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 composite nuclear fuel container for service in water cooled nuclear fission reactor plants having improved resistance to corrosion, and a method of producing same. The invention comprises each component of the fuel container being of specific compositions which have been heat treated to transform their microcrystalline structure in such a manner to optimize the corrosion resistance of each component of the fuel container.
Abstract:
A fuel assembly for a nuclear reactor comprising a fuel cladding tube of three-layer structure having an outer surface in contact with reactor water of the nuclear reactor, an inner surface layer in contact with the nuclear fuel, and an intermediate layer interposed between the outer surface layer and the inner surface layer. the outer surface layer is made of a Zr-based alloy containing Nb, Sn and Mo. The inner surface layer is made of pure zirconium. The intermediate layer is made of a high ductility alloy which is higher in ductility than the outer surface layer and is higher in strength than the inner surface layer.
Abstract:
The present invention pertains to zirconium base alloys containing about 0.1 to 0.6 weight percent tin; about 0.07 to 0.24 weight percent iron; about 0.05 to 0.15 weight percent chromium; and up to about 0.05 weight percent nickel. The balance of the alloy is zirconium with incidental impurities. The levels of the incidental impurity, oxygen, is controlled to a level of less than about 350 ppm. These alloys have been designed to minimize the adverse effects of pellet-clad interaction, when they are used as a liner bonded to the inside surface of water reactor nuclear fuel cladding. Specific cladding and fuel element designs according to the present invention are described.
Abstract:
A method of producing a cladding tube for nuclear fuel for a nuclear pressure water reactor includes forming a tube which at least principally consists of a cylindrical tube component of a zirconium-based alloy, where the alloying element, except for zirconium, which has the highest content in the alloy is niobium, wherein the niobium content in weight percent is between about 0.5 and about 2.4 and wherein no alloying element, except for zirconium and niobum, in the alloy, has a content which exceeds about 0.2 weight percent. The cladding tube is then annealed such that the tube component is partly but not completely recrystallized. The degree of recrystallization in the tube component is higher than about 40% and lower than about 95%. A fuel assembly for a nuclear pressure water reactor also has a plurality of such cladding tubes.
Abstract:
The invention relates to a flexible fitting solution for a portable device for inputting control signals to a peripheral unit, which includes a holding member designed to be attached to a hand of a user of the device so as to retain the device in a predetermined manner on the hand. The holding member comprises at least one elongated portion and at least one curved portion connected thereto. A side of the holding member that faces the user's hand when the device is attached to the hand includes a padding module. The padding module, in turn, includes at least one pad cushion adapted to fit the device to the user's hand. The padding module is removably fastened to the holding member by means of a set of fastening members, which secure the padding module in a fastened position. By selecting an appropriate padding module, each user of the device may obtain a good fit of the device onto his/her hand.
Abstract:
An underwater laser processing method is carried out by irradiating, through a laser beam irradiation apparatus, a laser beam having a high output, a short pulse and a visible wavelength to a surface of a structure immersed in a water to improve residual stress of a material of the surface of the structure and remove a crack or a CRUD thereof. The laser beam irradiation apparatus comprises a pulse laser device suspended into a water in which a metal material is accommodated from an upper side thereof for irradiating a laser beam having a visible wavelength to a processing position, a beam strength adjusting device for adjusting an output per 1 pulse of a laser beam generated by the pulse laser device and a mechanism for adjusting a spot diameter and a multiplexing ratio of an irradiated beam.
Abstract:
An underwater laser processing method is carried out by irradiating, through a laser beam irradiation apparatus, a laser beam having a high output, a short pulse and a visible wavelength to a surface of a structure immersed in a water to improve residual stress of a material of the surface of the structure and remove a crack or a CRUD thereof. The laser beam irradiation apparatus comprises a pulse laser device suspended into a water in which a metal material is accommodated from an upper side thereof for irradiating a laser beam having a visible wavelength to a processing position, a beam strength adjusting device for adjusting an output per 1 pulse of a laser beam generated by the pulse laser device and a mechanism for adjusting a spot diameter and a multiplexing ratio of an irradiated beam.
Abstract:
In order to provide an austenitic single crystal stainless steel having preferable stress corrosion cracking resistance, strength, and irradiation induced embrittlement resistance so as to extend the life of a nuclear reactor core structure, which is used under a high radiation dose environment, a method is employed, which comprises the steps of homogeneously dispersing carbides into a parent phase of the austenitic single crystal stainless steel by a two step solution heat treatment, and subsequently effecting an ageing heat treatment after rapid cooling for precipitating fine carbides. Austenitic single crystal stainless steel having preferable stress corrosion cracking resistance, strength, and irradiation induced embrittlement resistance can be provided, and the life of nuclear reactor core structure, which is used under a high radiation dose environment, can be extended.