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Irradiation Behavior Of Restrained And Vented Uranium 2 W O Zirconium Alloy
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Book Synopsis Irradiation Behavior of Restrained and Vented Uranium-2 W/o Zirconium Alloy by : J. A. Horak
Download or read book Irradiation Behavior of Restrained and Vented Uranium-2 W/o Zirconium Alloy written by J. A. Horak and published by . This book was released on 1962 with total page 46 pages. Available in PDF, EPUB and Kindle. Book excerpt: Twelve 0.22-in.-diameter fuel specimens containing a longitudinal central vent and clad with 0.010 in. of Type 304 stainless steel were irradiated to evaluate the effect of restraint and a central vent on fuel element stability. The cladding of 10 of the specimens contained porous end plugs to vent any released fission gas and thus to minimize the buildup of gas pressure within the stainless steel cladding. The specimens consisted of a 20% enriched uranium--2 wt% zirconium alloy core surrounded by a natural uranium--2 wt% zirconium alloy sleeve. Eight of the specimens were irradiated to burnups of the enriched core of 6.9 to 12.8% of all atoms (1.2 to 2.2 at.% of the duplex assembly) at maximum fuel temperatures ranging from 280 to 760 deg C. Most of the clad specimens exhibited negligible volume increases as a result of irradiation. Two specimens containing central vents but unclad were irradiated together with the clad specimens in an attempt to differentiate between the effects due to a central vent and the effects due to cladding. The central vent in itself did not appear to reduce the swelling characteristics of the alloy. Mechanical restraint appeared to have extended the useful operating temperatures of the metallic fuel alloy by at least 200 deg C and also greatly extended the burnup levels to which the fuel could be irradiated.
Book Synopsis Irradiation Behavior of Restrained and Vented Uranium-2 W/o Zirconium Alloy. Final Report-Programs 6.1.22 and 6.1.27 by :
Download or read book Irradiation Behavior of Restrained and Vented Uranium-2 W/o Zirconium Alloy. Final Report-Programs 6.1.22 and 6.1.27 written by and published by . This book was released on 1962 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Twelve 0.22-in.-diameter fuel specimens containing a longitudinal central vent and clad with 0.010 in. of Type 304 stainless steel were irradiated to evaluate the effect of restraint and a central vent on fuel element stability. The cladding of 10 of the specimens contained porous end plugs to vent any released fission gas and thus to minimize the buildup of gas pressure within the stainless steel cladding. The specimens consisted of a 20% enriched uranium--2 wt% zirconium alloy core surrounded by a natural uranium--2 wt% zirconium alloy sleeve. Eight of the specimens were irradiated to burnups of the enriched core of 6.9 to 12.8% of all atoms (1.2 to 2.2 at.% of the duplex assembly) at maximum fuel temperatures ranging from 280 to 760 deg C. Most of the clad specimens exhibited negligible volume increases as a result of irradiation. Two specimens containing central vents but unclad were irradiated together with the clad specimens in an attempt to differentiate between the effects due to a central vent and the effects due to cladding. The central vent in itself did not appear to reduce the swelling characteristics of the alloy. Mechanical restraint appeared to have extended the useful operating temperatures of the metallic fuel alloy by at least 200 deg C and also greatly extended the burnup levels to which the fuel could be irradiated. (auth).
Book Synopsis Irradiation Behavior of Uranium-fissium Alloys. EBR-II Project by : J. H. Kittel
Download or read book Irradiation Behavior of Uranium-fissium Alloys. EBR-II Project written by J. H. Kittel and published by . This book was released on 1971 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of uranium-fissium and uranium-fissium-zirconium alloys was irradiated in thermal test reactors to study the relationship of dimensional stability to alloy composition, thermal cycling, burnup, irradiation temperature, post-irradiation heating, and cladding restraint. None of the alloy compositions tested showed irradiation behavior superior to the uranium-5 wt./% fissium alloy that has been used as driver fuel in EBR-II since it began operation. This alloy is among those uranium-base alloys most capable of resisting high-temperature irradiation swelling. None of the alloys showed evidence of the reversion to the metastable gamma phase that has been observed in comparable uranium-molybdenum alloys. Swelling of uranium-fissium alloys was effectively restrained by most of the 0.009-inch thick cladding materials investigated. Local hydrostatic forces due to swelling of the fuel caused the fuel to extrude extensively out of small vent holes in the cladding. Little axial fuel movement occurred within the cladding, however, even when the upper fuel surface was entirely unrestrained.
Book Synopsis Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III by : J. H. Kittel
Download or read book Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III written by J. H. Kittel and published by . This book was released on 1959 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt: The fuel material specified for the Mark III core of EBR-I was uranium-2 wt. % zirconium alloy coextruded with Zircaloy-2 cladding. From previous work on swaged or rolled uranium-2 wt% zirconium alloy, it was anticipated that the extruded alloy would be dimensionally unstable under irradiation unless stabilized by suitable heat treatment. In order to determine an effective heat treatment, irradiation studies were made on both clad and unclad extruded uranium-2 wt.% zirconium alloy specimens at irradiation temperature estimated at 200 to 750 deg C. The irradiation specimens included material with three different heat treatments, selected on the basis of previous studies, and material transient melted in its cladding. For unclad specimens, it was found that the irradiation temperature strongly influenced the various irradiation growth rates resulting from different heat treatments. Growth rates of the clad specimens were relatively insensitive to either irradiation temperature or prior heat treatment. An exception was the transient-melted material, which shortened under irradiation. The cladding had only limited ability to restrain the swelling rates of specimens irradiated at the more elevated temperatures. Clad transient-melted material was found to be most resistant to high-temperature swelling under irradiation. The results of the present study combined with observations in earlier investigations resulted in a recommendation that the reference heat treatment for the core consist of gamma solution at 800 deg C followed by isothermal transformation at 690 deg C.
Book Synopsis IRRADIATION BEHAVIOR OF ZIRCALOY-2 CLAD THORIUM--URANIUM--ZIRCONIUM FUEL ELEMENTS. Interim Report by :
Download or read book IRRADIATION BEHAVIOR OF ZIRCALOY-2 CLAD THORIUM--URANIUM--ZIRCONIUM FUEL ELEMENTS. Interim Report written by and published by . This book was released on 1967 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :United States. Department of Commerce. Office of Technical Services Publisher : ISBN 13 : Total Pages :860 pages Book Rating :4.E/5 ( download)
Book Synopsis Keywords Index to U.S. Government Technical Reports (permuted Title Index). by : United States. Department of Commerce. Office of Technical Services
Download or read book Keywords Index to U.S. Government Technical Reports (permuted Title Index). written by United States. Department of Commerce. Office of Technical Services and published by . This book was released on 1962 with total page 860 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Irradiation Behavior of Uranium-fissium Alloys. EBR-II Project by : J. H. Kittel
Download or read book Irradiation Behavior of Uranium-fissium Alloys. EBR-II Project written by J. H. Kittel and published by . This book was released on 1971 with total page 46 pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of uranium-fissium and uranium-fissium-zirconium alloys was irradiated in thermal test reactors to study the relationship of dimensional stability to alloy composition, thermal cycling, burnup, irradiation temperature, post-irradiation heating, and cladding restraint. None of the alloy compositions tested showed irradiation behavior superior to the uranium-5 wt./% fissium alloy that has been used as driver fuel in EBR-II since it began operation. This alloy is among those uranium-base alloys most capable of resisting high-temperature irradiation swelling. None of the alloys showed evidence of the reversion to the metastable gamma phase that has been observed in comparable uranium-molybdenum alloys. Swelling of uranium-fissium alloys was effectively restrained by most of the 0.009-inch thick cladding materials investigated. Local hydrostatic forces due to swelling of the fuel caused the fuel to extrude extensively out of small vent holes in the cladding. Little axial fuel movement occurred within the cladding, however, even when the upper fuel surface was entirely unrestrained.
Author :United States. Department of Commerce. Office of Technical Services Publisher : ISBN 13 : Total Pages :990 pages Book Rating :4.3/5 (91 download)
Book Synopsis Keywords Index to U.S. Government Technical Reports by : United States. Department of Commerce. Office of Technical Services
Download or read book Keywords Index to U.S. Government Technical Reports written by United States. Department of Commerce. Office of Technical Services and published by . This book was released on 1962 with total page 990 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis "Pin-cushion" Irradiation Tests of Uranium and Its Zirconium Alloys by : S. H. Paine
Download or read book "Pin-cushion" Irradiation Tests of Uranium and Its Zirconium Alloys written by S. H. Paine and published by . This book was released on 1956 with total page 41 pages. Available in PDF, EPUB and Kindle. Book excerpt: An irradiation test of 1/16 inch diameter x 3/16 inch long U-235/ and U-235/-Zr alloy specimens, in the range 2 to 20 wt.% Zr, is described. The small pins were mounted in heat transfer blocks, or ''cushions,'' in such a way that surface roughening and length changes could be observed, recorded, and compared for five stabilizing heat treatments up to exposure levels of approximately 1% total atom burnup. Addition of 2% Zr was found to be very beneficial, but the behavior of the alloys underwent distinct changes in going from 2 to 4 wt.% Zr. Material, both as-cast and subsequently isothermally treated, changed from minus to plus elongation, whereas wrought materials, either gamma-slow cooled or isothermally treated, changed from plus to minus behavior. Data-quenched wrought material varied unpredictably. Addition of Zr up to 20% showed further improvement in stability for all treatments. Factors which may have contributed to this behavior are discussed.
Book Synopsis Influence of Heat Treatment on Irradiation-Induced Dimensional Changes in Some Uranium-Zirconium Alloys by : J. H. Kittel
Download or read book Influence of Heat Treatment on Irradiation-Induced Dimensional Changes in Some Uranium-Zirconium Alloys written by J. H. Kittel and published by . This book was released on 1957 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Specimens made of alloys containing 1, 2, and 3 weight percentage of zirconium in uranium were irradiated under conditions where the specimens were free from physical restraint. The specimens studied were from cast, wrought, and wrought and heat-treated materials. The heat treatments investigated included quenching from the gamma phase, quenching and tempering, isothermal transformations, and alpha phase annealing. All specimens were found to increase in length as a result of irradiation except a rolled and alpha-annealed group, which was observed to shorten. The rate of change of length with uranium atom burnup was found to depend strongly upon the prior metallurgical history of the specimens and upon their compositions.
Book Synopsis "PIN-CUSHION" IRRADIATION TESTS OF URANIUM AND ITS ZIRCONIUM ALLOYS. Metallurgy Program 6.1.1 by :
Download or read book "PIN-CUSHION" IRRADIATION TESTS OF URANIUM AND ITS ZIRCONIUM ALLOYS. Metallurgy Program 6.1.1 written by and published by . This book was released on 1956 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: ABS>An irradiation test of 1/16 in. diameter x 3/16 in. long U/sup 235/ and U/sup 235/-Zr alloy specimens, in the range 2 to 20 wt.% Zr, is described. The small pins were mounted in heat transfer blocks, or ''cushions, '' in such a way that suriace roughening and length changes could be observed, recorded, and compared for five stabilizing heat treatments up to exposure levels of approximately 1% total atom burnup. Addition of 2% Zr was found to be very beneficial, but the behavior of the alloys underwent distinct changes in going from 2 to 4 wt.% Zr. Material, both as-cast and subsequently isothermally treated, changed from minus to plus elongation, whereas wrought materials, either gamma-slow cooled or isothermally treated, changed from plus to minus behavior. Data-quenched wrought material varied unpredictably. Addition of Zr up to 20% showed further improvement in stability for all treatments. Factors which may have contributed to this behavior are discussed. (auth).
Book Synopsis PRELIMINARY STUDIES OF IRRADIATION DAMAGE TO URANIUM-ZIRCONIUM ALLOYS. by :
Download or read book PRELIMINARY STUDIES OF IRRADIATION DAMAGE TO URANIUM-ZIRCONIUM ALLOYS. written by and published by . This book was released on 1956 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis HIGH-TEMPERATURE IRRADIATION ON A ZIRCONIUM HYDRIDE-2 W/o URANIUM ALLOY. by :
Download or read book HIGH-TEMPERATURE IRRADIATION ON A ZIRCONIUM HYDRIDE-2 W/o URANIUM ALLOY. written by and published by . This book was released on 1959 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Enriched ZrH/sub 1.65/ -2 wt.% uranium specimens were irradiated under 1 atm. of hydrogen at center-line temperatures between 900 to 1400 deg F to uranium burnups of between 13 and 25 at.%. Specially designed irradiation capsules were used to provide the conditions of temperature and hydrogen atmosphere which were required. Each capsule was instrumented with five thermocouples so that ample temperature data could be obtained during the irradiation. Almost negligible density changes were produced in the material by the irradiation. Changes in length and diameter were of a degree which could fall within experimental error in measurement. Metallographic examination showed no change in microstructure which could be attributed to the effect of irradiation. (auth).
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1974 with total page 1436 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Uranium Dioxide written by J. Belle and published by . This book was released on 1961 with total page 744 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Advances in High Temperature Gas Cooled Reactor Fuel Technology by : International Atomic Energy Agency
Download or read book Advances in High Temperature Gas Cooled Reactor Fuel Technology written by International Atomic Energy Agency and published by . This book was released on 2012-06 with total page 639 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.
Book Synopsis The Safety of Nuclear Power Reactors (light Water-cooled) and Related Facilities by : U.S. Atomic Energy Commission
Download or read book The Safety of Nuclear Power Reactors (light Water-cooled) and Related Facilities written by U.S. Atomic Energy Commission and published by . This book was released on 1973 with total page 528 pages. Available in PDF, EPUB and Kindle. Book excerpt:
