Abstract: A nuclear fuel bundle has differential peak power limits for the edge or peripheral fuel rods and the interior fuel rods. Also, the magnitude of nuclear fuel in the edge or peripheral rods is decreased in comparison with the magnitude of nuclear fuel in the interior rods. The nuclear reactor can thus operate at higher power output by decreasing the margins between the power outputs and peak power limit of the interior rods, as well as by decreasing the margins between the power outputs of the edge or peripheral rods and the increased peak power limit of those rods. The outer or peripheral edge rods can also be enriched for enhanced power output.

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 sintered nuclear fuel body includes (U, Pu)O.sub.2 mixed crystals having a mean particle size in a range from 7.5 .mu.m to 50 .mu.m. This sintered nuclear fuel body has a high retention capacity for fission gas in a power reactor. In order to produce the sintered nuclear fuel body by sintering a body in a hydrogen-containing sintering atmosphere, a powered substance selected from the group consisting of aluminum oxide, titanium oxide, niobium oxide, chromium oxide, aluminum stearate, aluminum distearate and aluminum tristearate is added to the starting powder for the body. As an alternative or in addition, the body made from the starting powder is sintered during a holding period of 10 minutes to 8 hours at a sintering temperature of 1400.degree. C. to 1800.degree. C. in a hydrogen-containing sintering atmosphere, initially with an oxygen partial pressure of 10.sup.-10 to 10.sup.-20 bar and then from 10.sup.-8 to 10.sup.

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: 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: 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 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 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 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: 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: 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: Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Abstract: Particulates from a source fall into a tundish which discharges to a collection chamber. The build-up of particulates in the chamber is sensed, and a valve opened at the bottom of the chamber when a pre-determined level of the particulates is sensed in the chamber. A valve at the top of the chamber may be closed at the same time so that pressurized air can be injected into the chamber to expel the particulates through the bottom of the chamber through the valve. A filter circuit may be connected to the tundish to assist in removing particulates from the tundish.

Abstract: A removable and reusable cap for fuel rod ends to prevent grid damage during loading permits fuel end configurations to match utility's special tooling for fuel assembly repair and reconstitution. The elastomeric transitory fuel rod guidance ends expand over a flange and temporarily snap into place in a groove in the fuel rod end to provide a tapered guiding surface.

Abstract: An organometallic chromium compound in the gaseous phase is brought into contact with a substrate consisting of the inner surface of the zirconium alloy tubular cladding (5) of the fuel element, or the outer surface of the pellets of fuel material, the substrate being kept at a temperature between 300.degree. and 600.degree. C. The organo metallic compound may consist of chromium acetylacetonate. The process enables a chromium oxide coating to be obtained inside the tubular cladding (5) and/or on the outer surface of the nuclear fuel pellets. Pellet/cladding interaction is thus prevented or limited when this fuel element is used in the reactor.

Abstract: In order to detect leaky rods inside a nuclear fuel assembly (24), this assembly is placed inside a pool (36) so that the gases contained inside the rods (10) escape from the leaky rods. By heating the upper part of the assembly by means of a heated sleeve (42) the gases contained inside the leaky rods are expanded. Monitoring the diameter change of the rods (10) at the base of the assembly by means of a measuring device (44) makes it possible to distinguish the sealed rods whose sheaths elastically get out of shape from the leaky rods whose sheaths hardly warp at all.

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: 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 plumb nozzle for a nuclear fuel assembly having asymmetric loading of the fuel mass comprising a raised protuberance on the handling tool gripper finger engaging surface of the nozzle to compensate for skewing caused by the asymmetric fuel loading. The protuberance is positioned to be engaged by a gripper finger during handling, loading, or unloading of the fuel assemblies and extends downwardly a predetermined distance with respect to the other gripper finger engaging surfaces so that the fuel assembly hangs plumb.

Abstract: A nuclear fuel element suitable for use in fast breeder. The fuel element has a clad tube and a wire spacer wound round the clad tube. The clad tube of the fuel element is composed of a first clad tube and a second clad tube which are connected to each other through an intermediate plug. The first clad tube extends upwardly from the intermediate plug. The upper end of the first clad tube is closed by an upper end plug, while the lower end of the second clad tube is closed by a lower end plug. The first clad tube is charged with a multiplicity of fuel pellets, while the second clad tube defines a gas plenum therein. The spaces inside the first and second clad tubes are communicated with each other through vent holes formed in the intermediate plug. The wire spacer is wound round the first clad tube and is fixed at its one end to the end plug attached to the first clad tube and at its other end to the intermediate plug.

Abstract: A nuclear fuel cartridge comprises a sheath, and nuclear material inside the sheath. A continuous liquid permeable passageway extends from one end of the nuclear material to the other, and an annular end member is disposed at each end of the nuclear material. Inner spacing members also of annular form may be disposed in intermediate positions along the sheath between portions of the nuclear material.

Abstract: This invention teaches an encapsulated fuel unit for a nuclear reactor, such as for an enriched uranium fuel plate of thin cross section of the order of 1/64 or 1/8 of an inch and otherwise of rectangular shape 1-2 inches wide and 2-4 inches long. The case is formed from (a) two similar channel-shaped half sections extended lengthwise of the elongated plate and having side edges butted and welded together to define an open ended tube-like structure and from (b) porous end caps welded across the open ends of the tube-like structure. The half sections are preferably of stainless steel between 0.002 and 0.01 of an inch thick, and are beam welded together over and within machined and hardened tool steel chill blocks. The porous end caps preferably are of T-316-L stainless steel having pores of approximately 3-10 microns size.

Abstract: The present invention relates to an upper neutron protection device for a nuclear reactor assembly.This device comprises at least one container partly filled with a neutron-absorbing product and maintained within a wall located in the upper part of the assembly by at least one spacing plate.Application to fast neutron reactors.

Abstract: A nuclear fission electrical generator for extracting electrical power from fissionable material having semi-conductor characteristics without the necessity of an intervening thermal energy. Unwanted, randomly created electron-hole pairs are depleted from the semiconductor material. Phthalocyanine semiconductor are suitable, and uranium phthalocyanine is particularly suitable for use in this invention.

Abstract: Capillary liquid fuel elements, created by the method of confining a liquid fuel in horizontal capillary troughs, are employed to create the core of a nuclear reactor to generate useful heat energy. The reactor incorporates the inherent advantages of a liquid fuel reactor: high specific power, high fuel burnup, inherent safety, ease of control, and simple fuel preparation, processing, reprocessing, and handling. The reactor in addition, has advantages unavailable in other liquid fuel reactors: high breeding potential, low delayed neutron loss, low pumping power requirement, low fission material external holdup, direct fuel-coolant heat exchange capability, and low construction material cost.