[en] This presentation brings together the slides presented during the conference of the SESC/SPUA the 28 January 2005. The first part is devoted to the fuel behavior under irradiation, the second one to the main objectives of the program and the last part to the international collaboration. It provides information on the fuel behavior under irradiation, the fuel fabrication and performance, the SIROCCO program, the european contacts and the simulation programs. (A.L.B.)

[en] Two different modes exist for minor actinide transmutation: homogeneous route and heterogeneous route (UO₂ or inert matrix support). After twenty years of R and D focused on technical feasibility, an analysis can be performed based on the main outcomes obtained for transmutation fuel to show where we are in the fuel qualification process. Maturity of each transmutation fuel could be evaluated using a qualification scale. This scale is used for several types of fuel up to industrialisation stage. This method has been applied to these fuels and evaluated the work to be done in the coming years. This will be followed by an adjustment of the R and D programme until the final qualification step with different items: fabrication process, fuel element and sub-assembly design, measured properties, behaviour modelling and irradiation programme. The irradiation programme will include analytical, semi-integral and integral irradiations, in MTR or in SFR prototypes in normal conditions as well as for transients and accidental conditions. (authors)

[en] The PROFIL-R (fast spectrum) and PROFIL-M (moderated spectrum) experiments were performed between 2003 and 2008 in the French fast neutron reactor Phénix. These experiments consisted of the irradiation of pure isotope samples in a well-characterized neutrons flux in order to collect accurate information on the total capture integral cross sections of the principal heavy isotopes and some important fission products in the spectral range of fast reactor. This method can be used for all isotopes transformed by neutron capture into a stable or long-lived nuclide and is based on the measurement of the composition change induced by irradiation. Therefore, accurate and reproducible measurements of isotopic compositions and concentrations of the elements (actinides and fission products) before and after irradiation are required. The major difficulty for the analyses of these actinides and fission products is the low quantity of the initial powder enclosed in steel container (3 to 5 mg) and the very low quantities of products formed (several μg) after irradiation. We present the developments performed during the last few years by laboratories of the French Commission on Atomic Energy and Alternative Energies (CEA) in order to acquire very accurate and precise isotopic and elemental data on selected irradiated powders. Among the necessary developments are the conceptions of systems set in shielded hot-cells to open the steel containers and collect the full amount of powders, and the set-up of specific analytical methods for mass spectrometry measurements in order to obtain isotopic and elemental ratios at uncertainty of few per mil level. A synthesis of the results obtained and first preliminary interpretations will also be presented. (author)

[en] In the French HTR programme, CEA and AREVA/Framatome (now called as AREVA NP) conduct research and development on HTR fuel aiming in mastering the UO₂ Triso particle fuel fabrications technology, irradiating new fuels coming from the new French facilities, performing post-irradiation examinations on these fuels and developing codes predicting fuel performance and fission product transport. (author)

[en] In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10^-7. As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

[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] In the context of the French nuclear waste management act of 1991, two options have been considered following enhanced reprocessing of spent fuel. The reference option is the transmutation by neutronic bombardment, the second option is to immobilize the separated radiotoxic elements in dedicated matrices ensuring long term stability. This paper summarizes the research and development activities carried out by CEA in France, concerning technetium. Nearly 90% of the technetium-99 present in spent fuel is separated in the UP2-800 reprocessing plant at La Hague, producing about 900 kg of Tc each year. The first part of the paper deals with the irradiation experiment ANTICORP 1 which aims at transmuting ⁹⁹Tc under the form of metallic rods, in the core of the Phenix reactor. The goal of ANTICORP 1 is to achieve a transmutation rate of ⁹⁹Tc in ¹⁰⁰Ru greater than 20%. The main options chosen for the design and the safety study of the experiment are described. In addition, the planning of the irradiation and the post-irradiation exams foreseen, are detailed. The second part of the paper describes the studies carried out to investigate technetium metal as a potential containment matrix. The corrosion rate of this metal under water was measured over one year of alteration at 25degC. (author)

No abstract available

[en] A key part of the IAEA 6th Coordinated Research Project on advances in high temperature reactor (HTR) fuel technology includes benchmarking of fuel performance models under normal and accident conditions. The normal operation and accident behaviour benchmarks have been structured in two phases. In the first phase, a series of simplified analytical benchmarking problems have been established for both normal and accident conditions as a way to 'calibrate' all of the codes and/or models. In the second phase, the codes and/or models will be used to calculate fuel behaviour in past and future irradiation experiments and heating tests. Current participants in the benchmark include England, France, Germany, Russia and the United States. This paper will present a status of this international code benchmarking activity. (author)

[en] The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project reached in 2015 the end of its conceptual design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phenix and SuperPhenix Sodium Fast Reactors. One of the biggest challenges of the last five years was to propose a consistent design for the reflectors and neutron shielding sub-assemblies surrounding the fuel core in order to fulfill ASTRID requirements of minimising the secondary sodium activity level. Heavy iterative studies on both core and sub-assemblies were necessary to propose and evaluate different solutions following a strict value analysis process considering neutron shielding performances, life duration, maturity levels, washing and manufacturing capability, and qualification needs. Evaluated options were reflectors sub-assemblies made of steel or MgO rods, and radial neutron shielding sub-assemblies made of B₄C or borated steel, with different configurations in the design and in the core layout. This paper presents the iterative engineering studies, conducted by CEA and performed by AREVA-NP, concerning the radial shielding sub-assemblies for ASTRID core, from the selection of possible solutions to a final consistent conceptual design. (author)