[en] Pure zirconium and four annealed α-zirconium-based alloys (Zr-1760 ppm wt O, Zr-1% Nb-430 ppm O, Zr-1% Nb-1800 ppm O, Zircaloy-4) have been studied by transmission electron microscopy after 500 keV Zr⁺ ion or 1 MeV electron irradiation performed at high temperature (≥ 400⁰C). The dislocation loops were found to all have b=1/3a<11anti 20> type Burgers vectors, and were preferentially in {10anti 10} type planes; in the case of electron-irradiated Zr-1760 ppm O, the larger loops were found to be of interstitial type. Alloying elements increase the loop density. The kinetics of loop growth was observed in-situ during 1 MeV electron irradiation between 400 and 700⁰C: oxygen was found to reduce considerably the growth speed of loops. In-situ annealing at 450 or 500⁰C after ion irradiation led to a large coalescence of loops in the case of pure zirconium, but modified only slightly the defect structure of the alloys. (orig.)

[en] During the last recent years investigations on structure and properties of high dense fuel types for Fast Breeder Reactors were performed. One of the most important properties is the thermal conductivity. As could be observed, microcracks in the material can influence the thermal conductivity up to about 20%. The occurrence of these microcracks is related to a local stoichiometry change in the fuel. The influence of extreme values of some important design parameters, such as: Fuel stoichiometry, pellet diameter, purity of the backfill gas, and sintering atmosphere, and their combinations on fuel pin behaviour was studied by pre-test modelling calculations. These calculations show, that a reduction of fuel stoichiometry from an oxygen-to-metal ratio of 1.97 to 1.93 results in a centerline temperature increase of about 8%. The influence of a change in the backfill gas composition from 90% He to 50% He is less pronounced. A larger pellet diameter leads, generally, to higher centerline temperatures during operation. (orig.)

[en] A series of tensile and strain controlled low-cycle fatigue tests were conducted at temperatures ranging from RT to 900⁰C on Hastelloy XR-II, which is one of the candidate alloys for applications in HTGR. Materials were tested in the solution annealed as well as in the pre-aged condition where aging consisted of isothermal exposure at 900⁰C for 1000 h. In those tests the effects of aging on tensile and fatigue properties were investigated. The ductility of the aged specimens was lower than that of the solution annealed ones under all test temperature conditions employed, and the tendency was more pronounced at and below 600⁰C. The fatigue lives of the aged specimens were shorter than those of the solution annealed ones at and below 700⁰C. The tendency became more pronounced under higher strain range conditions. The results were interpreted through the fractographic and metallographic features of the fatigued specimens. (orig.)

[en] Tensile specimens were prepared from 20/25/Nb stainless steel fuel pin cladding irradiated in a Commercial Advanced Gas-cooled Reactor (CAGR) at temperatures in the range 622-866 K and integrated fast neutron doses up to 16x10²⁴ n/m². The tests were performed in air at temperatures in the range 298-873 K at strain rates from 2x10^-5 s^-1 to 7.2 s^-1. The tensile properties varied with irradiation temperature, test temperature and strain rate. At lower irradiation temperature, strengthening produced by fast neutron damage was accompanied by reduced elongation. Strengthening was also observed at higher irradiation temperatures, possibly due to precipitation phenomena. The maximum irradiation embrittlement was observed in tests at 873 K at low strain rates between 2x10^-4 s^-1 and 2x10^-5 s^-1. The failure mode of embrittled specimens irradiated at higher temperatures was characterized by prematurely ruptured ductile fibres, rather than by intergranular cracking. (orig.)

[en] Gas release from the nuclear fuels UO₂ and UN out of pressurized closed pores produced by autoclave anneals has been studied by Thermal Desorption Spectrometry (TDS). Investigation of gas release during heating and cooling has indicated stress related mechanical effects leading to gas release. This release occurred in a narrow temperature range between about 1000 and 1500 K for UO₂, but it continued down to ambient temperature for UN. No burst release was observed above 1500 K for UO₂. (orig.)

[en] Letter-to-the-editors

[en] The heat capacities and the electrical conductivities of U_0.80Zr_0.20Mo_0.20 alloys were measured by means of direct heating pulse calorimetry in the temperature range from 300 to 1300 K and from 300 to 1150 K, respectively. Four peaks were observed in the heat capacity of U_0.80Zr_0.20 corresponding to four phase transitions. The transition temperatures were fairly in good agreement with those estimated from the phase diagram of the U-Zr system reported previously. Two peaks were also observed in the heat capacity of U_0.80Mo_0.20 corresponding to two phase transitions. The transition temperatures of U_0.80Mo_0.20 were in good agreement with those estimated from the phase diagram of the U-Mo system summarized previously. The thermal conductivity of U_0.80Zr_0.20 was calculated from the heat capacity obtained in this study and the thermal diffusivity data recently reported by Takahashi et al., and was very close to that of U_0.75Zr_0.25 experimentally determined by Leibowitz et al. The thermal conductivity of U_0.80Zr_0.20 calculated from the electrical conductivity in this study on the basis of the Wiedemann-Franz's law was a little lower than that calculated from the heat capacity and the experimental values of U_0.75Zr_0.25 by Leibowitz et al. The thermal conductivity of U_0.80Mo_0.20 was also calculated from the electrical conductivity on the basis of the Wiedemann-Franz's law, and fell between the experimental values of U_0.77Mo_0.23 and U_0.88Mo_0.12. (orig.)

[en] A reevaluation of phase equilibria in the binary U-Al system has been performed as part of the ASM-NBS Binary-Phase Diagram Evaluation Program. The resulting diagram includes several modifications to the recently assessed diagram by Chiotti et al. The current assessment also emphasizes the uranium-rich portion of the diagram. The consistency between available thermodynamic data for the system (Gibbs energies of formation of UAl₂, UAl₃, and UAl₄, and activity data for dilute solutions of uranium in liquid Al) and key points of the assessed diagram has been tested using the POTCOMP and FITBIN subroutines of the thermochemical computer code FACT. With the assumption of a regular solution model for Al-rich liquid, an ideal solution model for U-rich liquid, and Henrian models for U(β) and U(γ) solid solutions, there is reasonably good consistency between the thermo data and the assessed phase diagram. (orig.)

[en] The emf of galvanic cells Pt, C, [Te]_Pd-Te, TeO₂vertical stroke15 YSZvertical strokeO₂(P_O2=0.21 atm), Pt using 15 mass% yttria-stabilized zirconia (15 YSZ) as the solid electrolyte were measured for 15 different alloy compositions ranging from 5 to 64.3 at% Te over the temperature range 550 to 1000 K. Breaks in the plots of emf could be observed in the temperature interval of 740 to 780 K in the composition range 28 to 43 at% Te corresponding to peritectoid, eutectoid, eutectic and peritectic reactions. Combining these data with the emf expression for the cell Pt, C, Te(s or l), TeO₂(s)vertical stroke15 YSZvertical strokeO₂(P_O2=0.21 atm), Pt reported in the literature, values of log a_Te were computed. Typical expression for the log a_Te in the miscibility gap region between PdTe and PdTe₂ was calculated to be (log a_Te(PdTe/PdTe₂)±0.14)=2.3496-3001.1/T (613-995 K) with reference to liquid tellurium as the standard state. At 700 K, there seems to be a systematic increase in Δanti G_Te values for the compositions from 32 to 38 at% Te instead of a constant value expected from the phase diagrams. (orig.)

[en] Energy dispersive X-ray analysis revealed segregation of silicon and phosphorus to existing precipitates in martensitic DIN 1.4914 stainless steel during proton irradiation at 793 K to displacement doses of 0.64 dpa. The results can be quantitatively described by layers of about 7 nm silicon on M₂₃C₆ and about 4 nm phosphorus and ≤1 nm Si and Mn on NbC precipitates. (orig.)