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The ECRIX-H experiment aims to assess the feasibility of transmuting americium micro-dispersed in an inert magnesia matrix under a locally moderated neutron flux in the Phenix reactor. A first set of examinations demonstrated that pellet behaviour was satisfactory with moderate swelling at the end of the irradiation. Additional post-irradiation examinations needed to be conducted to confirm the high transmutation rate so as to definitively conclude on the success of the ECRIX-H experiment. This article presents and discusses the results of these new examinations. They confirm the satisfactory behaviour of the MgO matrix not only during the basic irradiation but also during post-irradiation thermal transients. These examinations also provide additional information on the behaviour of fission products both in the americium-based particles and in the MgO matrix. These results particularly validate the transmutation rate predicted by the calculation codes using several different analytical techniques. The fission rate is also determined.

The HELIOS irradiation experiment is the latest of a series of experiments on americium transmutation, and has been carried out in the framework of EURATOM's 6th Framework Programme (FP6) project EUROTRANS, which has been completed in March 2010. The transmutation of Minor Actinide (MA) is a fundamental step in order to be able to close the cycle of the nuclear fuel. Past experimental activities in the field of transmutation and testing of innovative nuclear fuel containing Am has proved that the release or trapping of helium is a key issue for the design of such kind of fuel or targets. Therefore, the main objective of the HELIOS experiment is the study of the in-pile behaviour of U-free fuels such as (Pu,Am,Zr,Y)O2 or (Am,Zr,Y)O2, CERCER (Am2Zr2O7+MgO), CERMET ((Pu,Am)O2+Mo) or ((Am,Zr,Y)O2+Mo) in order to gain knowledge on the role of the fuel microstructure and of the operating temperature on gas release and fuel swelling. The HELIOS irradiation experiment started in the HFR (High Flux Reactor) in Petten (The Netherlands) on the 29th of April 2010 and finished on the 19th of February after 9 reactor cycles (~241 full power days). Although only the Post Irradiation Examination (not performed yet) conducted on the CERCER and CERMET target fuel tested during the HELIOS irradiation experiment will show the performance of the fuel, the behaviour of such targets during the irradiation did not show any difficulties. It is possible to conclude that from an operational point of view, these kinds of fuel targets which have been developed mainly having in mind the possibility to burn MA in a in a subcritical nuclear system, did not show significant issues. This paper summarises the main experimental data obtained during the 9-cycle irradiation of the HELIOS experiment in the HFR., JRC.F.4-Safety of future nuclear reactors

The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO 1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.). The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO x -type phase was created within the initial AmO 1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.

The present paper describes laboratory studies performed on composite materials made of a spinel inert matrix in which uranium dioxide particles are dispersed. This study aims to improve the lifetime of the standard nuclear fuel. The work involves the modelling of the thermomechanical in-pile behavior of these materials, manufacturing tests and out-of-pile characterizations. The influence of some physical characteristics such as the particle size and the nature of the matrix–particle interface inside the composite is analysed. The results show a very good out-of-pile mechanical behavior of the composite material, especially a high ability to deform and excellent thermal gradient resistance.

Experimental capsules irradiated in the fast neutron reactor Phenix could not be instrumented with classical thermocouples to monitor the internal temperature of materials during irradiation. Thereby two different types of passive temperature monitors were designed to determine the maximal temperature reached inside an irradiation experiment of inert materials in Phenix called FUTURIX MI. These two passive monitors consist in 1) several fuses of pure metal or metallic alloys processing well known melting points, 2) SiC specimens for which the swelling caused by irradiation damages is directly linked to the operating temperature. The FUTURIX MI experiment equipped with these passive monitors, has been loaded in the reactor on May 2007 and removed on March 2009, after a 235 Equivalent Full Power Days (EFPD) irradiation. The technologies used to elaborate the temperature monitor are depicted in this paper.