[en] We have studied radiation damage in TaC _1-x at 21 K, by means of electrical resistivity measurements. Threshold displacement energies and specific resistivities of C and of Ta defects have been determined, as well as the C recombinations volume. From isochronal and isothermal recovery experiments versus composition x, electron energy and dose we have separated the recovery stages of C and of Ta defects (three stages at 80 K, 170 K and 275 K for C and one stage between 40 and 180 K for Ta). They are due to interstitial migration mechanisms. The first stage at 80 K results from close pair recombination with a given activation energy. The three stages structures of C defects desappears for x=0,20, with just one single stage (between 40 K and 250 K, 200% high). This can be related to modifications of the vacancy order parameters due to instertitial migration

[en] A novel neutron absorbent composite material comprises a homogeneous boron carbide (B₄C) matrix, in which are homogeneously dispersed sized clusters of HfB₂, TiB₂ and/or ZrB₂. Production of the above composite material involves homogeneously mixing B₄C powder of less than 5 microns average particle size with the sized boride clusters, and hot pressing (uniaxially or isostatically) the mixture to a density greater than 90% of theoretical density. It can be used in nuclear reactors such as PWRs and fast neutron reactors. The boride clusters improve the behaviour (cracking resistance) under irradiation of the B₄C, while maintaining the maximum possible neutron absorption efficiency. 4 figs

[en] In the case of a hypothetical loss of primary coolant accident (LOCA) in a light water reactor, the zirconium alloys fuel cladding would be oxidized in steam at high temperature, typically in the range 800–1200 °C. The monoclinic to tetragonal phase martensitic transition of zirconia occurs within this temperature range and complex phenomena possibly having an impact on the oxidation kinetics are then to be expected. In order to provide an accurate description of the structure and microstructure of the oxide layers, systematic X-ray diffraction analyses have been performed in-situ under oxidizing atmosphere at high temperature (between 800 and 1100 °C) on Zircaloy-4 and M5™ sheet samples. It was confirmed that the volume fraction of the tetragonal and monoclinic zirconia phases formed during oxide growth drastically depends on the oxidation temperature. For example, the few outer microns of the oxide are fully tetragonal above 1050 °C and contain only 20% of tetragonal phase at 800 °C. It was also shown that cooling after oxidation induces irreversible phase transitions within the oxide. As a consequence, both the structure and the microstructure of the growing oxide cannot be observed post-facto, neither at room temperature nor after reheating at the prior oxidation temperature. It has been deduced from microstructural analyses that the grain size of the tetragonal zirconia phase is nanometric, about 100 nm during oxidation at 1100 °C down to 20 nm after cooling down to room temperature. This small grain size allows the stabilization of the tetragonal phase. The lattice parameters of the monoclinic and tetragonal zirconia phases have been analyzed, during both high temperature oxidation and cooling. In both cases, it appears the ‘a’ and ‘b’ cell parameters of the monoclinic phase are strongly constrained by the tetragonal ‘a’ one. The structural characteristics of the oxide formed at high temperature on Zircaloy-4 and M5™ are quite similar. All those results can be interpreted in the frame of the classical description of the monoclinic–tetragonal martensitic transition of zirconia

[en] Neutron absorbing materials are used to control neutron flux inside reactor cores. The 2 most used materials in PWR and sodium-cooled fast reactors are respectively Ag-In-Cd (AIC) and B₄C. It appears that both materials, because of the damaging effect of neutron irradiation can not comply to initial specifications in terms of operating life. For instance the change in the chemical composition is dramatic: initially the AIC alloy contains Ag (85%), In (15%), Cd (5%), at the end of 8 cycles the composition becomes: Ag (68%), In (7%), Cd (18%) and Sn (7%) and this change implies changes in neutron absorbing properties and in mechanical and thermal properties which are more or less difficult to manage. Research programs led in late USSR and in France recommend for light water reactors a come back to hafnium or the use of a more exotic material: the dysprosium titanate (DyTi₂O₅). As for fast reactors, di-boride compounds like MB₂ where M stands for Ti, Zr or Hf, and Dy₂HfO₅ are being studied as new absorbing materials. (A.C.)

[en] Fusion devices high heat loading due to off-normal events (e.g., plasma disruption, slow transients and ELMs, which can occur during a transition from detached to attached divertor operation) requires high thermal conductivity materials. Therefore, carbon fiber composites (CFCs) with high thermal conductivity are favorable. In the framework of the European Fusion Technology program, a great effort has been made to develop Si doped CFCs. NS31 is a 3D CFC containing about 8-10 at.% of silicon. The previous results showed, that NS31 poses lower chemical erosion, lower tritium retention and higher resistivity to water/oxygen reaction in comparing with undoped CFCs. Off-normal simulation experiments were performed under two conditions: (a) 700 MW/m², 10 ms and (b) slow transient 20 MW/m², 2 and 4 s. NS31 behaved very stable even under these extremely severe conditions. In this paper, the detailed results of simulation experiment on high heat loading due to off-normal events are presented and consequences are discussed

[en] Zirconia produced by the oxidation of zirconium alloys in nuclear reactors exhibits a phase transition under ionic irradiation, simulating a neutron irradiation. To understand the mechanism responsible for this irradiation driven phase transition, different kinds of projectiles were used to irradiate pure monoclinic zirconia samples. The evolution of these irradiated samples as a function of dpa has been studied using grazing X-ray diffraction. The Rietveld method has been applied on collected X-ray diffraction diagrams to study the phase produced under irradiation and the kinetics of its formation. Even at high dpa values, only the monoclinic and tetragonal phases were used to simulate X-ray diffraction diagrams. No amorphisation of zirconia was observed. The evolution of unit cells and short range strains in both phases under irradiation leads us to think that the irradiation driven transition is martensitic. Supposing that the inelastic stopping power in sub-cascades is responsible for the irradiation driven phase transition, we propose a model based on the Landau-Ginzburg effective hamiltonian to explain both the m→t transition observed under irradiation and the t→m transition measured during isochronal annealing after irradiation

[en] The spinel MgAl₂O₄ is a ceramic with relatively high thermal conductivity and results on good behaviour under irradiation by neutrons had been published since twenty years. This material is therefore among the inert matrix studied for nuclear uses for the future: targets for the transmutation of the actinide nuclear wastes, and plutonium nuclear fuels without uranium. Unfortunately, unexpected important swellings (up to more than 20%) had been measured after some of our recent irradiations with neutrons and fission products in nuclear fission reactors. On the other hand, these results on swelling are not systematically reproducible. This study has been done to determine the influent parameters on the behaviour of the ceramics under irradiation and particularly on the swelling. The polished surface of different disks of sintered polycrystalline spinel have been irradiated at room temperature and 500 deg C, at different fluencies between 10¹¹ and 5.10¹⁶ ions/cm², with different heavy ions (from S to Pb) of energy up to more than 1 GeV. Disks of ZnAl₂O₄ spinel have been also irradiated in the same conditions for more accurate characterisation of the irradiated zone by grazing XRD. The influence of electronic stopping power and fluence on the deviation of the lattice parameter of the oxides will be analysed: the lattice parameter variation presents a maximum for dE/dx near 7 keV/nm. The conditions for amorphization will be discussed, and the damages produced by nuclear stopping power or electronic stopping power will be compared. Characterisation of the defects has been done by optical spectroscopy before and after irradiation, and the results will be reported. The influence of the spinel tested (purity, grain size, density and internal stress) on the behaviour under irradiation will be analyzed. Outline of mechanism of ceramic damage under irradiation will be presented. (authors)

[en] We have irradiated samples of TaCsub(0.99) with electrons of different energies. Damage rate and isochronal recovery curves lead us to assume two kinds of defects, each characterised by a different threshold energy and different recovery stages

[en] HfB₂, a solid poor in boron, was irradiated by thermal neutrons in an experimental reactor. Using a nuclear microprobe, we have tracked lithium atoms produced by the ¹⁰B(n,α)⁷Li reaction and compared the calculated and measured ⁷Li profiles in HfB₂ irradiated samples. This comparison shows that Li atoms do not diffuse during irradiation (323 K). The comparison of non-annealed and annealed irradiated HfB₂ plates clearly shows that lithium atoms do not migrate out of samples even at high temperatures (1273 K). These results associated to previous transmission electron microscopy (TEM) observations seem to show that lithium atoms are trapped by dislocation loops created by displacement cascades during neutron irradiation

[en] For the future generation of PWRs, improvements in efficiency, life time and safety of the control rods are considered. An increase of reactivity worth will be particularly requested in case of reactors highly loaded with MOX fuel assemblies or operating with a limited boron content in water. A preliminary research has been undertaken to improve the design of the absorber material. We present calculations and out of pile experiment results revealing the advantage of using a combination of boron and hafnium base materials. These results show that two types of absorbers should be studied and developed: boron 10 enriched B₄C pellets/hafnium cladding and boron 10 enriched HfB₂ pellets/hafnium cladding. Due to the neutron efficiency of the hafnium, a significant lowering of the irradiation damage of the pellets is obtained. The different aspects of the expected irradiation behaviour of such control rods are discussed. (author). 19 refs, 3 figs, 6 tabs