Abstract: An adduct of at least one isocyanatoalkyltrimethoxysilane with at least one flame retardant reactive with the isocyanatoalkyltrimethoxysilane is used in compositions and for coating plexiglass.

Abstract: In a method for heat treating a metal tube or pipe for a nuclear power plant, the tube or pipe being accommodated in a batch-type vacuum heat treatment furnace, when the tube or pipe is laid down on and is subjected to heat treatment on a plurality of metal cross beams arranged along a longitudinal direction of the tube or pipe, it is possible to suppress scratches to be formed on the outer surface of the tube or pipe and attributable to heat treatment, and to reduce the discoloration on the outer surface of the tube or pipe by holding the tube or pipe and the metal cross beams in indirect contact with each other by virtue of a heat resistant fabric having a thickness of 0.1 to 1.2 mm interposed in between.

Abstract: In a method for heat treating a metal tube or pipe for a nuclear power plant, the tube or pipe being accommodated in a batch-type vacuum heat treatment furnace, when the tube or pipe is laid down on and is subjected to heat treatment on a plurality of metal cross beams arranged along a longitudinal direction of the tube or pipe, it is possible to suppress scratches to be formed on the outer surface of the tube or pipe and attributable to heat treatment, and to reduce the discoloration on the outer surface of the tube or pipe by holding the tube or pipe and the metal cross beams in indirect contact with each other by virtue of a heat resistant fabric having a thickness of 0.1 to 1.2 mm interposed in between.

Abstract: The present invention relates generally to nuclear reactors, and more particularly, to nuclear reactors having fuel assemblies that employ support grids. A method of reducing friction and physical contact between a fuel rod and support grid in a nuclear fuel assembly is provided. The method includes applying a lubricant composition to the outer surface of the fuel rod during fuel assembly fabrication and removing the lubricant composition afterward.

Abstract: There are provided an austenitic stainless steel having high stress corrosion crack resistance, characterized by containing, in percent by weight, 0.030% or less C, 0.1% or less Si, 2.0% or less Mn, 0.03% or less P, 0.002% or less S, 11 to 26% Ni, 17 to 30% Cr, 3% or less Mo, and 0.01% or less N, the balance substantially being Fe and unavoidable impurities; a manufacturing method for an austenitic stainless steel, characterized in that a billet consisting of the said austenitic stainless steel is subjected to solution heat treatment at a temperature of 1000 to 1150° C.; and a pipe and a in-furnace structure for a nuclear reactor to which the said austenitic stainless steel is applied.

Abstract: The invention proposes a zirconium-based alloy also containing, by weight, apart from unavoidable impurities, from 0.02 to 1% of iron, from 0.8% to 2.3% of niobium, less than 2000 ppm of tin, less than 2000 ppm of oxygen, less than 80 ppm of carbon, from 5 to 35 ppm of sulphur and less than 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, being lower than 3.

Abstract: A method of surface-treating a reactor member for effectively removing a Cr-deficient layer and a work-hardened layer considered to be a cause of stress corrosion cracking (SCC) under low-stress conditions. The method of surface-treating a reactor member which is worked by bending (step 1) and then processed by a heat treatment (step 2), in which a work-hardened layer is formed by the bending, and in which a Cr-deficient layer is formed due to an oxide film attached by the heat treatment, uses at least one of: acid wash; grinding; electrolytic polishing; electro-discharge machining; surface cutting; surface deoxidation and softening; wet blasting; laser machining; or surface plating (step 3) to remove the work-hardened layer and the Cr-deficient layer from the reactor member or to prevent contact of the work-hardened layer and the Cr-deficient layer of the reactor member with a primary coolant.

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.

Abstract: A method of producing austenite steel for use in the radiation zone of a nuclear reactor. The method comprising the steps of forming the austenite steel with about 17% by weight chromium, about 9 to 11.5% by weight nickel, about 0.04% by weight carbon, and iron impurities whose content of silicon is about 0.1% by weight and whose total content of sulfur and phosphorous is less than 0.03% by weight; and exposing the austenite steel to a temperature treatment at a temperature less than 1150 .degree. C. to produce a fine grain lattice with grain diameter features under approximately 20 .mu.m.

Abstract: An alloy component having an oxide film formed on its surface. The oxide film has atoms of a metal incorporated therein by a process comprising immersing the alloy surface in high-temperature water in which compound containing the metal is dissolved. The metal has the property of increasing the corrosion resistance of the alloy when incorporated in the oxide film. The compound has the property that it decomposes in the high-temperature water to release atoms of the metal which incorporate in the oxide film.

Abstract: Zirconium-based components for nuclear reactors are disclosed whereby a rare earth or other metal which preferentially concentrates hydrogen when in contact with zirconium is dispersed as a second phase throughout a zirconium-based matrix. The morphology and distribution of the two-phase material is controlled such that the second phase, internal hydrogen absorbing particles, are provided as spherical particles, randomly distributed, to minimize the effects of the occluded hydrogen.

Abstract: The alloy has a base composition similar to that of a zirconium alloy of known type used for the manufacture of an element intended for use in the core of a nuclear reactor, such as a cladding tube, a guide tube, or another structural element of a fuel assembly. In addition, the alloy contains sulphur in a proportion by weight of between 8 and 100 ppm and preferably between 8 and 30 ppm.

Abstract: A chromium nickel silicon stainless steel alloy with improved wear resistance consists of, in weight percent, 19 to 22 chromium, 8.5 to 10.5 nickel, 5.25 to 5.75 silicon, 1.7 to 2.0 carbon, 8.0 to 9.0 niobium, 0.3 to 0.5 titanium and the balance iron plus impurities. The addition of titanium and increased amounts of niobium and silicon alter the microstructure of the stainless steel to form a duplex austenitic/ferritic microstructure which undergoes secondary hardening due to the formation of an iron silicon intermetallic phase.

Abstract: This invention provides austenitic steel having a superior resistance to neutron irradiation, a nuclear reactor and a nuclear fission reactor using such steel.In the present invention, austenitic steel containing at least one of Pd and Pt is used for a member which undergoes neutron irradiation in high-temperature water. More specifically, the steel consists essentially of not more than 0.02% C, not more than 1% Si, 0.5-15% Mn, 9-26% Cr, 8-20% Ni, 0.5-3% Mo, at least one kind of 0.1-16% selected from the group consisting of Pd and Pt, and not less than 50% Fe, and may further contain not more than 1.0% of at least one kind selected from the group consisting of Nb, Ti, Zr, Hf and V, and this is used for structural members in a core of a nuclear reactor and a nuclear fission reactor.

Abstract: An improved stainless steel composition and an improved stainless Ni-based alloy are disclosed for use in nuclear reactor environments. The improved stainless alloys include a stainless steel and a stainless nickel based alloy that are particularly well adapted for use in reactor components that are positioned adjacent to components formed from a zirconium based metal. The improved stainless steel is an austenitic stainless steel material that include less than approximately 0.2 percent manganese by weight, but does include sufficient austenitic stabilizer to prevent a martensitic transformation during fabrication or use in a reactor environment. In one preferred embodiment, the stainless alloy is formed without any significant amount of manganese.

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 nuclear reactor having structural members made of austenitic stainless steel which is corrosion-resistant in an environment of neutron irradiation, and can suppress stress corrosion cracking and embrittlement. At least one additive selected from the group consisting of Ti more than 0.2% by weight but not more than 0.6%, Zr more than 0.2% but not more than 1.14%, Hf more than 0.2% but not more than 2.24%, V more than 0.2% but not more than 0.64%, Nb more than 0.5% but not more than 1.17% and Ta more than 0.5% but not more than 2.27% is added to austenitic stainless steel containing Cr, Ni and so on, and said at least one exists in a solid-solution state. In addition, the steel has a wholly austenitic structure substantially free of carbide. In order to maintain the irradiation-induced segregation prevention effected by the addition element, the C content is limited to 0.01 to 0.008%, and the N content is limited to 0.001 to 0.0%.

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: Methods and apparatus for improving fretting resistance of zirconium alloy components formed into a shape for use in a nuclear reactor are disclosed in which at least a portion of the outer surface of a component is reacted with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of the foregoing at a temperature below about 700.degree. C. to form a wear resistant layer on the surface of the component.

Abstract: A corrosion resistant zirconium alloy for use as a nuclear reactor cladding material. This alloy consists essentially of, on a weight % basis 0.02-1.7 % Sn, 0.19-0.6% Fe, 0.07-0.4% Cr, 0.01-0.2% Ta, and optionally 0.05 up to less than 0.5% Nb, the balance being Zr and incidental impurities, provided that the content of the nitrogen in the form of incidental impurities is no more than 60 ppm.

Abstract: An austenitic steel excellent in resistance to neutron irradiation embrittlement which contains, by weight, not more than 0.03% carbon, not more than 1% silicon, 5 to 25% manganese, 15 to 26% chromium, and 10 to 20% nickel, the ratio of atomic volume of chromium to the average atomic volume of matrix of the steel being from 0.900 to 1.030. It is possible to add to the austenitic steel, besides the above-mentioned alloying elements, at least one element selected from the group consisting of niobium, titanium, zirconium, tantalum and vanadium which are effective in corrosion resistance and irradiation embrittlement under neutron irradiation in total amounts of not more than 1.0%. At least one of components composing the inside of a nuclear reactor or nuclear fusion reactor is made of the austenitic steel.

Abstract: A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

Abstract: A zirconium alloy for use in spacer grids for nuclear reactor fuel claddings, which consists essentially of, on a weight percent basis,2.5 to 10% Nb,0.01 to 1.5% of one or more components selected from the group consisting of Y and rare earth elements and oxides thereof,0.05 to 1% of one or more optional element selected from the group consisting of Fe, Cr, Mo and V, andthe remainder Zr and incidental impurities.

Abstract: A corrosion resistant zirconium alloy for use as a nuclear reactor cladding material. This alloy consists essentially of, on a weight % basis 0.02-1.15% Sn, 0.19-0.6% Fe, 0.07-0.4% Cr, and optionally 0.05 to not more than 0.5% Nb, the balance being Zr and incidental impurities, provided that the content of the nitrogen in the form of incidental impurities is no more than 60 ppm.

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: A workable, boron-containing, stainless steel alloy and an article formed therefrom are disclosed together with a process for manufacturing same. The alloy consists essentially of, in weight percent, about______________________________________ w/o ______________________________________ Carbon 0.10 max. Manganese 2.00 max. Silicon 1.00 max. Phosphorus 0.045 max. Sulfur 0.010 max. Chromium 16.00-22.00 Nickel 10.00-15.00 Molybdenum 0-3.0 Boron 0.2-2.0 Nitrogen 0.075 max. ______________________________________and the balance consisting essentially of iron. The as-worked alloy in accordance with the invention is characterized by having a boride particle areal density per weight percent boron (A.sub.N) defined by the relationshipA.sub.N .gtoreq.58,080-18,130 (%B).The as-worked alloy of the invention is further characterized by having a Charpy V-notch impact strength (CVN) defined by the relationshipCVN.gtoreq.85.917 x e.sup.-1.20297(%B).

Abstract: Method for extending service life of nuclear reactor components prepared from ductile, high strength crystalline alloys obtained by devitrification of metallic glasses. Two variations of the method are described: (1) cycling the temperature of the nuclear reactor between the operating temperature which leads to irradiation damage and a lThe U.S. Government has rights in this invention by virtue of Department of Energy, Office of Fusion Energy, Grant No. DE-AC02-78ER-10107.

Abstract: A sheath for a nuclear reactor control rod consists of chromium nickel stainless steel. It is protected against friction wear by nitridation over a depth of from 15 to 50 .mu.m obtained by an electric discharge in a nitrogen containing atmosphere.

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

Abstract: In a control rod construction for moderating and controlling a reaction of a nuclear reactor, a control rod structure is disclosed in which hafnium is welded to stainless steel. Typically, the control rod is confined for movement axially along its length into and out of the reactor. The hafnium is surrounded in the control rod structure by a sheath of stainless steel to inhibit movement relative to the control rod structure other than in the axial direction of control rod movement. The hafnium is friction welded to the stainless steel of the control rod so as to stress the weld in either tension or compression due to dynamic loading occurring from control rod movement. The stainless steel to which the hafnium is welded is in turn conventionally welded or mechanically connected to the remaining steel components of the control rod.

Abstract: A process for improving resistance of pressure vessel shells, tubesheets, tubes, pipes, pipe fittings and machine parts to stress corrosion cracking comprising heating at least those portions of such bodies subject to danger by stress corrosion cracking to a critical elevated themperature level, cooling at least the surface portions of the metal body subject to stress corrosion cracking and then permitting such treated portions of the metal body to come to ambient temperature.

Abstract: The present invention is concerned with a method of reducing radioactivity in a nuclear plant by preliminarily forming oxide films on the surfaces of metallic structural members to be in contact with high-temperature and high-pressure reactor water containing radioactive substances before said metallic members are exposed to said reactor water. The method is characterized by the steps of subjecting said structural members to a first-step oxidation treatment of heating said structural members in an environment of a high temperature, and further subjecting the thus treated structural members to a second step oxidation treatment of heating said treated structural members in an environment having a higher oxidizing capacity than that of said environment in said first-step oxidation treatment to form a denser oxide film than an oxide film obtained in said first step oxidation treatment. According to the present invention, radioactivity in the nuclear plant can be reduced remarkably.

Abstract: An austenitic stainless steel alloy, with improved resistance to radiation-induced swelling and helium embrittlement, and improved resistance to thermal creep at high temperatures, consisting essentially of, by weight percent: from 16 to 18% nickel; from 13 to 17% chromium; from 2 to 3% molybdenum; from 1.5 to 2.5% manganese; from 0.01 to 0.5% silicon; from 0.2 to 0.4% titanium; from 0.1 to 0.2% niobium; from 0.1 to 0.6% vanadium; from 0.06 to 0.12% carbon; from 0.01% to 0.03% nitrogen; from 0.03 to 0.08% phosphorus; from 0.005 to 0.01% boron; and the balance iron, and wherein the alloy may be thermomechanically treated to enhance physical and mechanical properties.

Abstract: Apparatus such as a nuclear fuel sub-assembly for use in a liquid alkali metal environment has bearing surfaces (4) coated with alumina or aluminide particles to reduce wear, fretting and friction. The particles are deposited by plasma or detonation gun using a nickel chromium alloy as a metallic bond.

Abstract: A nuclear fuel assembly containing a plurality of vertical fuel rods and possibly occasional vertical water-filled rods, which are surrounded by a fuel channel of substantially square cross-section and which are divided, by means of a vertical channel for water of substantially cruciform cross-section, into four sub-bundles. Each fuel rod contains a stack of circular-cylindrical pellets of a nuclear fuel arranged in a cladding tube. Each sub-bundle consists of, in all, twenty-five rods arranged in a lattice containing 5.times.5 rods. The cladding tube of each fuel rod has a wall thickness in the range 0.50 to 0.70 mm and an external diameter in the range 8.5 to 10.3 mm.

Abstract: Slender water displacer rods for use in water reactors are provided with rings of a wear resistant coating spaced along the length of the rod. Each coating contains Cr.sub.2 C.sub.3 and is metallurgically applied and bonded to the zirconium base alloy forming the outer portion of the rod by electrospark-deposition (ESD) technique.

Abstract: This is a high quality zirconium or hafnium sponge, a fuel element cladding liner having material of the quality of this sponge and a process utilizing this quality sponge to fabricate liner material for lined fuel element cladding. The sponge contains 250-about 350 ppm of oxygen impurity, 50-300 ppm of iron impurity, and a total of 500-1,000 ppm of impurity. This material is much purer than conventional sponge and approaches the quality of crystal bar.

Abstract: It has been found that modifying standard Zircaloy alloy processing techniques by limiting the working and annealing temperatures utilized after conventional beta treatment results in Zircaloy alloy product having superior high temperature steam corrosion resistance.

Abstract: A metallic composite material and nuclear components such as fuel cladding, rod guide thimbles, grids and channels made therefrom. The metallic composite material comprises 90-60 volume percent of a metal matrix of zirconium or a zirconium alloy containing homogeneously incorporated, throughout the matrix, 10-40 volume percent of silicon carbide whiskers.

Abstract: A process for improving resistance of control rod guide tube split pins in nuclear reactors to stress corrosion cracking comprising heating the split pin to a critical elevated temperature level, cooling at least the surface portions of the split subject to stress corrosion cracking and then permitting the split pin to come to ambient temperature.

Abstract: The nuclear fusion reactor of the present invention presents a new vacuum vessel for enclosing plasma particles where a reactor wall exposed to the above plasma particles has a piled structure. A plurality of heat-resisting ceramic tiles are metallurgically bonded to a metal-base body having a cooling means through a brazing material. The ceramic tiles are preferably composed of sintered silicon carbide of high density and containing a little beryllium oxide between the boundaries of crystal grains.

Abstract: In a process for producing a zirconium-based alloy in which after the zirconium-based alloy is subjected to hot plastic working, it is subjected to cold plastic working and then to annealing and these steps are repeated at least twice, the present invention provides a process for producing a zirconium-based alloy characterized in that after the alloy is subjected to solid solution treatment in which it is heated to a temperature within the range including the .alpha. phase and the .beta. phase of the alloy, or within the range of the .beta. phase and is then quenched, cold plastic working is effected at least twice.

Abstract: An austenitic stainless steel exhibiting improved resistance to corrosion caused by nitric acid is disclosed, which consists essentially of:C: not more than 0.03% by weight, Si: 2-6% by weight,Mn: 0.1-8% by weight, Cr: 20-35% by weight,Ni: 17-50% by weight, Mg: not more than 0.02%,at least one of Nb, Ti and Ta in the total amount of 8.times.C(%) or more, but 1.0% by weight or less,S: not more than 0.003% by weight,Oxygen: not more than 0.003% by weight,N: not more than 0.03% by weight,P: not more than 0.02% by weight,Fe and incidental impurities: balancethe composition further satisfying the following relationships;-10.ltoreq.Ni(Bal).ltoreq.-0.1Cr(%).gtoreq.(7/4)Si(%)+16.5Mg(%)-0.75.times.S(%)-1.5.times.Oxygen(%).gtoreq.0wherein, Ni(Bal)=30.times.C(%)+0.5.times.Mn(%)+Ni(%)+8.2-1.1.times.[1.5.times.Si(%) +Cr(%)].

Abstract: Disclosed is an iron-base alloy eminently suited for use as components in nuclear energy installations. The alloy normally contains, in percent by weight, about 20% chromium, about 10% nickel, about 5.5% silicon, about 1.5% carbon, about 8% niobium plus vanadium, about 0.05% nitrogen, less than 1% cobalt as an impurity and the balance iron plus normal impurities found in alloys of this class.

Abstract: A method and an apparatus for avoiding or reducing the interactions and their consequences from contact of hot liquid metals with concrete. At least for the region of the concrete structure near the surfaces which may come into contact with the hot liquid metal, there is employed a concrete containing aluminum oxide, magnesium oxide, zirconium oxide, or a mixture thereof, in a proportion of 70% by weight to 96.5% by weight.

Abstract: Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

Abstract: Described herein is a water reactor composite nuclear fuel rod cladding tube which includes two concentric metallic layers metallurgically bonded to each other. The outer layer is composed of a conventional zirconium base alloy having high strength and excellent aqueous corrosion resistance. The inner layer is composed of: a low oxygen, BCC crystal structure niobium base material, containing less than about 59 w/o zirconium; or a ferritic stainless steel. This inner layer material is characterized by the ability to prevent the propagation of cracks initiated during reactor operation due to pellet-cladding interaction.

Abstract: A fuel rod for a nuclear reactor comprises a cladding tube of a zirconium-based alloy, on the internal surface of which there is arranged a layer of zirconium containing 0.1-1 per cent by weight tin. The fuel rod contains a nuclear fuel preferably of uranium dioxide.

Abstract: Nuclear fuel assembly having four bundles of fuel rods, at least one fuel bundle being made with twenty symmetrically arranged fuel rod positions, the number of fuel rod positions in each of the remaining fuel rod bundles being at least sixteen and at most twenty-five.

Abstract: Manganese-iron base and manganese-chromium-iron base austenitic alloys designed to have resistance to neutron irradiation induced swelling and low activation have the following compositions (in weight percent): 20 to 40 Mn; up to about 15 Cr; about 0.4 to about 3.0 Si; an austenite stabilizing element selected from C and N, alone or in combination with each other, and in an amount effective to substantially stabilize the austenite phase, but less than about 0.7 C, and less than about 0.3 N; up to about 2.5 V; up to about 0.1 P; up to about 0.01 B; up to about 3.0 Al; up to about 0.5 Ni; up to about 2.0 W; up to about 1.0 Ti; up to about 1.0 Ta; and with the remainder of the alloy being essentially iron.