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Dispersion Type Materials For Fuel Elements Part I Uranium Mononitride And Uranium Silicide Dispersion Materials
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Book Synopsis DISPERSION-TYPE MATERIALS FOR FUEL ELEMENTS. PART I. URANIUM MONONITRIDE AND URANIUM SILICIDE DISPERSION MATERIALS. by :
Download or read book DISPERSION-TYPE MATERIALS FOR FUEL ELEMENTS. PART I. URANIUM MONONITRIDE AND URANIUM SILICIDE DISPERSION MATERIALS. written by and published by . This book was released on 1958 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The basic requirements for dispersion-type fuel elements and the basis for the selection of the components, i.e., melting temperature, density, and uranium-content of the uranium compounds, and melting temperature and absorption cross section of the matrix materials are described. Two types of dispersions are investigated: uranium nitride and uraniuin silicide. Their reactions with Zircaloy-2 and other matrix materials, such as molybdenum, niobium, Nichrome-V, Ti -Nb alloys, or vanadium, are shown in some characteristic photomicrographs. The corrosion rate of some of these dispersion-type materials in 600 deg C water is briefly discussed. (auth).
Download or read book Reactor Core Materials written by and published by . This book was released on 1958 with total page 938 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1974 with total page 730 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Uranium Carbides, Nitrides, and Silicides by :
Download or read book Uranium Carbides, Nitrides, and Silicides written by and published by . This book was released on 1965 with total page 204 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Fabrication and Properties of Hot-pressed Uranium Mononitride by : Edward O. Speidel
Download or read book Fabrication and Properties of Hot-pressed Uranium Mononitride written by Edward O. Speidel and published by . This book was released on 1963 with total page 74 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis The Nitridation Rates of Uranium-fissium Alloys by : J. P. LaPlante
Download or read book The Nitridation Rates of Uranium-fissium Alloys written by J. P. LaPlante and published by . This book was released on 1962 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: The nitridation rates of uranium-5% fissium alloys were investigated at about 200 to 650 deg C and at nitrogen concentrations in argon from about 0.5 to 100%. The effects of Na coatings and irradiation of the alloy were examined. The nitridation rates of the alloys followed a parabolic law, with activation energies between 14 and 16 kcal/mole. The ignition behavior of irradiated alloy in nitrogen-argon mixtures was investigated briefly.
Book Synopsis Thermal Compatibility Studies of Unirradiated Uranium Silicide Dispersed in Aluminum. [Reduced Enrichment for Research and Test Reactor]. by :
Download or read book Thermal Compatibility Studies of Unirradiated Uranium Silicide Dispersed in Aluminum. [Reduced Enrichment for Research and Test Reactor]. written by and published by . This book was released on 1984 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Powder metallurgy dispersions of uranium silicides in an aluminum matrix have been developed by the international Reduced Enrichment for Research and Test Reactors program as a new generation of proliferation-resistant fuels. A major issue of concern is the compatibility of the fuel with the matrix material and the dimensional stability of this fuel type. A total of 45 miniplate-type fuel plates were annealed at 400°C for up to 1981 hours. A data base for the thermal compatibility of unirradiated uranium silicide dispersed in aluminum was established. No modification tested of a standard fuel plate showed any significant reduction of the plate swelling. The cause of the thermal growth of silicide fuel plates was determined to be a two-step process: (1) the reaction of the uranium silicide with aluminum to form U(AlSi)3 and (2) the release of hydrogen and subsequent creep and pillowing of the fuel plate. 9 references, 4 figures, 6 tables.
Book Synopsis DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS. by :
Download or read book DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS. written by and published by . This book was released on 1959 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl4 or UAl3 compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm3), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO2 was successfully employed in the preparation of large quantities of UC2, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC2-Al compacts. The hydrogen appears to reduce the UC2 to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC2 revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth).
Download or read book Titanium Abstract Bulletin written by and published by . This book was released on 1960 with total page 378 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Prospects of Stable High-density Dispersion Fuels by :
Download or read book Prospects of Stable High-density Dispersion Fuels written by and published by . This book was released on 1987 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The majority of research and test reactors around the world employ aluminum fuel element designs that contain dispersed powders of uranium compounds as fuel. Specifically, two compounds are used: (1) uranium oxide (U3O) and (2) an uranium aluminide mixed phase composed of the intermetallic compounds UAl2, UAl3, and UAl4, all made with highly enriched uranium (HEU), i.e., 93% 235U. The reduction of 235U enrichment to below 20%, to so-called low enriched uranium (LEU), requires the use of higher density fuels for those applications where increased fuel loading is not feasible. Fuel dispersant loading is, in practice, limited to approximately 45 vol %. Fuel development in the Reduced Envichment Research and Test Reactors (RERTR) program has focused on uranium silicides (U3Si and U3Si2) as the most promising high-density fuels. The compounds of U6Fe and U6Mn as well as U3Si containing Cu were tested as part of the search for stable very-high-density fuels. The problem of breakaway swelling in high-density fuel compounds is attributed to radiation-induced amorphization of these compounds. Alloy additions are a possible means by which the crystal structure of very-high-density compounds can be strengthened and preserved to high irradiation doses. Tailoring metallurgical treatment during fabrication, to avoid thermodynamically weak compounds, appears promising for certain compound combinations. 5 refs., 2 figs.
Book Synopsis High Uranium Density Dispersion Fuel for the Reduced Enrichment of Research and Test Reactors Program by : Adam B. Robinson
Download or read book High Uranium Density Dispersion Fuel for the Reduced Enrichment of Research and Test Reactors Program written by Adam B. Robinson and published by . This book was released on 2007 with total page 118 pages. Available in PDF, EPUB and Kindle. Book excerpt: This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to decrease the use of high enriched uranium in research and test reactors around the world, new fuels with high uranium densities must be developed such that low enrichment fuel may be used in its place. Preliminary studies on uranium molybdenum alloys have shown promising results. A uranium molybdenum fuel phase dispersed in a zirconium matrix is proposed and examined in this thesis. Work described herein includes the successful fabrication of materials, preparation of samples, diffusion testing, fuel fabrication, and analysis of the resulting product. The fabrication results appear to be very good and all data collected indicates that this fuel type is fabricable and justifies irradiation testing.
Download or read book Bibliographical Series written by and published by . This book was released on 1964 with total page 1278 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Design and Fabrication of High Density Uranium Dispersion Fuels by :
Download or read book Design and Fabrication of High Density Uranium Dispersion Fuels written by and published by . This book was released on 1997 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: Twelve different uranium alloys and compounds with uranium densities greater than 13.8 g/cc were fabricated into fuel plates. Sixty-four experimental fuel plates, referred to as microplates, with overall dimensions of 76.2 mm x 22.2 mm x 1.3 mm and elliptical fuel zone of nominal dimensions of 51 mm x 9.5 mm, began irradiation in the Advanced Test Reactor on August 23, 1997. The fuel test matrix consists of machined or comminuted (compositions are in weight%) U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6 Ru, 10Mo-0.05Sn, U2Mo and U3Si2(as a control). The low enriched (235U
Book Synopsis Fabrication of U3O8 - Aluminum Dispersion Fuel Elements by Extrusion by : LC. Hymes
Download or read book Fabrication of U3O8 - Aluminum Dispersion Fuel Elements by Extrusion written by LC. Hymes and published by . This book was released on 1960 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: A description is given of the coextrusion of aluminum-canned U3O8- aluminum powder mixtures to produce dispersion type fuel elements. The effects of powder preparation, extrusion billet can composition, extrusion ratio, extrusion temperatures, and tri-uranium octi-oxide (U3O8) content of the powder mixture are illustrated. Use of density measurements of individual elements to determine their uranium-235 content is described.
Book Synopsis Development Status of Metallic, Dispersion and Non-Oxide Advanced and Alternative Fuels for Power and Research Reactors by : International Atomic Energy Agency
Download or read book Development Status of Metallic, Dispersion and Non-Oxide Advanced and Alternative Fuels for Power and Research Reactors written by International Atomic Energy Agency and published by . This book was released on 2003 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt: Summarises knowledge accumulated in fuel research since the beginning of the 1960s. This publication concentrates on the "advanced fuels" for the current different types of reactors, including metallic, carbide and nitride fuels for fast reactors, so-called ""cold"" fuels and fuels to burn excess ex-weapons plutonium in thermal power reactors.
Book Synopsis FABRICATION OF DISPERSED URANIUM FUEL ELEMENTS USING POWDER-METALLURGY TECHNIQUES. by :
Download or read book FABRICATION OF DISPERSED URANIUM FUEL ELEMENTS USING POWDER-METALLURGY TECHNIQUES. written by and published by . This book was released on 1957 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Fabrication techniques for producing dispersion fuel elements with cores of 30 volume per cent of UC, U2Ti, U3Si, or U6Ni dispersed in Zircaloy 2 and 30 volume per cent of UC or UN dispersed in Type 18-8 stainless steel have been investigated. Roll-clad plate-type elements of all these compositions may be fabricated by powdermetallurgy methods in such a manner that good core-tocladding bonds and cores with uniform dispersions of discrete uranium- compound particles are obtained. From the standpoint of fabricability, elements containing UC in Zircaloy 2, UC in stainless steel, and UN in stainless steel are the most promising. The UN in stainless steel has the best corrosion resistance in 680 deg F degassed water; however, UC in stainless steel has the best resistance to corrosion in 700 deg F NaK. (auth).
Download or read book Energy Research Abstracts written by and published by . This book was released on 1991 with total page 678 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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