Showing papers on "Zirconium alloy published in 2000"
TL;DR: In this paper, two alloys, having different oxidation behavior (Zy4 and Zr-1NbO), have been investigated during oxidation at high temperature (743 K) and low oxygen pressure (10 kPa) by in situ X-ray diffraction (XRD).
125 citations
01 Jan 2000
115 citations
TL;DR: In this paper, a finite element model is presented for hydrogen embrittlement of zirconium alloys, which takes into account the coupled processes of hydrogen diffusion, non-mechanical energy flow, hydride precipitation and fracture.
73 citations
TL;DR: In this article, the effect of hydride precipitation on the mechanical behavior and on the rupture mechanism of ZIRCALOY-4 sheets at room temperature was investigated, in the range from 200 to 1200 wt ppm hydrogen and for different stress triaxialities.
Abstract: The mechanical behavior of ZIRCALOY-4 sheets is investigated at room temperature. The effect of hydride precipitation on the mechanical behavior and on the rupture mechanism is also studied, in the range from 200 to 1200 wt ppm hydrogen and for different stress triaxialities. It is shown that the material exhibits a strong anisotropy due to its pronounced texture, and that its mechanical properties depend on the strain rate. Hydride precipitation appears to have no effect on the anisotropy or on the strain-rate sensitivity, in the range from 10−4 to 10−2 s−1. The main effect of hydrogen is the reduction of the ductility and of crack resistance. The ductile rupture mechanism is studied, focusing on the stage of damage nucleation by hydride fracture. Observations during scanning electron microscopy (SEM) in situ tests show that hydrides allow the transmission of slip, which occurs in ZIRCALOY-4 grains. Hydrides can also deform, together with surrounding zirconium matrix. Damage appears after a plastic-strain yield of about 15 to 25 pct. Fracture occurs first on intergranular hydrides. Fracture of transgranular hydrides is observed only prior to failure, for higher plastic strains.
72 citations
Patent•
08 Dec 2000
TL;DR: Methods for manufacturing titanium and zirconium alloys, particularly including such alloys in powder form by continuous vapor phase chloroaluminothermic reduction, were described in this article.
Abstract: Methods for manufacturing titanium and zirconium alloys, particularly including such alloys in powder form by continuous vapor phase chloroaluminothermic reduction
62 citations
TL;DR: In this paper, the authors analyzed long-term Zircaloy autoclave corrosion data to develop new Zircaleoy corrosion correlations and found that the post-transition corrosion data can best be described by two successive linear equations, each active over a different range.
61 citations
TL;DR: In this paper, the effect of Fe addition on hydrogen storage characteristics of Ti 0.16 Zr 0.05 Cr 0.22 V 0.57 alloy has been studied at 303 K.
61 citations
51 citations
01 Jan 2000
51 citations
TL;DR: In this paper, the incorporation of precipitates β-Nb and Zr(Fe, V) 2 into the oxide films formed on advanced Zr-based alloys (Zr 1%Nb, Zr 0.5%Sn, 0.6%Fe 0.3%V) was studied by transmission electron microscopy, on cross-sectional thin foils.
47 citations
TL;DR: In this paper, the effects of yttrium ion implantation on the aqueous corrosion behavior of zircaloy-4 were studied using X-ray photo-emission spectroscopy (XPS).
Abstract: In order to study the effects of yttrium ion implantation on the aqueous corrosion behavior of zircaloy-4, specimens were implanted with yttrium ions using a MEVVA source at an energy of 40 keV, with a dose range from 1 × 1016 to 1 × 1017 ions/cm2 at about 150°C. Transmission electron microscopy (TEM) was used to obtain the structural character of the yttrium-implanted zircaloy-4. The valence of the yttrium ions in the surface layer was analyzed by X-ray photoemission spectroscopy (XPS). Three-sweep potentiodynamic polarization measurement was employed to evaluate the aqueous corrosion behavior of zircaloy-4 in a 1 N H2SO4 solution. It was found that a significant improvement was achieved in the aqueous corrosion resistance of zircaloy-4 compared with that of the as-received zircaloy-4. The mechanism of the corrosion resistance improvement of the yttrium-implanted zircaloy-4 is probably due to the addition of the yttrium oxide dispersoid into the zirconium matrix.
TL;DR: In this paper, the microstructure of stainless steel-zirconium alloys containing 2 wt% technetium was characterized using a combination of microscopy, spectroscopy, diffraction, and chemical analysis techniques.
15 Feb 2000-Materials Science and Engineering A-structural Materials Properties Microstructure and Processing
TL;DR: In this article, the microstructure and mechanical properties of the as-solidified and extruded alloys were evaluated and it was shown that the addition of a quaternary element resulted in the formation of very fine particles that contributed to the improved mechanical properties.
Abstract: Rapidly solidified Al–20Si–5Fe–2X (X=Cr, Zr, or Ni) alloys produced by gas atomization were degassed under different vacuum levels before consolidating them by hot extrusion. The microstructure and mechanical properties of the as-solidified and extruded alloys were evaluated. It was shown that the addition of a quaternary element resulted in the formation of very fine particles that contributed to the improved mechanical properties of the alloys, especially at elevated temperatures. Of these additions, nickel was found to be the most effective in increasing the UTS of the alloy both at room and elevated temperatures. Degassing the rapidly solidified powders under better vacuum improved the properties further.
Patent•
28 Sep 2000TL;DR: In this article, a zirconium alloy for nuclear reactor components was proposed, which contained 0.02 to 1 % iron, 0.8 to 2.3 % niobium, less than 2000 ppm of tin, more than 100 ppm of carbon, from 5 to 35 ppm of sulphur and less than 0.25 % of chromium and/or vanadium.
Abstract: The invention concerns a zirconium alloy for nuclear reactor components, containing, besides the unavoidable impurities, 0.02 to 1 % iron, 0.8 to 2.3 % niobium, less than 2000 ppm of tin, less than 2000 ppm of oxygen, less than 100 ppm of carbon, from 5 to 35 ppm of sulphur and less than a total of 0.25 % of chromium and/or vanadium. The ratio of niobium content at less than 5 % over the iron content, optionally completed with chromium and/or vanadium content, is more than 2.5.
TL;DR: In this paper, the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes was estimated by finite element method (FEM) analysis.
Abstract: In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and ...
TL;DR: In this paper, the authors modified the steady-state model of delayed hydride cracking to take account of a temperature gradient in the material and predicted an increase in the crack velocity in a positive temperature gradient (crack-tip cooler than surroundings) and a reduction in a negative temperature gradient.
TL;DR: In this article, an autoclave corrosion test was carried out in 400°C steam for 300 days on the Zr-0.5Nb-1.0Sn−0.25Cr alloy, which had been given 18 different accumulated annealing parameters.
TL;DR: The compositional dependence of the soft magnetic properties of the nanocrystalline Fe-Zr-Nb-B alloys has been investigated in this article, where the authors showed that the Fe85.5Zr2Nb4B8.5 alloy exhibits a very low core loss of 0.09 W/kg at 1.4 T and 50 Hz, which is extremely lower than that of Fe-Si-B amorphous alloys.
Abstract: The compositional dependence of the soft magnetic properties of the nanocrystalline Fe–Zr–Nb–B alloys has been investigated. The magnetostriction (λs) and the grain size of the (Fe90Zr7B3)1−x(Fe84Nb7B9)x alloys, which are two typical ternary alloys mixed with the best soft magnetic properties, show intermediate values between those of the Fe90Zr7B3 with negative λs and the Fe84Nb7B9 with positive λs. However, the soft magnetic properties of the Fe–(Zr, Nb)7–B alloys are inferior to those of the Fe90Zr7B3 and the Fe84Nb7B9 alloys. The best soft magnetic properties have been obtained at Zr+Nb=6 at. %. The Fe85.5Zr2Nb4B8.5 alloy shows a high μe of 60 000 at 1 kHz, a high Bs of 1.64 T, and zero λs, simultaneously. The alloy also exhibits a very low core loss of 0.09 W/kg at 1.4 T and 50 Hz, which is extremely lower than that of Fe–Si–B amorphous alloys. The nanocrystalline Fe–Zr–Nb–B alloys with improved soft magnetic properties are therefore suitable for pole transformers.
TL;DR: In this paper, the morphological, micro-chemical and micro-mechanical characteristics of the thick zirconium oxide layer formed on a zircaloy-4 clad after a standard water corrosion test (ASTM G2-67) were studied.
TL;DR: In this paper, high energy mechanical milling of Mg powders with either crystalline ZrNiCr and ZrNIC 0.6Cr 0.4 powders or mechanically milled amorphous Zr NiCr and NNi1.6C0.4 powder is shown to be an effective hydrogen absorption and desorption catalyst.
TL;DR: In this article, the authors have conducted diffusion experiments at 923 K using two couples: U-13 at%Pu-22 at%Zr/Fe and U-22at%Pus- 22 at% Zr /Fe, and examined the influence of the Pu content in the fuel alloy on the phases formed in the reaction zones.
Abstract: In metallic U-Pu-Zr fuel, metallurgical reactions occur between the fuel slug and the cladding, and a phase of which the melting point is relatively low is formed in the reaction zone. If a liquid phase is formed, it can degrade cladding integrity. The potential for liquid phase formation near the cladding, therefore, should be excluded during normal reactor operation. In order to clarify the mechanism of liquefaction, the authors have conducted diffusion experiments at 923 K using two couples: U-13 at%Pu-22 at%Zr/Fe and U-22at%Pu- 22at%Zr/Fe, and examined the influence of the Pu content in the fuel alloy on the phases formed in the reaction zones. The liquid phase has been observed in the U-22 at%Pu-22 at%Zr/Fe couple. An assessment of the diffusion paths for these couples has indicated that the Pu content in the (U, Pu)6Fe-type phase in the reaction zone is a crucial factor in determining the conditions that lead to liquefaction. The Pu content in the (U, Pu)6Fe-type phase increases with that in the ini...
TL;DR: The substitution of 1.5 at% Zr by Ti, Ta or Mo in the basic Fe56Co7Ni7Zr7.5B20 alloy has the effect of increasing the glass-forming ability, as well as increasing the Ms from 1.51 to 2.08 μ Ω m and decreasing the Hc from 9.5 to 6.1 A/m as discussed by the authors.
01 Jan 2000
Patent•
08 Mar 2000TL;DR: In this article, a niobium-containing zirconium alloy for use in nuclear fuel cladding is presented, which is characterized as comprising an alloy composition as follows.
Abstract: The invention presented herein relates to a niobium-containing zirconium alloy for use in nuclear fuel cladding. The Zr alloy of this invention with superior corrosion resistance is characterized as comprising an alloy composition as follows: 1) niobium (Nb), in a range of 0.8 to 1.2 wt. %; one or more elements selected from the group consisting of iron (Fe), molybdenum (Mo), copper (Cu) and manganese (Mn), in a range of 0.1 to 0.3 wt. %, respectively; oxygen (O), in a range of 600 to 1400 ppm; silicon (Si), in a range of 80 to 120 ppm; and the balance being of Zr, 2) Nb, in a range of 1.3 to 1.8 wt. %; tin (Sn), in a range of 0.2 to 0.5 wt. %; one element selected from the group consisting of Fe, Mo, Cu and Mn, in a range of 0.1 to 0.3 wt. %; O, in a range of 600 to 1400 ppm; Si, in a range of 80 to 120 ppm; and the balance being of Zr, 3) Nb, in a range of 1.3 to 1.8 wt. %; Sn, in a range of 0.2 to 0.5 wt. %; Fe, in a range of 0.1 to 0.3 wt. %; one element selected from the group consisting of chromium (Cr), Mo, Cu and Mn, in a range of 0.1 to 0.3 wt. %; O, in a range of 600 to 1400 ppm; Si, in a range of 80 to 120 ppm; and the balance being of Zr, and 4) Nb, in a range of 0.3 to 1.2 wt. %; Sn, in a range of 0.4 to 1.2 wt. %; Fe, in a range of 0.1 to 0.5 wt. %; one element selected from the group consisting of Mo, Cu and Mn, in a range of 0.1 to 0.3 wt. %; O, in a range of 600 to 1400 ppm; Si, in a range of 80 to 120 ppm; and the balance being of Zr.
TL;DR: In this paper, thermal migration of hydrogen under temperature gradient has been estimated according to traditional diffusion theory for δ-ZrHx+U fuel, and it has been shown that hydrogen profiles at steady are stable.
Abstract: To evaluate irradiation behavior of δ-ZrHx+U fuel, thermal migration of hydrogen under temperature gradient has been estimated according to traditional diffusion theory Hydrogen profiles at steady
01 Jan 2000
TL;DR: In this paper, the fracture of hydrides in zirconium alloys is considered and the fracture toughness decreases with T-stress, which is attributed to the interaction of the K-T field with hydride expansion, during precipitation.