Abstract: A part of a light-water reactor, for example, a cladding for a light-water reactor has at least a portion made of an intermetallic compound, such as Ni.sub.3 Al, Ni.sub.2 Al.sub.3, TiAl, Ti.sub.3 Al, Pt.sub.2 Si, PtSi, FeAl.sub.2, CoAl, and MoSi.sub.2, thereby having limited irradiation degradation by fast neutrons, improved ductility and usability at high temperatures the intermetallic compound is produced by a process comprising a step of irradiating the intermetallic compound with at least one selected from the group consisting of a neutron, a light ion, and an electron. The intermetallic compound may contain a twin and a third additional element segregatedly present or forms a second phase at or near to a grain boundary.

Abstract: An alloy having improved resistance to hydriding as well as good corrosion resistance, adequate strength, fabricability and irradiation growth. These properties make the new alloy exceptionally suited for use in boiling water reactor components such as fuel channels, fuel cladding and fuel rod spacers. The class of alloys includes zirconium-based alloys having about 0.3-1.2% tin, about 0.8-1.4% chromium, about 0.05-0.7% iron, and the remainder substantially zirconium.

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: The invention relates to a fuel assembly (1) and a spacer (7) for a nuclear reactor wherein the spacer comprises a plurality of cells (9) for retaining and mutually fixing parallel elongated elements (3, 8) extending through the cells to form a bundle in the fuel assembly. The cells are tubular and internally provided with four elongated, non-independently resilient supports (10), formed in the sleeve material, or with two elongated, non-independently resilient supports (10), formed in the sleeve material, and two independently resilient supports (16), wherein the supports are intended to provide an all-sided positioning of the elongated element. (FIG.

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: The girth welds joining the end plugs and the hollow tubes of nuclear fuel rods are inspected automatically using a technique that averages reflectance values, compares the reflectance values to standards defined as proportions of the average, and counts adjacent pixels outside the standards to analyze for defects exceeding a minimum defect size. The minimum defect size is checked by counting the adjacent pixels in mutually perpendicular directions, such as rows and columns in the collected matrix of pixel data. The maximum count in the two directions can be different, for profiling the maximum acceptable defect as to the direction of its extension relative to the tube and/or weld. Tubes are inspected prior to surface treatments that may conceal defects.

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 tubular cladding (1) comprises an internal surface of substantially prismatic shape, having successive faces. An ultrasonic examination head (3) makes it possible to emit ultrasound waves in the form of pulses from the outside of the cladding (1), so that the ultrasound waves pass through the wall of the cladding (1) and scan the wall over its entire periphery. The frequency of the ultrasound pulses is adjusted as a function of the circumferential scanning speed of the cladding (1). An ultrasound signal reflected by the internal surface of the cladding (1) is recorded and the reflected signal is analyzed in order to examine the number and the amplitude of the faces and the variation in thickness of the tubular cladding (1) over one revolution.

Abstract: Zircaloy 2 and zircaloy 4 are zirconium alloys which are permitted and tried and tested in nuclear engineering and which have constituents with fixed concentration ranges. The properties, especially corrosion resistance, mechanical stability and sensitivity to pellet-cladding interaction of those alloys are subject to pronounced spreads of unknown origin. According to the invention, the tin content is between 1.4 and 1.8% by weight, the Fe content between 0.1 and 0.25% by weight, the Cr content between 0.1 and 0.3% by weight, the Si content between 0.05 and 0.02% by weight, the O content between 0.05 and 0.11% by weight, the C content below 0.02% by weight and the Ni content below 0.08% by weight. This restriction of the permissible concentration ranges ensures that the material properties are spread only within a narrow favorable range. A liner made from zirconium with an iron constituent of between 0.2 and 0.8% by weight is proposed for the inner lining of a fuel-rod sheathing tube.

Abstract: An improved method for assembling bundles of nuclear fuel rods which reduces damage to the fuel cladding surface as the fuel rods are passed through an array of spacers. In particular, the invention prevents or greatly reduces the severity of longitudinal scratches on the fuel rods. The method uses thin-walled metallic tubes having a wall thickness of 0.003 to 0.006 inch and a yield strength in excess of 140,000 psi. The predetermined tube thickness ensures that the tube remains elastic. The tube deflects the cell springs when inserted in the spacer cells and reduces the load exerted on the fuel rods.

Abstract: A debris resistant fuel rod sleeve that is received over the lower end of a fuel rod. The sleeve extends above the top of the lowermost spacer grid. Openings are spaced apart around the circumference of the sleeve to correspond to the location of hard stops in the spacer grid. The hard stops are received in the openings and retain the sleeves in position during operation and during reconstitution or recaging if necessary. The outboard side of the peripheral sleeves may be provided with top and bottom lead-in features to prevent hang-ups.

Abstract: A cladding tube having a cross-section and including (1) a zirconium alloy outer circumferential substrate having an inner surface and having one or more alloying elements, (2) a zirconium barrier layer bonded to the inner surface of the outer circumferential substrate and being alloyed with the one or more alloying elements, and (3) a zirconium alloy inner circumferential liner bonded to the inner surface of the zirconium barrier layer. The barrier layer will have a concentration profile including a diffusion layer extending from the barrier layer's inner surface (facing nuclear fuel) to the barrier layer's interior (between the barrier layer's inner and outer surfaces). At the interior edge of the diffusion layer, there will be substantially no alloying elements. At the outer edge of the diffusion layer (the barrier layer's inner surface), the maximum concentration of alloying elements will occur.

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 method is provided for preparing a cladding tube having an outer substrate, an intermediate zirconium barrier layer, and an inner liner. The method includes the following steps: (a) bonding an inner liner alloy sheath exterior circumferential surface to a zirconium sheath interior circumferential surface to form a barrier/inner liner sheath, and (b) bonding the exterior surface of the zirconium sheath on the barrier/inner liner sheath to the interior circumferential surface of an outer substrate alloy tube to form the cladding tube. Alternatively, the method includes the following steps: (a) bonding the zirconium sheath exterior circumferential surface to the outer substrate alloy tube interior circumferential surface to form a substrate tube/barrier sheath, and (b) bonding the exterior circumferential surface of the inner liner alloy sheath to the interior circumferential surface of the zirconium sheath of the substrate tube/barrier sheath to form said cladding tube.

Abstract: The present invention provides a cladding having an outer circumferential substrate, a zirconium barrier layer metallurgically bonded to the inside surface of the substrate and an inner circumferential liner metallurgically bonded to the zirconium barrier. The inner circumferential liner is more ductile than conventional Zircaloy. The low ductility of the inner circumferential liner is obtained, for example, by using a zirconium alloy containing a low tin content (e.g. less than 1.2% by weight) and/or a low oxygen content (e.g. less than 1000 ppm). The inner circumferential liner is less than about 25 micrometers thick.

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: Zirconium or zirconium alloy components of a cylindrical shape are cleaned with an ice blasting process to produce a defect-free bondline in multi-layered tubing suitable for nuclear fuel cladding and the chemical processing industry. The ice blasting process improves the integrity of the metallurgical bond by removing foreign contamination that can initiate non-bonding defects.

Abstract: A cladding tube for use in holding fissionable material in a water cooled nuclear reactor is provided. The cladding tube includes inner and outer circumferential regions, and including (1) a substrate defining the outer circumferential region, and (2) an inner liner defining the inner circumferential region. The inner liner is made from a zirconium alloy having at least one alloying element which promotes hydrogen absorption in a concentration of between about 1 and 15 weight percent. The hydrogen absorption promoting element can be one of the following: nickel, chromium, iron, zinc, vanadium, gallium, yttrium, palladium, platinum, or aluminum.

Abstract: The present invention relates to a process of decontaminating zirconium-based alloy claddings tubes used in nuclear reactor fuel rods. The process involved the use of a permanganate in a dilute acid solution. The process renders the alloy suitable for uncontrolled release into a non-radioactive environment.

Abstract: A method is provided for forming a three-layer cladding tube having an outer substrate, a zirconium barrier layer, and an inner liner having alloying elements, in which the zirconium barrier layer (located between an outer substrate and inner liner) is at least partially alloyed with alloying elements that impart resistance to corrosion. The barrier layer has a diffusion layer extending from its inner surface (facing the fuel) to the barrier layer's interior (the interior being defined between the barrier layer's inner and outer surfaces). At the interior edge of the diffusion layer, there will be substantially no alloying elements beyond those normally present in zirconium. The methods of forming such structure include a diffusion anneal of a three-layer cladding in the range of 650.degree.-1000.degree. C. for times between about 1 minute and 20 hours. This anneal drives some of the alloying elements from the inner liner into the zirconium barrier layer to form the diffusion layer.

Abstract: A cladding is provided for use in housing fissionable material in water cooled nuclear fission reactors. The cladding has inner and outer surfaces and includes (1) a cross-section of a Zirconium-based alloyed matrix, and (2) alloying elements in sufficient concentration to form precipitates disposed in the matrix. The cladding includes no more than 20 parts per million nitrogen by weight and is typically a modified Zircaloy-2 or Zircaloy-4. Metallurgically bonded to the inner region of the cladding is a zirconium barrier layer.

Abstract: A cladding tube is provided having an axis and a cross-section perpendicular to the axis. The cross-section includes (1) an outer circumferential substrate having an inner surface, (2) a zirconium barrier layer bonded to the inner surface of the outer circumferential substrate, and (3) an inner circumferential liner bonded to the inner surface of the zirconium barrier layer. The inner circumferential liner includes a plurality of facets aligned substantially in parallel with the cladding tube axis. The facets--which define the geometry of the cladding interior--facilitate mixing among the gases in the cladding interior. Also provided is a fuel element including (1) a cladding tube having an faceted inner liner as described above, (2) nuclear fuel material disposed within said cladding tube in a fuel region, and (3) a plenum or plenums located in one or both end regions (beyond the nuclear fuel region) within the cladding tube.

Abstract: A nuclear fuel rod cladding for a water moderated and nuclear reactor comprises an inner portion of Zircaloy 4 and an outer portion of a zirconium-based alloy which contains by weight, besides zirconium and unavoidable impurities:0.35% to 0.65% tin0.18% to 0.25% iron0.07% to 0.13% chromium, and0.19% to 0.23% oxygen,with the sum of the iron, chromium, tin, and oxygen contents being less than 1.26% by weight. It may alternatively or additionally comprise 0.80% to 1.20% by weight niobium and then the oxygen content is in the range 0.10% to 0.16% by weight. The thickness of the outer layer is 10% to 25% of the total thickness of the cladding. In a modification, up to 0.5% of iron, chromium, niobium is replaced by an equivalent content of vanadium.

Abstract: A multi-layer, rolled metal foil is laser tackwelded to form a tube which can be inserted in the intermediate space between the fuel body and cladding of a metal alloy fissionable fuel element. The rolled foil has at least three foil layers at the point of tack welding. The laser welding penetration is adjusted so that the foil weld is maintained at a thickness of at least two foil layers, but does not penetrate all of the layers. The weld is designed to fail in response to fuel or blanket alloy swelling during irradiation. After weld failure, the overlapping layers slip and the multi-layer foil unrolls as the fuel swells, providing a continuous, unbroken barrier between the fuel or blanket alloy and cladding which masks defects in the barrier due to weld failure.

Abstract: A fuel assembly, where crystallographic orientations of a channel box are brought into a random distribution; and cladding tubes, spacers and a channel box are made from highly corrosion-resistant, Fe--Ni, zirconium-based alloy, hardened in the (.alpha.+.beta.) phase or .beta.-phase temperature region, has an average discharge burnup level of 50 to 550 GWd/t.

Abstract: A low pressure drop easy load fuel assembly rod end cap 40 for pressurized water reactor fuel assembly rods 10 has a solid Zircaloy radiused tip 42. The tip 42 blends into a longitudinal profile defined by a smooth convex curve 44,46 which blends into a substantially cylindrical surface portion 48 of a diameter substantially equal to the diameter of the fuel cladding tube or other shank 50.

Abstract: A fuel rod for a nuclear reactor is described. The fuel rod contains a zirconium alloy cladding tube, with a thin coating of a burnable poison which has been plasmaarc sprayed on the inside diameter of the cladding tube. Also described is a plasma-arc spraying device and a method for applying a coating to the inside of a small-diameter metal tube.

Abstract: An Fe-Cr-Mn alloy is disclosed which has the following composition by wt% and corrosion resistance of which is improved and deterioration in its strength is prevented at grain boundaries due to irradiation of high-energy particles such as neutrons: 5 to 40% of Mn, 5 to 18% of Cr, 2.0 to 12% of Al and the balance of Fe except for unavoidable impurities. In the alloy according to the present invention, Al is added to an Fe-Cr-Mn alloy by a restricted quantity as a main component element. As a result of the addition of Al, an alloy can be obtained in which lowering of concentration of Cr at grain boundaries due to irradiation of high-energy particles such as neutrons can be prevented or concentration of the solutes can be raised.

Abstract: A method for enhancing the wear and corrosion resistance of a tubular nuclear fuel assembly component (40), comprising the step of coating the component with a corrosion and wear resistant material by an anodic arc plasma deposition process (70). The coating is preferably a nitride reactively formed during the plasma deposition process. The component is preferably a nuclear fuel rod cladding tube and the coating material is one of ZrN or TiN.

Abstract: A structural part for a nuclear reactor fuel assembly includes a zirconium alloy material having at least one alloy ingredient selected from the group consisting of oxygen and silicon, a tin alloy ingredient, at least one alloy ingredient selected from the group consisting of iron, chromium and nickel, and a remainder of zirconium and unavoidable contaminants. The zirconium alloy material has a content of the oxygen in a range of substantially from 700 to 2000 ppm, a content of the silicon of substantially up to 150 ppm, a content of the iron in a range of substantially from 0.07 to 0.5% by weight, a content of the chromium in a range of substantially from 0.05 to 0.35% by weight, a content of the nickel of substantially up to 0.1% by weight, and a content of the tin in a range of substantially from 0.8 to 1.7% by weight.

Abstract: A fuel element (10) for a nuclear reactor having a zirconium-tin alloy cladding tube (20), with a thin coating (30) of a burnable poison consisting of an enriched boron particles sealed in a boron-containing glass or glass compound deposited from a liquid sol-gel.

Abstract: A burnable absorber controls axial power peaking or moderator temperature coefficient while additional elements are added to improve strength and/or corrosion resistance in a zirconium alloy containing erbium in a range of from about 0.05 to 2 wt. % selected from the group consisting of a naturally occurring distribution of isotopically enriched erbium-167 and a combination thereof; in a range of from a measurable amount up to 1.4% tin; from 0.2 to 0.5 wt. % iron; from 0.07 to 0.25 wt. % chromium; in a range of from a measurable amount up to 0.6 wt. % niobium; in a range of from a measurable amount up to 0.5 wt. % vanadium; 50-120 ppm silicon; 1000-2200 ppm oxygen and a balance of zirconium. Alternatively, the erbium can be replaced by gadolinium in a range of from about 0.05 to 6 wt. % selected from the group consisting of a naturally occurring distribution of gadolinium isotopes, isotopically enriched gadolinium-157 and a combination thereof.

Abstract: A corrosion resistant metallic coating (60) of zirconium nitride is applied to the cladding tube (40) of a nuclear fuel rod (20). The zirconium nitride is reactively deposited on a zirconium-alloy cladding tube by a cathodic arc plasma deposition process. The zirconium nitride coating provides superior wear test results and enhances the corrosion resistance of the cladding tube.

Abstract: A zirconium-base alloy including gallium in amounts effective to improve creep strength of the alloy. The Zr-base alloy can include up to 1 wt. % Ga, up to 0.5 wt. % oxygen, up to 1 wt. % Sn and up to 1 wt. % total of Fe, Cr and V. For instance, the alloy can include 0.5-0.5 wt. % Ga, 0.1-0.7 wt. % Sn, 0.1-0.5 wt. % Fe, 0.15-0.4 wt. % V, 0-0.5 wt. % Cr, 0.1-0.25 wt. % oxygen, balance Zr and unavoidable impurities. The alloy is particularly useful as a component of a fuel assembly such as a fuel tube.

Abstract: A fuel assembly for a boiling water reactor includes an elongated fuel assembly case, mutually parallel fuel rods having longitudinal axes and being disposed in the case, and longitudinally extending inner walls having lateral surfaces facing toward the fuel rods. At least some of the lateral surfaces have grooves formed therein extending perpendicular to the longitudinal axes of the fuel rods. A boiling water reactor includes a boiling water circuit having a pressure vessel and a steam turbine. A plurality of the fuel assemblies are disposed mutually parallel in the pressure vessel. The fuel rods are bathed by boiling water flowing around them in the longitudinal direction of the fuel rods.

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 tin, 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 metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

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.

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.

Abstract: A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

Abstract: A fuel element for a nuclear reactor having a zirconium-tin alloy cladding tube, with a thin coating of particles of enriched boron-containing compound burnable poison particles, deposited from a liquid sol-gel which includes a glass binder material.

Abstract: A method of producing enhanced radial texture in zirconium alloy tubing suitable for use in forming cladding for nuclear fuel rods is provided. The tubing production method described herein employs a combination of mechanical expansion and heat treatment steps in the final stage of tubing formation to produce a single peak radial texture in the tubing, thereby imparting enhanced resistance to pellet-cladding-interaction to the finished tubing. The tubing is preferably processed to a diameter within less than about 10 to 20% of the desired final diameter, annealed, and expanded less than about 10 to 20% to the desired final diameter, thereby producing a unique radial texture in the finished tubing. In an alternative method, the finally expanded tubing is subjected to a final recrystallization anneal to produce a significantly enhanced split radial texture.

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.

Abstract: A fuel rod for a nuclear reactor fuel assembly includes a cladding tube having an outer surface and a given total wall thickness. Nuclear fuel is disposed in the cladding tube. The cladding tube is formed of a first zirconium alloy which may have alloy components of from 1.2 to 2% by weight of tin, 0.07 to 0.2% by weight of iron, 0.05 to 0.15% by weight of chromium, 0.03 to 0.08% by weight of nickel, 0.07 to 0.15% by weight of oxygen, and a total percent by weight for the alloy components of iron, chromium and nickel in a range of from 0.18 to 0.38% by weight. The first zirconium alloy may also have alloy components of from 1.2 to 2% by weight of tin, 0.18 to 0.24% by weight of iron, 0.07 to 0.13% by weight of chromium, 0.10 to 0.16% by weight of oxygen, and a total percentage by weight for the components of iron and chromium in a range of from 0.28 to 0.37% by weight.

Abstract: A nuclear reactor fuel assembly includes a fuel rod containing nuclear fuel in a cladding tube formed of an iron-containing zirconium alloy. A fuel assembly skeleton to which the fuel rod is attached has a structural part formed of the iron-containing zirconium alloy. The iron-containing zirconium alloy has an oxygen content of from 0.1 to 0.16% by weight and contains alloy components of from 0 to 1% by weight of niobium, 0 to 0.8% by weight of tin, at least two metals from the group consisting of iron, chromium and vanadium having from 0.2 to 0.8% by weight of iron, 0 to 0.4% by weight of chromium and 0 to 0.3% by weight of vanadium, a total percent by weight of iron, chromium and vanadium of from 0.25 to 1% by weight, and a total percent by weight for niobium and tin in the range from 0 to 1% by weight.

Abstract: The present invention relates to a method of manufacturing tubes for a nuclear fuel element sheath, comprising a plurality of successive cold rolling and annealing steps including a final .beta. phase heat treatment consisting of maintaining said tube at a temperature between 950.degree. C. and 1250.degree. C. for a time duration sufficient for obtaining a homogeneous .beta. phase within and throughout the whole thickness of said tube, and rapidly cooling said tube to ambient temperature for retaining said .beta. phase throughout the whole thickness of said tube during subsequent cold processing steps including final cleaning of the tube.

Abstract: Disclosed are a cladding tube for a nuclear fuel and a nuclear fuel element incorporating the cladding tube. The cladding tube consists of an inner zirconium liner layer and an outer zirconium alloy layer. The cladding tube has at least one of the following features: (I) the ratio a/b of the oxygen content a to iron content b in the zirconium liner layer is greater than 1.0, (II) the zirconium liner layer is made of a zirconium into the matrix of which impurities are dissolved, and (III) the second phase particles having microscopic sizes and dispersed in the inner surface of the zirconium liner layer and/or the outer surface of the zirconium alloy layer have been removed substantially. Owing to these features, undesirable stress corrosion cracking and local corrosion are remarkably suppressed in the cladding tube and the nuclear fuel element of the invention.

Abstract: A chromium-nickel austenitic stainless steel alloy composition, including specific proportions of carbon with a combination of niobium and tantalum.

Abstract: A zirconium-based alloy with a high corrosion resistance, consisting essentially of 1 to 2 wt % Sn, 0.20 to 0.35 wt % Fe, 0.03 to 0.16 wt % Ni and the balance substantially Zr. The Fe/Ni content ratio of the alloy ranges between 1.4 and 8. The structure of the alloy has fine intermetallic compound of Sn and Ni is precipitated within the zirconium crystal grain of .alpha.-phase. The alloy may further contain 0.05 to 0.15 wt % Cr. This alloy exhibits reduced hydrogen absorption rate and suffers from no nodular corrosion, so that it can suitably be used as a material of nuclear fuel cladding tubes. The nuclear fuel cladding tube made of this alloys exhibits extended service life when used in a nuclear reactor of high degree of burn-up.

Abstract: An improved nuclear fuel rod includes a plurality of cylindrical nuclear fuel pellets being disposed end-to-end in a stack, and an elongated cylindrical cladding tube providing a hermetically sealed chamber. The cladding tube includes a cylindrical wall having inner and outer spaced surfaces and a thickness "x". The stack of fuel pellets are contained in the chamber and spaced radially inwardly from the wall. The thickness "x" of the cladding tube wall between the inner and outer surfaces is the thickness required to generate sufficient heat internally of the wall between its inner and outer surfaces to satisfy the following relationship of a predetermined total fuel rod radiation intensity output, I, to a predetermined fuel rod radiation intensity generated by the fuel pellets contained in the fuel rod, I.sub.0 :I=I.sub.0 e.sup.-ux, where "u" is the attenuation coefficient which varies with cladding material and radiation type.

Abstract: A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.