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AbstractAbstract
[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.)
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International conference on fundamental mechanisms of radiation-induced creep and growth; Hecla Island, Manitoba (Canada); 22-25 Jun 1987
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ALLOYS, ATOMIC IONS, BEAMS, CHARGED PARTICLES, CHROMIUM ADDITIONS, CORROSION RESISTANT ALLOYS, CRYSTAL DEFECTS, CRYSTAL STRUCTURE, CRYSTALS, ELECTRON MICROSCOPY, ELEMENTS, ENERGY RANGE, HEAT RESISTING ALLOYS, HEAT TREATMENTS, IONS, IRON ADDITIONS, KEV RANGE, LEPTON BEAMS, LINE DEFECTS, METALS, MICROSCOPY, PARTICLE BEAMS, RADIATION EFFECTS, TIN ALLOYS, TRANSITION ELEMENTS, ZIRCALOY, ZIRCONIUM BASE ALLOYS
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[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.)
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3. conference on characterization and quality control of nuclear fuels; 3. Konferenz ueber Charakterisierung und Qualitaetskontrolle von Kernbrennstoffen; Karlsruhe (Germany, F.R.); 25-27 May 1987
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[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.)
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AGING, ALLOY-NI50CR22FE18MO9, FAILURES, FATIGUE, FRACTOGRAPHY, FRACTURES, HEAT RESISTING ALLOYS, HIGH TEMPERATURE, HTGR TYPE REACTORS, MECHANICAL TESTS, MEDIUM TEMPERATURE, MICROSTRUCTURE, SCANNING ELECTRON MICROSCOPY, STRAINS, STRESSES, TEMPERATURE DEPENDENCE, TENSILE PROPERTIES, VERY HIGH TEMPERATURE
ALLOYS, CHROMIUM ALLOYS, CORROSION RESISTANT ALLOYS, CRYSTAL STRUCTURE, ELECTRON MICROSCOPY, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, HASTELLOYS, IRON ALLOYS, MATERIALS TESTING, MECHANICAL PROPERTIES, MICROSCOPY, MOLYBDENUM ALLOYS, NICKEL ALLOYS, NICKEL BASE ALLOYS, REACTORS, TESTING, TUNGSTEN ADDITIONS
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[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.)
Original Title
steel composition in wt%: 25% Ni; 20% Cr; 0,6-0,8% Nb
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AGR TYPE REACTORS, CHROMIUM-NICKEL STEELS, CLADDING, ELONGATION, FAST NEUTRONS, FRACTURES, FUEL PINS, HIGH TEMPERATURE, IRRADIATION, MECHANICAL TESTS, MEDIUM TEMPERATURE, NIOBIUM ADDITIONS, PHYSICAL RADIATION EFFECTS, SCANNING ELECTRON MICROSCOPY, STRAIN RATE, STRESSES, SURFACES, TEMPERATURE DEPENDENCE, TENSILE PROPERTIES
ALLOYS, BARYONS, CARBON ADDITIONS, CHROMIUM ALLOYS, DEFORMATION, DEPOSITION, ELECTRON MICROSCOPY, ELEMENTARY PARTICLES, ENRICHED URANIUM REACTORS, FAILURES, FERMIONS, FUEL ELEMENTS, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, HADRONS, HIGH ALLOY STEELS, IRON ALLOYS, IRON BASE ALLOYS, MATERIALS TESTING, MECHANICAL PROPERTIES, MICROSCOPY, NEUTRONS, NICKEL ALLOYS, NUCLEONS, RADIATION EFFECTS, REACTOR COMPONENTS, REACTORS, STAINLESS STEELS, STEELS, SURFACE COATING, TESTING
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[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.)
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[en] Letter-to-the-editors
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ALKALI METALS, ALLOYS, AUSTENITIC STEELS, CARBON ADDITIONS, CHEMICAL REACTIONS, CHROMIUM ALLOYS, CHROMIUM-NICKEL STEELS, CORROSION RESISTANT ALLOYS, ELEMENTS, HEAT RESISTING ALLOYS, HIGH ALLOY STEELS, INFORMATION, IRON ALLOYS, IRON BASE ALLOYS, METALS, MOLYBDENUM ALLOYS, NICKEL ALLOYS, STAINLESS STEELS, STEELS
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[en] The heat capacities and the electrical conductivities of U0.80Zr0.20Mo0.20 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 U0.80Zr0.20 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 U0.80Mo0.20 corresponding to two phase transitions. The transition temperatures of U0.80Mo0.20 were in good agreement with those estimated from the phase diagram of the U-Mo system summarized previously. The thermal conductivity of U0.80Zr0.20 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 U0.75Zr0.25 experimentally determined by Leibowitz et al. The thermal conductivity of U0.80Zr0.20 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 U0.75Zr0.25 by Leibowitz et al. The thermal conductivity of U0.80Mo0.20 was also calculated from the electrical conductivity on the basis of the Wiedemann-Franz's law, and fell between the experimental values of U0.77Mo0.23 and U0.88Mo0.12. (orig.)
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7. international symposium on thermodynamics of nuclear materials (STNM-7) in conjunction with the combined TMS fall meeting and world materials congress (WMC); Chicago, IL (USA); 24-30 Sep 1988
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[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.)
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7. international symposium on thermodynamics of nuclear materials (STNM-7) in conjunction with the combined TMS fall meeting and world materials congress (WMC); Chicago, IL (USA); 24-30 Sep 1988; CONTRACT W-7405-ENG-48
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[en] The emf of galvanic cells Pt, C, [Te]Pd-Te, TeO2vertical stroke15 YSZvertical strokeO2(PO2=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(PO2=0.21 atm), Pt reported in the literature, values of log aTe were computed. Typical expression for the log aTe in the miscibility gap region between PdTe and PdTe2 was calculated to be (log aTe(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 GTe values for the compositions from 32 to 38 at% Te instead of a constant value expected from the phase diagrams. (orig.)
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7. international symposium on thermodynamics of nuclear materials (STNM-7) in conjunction with the combined TMS fall meeting and world materials congress (WMC); Chicago, IL (USA); 24-30 Sep 1988
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[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.)
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