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Irradiation Behavior Of Uranium Fissium Alloys Ebr Ii Project
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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.
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 The Effects of Irradiation on Uranium-plutonium-fissium Fuel Alloys by : J. A. Horak
Download or read book The Effects of Irradiation on Uranium-plutonium-fissium Fuel Alloys written by J. A. Horak and published by . This book was released on 1962 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt: A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fuel-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. It was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in zirconium or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pour-cast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated.
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 THE EFFECTS OF IRRADIATION ON URANIUM-PLUTONIUM-FISSIUM FUEL ALLOYS. Final Report on Metallurgy Program 6.5.5 by :
Download or read book THE EFFECTS OF IRRADIATION ON URANIUM-PLUTONIUM-FISSIUM FUEL ALLOYS. Final Report on Metallurgy Program 6.5.5 written by and published by . This book was released on 1962 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fue1-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. lt was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in Zr or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pourcast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated. (auth).
Book Synopsis Irradiation Swelling of Uranium and Uranium Alloys by : Gordon G. Bentle
Download or read book Irradiation Swelling of Uranium and Uranium Alloys written by Gordon G. Bentle and published by . This book was released on 1961 with total page 76 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Irradiation Behavior of Uranium Carbide Fuels by : D. I. Sinizer
Download or read book Irradiation Behavior of Uranium Carbide Fuels written by D. I. Sinizer and published by . This book was released on 1962 with total page 52 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 High Purity Uranium by : R. D. Leggett
Download or read book Irradiation Behavior of High Purity Uranium written by R. D. Leggett and published by . This book was released on 1963 with total page 66 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Small-scale Demonstration of the Melt Refining of Highly Irradiated Uranium-fissium Alloy by : V. G. Trice
Download or read book Small-scale Demonstration of the Melt Refining of Highly Irradiated Uranium-fissium Alloy written by V. G. Trice and published by . This book was released on 1963 with total page 50 pages. Available in PDF, EPUB and Kindle. Book excerpt: The behavior of fission products was consistent with the earlier results. Fission product removals were over 99 per cent for krypton, xenon, iodine, cesium, barium, and strontium, over 95 percent for yttrium, rare earths, and tellurium, and zero for the noble metals.
Book Synopsis Irradiation of Uranium-fissium Alloys and Related Compositions by : K. F. Smith
Download or read book Irradiation of Uranium-fissium Alloys and Related Compositions written by K. F. Smith and published by . This book was released on 1957 with total page 62 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis IRRADIATION OF URANIUM-FISSIUM ALLOYS AND RELATED COMPOSITIONS. Work Performed by :
Download or read book IRRADIATION OF URANIUM-FISSIUM ALLOYS AND RELATED COMPOSITIONS. Work Performed written by and published by . This book was released on 1957 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation results in the range near 1/2% burnup and 500 to 600 deg C are presented for U-fissium amd U-Mo alloys. Under these conditions both classes of alloys show quite low growth coefficients and volume increases, with a few exceptions. Water queaching either alloy from 850 deg C is shown to be unsatisfactory. The effect of an axial hole for relief of fission gases appears to be inconclusive. Surface condition of irradiated U-base alloys appears to be not quite as good as that for U-20 wt.% Pu base alloys. (auth).
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1975 with total page 764 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis The Effects of Irradiation on the Tensile Properties of Uranium by : R. E. Hueschen
Download or read book The Effects of Irradiation on the Tensile Properties of Uranium written by R. E. Hueschen and published by . This book was released on 1955 with total page 60 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis PREPARATION OF ALLOY FOR FIRST CORE LOADING OF EBR-II. by :
Download or read book PREPARATION OF ALLOY FOR FIRST CORE LOADING OF EBR-II. written by and published by . This book was released on 1961 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The alloy used for the fabrication of the fuel pins for the first core loading of the second Experimental Breeder Reactor (EBR-II) was prepared in the prototype equipment developed for the melt-refining processing of the irradiated EBR-II fuel. One hundred and twenty-five 10-kg ingots were made, of which 40 were unenriched uranium-fissium alloy and 85 were enriched uranium-fissium alloy. In addition, nineteen 10-kg batches of unenriched uranium-fissium scrap and forty- seven 10-kg batches of enriched uraniumfissium alloy scrap were melted for consolidation into ingots. The average yield for the alloy preparation runs was 96.5% and for the scrap remelt runs was 93%. The chemical and isotopic compositions of the ingots produced were all within specifications (95 plus or minus 1.0 wt% uranium, of which 48.1 plus or minus 1.2 wt% is U/sup 235/). (auth).
Book Synopsis Irradiation of U-Mo Base Alloys by : M. P. Johnson
Download or read book Irradiation of U-Mo Base Alloys written by M. P. Johnson and published by . This book was released on 1964 with total page 38 pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of experiments was designed to assess the suitability of uranium-molybdenum alloys as high-temperature, high-burnup fuels for advanced sodium cooled reactors. Specimens with molybdenum contents between 3 and 10% were subjected to capsule irradiation tests in the Materials Testing Reactor, to burnups up to 10,000 Mwd/MTU at temperatures between 800 and 1500 deg F. The results indicated that molybdenum has a considerable effect in reducing the swelling due to irradiation. For example. 3% molybdemum reduces the swelling from 25%, for pure uranium. to 7% at approximates 3,000 Mwd/MTU at 1270 deg F. Further swelling resistance can be gained by increasing the molybdenum content, but the amount gained becomes successively smaller. At higher irradiation levels, the amount of swelling rapidly becomes greater, and larger amounts of molybdenum are required to provide similar resistance. A limit of 7% swelling, at 900 deg F and an irradiation of 7,230 Mwd/ MTU, requires the use of 10% Nonemolybdenum in the alloy. The burnup rates were in the range of 2.0 to 4.0 x 10p13s fissiom/cc-sec. Small ternary additions of silicon and aluminum were shown to have a noticeable effect in reducing swelling when added to a U-3% Mo alloy base. Under the conditions of the present experiment, 0.26% silicon or 0.38% aluminum were equivalent to 1 to 1 1/2% molybdenum. The Advanced Sodium Cooled Reactor requires a fuel capable of being irradiated to 20,000 Mwd/MTU at temperatures up to 1500 deg C in metal fuel, or equivalent in ceramic fuel. It is concluded that even the highest molybdenum contents considered did not produce a fuel capable of operating satisfactorily under these conditions. The alloys would be useful, however, for less exacting conditions. The U-3% Mo alloy is capable of use up to 3,000 Mwd/MTU at temperatures of 1300 deg F before swelling becomes excessive. The addition of silicon and aluminum would increase this limit to at least 3,000 Mwd/MTU, and possibly more if the
