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Author :
Publisher :
ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Irradiation Behavior of Uranium-molybdenum Dispersion Fuel written by and published by . This book was released on 2001 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : D. I. Sinizer
Publisher :
ISBN 13 :
Total Pages : 52 pages
Book Rating : 4.3/5 (91 download)
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:
Author : M. P. Johnson
Publisher :
ISBN 13 :
Total Pages : 38 pages
Book Rating : 4.3/5 (91 download)
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
Author : J. H. Kittel
Publisher :
ISBN 13 :
Total Pages : 0 pages
Book Rating : 4.:/5 (865 download)
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.
Author : International Atomic Energy Agency
Publisher :
ISBN 13 : 9789201157201
Total Pages : 144 pages
Book Rating : 4.1/5 (572 download)
Download or read book Material Properties of Unirradiated Uranium-Molybdenum (U-Mo) Fuel for Research Reactors written by International Atomic Energy Agency and published by . This book was released on 2020-10-12 with total page 144 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication presents the material properties of all unirradiated Uranium-Molybdenum (U-Mo) fuel constituents that are essential for fuel designers and reactor operators to evaluate the fuel's performance and safety for research reactors. Many significant advances in the understanding and development of low enriched uranium U-Mo fuels have been made since 2004, stimulated by the need to understand irradiation behavior and early fuel failures during testing. The publication presents a comprehensive overview of mechanical and physical property data from U-Mo fuel research
Author : Gordon G. Bentle
Publisher :
ISBN 13 :
Total Pages : 76 pages
Book Rating : 4.3/5 (91 download)
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:
Download or read book A Physical Description of Fission Product Behavior Fuels for Advanced Power Reactors written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO2 power reactor fuel and the Dispersion Analysis Research Tool (DART) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the [alpha]-, intermediate- and [gamma]-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.
Author : Adam B. Robinson
Publisher :
ISBN 13 :
Total Pages : 118 pages
Book Rating : 4.:/5 (835 download)
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 IRRADIATION STUDIES OF URANIUM-10 W/o MOLYBDENUM FUEL ALLOY. written by and published by . This book was released on 1961 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Bare and zirconium-clad uranium-10 wt% molybdenum specimens were irradiated in NaK-filled capsules in the MTR. Irradiation conditions varied to include central-core temperatures ranging from 300 to over 1200 deg F, fuel burnups ranging from 0.36 to over 3.0 total at.% and fission rates in the range of 0.35 to 1.9 x 10/sup 14/ fissions/(sec)(cm/sup 3/) of alloy. Other parameters studied included the effects of heat treatment, changes in composition, different fabrication techniques, and changes in cladding thickness on the behavior of the fuel alloy. The objective of the irradiations was to determine the behavior of the fuel alloy under conditions approaching as closely as possible those to be encountered in the Enrico Fermi Atomic Power Plant as they were known at the time. The results indicated that the volume of the fuel alloy would increase conservatively at a rate of about 3.0% per at.% burnup as long as the critical temperature of 1000 to 1100 deg F was not exceeded and the gamma phase of the alloy did not transform during irradiation. If the critical temperature was exceeded, the alloy swelled until rupture or complete disintegration occurred. The occurrence of transformation during irradiation was noted at burnups in the range of 2.5 total at.% at fission rates of 1.5 x 10/sup 14/ to 1.9 x 10/sup 14/ fissions/(sec)(cm/sup 3/) and temperatures of 800 to 1000 deg F. The alloy was normally maintained in the gamma phase during irradiation, even at temperatures below 1100 deg F and at fission rates in the range of 0.7 x 10/sup 14/ fissions per sec. Transformation during irradiation was accompanied by excessive swelling of the alloy. (auth).
Author : R. D. Leggett
Publisher :
ISBN 13 :
Total Pages : 66 pages
Book Rating : 4.:/5 (319 download)
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:
Author : J. A. Horak
Publisher :
ISBN 13 :
Total Pages : 46 pages
Book Rating : 4.3/5 (91 download)
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.
Author : National Research Council
Publisher : National Academies Press
ISBN 13 : 0309130395
Total Pages : 220 pages
Book Rating : 4.3/5 (91 download)
Download or read book Medical Isotope Production Without Highly Enriched Uranium written by National Research Council and published by National Academies Press. This book was released on 2009-06-27 with total page 220 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book is the product of a congressionally mandated study to examine the feasibility of eliminating the use of highly enriched uranium (HEU2) in reactor fuel, reactor targets, and medical isotope production facilities. The book focuses primarily on the use of HEU for the production of the medical isotope molybdenum-99 (Mo-99), whose decay product, technetium-99m3 (Tc-99m), is used in the majority of medical diagnostic imaging procedures in the United States, and secondarily on the use of HEU for research and test reactor fuel. The supply of Mo-99 in the U.S. is likely to be unreliable until newer production sources come online. The reliability of the current supply system is an important medical isotope concern; this book concludes that achieving a cost difference of less than 10 percent in facilities that will need to convert from HEU- to LEU-based Mo-99 production is much less important than is reliability of supply.
Author : Rehab Abdel Rahman
Publisher : BoD – Books on Demand
ISBN 13 : 9535124471
Total Pages : 174 pages
Book Rating : 4.5/5 (351 download)
Download or read book Nuclear Material Performance written by Rehab Abdel Rahman and published by BoD – Books on Demand. This book was released on 2016-06-29 with total page 174 pages. Available in PDF, EPUB and Kindle. Book excerpt: Assessing and improving nuclear material performance is a crucial subject for the sustainability of the nuclear energy and radioactive isotope supplies. This book aims to present research efforts used to identify nuclear materials performances in different areas. The contributions of esteemed international experts have covered important research aspects in fission and fusion technologies and naturally occurring radioactive materials management. The authors introduced current and anticipated trends toward better performances and mitigating challenges for commercial application of innovative technologies, biological remediation of mine effluents, nuclear fuel performance in power and research fission reactors, gamma ray spectrometer calibration, and recent advances in understanding the performance of tungsten composite in fusion reactor environment.
Author : J. A. Horak
Publisher :
ISBN 13 :
Total Pages : 40 pages
Book Rating : 4.3/5 (91 download)
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.
Author : Brianna Coulson
Publisher :
ISBN 13 :
Total Pages : 76 pages
Book Rating : 4.:/5 (847 download)
Download or read book Two- and Three-dimensional Thermal Analyses of Uranium/molybdenum Dispersion Fuel Microstructures written by Brianna Coulson and published by . This book was released on 2012 with total page 76 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :
Publisher :
ISBN 13 :
Total Pages : 9 pages
Book Rating : 4.:/5 (683 download)
Download or read book Progress in Developing Very-high-density Low-enriched-uranium Fuels written by and published by . This book was released on 1999 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-MO alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm3. Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 C in 8-g U/cm3 fuel.
Author : Yue Jin
Publisher : Frontiers Media SA
ISBN 13 : 283250518X
Total Pages : 153 pages
Book Rating : 4.8/5 (325 download)
Download or read book Numerical Methods and Computational Sciences Applied to Nuclear Energy written by Yue Jin and published by Frontiers Media SA. This book was released on 2022-11-11 with total page 153 pages. Available in PDF, EPUB and Kindle. Book excerpt:
