Showing papers on "Zirconium alloy published in 1990"

Experimental studies of U-Pu-Zr fast reactor fuel pins in the experimental breeder reactor-ll

Abstract: Argonne National Laboratory’s Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.

Zirconium oxide and nitride coated prothesis for reduced microfretting

Abstract: Orthopedic implants of zirconium or zirconium-based alloy coated with blue or blue-black zirconium oxide or zirconium nitride to provide low friction, highly wear resistant coatings especially useful in artificial joints, such as hip joints, knee joints, elbows, etc. The invention zirconium oxide or nitride coated prostheses are also useful to reduce microfretting wear in multi-component surgical implants subject to such wear. Further, the coatings provide a barrier against implant corrosion caused by ionization of the metal prosthesis. Such protection can be extended by the use of oxidized or nitrided porous coatings of zirconium or zirconium alloy beads or wire mesh into which bone spicules may grow so that the prosthesis may be integrated into the living skeleton.

Effect of mechanical alloying beyond the completion of glass formation for Ni-Zr alloy powders

Abstract: Elemental powders of nickel and zirconium were mechanically alloyed over a wide concentration range 10 to 90 at % Zr. The amorphous single phase was formed over the range 20 to 80 at % Zr. The effect of the excessive mechanical alloying on the glass formation was studied by continuing ball-milling beyond the completion of the glass formation for the powders with the average compositions Ni30Zr70, Ni50Zr50 and Ni70Zr30. A partial crystallization took place in all three cases and its initiation was the fastest in Ni30Zr70 and was delayed with decreasing zirconium content. The critical factor for triggering the crystallization was attributed to the oxygen contamination for the zirconium-rich Ni30Zr70 powders and to the reduction in glass-forming ability for the nickel-rich Ni70Zr30 powders. The latter conclusion is drawn from the facts that the impurity concentrations arising from the debris of the stainless steel balls and the vial are gradually accumulated with increasing milling time and that the effective zirconium concentration is reduced below the critical concentration of approximately 20 at % as a result of alloying with the elements iron, chromium and nickel in the stainless steel.

Iron aluminide alloys with improved properties for high temperature applications.

Abstract: An improved iron aluminide alloy of the DO3 type that has increased room temperature ductility and improved high elevated temperature strength. The alloy system further is resistant to corrosive attack in the environments of advanced energy corrosion systems such as those using fossil fuels. The resultant alloy is relatively inexpensive as contrasted to nickel based and high nickel steels currently utilized for structural components. The alloy system consists essentially of 26-30 at.% aluminum, 0.5-10 at.% chromium, 0.02-0.3 at.% boron plus carbon, up to 2 at.% molybdenum, up to 1 at.% niobium, up to 0.5 at.% zirconium, up to 0.1 at.% yttrium, up to 0.5 at.% vanadium and the balance iron.

The Influence of Tungsten on the Pitting of Aluminum Films

Abstract: Many of the alloying additions responsible for the corrosion resistance of stainless steels can also be used to enhance the passivity of aluminum, provided that these elements remain in solid solution in the alloy. Passivity enhancing elements such as Cr, Mo, Ta, Zr, and W typically exhibit very low solubility limits in aluminum, below 1 atomic percent, and at these concentrations exert little influence on corrosion behavior. However, these solubility limits can be increased with a concomitant enhancement in corrosion performance if the alloys are produced using a rapid solidification process. One such process is co-sputter deposition of two metals, such as Al and W, simultaneously to form a binary alloy of W in solid solution with Al. Co-sputter deposition is routinely used to produce compound semiconductor films. The authors report on shifts in pitting potentials and on the effect of a transition element (W) on pitting resistance of Al.

Corrosion resistant zirconium alloys containing copper, nickel and iron

Abstract: Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor consist essentially of by weight percent about 0.5 to 2.0 percent thin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. Nuclear fuel elements for use in the core of a nuclear reactor have improved corrosion resistant cladding made from these zirconium alloys or composite claddings have a surface layer of the corrosion resistant zirconium alloys metallurgically bonded to the outside surface of a Zircaloy alloy tube. The claddings may contain an inner barrier layer of moderate purity zirconium metallurigcally bonded on the inside surface of the cladding to procide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Alloy preparation of hydrogen storage material

Abstract: A method for the preparation of a highly alloyed metal hydride, hydrogen storage alloy material including titanium, zirconium, vanadium, nickel and chromium. The hydrogen storage alloy material is prepared by vacuum induction melting electrochemically operative amounts of the materials in a nigh density, high purity graphite crucible, under an inert gas atmosphere.

Corrosion resistant zirconium alloy

Abstract: An improved corrosion resistant ductile modified zirconium alloy for extended burnups in water-moderated nuclear reactor core structural components, fuel cladding and analogous corrosive environment uses is provided. It comprises: measurable amounts of niobium in a range up to 0.6 percent by weight, measurable amounts on antimony in a range up to 0.2 percent by weight, measurable amounts of tellurium in a range up to 0.2 percent by weight, tin in the range 0.5 to 1.0 percent by weight, iron in the range 0.18 to 0.24 percent by weight, chromium in the range 0.07 to 0.13 percent by weight, oxygen in the range of from 900 to 2,000 ppm, nickel in an amount less than 70 ppm, carbon in an amount less than 200 ppm and the balance zirconium and minor amounts of impurities. The alloy structure is substantially alpha phase with some precipitated second phase particles which are preferably within the size range of 1200 to 1800 angstroms. Bismuth may be substituted for part of either or both of the elements antimony or tellurium in a range up to 0.2 percent by weight of bismuth.

Rod for a fuel assembly of a nuclear reactor resisting corrosion and wear

Abstract: The fuel rod comprises a sheath having an inner tubular layer and an outer surface layer composed of zirconium alloys which differ from each other. The surface layer, whose thickness is between 10 to 25% of the total thickness of the wall of the sheath, is constituted by a zirconium-base alloy containing by weight 0.35 to 0.65% tin, 0.20 to 0.65% iron, 0.09 to 0.16% oxygen and niobium in a proportion of 0.35 to 0.65% or vanadium in a proportion of 0.25 to 0.35%. The inner layer may be constituted by an alloy such as Zircaloy 4 or a zirconium-niobium alloy.

2. Effects of dehydrogenation of 2-propanol on the structure and catalytic activity of an amorphous copper-zirconium alloy sample

Abstract: The effects of the dehydrogenation of 2-propanol on the structure of an amorphous Cu61Zr39 alloy sample were analysed by means of scanning electron microscopy (SEM), energy dispersive X-ray microanalysis (EDAX), and Auger electron spectroscopy (AES). The inner and outer sides of the as-received amorphous ribbon differed considerably, indicating copper and zirconium enrichment, respectively. During a 24 h reaction dramatic changes in surface morphology and a substantial copper enrichment took place on both sides, resulting in very similar surfaces.

Residual grain-interaction stresses in zirconium alloys

Abstract: Residual grain-interaction stresses develop during the thermomechanical treatment of Zr alloys due to the anisotropy of the mechanical and thermal properties of the hep lattice. The origin and characteristics of this type of residual stress are described in conjunction with underlying physical principles employed to measure grain-interaction strains by means of neutron diffraction. The effect of thermal treatments, deformation, and irradiation on the evolution of residual grain-interaction strains are reviewed, and the most up-to-date experimental results are presented. The effects of grain-interaction stresses on the plasticity and in-reactor deformation of ZIRCALOY-2 will be discussed.

Autoclave corrosion test for zirconium alloys

Abstract: An improved short-term autoclave (10) test for ex-reactor evaluation of in-reactor corrosion resistance of zirconium alloy members (16) for use in pressurized water reactors and pressurized heavy water reactors by: providing a heat flux to initiate hydride precipitation close to the metal-oxide interface of the tube outside surface by means of a resistance heater (20) and a directed flow of aqueous coolant in the water phase to only the outside surface which is one of two specimen (16) opposed surfaces; providing the inside surface with access to the coolant from the flow but not in the flow; reducing the distance between the opposed surfaces of specimen (16); regulating coolant flow from inlet port (12) to outlet port (14) and power of autoclave (10) to optimum saturation condition of 360° C.± 5° C. and pressure 2700 p.s.i.± 100 p.s.i.; providing a coolant chemistry which simulates a rector's; removing most of the oxygen from the coolant.

Corrosion resistant composite claddings for nuclear fuel rods

Abstract: A nuclear fuel element for use in the core of a nuclear reactor is disclosed having an improved corrosion resistant cladding. The cladding is comprised of zirconium alloys containing in weight percent 0.5 to 2.0 percent tin, or 0.5 to 2.5 percent bismuth, or 0.5 to 2.5 percent bismuth and tin, and about 0.5 to 1.0 percent of a solute composed of a member selected from the group consisting of molybdenum, niobium, tellurium and mixtures thereof, and the balance zirconium. Composite claddings are disclosed having a surface layer of one of the corrosion resistant zirconium alloys metallurgically bonded to a Zircaloy alloy tube. Claddings may contain an inner barrier layer of a moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

High coercivities in Sm-Fe-Tm magnets

Abstract: High coercivities have been obtained in three Sm-Fe-TM phases (TM=V, Ti, Zr) prepared by mechanical alloying and a subsequent heat treatment at relatively low temperatures. In Sm-Fe-V with the ThMn/sub 12/ crystal structure, coercivities up to 11.8 kOe were achieved; in Sm-Fe-Zr with the PuNi/sub 3/ crystal structure, coercivities up to 14.8 kOe were achieved; and in omega -phase Sm-Fe-Ti, giant coercivities up to 64 kOe (at room temperature) were achieved. The temperature dependence of the coercivity also differs favorably from that of Nd-Fe-B. >

Silicon grain refinement of zirconium

Abstract: Substantially pure zirconium for use as a cladding material for nuclear fuel elements containing between about 40 ppm to about 120 ppm silicon and containing less Fe than its solubility limit in the zirconium.

Zironium alloy for use in pressurized nuclear reactor fuel components

Abstract: A zirconium alloy that has sufficient corrosion resistance, strength and stress relaxation property for use as a component of a pressurized water nuclear reactor fuel assembly is disclosed This alloy consists essentially of 02 to 091% Sn, 018 to 06% Fe, 007 to 04% Cr, one or both of 005 to less than 05% and 001 to 02% Ta, one or both of 005 to 1% V and 005 to 1% Mo, with the balance being Zr and incidental impurities, all percentages being based on weight

Zirconium based alloy material for light water reactor applications

Abstract: This is an alloy comprising, by weight percent, 0.5-2.0 niobium, 0.7-1.5 tin, 0.07-0.14 iron, and 0.03-­0.14 of at least one of nickel and chromium, and at least 0.12 total of iron, nickel and chromium, and up to 220 ppm C, and the balance essentially zirconium. Preferably, the alloy contains 0.03-0.08 chromium, and 0.03-0.08 nickel. The alloy is also preferably subjected intermediate recrystallization anneals at a temperature of about 1200-­1300°F, and to a beta quench two steps prior to final size.

Tensile properties and plastic deformation modes of? Zr-Nb alloys

Abstract: Tensile properties and plastic deformation modes ofβ zirconium-niobium alloys were investigated at 290 and 77 K in the wide composition range from metastable to stableβ phase. Three types of plastic deformation modes, {332}〈113〉 twinning, {112}〈111〉 twinning and slip, were observed depending on alloy composition and temperature. {332}〈113〉 twinning, which occurs in metastableβ zirconium alloys, is related to the stability ofβ phase toω decomposition and leads to low yield stress and large elongation. On the other hand, {112}〈111〉 twinning, which appears in stableβ zirconium alloys, results from high critical stress for slip due to solution hardening and high Peierls stress and does not affect tensile properties significantly. The results obtained for zirconium-alloys are similar to those forβ titanium alloys, strongly suggesting that {332}〈113〉 twinning is an important plastic deformation mode which is common toβ phase alloys containing athermalω phase.

SCDAP/RELAP5/MOD2 code manual

Abstract: This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.

A replica technique for extracting precipitates from zirconium alloys for transmission electron microscopy analysis.

Abstract: A reliable two-stage carbon replica technique has been developed to extract precipitates from zirconium alloys. Using this technique, all precipitating phases can be extracted from Zircaloy-2, Zr-Cr-Fe, and Zr-Nb-Fe alloys. Precipitate identification using EDS X-ray analysis and convergent beam electron diffraction was greatly facilitated in comparison to thin foils. In addition, the sensitivity for the detection of trace elements in particles was increased using extraction replicas. The chemical compositions of the precipitates as determined from both replica and thin foils were in excellent agreement.

A method of manufacturing cladding tubes for fuel rods for nuclear reactors

Abstract: The invention relates to a method of making cladding tubes of zirconium alloys. According to the invention an improved resistance towards nodular corrosion under operating conditions in boiling water nuclear reactors can be achieved by carrying out a beta-quenching step, prior to the final cold rolling step, by heating 5-50% of the outer portion of the tube wall to a temperature in the beta-phase region and therafter rapidly cooling the tube.