[en] Techniques were defined, tested and validated to determine the amount of ¹⁴C and ¹²⁹I in spent fuel. Two types of spent fuel were examined: BWR UOX fuel with a burnup of 41.4 GWd/tHMi cooled for 9 years, and MOX fuel with 5.6 wt% Pu/(U+Pu) enrichment and a burnup of 25 GWd/tHMi, cooled for 13 years. The measured values for the UOX and MOX samples were, respectively, 1371 and 1066 Bq ¹²⁹I/t-oxide, 17212 and 6660 Bq ¹⁴C/t-oxide. (authors)

[en] After 4 years work project, a new shielded cells facility is set into service in Atalante Marcoule under the code name C11/C12. It is devoted to the head-end high activity reprocessing studies. Physically. C11/C12 is constituted of eleven working places behind one meter of both concrete and leaded glasses biologic wall. The conception of C11-C12 was decided and organized to permit a large variety of experiments. These equipments permit among other studies on the following steps. Fuel and target mechanical treatment. Thermal treatments under inert gas flow. Dissolution using several media like nitric acid or complexing agent. Solid-liquid separation by filtration or centrifugation. Elements separation using liquid-liquid extraction apparatus like mixer-settlers or centrifugal contactors. Elements separation using solid-liquid chromatography. From June 99 to april 2001, the principal experiments realised were consisted to: June-august 99, MOX fuel dissolution. September 99: Liquid-liquid separation of U, Pu from fission products and minor actinides by the Purex process. October 99: Test of- DIAMEX process (selective separation of trivalent elements versus other fission products). June 00: Test of SESAME process (selective extraction of americium versus curium) September 00: Test of SANEX II process. November 00: Test of DIAMEX process in presence of complexing reagents to improve selectivity versus Pd. April 01: Test of CALIXARENE process for selective extraction of Cs from DIAMEX raffinate. April 01: dissolution of several pieces of irradiated fuel for metallurgical cladding tests at the CEA/LECI. The next years will be devoted to increase and diversity experiments in respect with quality and waste management, reducing the size of apparatus and amount of needed radioactive materials. (Author)

[en] This paper describes the study of iodine trapping performed on a test rig called SIRROCO on a scale 1 cartridge filled with adsorbent AC 6120 and the tests on samples done in parallel to compare the removal efficiencies of the industrial filter with those of the sorbent. Influence of the major parameters encountered in French operating conditions are discussed. The sorbent tests have to be further pursued, particularly those concerning influence of number

[en] Techniques were defined, tested and validated to determine the ¹⁴C and ¹²⁹I concentrations in spent fuel. The extended uncertainly on the values measured for these two radioisotopes was estimated initially at about 14%. Two types of spent fuel were examined: BWR UOX fuel with a burnup of 41.4 GWd/tHMi cooled for 9 years, and MOX fuel with 5.6 wt% Pu/(U+Pu) enrichment prior to irradiation and a burnup of 25 GWd/tHMi, cooled for 13 years. The measured values for the UOX and MOX fuel samples, respectively, were: 1371 and 1066 MBq ¹²⁹I/t_oxide, 17212 and 6660 MBq ¹⁴C/t_oxide. The first tests on surrogate solutions have been carried out to measure ³⁶Cl. (author)

[en] The definition of the U-Mo fuels reprocessing process is part of the French qualification program. The proposed scheme consists, first of all, in a specific dissolution in a nitric acid medium. The obtained solution is then diluted in standard UOx fuel dissolution solutions before treatment by the PUREX process. Despite the experience acquired by the reprocessing of U-Mo alloys containing from 0.5 to 1 weight % of molybdenum, it was necessary to obtain additional data in relation to the RTR fuels (7 % of Mo) and the chosen process: solubility of Mo in terms of Al, U and H⁺ concentrations, dissolution rate versus the temperature and the nitric acid concentration. Only the aluminium concentration was found to have a significant effect on molybdenum solubility. The results have led to the definition of the amount of fuel by dissolution batch. The experiments carried out on pieces of fuel plates, show a complete dissolution and a meat dissolution rate ranging from 200 to 250 mg.cm^-2.h^-1 (T =107⁰C), the acidity effect being insignificant. Finally, no problem was brought up with first experiments carried out on heated samples, in order to produce the interaction products between U-Mo alloy and Al. The results acquired to date have permitted us to define dissolution conditions which are now qualified with irradiated samples. (author)

[en] This study is devoted to evaluation of a new innovative micro structured fuel for future pressurized water reactor. This fuel would have potential to increase the safety margins, lowering fuel temperatures by adding a small fraction of a high conductivity second phase material in the oxide fuel phase. The behavior of this fuel in a standard rod has been modeled with finite element codes and was assessed for different aspects of the cycle as neutronic studies, thermal behavior, reprocessing and economics. Feasibility of fuels has been investigated with the fabrication and characterizations of the microstructure of composite fuels with powder metallurgy and HIP processes. First, a CERCER (Ceramic = UO₂- Ceramic matrix made of silicon carbide, SiC) fuel type has been investigated, the advantages of a ceramic being generally its transparency to neutrons and its high melting temperature. A first design of kernel type fuel was first chosen with a gap between the UO₂ particles and the second phase material in order to avoid mechanical interaction between the two components. Due to lowering thermal conductivity of SiC under irradiation, this CERCER fuel did not allow a temperature gain compared to current fuel. No ceramic material seems to exhibit all required properties. Even beryllium oxide (BeO), which conductivity does not decrease with irradiation according to the literature, induces difficulties with (α, n) reactions and toxicity. The study then focused on Cermet fuels (Ceramic-Metal). The metal matrix must be transparent to neutrons and have a good thermal conductivity. Several materials have been considered such as zirconium alloys, austenitic and ferritic stainless steals and chromium based alloys. The heterogeneous composite fuels were modeled using the 3D/CASTM finite element code. From an economical and neutron point of view, it was important to keep a low fraction of metal phase, i.e. less than 10 % of Zr for example. However, the fuel temperatures calculations show that 10% vol. fraction of metallic matrix is a minimum to have a substantial gain in fuel temperatures in nominal and LOCA conditions. A higher matrix fraction (15% of metal) could even simplify the safety system for accident mitigation. Calculations of the hydrogen risk in case of an accident show that the increase of hydrogen production can be limited to 40 to 50% more than the UO₂ standard if the amount of metal fraction is less than 10% vol. Two processes have been chosen to fabricate Cermet fuel with low fraction of metallic matrix: the first one by powder metallurgy (tested with UO₂ - 20% vol. stainless steel), the second one using HIP of coated CVD (tested with ZrO₂ with yttrium - 10 % vol. Mo). The downstream cycle was studied: starting with the classical PUREX process, the head-end of the reprocessing process as well of the amount of the generated waste would be impacted depending on the solubility of the different materials of the composite fuel in the nitric acid. Finally, a global evaluation is given. (authors)

[en] U-Mo fuels, containing up to 10 mass % of molybdenum, are promising low enriched uranium fuels (less than 20 mass % of ²³⁵U). Although their reprocessability is ascertained, more precise reprocessing conditions are now under studies. The process would consist at first in a specific dissolution in nitric acid media. The obtained solution could then be diluted into standard UOx type fuels dissolution solution. Uranium and plutonium could be selectively recovered from this feed solution using adapted PUREX process (liquid/liquid extraction using tributyl phosphate). Tests on U-Mo powder are the first step of the planned studies to define dissolution conditions. Dissolution experiments on inactive pieces of UMo fuels will then be conducted according to the previously defined conditions. At last, complete reprocessing test of U-Mo spent fuels (irradiated inside a French reactor) - that means dissolution, purification of U and Pu, fission products solution concentration and waste management will be carried out. All this program will be implemented on the French Atomic Energy Commission Center CEA/VALRHO - Marcoule by the end of 2005. (author)

[en] This article draws a state of knowledge of the dissolution of uranium dioxide in nitric acid media. The chemistry of the reaction is first investigated, and two reactions appear as most suitable to describe the mechanism, leading to the formation of monoxide and dioxide nitrogen as reaction by-products, while the oxidation mechanism is shown to happen before solubilization. The solid aspect of the reaction is also investigated: manufacturing conditions have an impact on dissolution kinetics, and the non-uniform attack at the surface of the solid results in the appearing of pits and cracks. Last, the existence of an autocatalytic mechanism is questioned. The second part of this article presents a compilation of the impacts of several physico-chemical parameters on the dissolution rates. Even though these measurements have been undertaken under a broad variety of conditions, and that the rate determining step of the reaction is usually not specified, general trends are drawn from these results. Finally, it appears that several key points of knowledge still have to be clarified concerning the dissolution of uranium dioxide in nitric acid media, and that the macroscopic scale which has been used in most studies is probably not suitable. (authors)

[en] The aim of this report is to evaluate the technical feasibility of long-lived wastes transmutation in different type of reactors and their associated cycles. This feasibility depends both on the type of waste and on the type of reactor. It is performed through scenario studies which allow to evaluate the overall steps of the fuel cycle (reactor, fabrication, storage, reprocessing) and which include the detailed studies of changes in cores design and management induced by transmutation, the impacts on fuel cycle facilities, and on reprocessing and fabrication processes. Previous scenario studies have permitted to underline the advantages and drawbacks of the different strategies. The scenarios considered in this document cover the overall options foreseeable today: a PWR-based scenario for the recycling of plutonium and americium in homogeneous mode based on the MOX UE Am assembly concept from 2020 onward; a 4. generation reactor-based scenario with fast spectrum and self recycling of actinides from 2035 onward; and a scenario where minor actinides are recycled in a specific cycle in association with subcritical systems. The document comprises also a specific chapter about the technical feasibility of the transmutation fuel which covers the overall aspects of the fuel cycle to be considered. (J.S.)

[en] subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)