai, O.Shirai, Y.Arai (2001), “Development of Plutonium recovery process by molten salt electrorefining with liquid cadmium cathode”, OECD/NEA, 6th international information exchange meeting on actinide and fission product partitioning and transmutation, December, 2001.
[4] K.Kinoshita, M.Kurata, T.Inoue (2000), “Estimation of material balance in pyrometallurgical partitioning process of transuranic elements from high-lebel liquid waste”, J. Nucl. Sci. Technol., 37, 75.
[5]  T.Usami, T.Kato, M.Kurata, T.Inoue, H.E.Sims, A.Jenkins (2001), “Lithium reduction of MOX pellets”, Actinides 2001, 8P26, 4-9 November, Hayama. 
[6] T.Koyama, K.Kinoshita, T.Inoue, R.Malmbeck, M.Ougier, J-P.Glatz, L.Koch (2001), “Demonstration of pyrometallurgical processing for metal fuel and HLW”, proceedings of GLOBAL2001, September, Paris.

History and current status of French dry technologies developments

Philippe Brossard

Bernard Boullis and Philippe Brossard
Radiochemistry and Processes Department, CEA Rhône Valley Center
Nuclear Energy Division of CEA

Non aqueous technologies have a long history of research work. Speaking of France, the investigation of a volatilization process to treat irradiated fuels was initiated in the late 50’s and, in collaboration with Russia, the process was developed up to pilot scale in the early seventies.  Due to technical difficulties and a fierce competition with aqueous processes (PUREX) it was not chosen to be implemented in the COGEMA plant of La Hague and studies in the field were more or less shelved.
A renewed interest appears when the issue of highly radioactive waste management raised about 10 years ago.  Indeed, in December 1991, the French Parliament carried an Act specifying the major areas of investigation to determine operational solutions for the back-end of the fuel cycle.  One of the areas should considers separation/transmutation solutions capable of significantly reducing the mass and radio‑toxicity of the long-lived radionuclides that constitute the principal long-term risk.  In that frame it became evident that pyrochemical processes possess unique features that make them best suited to treat long-lived radionuclides transmutation targets. The capability of pyrochemical processes to deal with highly irradiated materials could represent an advantageous alternative to hydrometallurgical processes, which could in some cases prove less compatible with the high burn‑up and short cooling times proposed under some scenarios. The more distant future options, such as dedicated fuel cycles or transmutation in molten salt reactors, provide further impetus for exploring these processes.  
More recently, new concepts for advanced nuclear systems could also benefit from some features of pyrochemical cycles: in situ reprocessing, combined management of all the transuranic nuclides (plutonium and the minor actinides), tight integration with refabrication processes, minimal inventory of radio‑toxic materials confined in the cycle.
For all these potential applications, the main objective for processing irradiated fuels or transmutation targets lies in the ability to quantitatively recover the actinides for recycling.  This key point is at the center of the CEA R&D program in pyrochemistry. The current work focuses primarily at laboratory-scale research without seeking significant technological developments except those that might appear necessary to obtain a preliminary assessment of the operating conditions for the processes investigated. This approach addresses a broad spectrum of possible applications with the following objectives:
•	consolidate our knowledge of the concepts already described or tested elsewhere,
•	explore any promising options,
•	proceed with a preliminary selection of concepts for more thorough investigation of the aspects affecting their possible implementation,
•	perform one or more demonstrations on a few tens of grams of representative irradiated material as exhaustively as possible.
A related objective is to acquire greater basic knowledge of the chemistry of the elements in the proposed media (molten salts), and of process engineering considerations in order to prepare for possible subsequent development of specific applications.