[en] Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre (BARC), Tarapur, India is an industrial scale fuel manufacturing facility for fabricating (U-Pu) mixed oxide fuel for different types of reactors. MOX fuel pins for an experimental fuel sub assembly for irradiation testing in Fast Breeder Test Reactor (FBTR) at Kalpakkam were fabricated at our facility. AFFF has now taken up manufacturing of MOX fuel pins for the first core of PFBR. A comprehensive quality control plan was prepared based on the specification and advanced process and quality control procedures were adopted in order to meet the stringent quality requirement. A number of techniques were developed for fabricating fuel of required quality for FBTR and PFBR. The technology for making annular MOX pellets was developed using multistation rotary presses for the compaction of the pellets. Automation has been introduced wherever possible at different stages of fuel fabrication. Laser welding technique has been developed for welding the end plugs and a large number of bottom end plug welds were made using laser. A number of advanced quality control techniques were also developed to increase the confidence of MOX fuel produced. This paper outlines the details of fabrication of MOX fuel for PFBR and FBTR and describes the new techniques developed for fabrication, quality control/process control for MOX fuel for fast reactors. (author)

[en] The sodium cooled fast reactors programme in India is highlighted. Indigenously developed unique Pu-rich mixed carbide fuel containing 70% PuC and 30% UC, is used in Fast Breeder Test Reactor (FBTR). The clad and wrapper materials for FBTR are 20% CW 316M and 20% CW 316L respectively. The Mark-I core has seen a burnup of 165 000 MW-d-t^-1. One of the important achievements was closing of the fuel cycle of FBTR. The FBTR fuel discharged at 150 000 MW-d-t^-1 has been successfully reprocessed. A 500 MW(e) Prototype Fast Breeder Reactor (PFBR) uses mixed oxide with natural and depleted uranium. The core has two enrichment zones with 21% and 28% PuO2. For the core components, 20% cold worked D9 material (15%Cr-15%Ni with Ti and Mo) is used to have better irradiation resistance. The PFBR test fuel subassembly irradiated in FBTR has attained the burnup of 112 GW-d-t^-1. With advanced structural materials for clad and wrapper, a burnup of 20 at.% is envisaged for mixed oxide fuels. Roadmap has been conceived for the development of fuels and structural materials and test facilities for enhancing burnup gradually to 200 GW-d-t^-1, subsequently 250 GW-d-t^-1. Directed research is under way to develop metallic fuels for achieving high breeding ratio (1.4-1.5) and high burnup of around 25 at.% which will be employed in the long run in commercial FBRs. The economic advantages of high burnups, international experience on achieving high burnup, Indian roadmap of achieving high burnup fuels and materials, experiences of carbide fuel used in FBTR, comprehensive capability of Indira Gandhi Centre for Atomic Research (IGCAR) in the field of design, R and D particularly in the fields of mechanics and thermal hydraulics analysis of fuel pins and subassembly and post irradiation examination techniques are presented in this paper. (author)

[en] The innovative fast spectrum experimental facility MYRRHA is being developed at the Belgian Nuclear Research Center SCK-CEN. MYRRHA, a flexible fast spectrum research reactor (50-100 MW(th)), is conceived as an accelerator driven system (ADS) demonstrator, able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying core with MOX fuel, cooled by liquid lead-bismuth. The project started in 1997 and the reactor is forecasted to be fully operational around 2022-2023. The driver fuel for the MYRRHA core is one of its key components. Of special concern is the corrosion behaviour of the cladding materials (15-15Ti or T91 steels) in a lead-bismuth environment with the power profile envisaged for MYRRHA. An overview of the MYRRHA design and the R and D effort ongoing or planned is presented. (author)

[en] The results of fabrication and post-irradiation examination of helium-bonded fuel pins with U^0,55Pu_0,45N and U_0,4Pu_0,6N and claddings made of austenitic ChS-68 CW steel irradiated in the BOR-60 reactor are given. The maximum burnup is 9.4 at.% at maximum linear rating around 430 W-cm^-1 for U_0,55Pu_0,45N and 12.1 at.% at 540 W.cm^-1 for U_0,4Pu_0,6N. All fuel pins are tight. The BORA-BORA results confirm the possibility to provide at least 12 at.% of burnup for He-bonded nitride pins. (author)

[en] The current RIAR activity is manufacturing, in-pile tests and PIE of granulated vibropac MOX fuel rods and FAs. Fuel of various origins (electrochemical granulation in molten salts, granulation by sol-gel process, grains after pellet fuel crushing) can be used for manufacturing of vibropac fuel rods. Plutonium of various grade and purification degree (from FPs) has been used for production of this fuel type (weapon-grade plutonium (WG); power generating plutonium (RG), including plutonium with low degree of purification from FPs and with MA additives). The total amount of manufactured fuel rods for BOR-60, BN-350, BN-600 reactors is more than 22 000. Parallel with scale fabrication of fuel rods and FA, RIAR develops procedures of granulated oxide fuel production and processes of fuel rods fabrication for fast reactors of the next generation. There are fuel rods containing recycled PuO₂, residual fission products (up to 8 wt%), recycled MOX-fuel, MA additives NpO₂, Am_xO_y (2-5 wt%), high PuO₂ content (up to 45 wt%), etc. The fuel rods are tested in the BOR-60 reactor mostly in dismountable FAs that make it possible to examine the irradiated fuel rods at an intermediate stage by using nondestructive methods. (author)

[en] Recent developments of the TRANSURANUS code concerning the modelling and simulation of nuclear fuel rods for fast breeder reactors (FBR) are presented. The work deals with the following main topics: extension of the burnup model (TUBRNP) for fast spectrum, plutonium and oxygen redistribution, normal grain growth and helium release. The extended version of TUBRNP allows now to perform nuclide analysis also for FBRs. This is actually an important task for fuel performance codes since it allows describing the major and minor actinides transport and evolution across the fuel pellet, which represents a key issue for the behaviour of FBR fuel pins. To this purpose, the models of transport phenomena involving plutonium and oxygen redistribution have been revised and improved. Based on the multiscale philosophy (in particular, making use of neutron transport calculation results and experimental data coming from out-of-pile experiments), neutron cross sections and fission yields for FBRs were included and oxygen and plutonium diffusion coefficients were upgraded. As concerns helium production and release, which is an important issue during irradiation and storage, a simplified model was also implemented that makes use of experimental data from separate effect experiments. Further code improvements for FBR that are underway are outlined and discussed. (author)

[en] The French Generation IV prototype ASTRID should be operated at the same temperatures as PHENIX and SUPERPHENIX reactors. Thus, feedback experience from PHENIX which operated during 35 years, give us guidance for the choice of materials for ASTRID prototype and in future for SFR cores. This paper described the behaviour induced by irradiation of materials, from austenitic steels to 9Cr martensitic steels and ODS alloys. The microstructure at the origin of this behaviour of steels used for fuel pin cladding and wrapper tubes was also described in this paper. Improvements in swelling resistance of materials, from SA 316, first material used in PHENIX core, to CW 15-15Ti steels and 9Cr martensitic steels, have led to a gain of a factor 3 in the life time of standard fuel subassemblies in reactor. (author)

[en] Closing the nuclear fuel cycle and the deployment of fast reactors will be a major step in increasing the sustainability of uranium resources, but highest safety standards must be applied. The Joint Research Centre (JRC) has over 40 years of experience in studying the safety of fuels for fast reactors and the necessary reprocessing cycles. The majority of the early programmes focussed on advanced driver fuels, in particular carbides and nitrides. More recently attention has turned to advanced oxide and metal alloy fuels bearing minor actinides (MA) for either homogenous or heterogeneous MA recycling. For the former fuel form small additions of MA to MOX fuels are considered. For the latter, targets based on UO2 or the so called inert matrix fuel (IMF) concept (largely considered for accelerator driven systems (ADS)) have been studied and safety aspects investigated. Innovative liquid conversion processes, beyond the state of the art, have been used to manufacture MA bearing fuels. Advanced techniques have been developed and used for the determination of their properties, such as thermal conductivity, phase diagrams, vaporisation behaviour, etc., before and after irradiation. A number of irradiation performance tests and reprocessing runs have been performed, with the SUPERFACT irradiation experiment being a major milestone. Though significant advances have been made, by no means can such fuels be considered as qualified. Within this presentation results from past and present programmes will be presented and prospects for the future discussed. (author)

[en] The BN-600 is the most powerful power fast reactor in the world. Uranium oxide fuel of three enrichments on U-235 is used. The FA lifetime is 560 effective full power days (EFPDs), two reloadings per year with average duration between reloadings, 140 EFPDs. The maximum damage dose rate is 41 dpa per year. Sodium temperature range in the core is 368-550^oC. The present unique combination of irradiation conditions is extremely attractive to support the irradiation of materials and items for the testing purposes. The BN-600 is the power reactor with assigned commercial parameters, and the tests should not essentially influence its operation mode. Besides it is necessary to take into account normative safety rules and limits on allowed perturbations in the reactivity margin and the heat release distribution. At preparation of tests licensing the nuclear and radiation safety justification should be supported with theoretical and experimental results. The paper describes the BN-600 irradiation capabilities, irradiation test experience and requirements for the irradiation tests arrangement. (author)

[en] This paper briefly descripts the fuel development status for a fast reactor in China. MOX fuel is considered for the usage in the China Experimental Fast Reactor (CEFR) and in the future China Demonstration Fast Reactor (CDFR) at the first stage, and metallic fuel at the second stage. As CEFR will be put into operation after physical and power tests, irradiation technologies are being progressed and some irradiation experiments are on the plan. The introduction about irradiation technologies developed and tests under consideration are also included. (author)