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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Irradiation Testing of Intermediate and High-density U-Mo Alloy Dispersion Fuels to High Burnup written by and published by . This book was released on 2000 with total page 5 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
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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Prototypic Irradiation Testing of High-density U-Mo Alloy Dispersion Fuels written by and published by . This book was released on 2001 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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ISBN 13 :
Total Pages : 9 pages
Book Rating : 4.:/5 (684 download)
Download or read book Irradiation Testing of High Density Uranium Alloy Dispersion Fuels written by and published by . This book was released on 1997 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U2Mo, or U3Si2. These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.
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ISBN 13 :
Total Pages : 11 pages
Book Rating : 4.:/5 (683 download)
Download or read book Postirradiation Examination of High-density Uranium Alloy Dispersion Fuels written by and published by . This book was released on 1998 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt: Two irradiation test vehicles, designated RERTR-1 and RERTR-2, were inserted into the Advanced Test Reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn; the intermetallic compounds U2Mo and U3Si2 were also included in the fuel test matrix. These fuels are included in the experiments as ''microplates'' (76 mm x 22 mm x 1.3 mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature ((approximately) 100 C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively, at calculated peak fuel burnups of 45 and 71 at.%-U235. Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments.
Download or read book IRRADIATION OF U-Mo BASE ALLOYS. written by and published by . This book was released on 1964 with total page 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 1013 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 alloy were heat treated to provide a fine dispersion of second phase. (auth).
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ISBN 13 :
Total Pages : 16 pages
Book Rating : 4.:/5 (684 download)
Download or read book Selection and Microstructures of High Density Uranium Alloys written by and published by . This book was released on 1997 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: Twelve uranium alloys have been selected for incorporation into very high density aluminum dispersion fuel plates for irradiation testing. These alloys are (nominally) U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-2Mo-1Nb-1Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-6Mo-0.6Si, and U-10Mo-0.05Sn. The rationale for selection of these fuels based on gamma phase stability, reports of good irradiation performance, and high uranium density will be discussed. The microstructures of these fuels were examined by SEM/EDS and XRD at three storage during the powder fabrication process. Microstructures of selected alloys are discussed.
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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Observation on the Irradiation Behavior of U-Mo Alloy Dispersion Fuel written by and published by . This book was released on 2000 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels.
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ISBN 13 :
Total Pages : 10 pages
Book Rating : 4.:/5 (683 download)
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
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ISBN 13 :
Total Pages : 7 pages
Book Rating : 4.:/5 (727 download)
Download or read book Predicted Irradiation Behavior of U Sub 3 O Sub 8 -Al Dispersion Fuels for Production Reactor Applications written by and published by . This book was released on 1990 with total page 7 pages. Available in PDF, EPUB and Kindle. Book excerpt: Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U3O--Al dispersion fuels. The U3O--Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U3O8--Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U3O8--Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U3O8--Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U3O8--Al performance over a wide range of irradiation conditions. 8 refs., 8 figs., 1 tab.
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
Download or read book Irradiation Performance of U-Mo Alloy Based 'Monolithic' Plate-Type Fuel - Design Selection written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.
Author : IAEA
Publisher : International Atomic Energy Agency
ISBN 13 : 9201205201
Total Pages : 118 pages
Book Rating : 4.2/5 (12 download)
Download or read book Impact of Fuel Density on Performance and Economy of Research Reactors written by IAEA and published by International Atomic Energy Agency. This book was released on 2021-04-22 with total page 118 pages. Available in PDF, EPUB and Kindle. Book excerpt: Research reactor fuel technology continues to evolve, driven in part by international efforts to develop high density fuels to enable the conversion of more reactors from highly enriched uranium (HEU) to low enriched uranium (LEU) fuels. These high density fuels may offer economic benefits for research reactors, despite being more expensive initially, because they offer the prospect of higher per-assembly burnup, thus reducing the number of assemblies that must be procured, and more flexibility in terms of spent fuel management compared to the currently qualified and commercially available LEU silicide fuels. Additionally, these new fuels may offer better performance characteristics. This publication provides a preliminary evaluation of the impacts on research reactor performance and fuel costs from using high density fuel. Several case studies are presented and compared to illustrate these impacts.
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ISBN 13 :
Total Pages : 81 pages
Book Rating : 4.:/5 (251 download)
Download or read book Materialy Postojannoj stratigrafičeskoj Komissi po karbonu SSSR. written by and published by . This book was released on 1971 with total page 81 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : S. A. Rabin
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ISBN 13 :
Total Pages : 56 pages
Book Rating : 4.3/5 (91 download)
Download or read book Fabrication Development of U-Mo-UO2 and U-Mo-UC Dispersion Fuels for the Enrico Fermi Fast-breeder Reactor written by S. A. Rabin and published by . This book was released on 1963 with total page 56 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Irradiation Testing of Actinide Transmutation Fuels in the Advanced Test Reactor written by and published by . This book was released on 2001 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: The first irradiation experiment to evaluate the technical feasibility of proposed acitnide transmutation fuels for the US. Accelerator Transmutation of Waste program is currently under design. The goal of this irradiation experiment is to obtain initial irradiation performance data on candidate transmutation fuel concepts. The candidate fuels include non-fertile variations of (1) metallic alloys, (2) nitrides, (3) oxides, and (4) metal-matrix dispersion fuels. These fuels will be irradiated in the form of rodlets in the Advanced Test Reactor in Idaho beginning in September 2002. it is expected that postirradiation examinations will be performed on these fuels at the (almost equal to) 7 and 20 at.-% burnup levels. This paper presents the design of the irradiation test vehicle and the fuel rodlets; the test matrix of fuel variations, the target test conditions; and the planned postirradiation examinations.
