Abstract: The present invention provides a fuel cladding pipe made of Zr alloy for nuclear reactor having a high degree of strength and possessing superior resistance to corrosion. The fuel cladding pipe of the present invention is formed of a Zr alloy containing Nb in the range of 1.35˜2.15% by weight, with the remainder comprising Zr and inevitable impurities, or is formed of a Zr alloy containing Nb in the range of 0.4˜2.15% and Cr in the range of 0.01˜0.7%, with the remainder comprising Zr and inevitable impurities.

Abstract: The present invention provides a creep resistant zirconium alloy comprising a coarse grained lath alpha microstructure. The microstructure can include small second phase precipitates. The small second phase precipitates can have a diameter less than 0.15 &mgr;m. The microstructure can be partially recrystallized. The microstructure is an acicular structure and can include a lath spacing within the range from about 0.5 to about 3.0 &mgr;m. The microstructure is an acicular structure and can include a lath spacing within the range from about 0.5 to about 3.0 &mgr;m. The present invention provides a nuclear fuel cladding comprising an annular layer of the creep resistant zirconium alloy.

Abstract: A fuel element for a pressurized water reactor is described. The fuel element contains a laterally open skeleton having control-rod guide tubes each with a first end and a second end, spacers fastened to the control-rod guide tubes, a fuel element head disposed at the first end of the control-rod guide tubes, and a fuel element foot disposed at the second end of the control-rod guide tubes. Gastight cladding tubes are inserted into the skeleton and each is filled with a column of fuel pellets. At least some of the gastight cladding tubes have a multilayer wall. The multilayer wall is formed of a mechanically stable matrix containing a first zirconium alloy disposed in a middle of the multiplayer wall; and a thinner protective layer of a second zirconium alloy alloyed to a lesser extent than the first zirconium alloy. The thinner protective layer is bound metallurgically to the matrix and is disposed on an inside of the matrix facing the fuel pellets.

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: An improved fuel element for use in a nuclear reactor comprised of a central core of nuclear material, which is surrounded by a composite cladding. The cladding has an outer metallic tubular portion comprised of well-known cladding alloys used for such purposes. Metallurgically bonded to the outer metallic tubular portion is a commercially pure zirconium microalloyed with a controlled quantity of iron. The zirconium microalloyed with iron produced an inner metallic barrier having a beneficial balance between stress corrosion crack resistance and corrosion resistance while retaining other beneficial properties of pure zirconium, such as ductility.

Abstract: A method of jacketing a uranium slug to an aluminum container comprising applying a coating to the exterior of the container, the coating consisting of colloidal graphite in water, permitting the coating to dry, applying an alloy of aluminum and silicon to the interior surface of the container at a temperature between 588° C. and 594° C., inserting the slug into the container in complete contact with the alloy, and quenching the assembly.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

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 corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A fuel assembly for a pressurized water reactor having a fuel rod with a high strength cladding tube including an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: The invention relates to a nuclear fuel element for pressurized-water reactors comprising a fuel cladding of a zirconium alloy composed of two parts. The inner part constitutes a supporting part of a conventional zirconium and the outer part a corrosion-resistant layer of zirconium with 0.65-0.95% Sn and 0.4-0.5% Fe. The difference in Sn content between the inner and outer parts must not exceed 0.7%. The cladding is finally heat-treated within the range 450 to 510.degree. C.

Abstract: A fuel assembly for a pressurized water reactor having a fuel rod with a high strength cladding tube including an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

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: 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: Fuel rod 2 has a fuel cladding tube 3, and upper end plug 5 and a lower end plug 6 welded to the ends of fuel cladding tube 3. He gas 1 and U pellets 4 are filled in the fuel cladding tube 3. The He gas pressure 1 is about 8 atm, and the average crystal grain size of the U pellet 4 is in the range of 30-60 .mu.m. The fuel rod can be manufactured at a lower manufacturing cost, while maintaining the fuel integrity the same as in a conventional fuel rod.

Abstract: A composite zirconium alloy component that exhibits a reduced propensity for irradiation growth and enhanced corrosion resistance when in a radiation field. The component is formed to have an inner core and two oppositely-disposed outer layers that are metallurgically bonded to the inner core. The inner core and the outer layers are formed from two different zirconium alloys, with the inner core alloy contributing to the ability of the component to resist dimensional distortions as a result of irradiation growth, while the outer layer alloy contributes to the ability of the component to resist corrosion. As such, the alloys are combined within the component in a manner that exploits the different corrosion and irradiation-resistance properties of the alloys.

Abstract: A high strength cladding tube for nuclear fuel for a nuclear reactor having an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A fuel assembly for a pressurized water reactor having a fuel rod with a high strength cladding tube including an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A nuclear fuel rod for a pressurized water reactor having an elongated hollow metallic tubular cladding for containing nuclear fuel, the tubular cladding having an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: In reactor water control for a BWR power plant during an operation cycle after loading new fuel rods in the BWR power plant, an operation is performed which accelerates deposition of crud on the fuel rods until the deposition amount of the crud on the fuel rods reaches a predetermined target value within the operation cycle concerned. After the deposition amount of the crud has reached the predetermined target value, the crud deposition accelerating operation is terminated, whereby even when loading new fuel rods such as Zr liner type fuel rods having a surface on which metal ion deposition has been difficult, radioactivity in the reactor water is efficiently reduced.

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 process for making a cladding for a nuclear fuel rod is disclosed which includes the following steps. First, at least two layers of zirconium-based metals are spray deposited on a generally cylindrical substrate in a manner that provides a rough, cylindrically-shaped, porous billet. The zirconium-based metal for each layer is different from an adjacent layer and is selected to provide a desired combination of properties. Next, the substrate is removed from the spray deposited material, thereby providing an elongated tube hollow. The spray deposited material is then consolidated in a known manner to remove the porosity, densify the tube hollow to substantially full density, and reduce the cross-sectional dimension of the tube hollow to a desired size.

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: 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: 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: A method for fabricating a composite cladding comprised of a moderate-purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube which improves corrosion resistance. The improved corrosion resistance of the liner is accomplished by suitable heat treatment of the Zircaloy-zirconium composite cladding to allow diffusion of alloying elements, notably Fe and Ni, from the Zircaloy into the zirconium, in particular, to the inner surface of the zirconium liner. This diffusion anneal reduces the undesirable tendency of zirconium liner to oxidize rapidly.

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: 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: 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 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: 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: 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 wear resistant coating (50) for a component of a nuclear reactor fuel assembly or control assembly comprising a ceramic material (52) which is premixed with a glass (54). The coating is applied to components of the nuclear reactor vessel, or portions thereof, which are subject to the greatest wear due to friction or fretting. Also disclosed herein is a method for enhancing the wear resistance of a portion of a metal component of a nuclear reactor by coating a surface of the component with a wear-resistant coating formed from diamond, metal nitride, or a composite of a ceramic material and a glass.

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 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 barrier tube of non-circular cross section is arranged in the space between the stainless steel cladding and metal fuel slugs of a liquid metal reactor. The non-circular shape of this barrier design promotes improved thermal bonding in both the cladding/barrier and barrier/fuel interface regions. The non-circular barrier results in three areas with liquid metal thermal bond gaps having an angle which approaches that for a metal fuel pin without a barrier. Thus, the degree of circumferential unbonding will be less than that which results in localized fuel melting during radiation.

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: 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: Nuclear fuel rod tubes of zirconium alloy are heat treated in an induction furnace to produce a protective oxide coating two to fifteen microns in thickness. The furnace is only slightly larger than the tubes and receives the endmost eight inches of the tube. The furnace is controllable in zones along the tube. To calibrate the furnace to produce the desired temperature profile, typically a flat profile at a temperature between 650.degree. and 750.degree. C..+-.1.5.degree. C., a temperature calibration probe is provided with spaced thermocouples for sensing the temperature developed in the probe at each of the zones when heated. The probe is made of inconel 600 stainless or the like, and is dimensioned and shaped to correspond closely to the dimensions of the fuel rod tubes, including having a closed chamfered end. At the opposite end the probe protrudes from the furnace, where the thermocouple leads are terminated. The leads pass through a potting compound in the probe, such as magnesium oxide.

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 method for coating the inside surface (18) of tubular components (16) of a nuclear fuel assembly, including the steps of supporting the component within a vacuum chamber (12), positioning a source rod (20) having a field emitter structure (22) within the component, the structure being formed of material to be coated on the surface, and inducing an electrical current flow (26) through the rod sufficient to evaporate at least a portion of the emitter structure, whereby the evaporated material of the emitter structure is deposited on and adheres to the surface as a coating. Optionally, the vacuum chamber is backfilled with a reactive gas (38), and the material evaporated from the emitter structure chemically reacts with the gas before adhering to the surface. The reactive gas can be one of nitrogen, oxygen, or carbon plasma and the coating adhered to the component can be one of a nitride, oxide, or carbide, respectively.

Abstract: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing on a weight percentage basis tin in a range of 0.4 to 1.0 percent and typically 0.5; iron in a range of 0.3 to 0.6 percent, and typically 0.46 percent; chromium in a range of 0.2 to 0.4 percent, and typically 0.23 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1200 to 2500 ppm, typically 1800 to 2200 ppm. The high oxygen level assists in reducing hydrogen uptake of the alloy compared to Zircaloy-4, for example.

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 method for sputtering coating the inside surface of a fuel assembly tubular component (10,200,400,700) with absorber material such as a burnable poison or getter. The steps include supporting the e.g., cladding tube (10) in a fixture (12,14) and supporting a source tube (100) of e.g., burnable poison material coaxially within the cladding tube, thereby defining a cylindrical annulus (24) between the tubes. The annular space is evacuated and backfilled with an inert working gas (26) such as argon, to a pressure sufficient to sustain a plasma discharge. The cladding tube is positively biased (36) as an anode, and the source tube is negatively biased (34) as a cathode, such that a plasma of the working gas is established in the annular space. A circumferential magnetic field is generated around the source tube to confine and shape the plasma whereby the source tube is bombarded with ions from the plasma substantially uniformly over the length of the source tube.