Abstract: The present invention relates to a method for manufacturing zirconium-based alloys containing niobium with superior corrosion resistance for use in nuclear fuel rod claddings. The method of this invention comprises melting of the alloy, ?-forging, ?-quenching, hot-working, vacuum annealing, cold-working, intermediate annealing and final annealing, whereby the niobium concentration in the ?-Zr matrix decreases from the supersaturation state to the equilibrium state to improve the corrosion resistance of the alloy. Such zirconium-based alloys containing niobium are usefully applied to nuclear fuel rod cladding of the cores in light water reactors and heavy water reactors.

Abstract: The invention proposes a zirconium-based alloy also containing, by weight, apart from unavoidable impurities, from 0.02 to 1% of iron having from 0.8% to 2.3% of 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 from 0.01% to 0.25% in total of chromium and/or vanadium, the ratio R of the niobium content, less 0.5%, to the iron content, optionally supplemented by the chromium and/or vanadium content.

Abstract: The present invention relates to a zirconium alloy having excellent corrosion resistance and mechanical properties and a method for preparing a nuclear fuel cladding tube by zirconium alloy. More particulary, the present invention is directed to a zirconium alloy comprising Zr-aNb-bSn-cFe-dCr-eCu (a=0.05-0.4 wt %, b=0.3-0.7 wt %, c=0.1-0.4 wt %, d=0-0.2 wt % and e=0.01-0.2 wt %, provided that Nb+Sn=0.35-1.0 wt %), and to a method for preparing a zirconium alloy nuclear fuel cladding tube, comprising melting a metal mixture comprising of the zirconium and alloying elements to obtain ingot, forging the ingot at &bgr; phase range, &bgr;-quenching the forged ingot at 1015-1075° C., hot-working the quenched ingot at 600-650° C., cold-working the hot-worked ingot in three to five passes, with intermediate vacuum annealing and final vacuum annealing the worked ingot at 460-540° C.

Abstract: The invention concerns a method for making tubes designed for a nuclear fuel pencil case or guide tube which consists in forming a bar in a zirconium based alloy containing equally 0.3 to 0.25 wt. % of the total iron, of chromium or vanadium, 0.8 to 1.3 wt. % of niobium, less than 2000 ppm of tin, 500 to 2000 ppm of oxygen, less than 100 ppm of carbon, 5 to 30 ppm of sulphur and less than 50 ppm of silicon. The bar is soaked in water after being heated at a temperature between 1000° C. and 1200° C. A blank is spun after being heated at a temperature between 600° C. and 800° C. The blank is cold-rolled, in at least three of four passes, to obtain a tube with intermediate heat treatments between 560° C. and 620° C. and s final heat treatment is carried out between 560° C. and 620° C. in inert atmosphere or under vacuum.

Abstract: This self-compensating spiral spring for a mechanical balance-spiral spring oscillator for a watch or clock movement or other precision instrument, made of an Nb—Hf paramagnetic alloy possessing a thermal coefficient of Young's modulus (TCE), such that it enables the following expression to be substantially equal to zero: 1 E ⁢ ⅆ E ⅆ T + 3 ⁢ α s - 2 ⁢ α b where: E: Young's modulus of the spiral spring of the oscillator; 1 E ⁢ ⅆ E ⅆ T = T ⁢   ⁢ C ⁢   ⁢ E

Abstract: A fuel channel box manufacturing method processes a fuel channel box of a zirconium-base alloy by a beta-quench treatment that heats the fuel channel box by a heating coil. The distance between the heating coil and the opposite walls of the fuel channel box is controlled so that temperature differences between the opposite walls of the fuel channel box are reduced. The fuel channel box of a zirconium-base alloy is heated at a temperature in a temperature range including &bgr;-phase temperatures so that temperature difference between the opposite walls is 50° C. or below. The fuel channel box manufacturing method is capable of manufacturing a fuel channel box that is not subject to significant irradiated bow even if a deflection is produced therein originally when the same is manufactured and of manufacturing the fuel channel box at a high production efficiency.

Abstract: An alloy is described that is capable of forming a metallic glass at moderate cooling rates and exhibits large plastic flow at ambient temperature. Preferably, the alloy has a composition of (Zr, Hf)a TabTicCudNieAlf, where the composition ranges (in atomic percent) are 45≦a≦70, 3≦b≦7.5, 0≦c≦4, 3≦b+c≦10, 10≦d≦30, 0≦e≦20, 10≦d+e≦35, and 5≦f≦15. The alloy may be cast into a bulk solid with disordered atomic-scale structure, i.e., a metallic glass, by a variety of techniques including copper mold die casting and planar flow casting. The as-cast amorphous solid has good ductility while retaining all of the characteristic features of known metallic glasses, including a distinct glass transition, a supercooled liquid region, and an absence of long-range atomic order. The alloy may be used to form a composite structure including quasi-crystals embedded in an amorphous matrix.

Abstract: Disclosed herein are zirconium-base alloys excellent in both corrosion resistance and hydrogen absorption property, useful as materials for nuclear reactors. Such a zirconium-base alloy for nuclear reactors comprises 0.5-2 wt. % Sn, 0.07-0.6 wt. % Fe, 0.03-0.2 wt. % Ni, 0.05-0.2 wt. % Cr, and the balance being zirconium and unavoidable impurities, wherein the Fe content (X wt. %) of the zirconium-base alloy and the mean size (Y nm) of precipitates in the zirconium-base alloy are present in a region on the x (Fe content X) and y (mean precipitate size) rectangular coordinates, surrounded by the following five lines: i) Y=−444×X+154, ii) Y=910×X−46, iii) Y=0, iv) Y=300, and v) X=0.6.

Abstract: A method for producing a component intended to be subjected to an increased radiation in a corrosive environment, wherein the component includes an alloy which at a first, high temperature has a BCC-structure and which at a second, lower temperature has an HCP-structure, wherein the alloy includes at least one alloying element which has a low solubility in the HCP-structure and wherein the alloy is rapidly cooled from the first to the second temperature while secondary phase particles, which include said alloying element and which contribute to improve corrosion properties are separated in the HCP-structure. The rapid cooling is performed in two stages, wherein the first stage includes a rapid cooling at a relatively lower intensity and the second, subsequent stage includes a rapid cooling at a relatively higher intensity.

Abstract: The present invention relates to a method for manufacturing zirconium-based alloys containing niobium with superior corrosion resistance for use in nuclear fuel rod claddings. The method of this invention comprises melting of the alloy, &bgr;-forging, &bgr;-quenching, hot-working, vacuum annealing, cold-working, intermediate annealing and final annealing, whereby the niobium concentration in the &agr;-Zr matrix decreases from the supersaturation state to the equilibrium state to improve the corrosion resistance of the alloy. Such zirconium-based alloys containing niobium are usefully applied to nuclear fuel rod cladding of the cores in light water reactors and heavy water reactors.

Abstract: The present invention relates to a method for making flat, thin elements which consist of: producing a zirconium alloy blank also containing, besides the inevitable impurities, 0.8 to 1.3% of niobium, 1100 to 1800 ppm of oxygen, and 10 to 35 ppm of sulfur; carrying out a &bgr; hardening and hot rolling to obtain a blank and performing on it at least three cold rolling passes with intermediate annealing heat treatments. One of the intermediate heat treatments is performed for a duration of at least 5 hours at a temperature less than 560° C. and all the optional treatments subsequent to the long treatment are carried out at a temperature of less than 620° C. for not more than 15 minutes.

Abstract: A ceramic hard layer is formed on the surface of a formed article by subjecting the formed article of a substantially amorphous alloy to a heat treatment in an atmosphere containing a reactive gas under the conditions of temperature and time falling within the amorphous region in the isothermal transformation curve (TTT curve) of the alloy. The heat treatment is carried out, for instance, in an atmosphere containing oxygen and/or nitrogen at a concentration of not less than 1 ppm or in the air at a temperature of not less than the lowest temperature required for the oxidation or nitriding reaction of at least one component element of the matrix material. By this heat treatment, it is possible to produce a ceramic hard layer on the surface of the article in such a manner that the content of oxide and/or nitride gradually decreases in the depth direction.

Abstract: Disclosed is a method for manufacturing a tube and a sheet of niobium-containing zirconium alloys for the high burn-up nuclear fuel. The method comprises melting Nb-added zirconium alloy to ingot; forging the ingot at &bgr; phase range; &bgr;-quenching the forged ingot after solution heat-treatment at 1015-1075° C.; hot-working the quenched ingot at 600-650° C.; cold-working the hot-worked ingot in three to five passes, with intermediate vacuum annealing; and final vacuum annealing the cold-worked ingot at 440-600° C., wherein temperatures of intermediate vacuum annealing and final vacuum annealing after &bgr;-quenching are changed so as to attain the condition under which precipitates in the alloy matrix are limited to an average diameter of 80 nm or smaller and the accumulated annealing parameter (&Sgr; A) is limited to 1.0×10−18 hr or lower.

Abstract: Cladding tube for a fuel rod for a boiling water reactor fuel element, and its production. The cladding tube is composed practically homogeneously of the constituents of zircaloy and, with the ductility parameter &ggr;=3{square root over ((kd))}/(fr)2≦3.5 (&ggr;=ductility parameter, KD=mean grain diameter; fr=Kearns factor), has an elongation at break of at least 20%, set by low-temperature treatment of an extruded tube blank. The starting body used for the extrusion has a defined distribution of precipitated secondary particles which is produced by &bgr;-quenching and differs in the areas which form the inner surface and outer surface of the extruded tube. At the inner surface, the particles have a greater diameter and are at a greater average distance apart, this distribution being described by the “spacing”.

Abstract: A titanium alpha-beta alloy having enhanced notch toughness comprises titanium, aluminum, and vanadium and is characterized by a microstructure having equiaxed alpha grains whose volume fraction is about 75 to 85 percent, a maximum grain size of the microstructure not exceeding about 10 &mgr;m, and with the volume fraction of primary alpha grains not exceeding about 2 percent.

Abstract: The present invention relates to a method for producing tubing products from binary and multicomponent zircon alloys, consisting in obtaining from an ingot previously submitted to a beta distortion process p source preparation, hot forming said source preparation at the subsistence temperature of zirconium alpha+beta to get a tubular blank hardenning the latter at a temperature higher by 30 to 60° to the temperature at which the alloy passes from an intermediary state zirconium alpha+beta to the beta state of zirconium, mechanically processing, following by a tempering process at the subsistence temperature of alpha zirconium until the finished product is obtained, the cold distortion of the tubular product being effected with the following global stretching: &mgr;&Sgr;>100 for the finished products and &mgr;&Sgr;<50 for the TREX or SUPER TREX-type semi-finished tubular products, whereby at the first stage of the cold distortion the stretching &mgr;<2.

Abstract: An alloy of zirconium and niobium that includes erbium as a consumable neutron poison, its method of preparation and a component comprising said alloy are provided. This invention relates to an alloy of zirconium and niobium that includes erbium as a consumable neutron poison. The invention also relates to a method for the preparation and conversion of said alloy and a component comprising said alloy. Such an alloy is particularly intended for the manufacture of cladding and/or other elements or structural components of fuel assemblies for nuclear reactors using water as coolant.

Abstract: A high-strength amorphous alloy represented by the general formula: XaMbAlcTd (wherein X is at least one element selected between Zr and Hf; M is at least one element selected from the group consisting of Ni, Cu, Fe, Co and Mn; T is at least one element having a positive enthalpy of mixing with at least one of the above-mentioned X, M and Al; and a, b, c and d are atomic percentages, provided that 25≦a≦85, 5≦b ≦70, 0<c≦35 and 0<d≦15) and having a structure comprising at least having an amorphous phase. The amorphous alloy is produced by preparing an amorphous alloy having the above-mentioned composition and containing at least an amorphous phase, and heat-treating the alloy in the temperature range from the first exothermic reaction-starting temperature (Tx1: crystallization temperature) thereof to the second exothermic reaction-starting temperature (Tx2) thereof to decompose the amorphous phase into a mixed phase structure consisting of an amorphous phase and a microcrystalline phase.

Abstract: A method for increasing the resistance of zirconium alloy tubing to nodular corrosion by applying a protective anneal at a temperature within a clearly defined temperature range. Also, a zirconium alloy tubing having such protective anneal is disclosed. The protective anneal comprises heating exposed surfaces of zirconium tubing to a temperature range bounded at its lower limit by the temperature T.sub.c, T.sub.c being the temperature which at equilibrium conditions a critical concentration of solute exists in .alpha.-matrices of the zirconium alloy to resist nodular corrosion, and bounded at its upper limit by the maximum temperature at which precipitates exist in association with the .alpha. and .beta. matrices in the particular zirconium alloy. In respect of Zircalloy-2 containing zirconium and the following metals by weight, namely 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium, and 0.05-0.08% nickel, the lower temperature limit T.sub.

Abstract: The invention provides a method for making Zr alloy nuclear reactor fuel cladding having excellent corrosion resistance and creep properties. The method includes performing hot forging, solution heat treatment, hot extruding, and repeated cycles of annealing and cold rolling of a Zr alloy including, by weight, 0.2 to 1.7% Sn, 0.18 to 0.6% Fe, 0.07 to 0.4% Cr and 0.05 to 1.0% Nb, with the remainder being Zr and incidental impurities, and the incidental nitrogen impurity content being 60 ppm or less, and then performing final stress relief annealing thereon. The annealing is performed at a temperature of 550.degree. C. to 850.degree. C. for 1 to 4 hours such that the accumulated annealing parameter .SIGMA.Ai=.SIGMA.ti.multidot.exp(-40,000/Ti) satisfies relationships -20.ltoreq.log.SIGMA.Ai.ltoreq.-15, and -18-10.multidot.X.sub.Nb .ltoreq.log.SIGMA.Ai.ltoreq.-15-3.75.multidot.(X.sub.Nb -0.

Abstract: A zirconium alloy tube for forming the whole or the outer portion of a nuclear fuel pencil housing or a nuclear fuel assembly guide tube. The zirconium alloy contains 0.8-1.8 wt. % of niobium, 0.2-0.6 wt. % of tin and 0.02-0.4 wt. % of iron, and has a carbon content of 30-180 ppm, a silicon content of 10-120 ppm and an oxygen content of 600-1800 ppm. The tube may be used when recrystallized or stress relieved.

Abstract: A cast article is heat treated by vacuum annealing at an annealing temperature above the alpha/beta transition temperature of the cast article. Further, the cast article is subjected to a hot isostatic pressing step and a post-weld stress relief processing step. The vacuum annealing step is executed prior to the hot isostatic pressing step and the hot isostatic pressing step is executed prior to the post-weld stress relief processing step. The resulting impact resistance of the heat treated cast article, as measured by the Charpy test, represents an increase of approximately 50-200%, as compared to the untreated cast article. Further, even where the article is subjected to a post-HIP anneal the impact resistance increases by approximately 10-50%.

Abstract: A tube of zirconium-based alloy for constituting all or a portion of a cladding or guide tube for a nuclear fuel assembly. The tube is made of an alloy containing, by weight, 1.0-1.7% of tin, 0.55-0.80% of iron, 0.20-0.60% total of chromium and/or vanadium, and 0.10-0.18% of oxygen, with 50-200 ppm of carbon and 50-120 ppm of silicon. The alloy further contains only zirconium and unavoidable impurities, and it is completely recrystallized.

Abstract: A self-compensating spring for a balance-spring for a balance-spring/balance assembly of a mechanical oscillator of a horological movement or of any other precision instrument, made of a paramagnetic Nb--Zr alloy containing between 5% and 25% by weight of Zr, obtained by cold rolling or cold drawing, and having a Young's modulus whose temperature coefficient (TCY) is adjustable by precipitation of Zr-rich phases in the Nb--Zr solid-solution. It contains at least 500 ppm by weight of an interstitial doping agent at least partly formed of oxygen.

Abstract: The present invention relates to a method of manufacturing tubes of zirconium based alloys for usage in nuclear reactors. According to the invention one can achieve good resistance towards general corrosion, improved transversal creep strength and reduced irradiation induced growth by subjecting a .beta.-quenched tube a vacuum anneal in the .alpha.-phase range at a temperature and a time sufficient to obtain an annealing parameter value A in the range 3.4 .times.10.sup.-16 to 3.4 .times.10.sup.-13.

Abstract: The present invention provides a zirconium based alloy member which has very small deformation of elongation and bow occurring due to irradiation growth, a method of manufacturing it, and particularly an channel box for an atomic reactor fuel assembly. A zirconium based alloy plate member having a width of not less than 100 mm and a long length, containing not more than 5 wt % Nb and/or not more 5 wt % Sn, the member having (0001) orientation (Fl value) of hexagonal Zr with respect to longitudinal direction ranging from 0.20 to 0.35, the difference in Fl between the middle and the end being not more than the value calculated from (0.0935.times.Fl-0.00585) and an amount of bow at neutron irradiation of 35 GWd/t which bow occurs in the channel box for a reactor being not more than 2.16 mm.

Abstract: Zirconium alloys for use in an aqueous environment subject to high fluence of a water reactor and characterized by improved corrosion resistance, consisting essentially of 0.3 to 1.8 weight percent tin, 0.1 to 0.65 weight percent iron, the balance of alloys being essentially nuclear grade zirconium with incidental impurities and having a microstructure of Zr.sub.3 Fe second phase precipitates distributed uniformly intragranularly and intergranularly to form radiation resistant second phase precipitates in the alloy matrix.

Abstract: Zirconium alloys for use in an aqueous environment subject to high fluence of a water reactor and characterized by improved corrosion resistance, consisting essentially of from 0.5 to 3.25 weight percent niobium, from 0.3 to 1.8 weight percent tin, the balance of alloys being essentially nuclear grade zirconium with incidental impurities and having a microstructure of beta niobium second phase precipitates distributed uniformly intragranularly and intergranularly to form radiation resistant second phase precipitates in the alloy matrix.

Abstract: A process for fabricating nuclear fuel rod cladding tube comprising beta quenching a zirconium alloy billet consisting essentially of from 0.5 to 3.25 weight percent niobium, from 0.3 to 1.8 weight percent tin, the balance of the alloy being essentially nuclear grade zirconium with incidental impurities by heating to a temperature in the beta range above 950.degree. C. and rapidly quenching the billet to a temperature below the .alpha. plus .beta. to .alpha. transformation temperature to form a martensitic structure; extruding the beta-quenched billet at a temperature below 600.degree. C. to form a hollow; annealing the hollow by heating at a temperature up to 590.degree. C.; pilgering the annealed hollow; and final annealing the pilgered annealed hollow to a temperature up to 590.degree. C.

Abstract: A process for fabricating nuclear fuel rod cladding tube comprising beta quenching a zirconium alloy billet consisting essentially of 0.3 to 1.8 weight percent tin, 0.1 to 0.65 weight percent iron, the balance of the alloy being essentially nuclear grade zirconium with incidental impurities by heating to a temperature in the beta range greater than about 1000.degree. C. and rapidly quenching the billet to a temperature below the .alpha. plus .beta. to a transformation temperature to form a martensitic structure; extruding the beta-quenched billet at a temperature between 600.degree. and 750.degree. C. to form a hollow; annealing the hollow by heating at a temperature up to about 700.degree. C.; pilgering the annealed hollow; and final annealing the pilgered annealed hollow to a temperature up to about 700.degree. C. to form the nuclear fuel rod cladding tube comprising the alloy having a microstructure of Zr.sub.

Abstract: An alloy material 4 received in a melting hearth 1 is melted by high-density energy supplied from a heat source 5. The molten alloy is transferred to a forced-cooled die 3 having a cavity 2 defining the profile of a product, and quenched to an amorphous state. The alloy has the composition represented by the general formula of Zr.sub.100-a-b-c A.sub.a B.sub.b C.sub.c (wherein the mark A represents one or more elements selected from Ti, Hf, Al and Ga, the mark B represents one or more elements selected from Fe, Co, Ni and Cu, the mark C represents one or more elements selected from Pd, Pt, Au and Ag, and the marks a-c represent the atomic ratios of respective elements A-C under the conditions of a=5-20, b=15-45, c.ltoreq.10 and a+b+c=30-70. The differential temperature region .DELTA.T (=T.sub.x -T.sub.g) in the supercooled liquid phase of the Zr alloy represented by the difference between the crystallization point T.sub.x and the glass transition point T.sub.g, is preferably 100 K or more.

Abstract: A method of manufacturing a tube suitable for use as sheath of a nuclear fuel rod zirconium based alloy. The alloy contains 0.18% to 0.25% iron, 0.07% to 0.13% chromium, 0.35% to 1.70% tin, all percentages by weight, 900 ppm to 2300 ppm oxygen, 80 ppm to 200 ppm carbon, and 50 ppm to 120 ppm silicon. A drawn blank is subjected to successive passes of metallurgical treatment and of heat treatment. The Fe/Cr ratio by weight in the alloy lies in the range 1.6 to 3. The heat treatment includes, after at least one rolling pass, a plurality of annealing passes under an inert atmosphere such that the sum .SIGMA.A is greater than 1.5.10.sup.-17, and advantageously in the range 2.10.sup.-17 to 2.10.sup.-16. The first rolling pass is performed on the tube coming from the drawing operation, without an intermediate annealing. .SIGMA.A designates the sum of the products of the times t (in hours) multiplied by the exponential of --Q/RT, where T is the temperature in K.

Abstract: A tube for constituting all or part of a cladding or guide tube for a fuel assembly is made from an alloy containing, by weight, 1.4% to 0.6% of tin, 0.5% to 0.8% of iron, 0.35% to 0.50% of vanadium, and 0.10% to 0.18% of oxygen, 100 ppm to 180 ppm of carbon and 50 ppm to 120 ppm of silicon. The alloy contains nothing else apart from zirconium and unavoidable impurities and it is completely recrystallized.

Abstract: A zirconium-based alloy with a reduced ahoy content is described that has resistance to both uniform and nodular corrosion comparable to present zirconium-based alloy compositions, such as Zircaloy-2. The alloy represents in essence a modified or diluted Zircaloy-2 or Zircaloy-4. The alloys of this invention are also expected to have improved uniform corrosion resistance at under high burn-up conditions The alloy comprises 0.05-0.09 weight percent of iron, 0.03-0.05 weight percent of chromium, 0.02-0.04 weight percent of nickel, 1.2-1.7 weight percent of tin and 0-0.15 weight percent oxygen, with a balance of zirconium. The iron chromium and nickel alloying elements form precipitates in the alloy matrix. The alloy is suitable for use as a cladding material for a fuel element housing fissionable nuclear materials in water cooled nuclear fission reactor.

Abstract: In a zirconium-alloy fuel element cladding, a method for generating regions of coarse and fine intermetallic precipitates across the cladding wall is provided. The method includes steps of specific heat treatments and anneals that coarsen precipitates in the bulk of the cladding. The method also includes at least one step in which an outer region (exterior) of the cladding is heated to the beta or alpha plus beta phase, while an inner region (interior) is maintained at a temperature at which little or no metallurgical change occurs. This method produces a composite cladding in which the outer region comprises fine precipitates and the inner region comprises coarse precipitates.

Abstract: A method for manufacturing a delayed hydride cracking resistant zirconium ahoy (Zircaloy-2, Zircaloy-4, Zr-2.5% Nb, pure Zr, etc.) pressure tube includes the steps of making a seamless pressure tube having a diameter smaller than the final size by extrusion or drawing, and then expanding the tube at a temperature below 600.degree. C. by cross rolling.

Abstract: The invention relates to a fuel element for a pressurized-water reactor comprising a number of fuel rods arranged in a square lattice and wherein certain positions in the lattice contain guide tubes for receiving control rods, wherein the fuel element can be used for higher burnup by manufacturing the guide tubes of the element such that their irradiation-induced growth is reduced and their corrosion resistance is increased compared with conventionally manufactured guide tubes. This is achieved by finally heat-treating the guide tubes in two steps, the first step being a beta quenching and the second step a heat treatment in the alpha-phase region.

Abstract: A process for the fabrication of zirconium alloy sheet specifically intended for the manufacture of structural elements for boiling water reactors, which includes the following steps:a) producing in a vacuum an ingot having the composition of the desired alloy;b) forging and hot rolling the ingot;c) quenching of the blank thus obtained after reheating in the beta range;d) hot rolling after heating;e) heat treatment in the alpha range;f) at least one cycle of cold rolling followed by a heat treating in the alpha range; andg) final cold rolling followed by subcritical annealing in the alpha range;where the hot rolling of the sheet after quenching from the beta range is carried out in an initial direction, then in a direction perpendicular to the initial direction.

Abstract: A method for producing a refractory metal foil from a workpiece fabricated of a refractory metal material includes the step of heat treating the workpiece for a first time period of approximately 1.5 hours at a first temperature of approximately 2,650.degree. Fahrenheit and subsequently for a second time period of approximately 2.0 hours at a second temperature of approximately 2,200.degree. Fahrenheit. The next step is forming the workpiece into a sheet of refractory metal material having a first sheet thickness of approximately one inch. The next step is heat treating the sheet of refractory metal material for a third time period of approximately 2 hours at a third temperature of approximately 2,200.degree. Fahrenheit. The next step is compressing the sheet of refractory metal material into a second sheet thickness of approximately 0.040 inches. The next step is annealing the sheet of refractory metal material for a first annealing time period for approximately 2.

Abstract: The method serves to manufacture tubes for constituting sheaths for nuclear fuel rods. A bar is made out of a zirconium-based alloy containing 50 ppm to 250 ppm iron, 0.8% to 1.3% by weight niobium, less than 1600 ppm oxygen, less than 200 ppm carbon, and less than 120 ppm silicon. The bar is heated to a temperature in the range 1000.degree. C. to 1200.degree. C. and is quenched in water. A blank is extruded after heating to a temperature in the range 600.degree. C. to 800.degree. C. and cold-rolled in at least four passes in order to obtain a tube, with intermediate heat treatment being performed between passes at temperatures in the range 560.degree. C. to 620.degree. C. A final heat treatment is performed at a temperature in the range 560.degree. C. to 620.degree. C., all of the heat treatments being performed under an inert atmosphere or a vacuum.

Abstract: A product having at least an exterior surface formed by a Zr alloy comprising the following addition elements (% by weight): Sn 0.40 to 1.70, Fe 0.05 to 0.25, Cr 0.03 to 0.16, Ni less than 0.08; the following maximum amounts of impurities (ppm): Al 75, B 0.5, Cd 0.5, C 270, Co 20, Cu 50, Hf 100, H.sub.2 25, Mg 20, Mn 50, Mo 50, N.sub.2 65, Si 120, Ti 50, W 10, U total 3.5; with the balance being O.sub.2 and Zr, wherein the Zr alloy has an Ni content which is greater than 70 ppm and less than 300 ppm.

Abstract: A method of manufacturing nuclear fuel elements which include fuel rods whose cladding tubes are provided with an internal liner layer to obtain PCT resistance in the nuclear fuel element involves carefully choosing parameters for heat treatment of the inner component even from the machining of an ingot of the inner component. The internal layer of zirconium or a zirconium alloy, suitable as inner layer in a PCI-resistant cladding, from the fabrication of an ingot of the inner component up to the completion of a cladding tube, including forging, rolling, extrusion, heat treatment and final heat treatment, is manufactured in such a way that the temperature in the inner component never exceeds the temperature when an incipient phase transformation to beta phase takes place.

Abstract: In a zirconium-alloy fuel element cladding, a method for generating regions of coarse and fine intermetallic precipitates across the cladding wall is provided. The method includes steps of specific heat treatments and anneals that coarsen precipitates in the bulk of the cladding. The method also includes at least one step in which an outer region (exterior) of the cladding is heated to the beta or alpha plus beta phase, while an inner region (interior) is maintained at a temperature at which little or no metallurgical change occurs. This method produces a composite cladding in which the outer region comprises fine precipitates and the inner region comprises coarse precipitates.

Abstract: The invention concerns a process for the production of a tubular zircaloy 2 blank which is internally clad with zirconium for use in producing composite cladding tubes for nuclear fuel. The internal zirconium cladding is rendered suitable for ultrasound monitoring of its thickness by an appropriate thermomechanical treatment which takes place during one or more of the production steps for said blank. The aim is to adjust the grain size to an ASTM index of between 9 and 12 for the zircaloy 2 and between 6 and 10 for the unalloyed zirconium, retaining a grain size difference between the zircaloy 2 and the unalloyed zirconium of at least 2 ASTM index numbers.

Abstract: A fuel assembly avoiding the generation of irradiation damage, a Zr alloy used for the same, and a manufacturing method thereof. According to one embodiment, a super-saturated solid-solution Zr alloy powder having a crystal grain size in the range of 1000 nm or less and containing Fe, Ni and Cr is prepared by mechanical alloying, and the alloy powder is subjected to HIP, hot-working, cold-working and final heat-treatment.

Abstract: The material, according to the present invention, based on a Zr contains also the following components, on a weight percent basis: 0.5 to 1.5 Nb, 0.9 to 1.5 Sn, 0.3 to 0.6 Fe, 0.005 to 0.2 Cr, 0.005 to 0.04 C, 0.05 to 0.15 O, 0.005-0.15 Si, Zr being the remainder. The proposed product used in the core of nuclear reactors and made from the aforementioned material, features a microstructure which is essentially a zirconium matrix strengthened with particles of the stannfferous and ferriferous intermetallides; at least 60 volume percent of a total amount of the ferriferous intermetallides falls on the particles of the intermetallides of the type of Zr(Nb,Fe).sub.2, Zr(Fe,Cr,Nb), and (Zr,Nb).sub.3 Fe, with the interparticle distance from 0.20 to 0.40 m.

Abstract: A Zircaloy cladding having an outer region comprising fine precipitates and inner region comprising coarse precipitates is provided. The outer region comprises about 10% and the inner region comprises about 90% of the cladding wall thickness. Such Zircaloy tubing is resistant to propagation of cracks and at the same time resistant to corrosion in boiling water reactors (BWR). Resistance to damage caused by the pellet-cladding-interaction can be achieved by standard application of a zirconium or zirconium-alloy liner on the tubing inside surface.

Abstract: A fuel rod has a cladding including a thicker inner layer and a thin outer layer being metallurgically bound thereto. In view of the conditions prevailing on the inside of the cladding tube and the mechanical properties of the entire cladding tube, the inner layer is formed of zircaloy having a comparatively high Sn content and a low Fe and Fe+Cr content. The outer layer also contains virtually only zircaloy constituents, but in view of corrosion, H2 take-up and sensitivity to Li dissolved in the cooling water, the Fe and Fe+Cr content is greater than or at most equal to that of the inner layer, the chosen Sn content is less than 1.3% and the chosen Sn+Fe+Cr content is more than 1.0%. Low failure rates of the cladding tube are thereby achieved even for long service lives.

Abstract: A process for the manufacture of a sheet, band or strip of zirconium alloy or hafnium alloy, comprising hot roughing of an ingot into a blank, then hot rolling of this blank in the alpha domain, this hot rolling including several passes which follow one or more reheatings in a furnace at one or more temperatures (TR), each of which improves the formability of this blank, then cold rolling which consists of one or more cycles of rolling pinch passing/heat treatment. The hot rolling pass following the last furnace reheating is followed by reheating of the hot rolled blank by at least 100.degree. C. at more than 4.degree. C./s by infrared heating means with a wavelength between 0.8 and 5 micrometers. The hot rolling is then continued until the thickness at the end of the hot rolling is less than or equal to 0.8 times the thickness of the hot rolled blank. The product obtained is used for the manufacture of zirconium or hafnium alloy parts for water-type nuclear reactors.

Abstract: A process for the manufacture of a flat product of zirconium alloy having a width between 50 and 600 mm from a flat cold rolled product. The cold rolled product is heated to the beta range and maintained in the beta range using an infra-red heating device with a wavelength of between 0.8 and 5 microns, the maximum temperature difference over the width of the product being less than 35.degree. C. The product is then rapidly cooled and optionally annealed to obtain a flat product, which can be a strip, sheet or band.