Postirradiation Examination of High-U-loaded Low-enriched U3O, UAl2, and U3Si Test Fuel Plates
Postirradiation Examination of High-U-loaded Low-enriched U3O, UAl2, and U3Si Test Fuel Plates

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Published: 1984

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The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U3O, UAl2 and U3Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U3Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup.

Post-irradiation Examination of U3Si{sub 1.6}-Al Dispersion Fuel Element LC04
Post-irradiation Examination of U3Si{sub 1.6}-Al Dispersion Fuel Element LC04

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Published: 1993

Total Pages: 16

ISBN-13:

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A low-enrichment uranium (LEU) irradiation test (SIMONE project) was performed in the 45 MW Petten High-flux Reactor (HFR) as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. An LEU U3Si{sub 1.6}-Al dispersion fuel element with a fuel loading of 5.5 g cm−3 was irradiated and performed well to a total burnup of 66% and a peak burnup of 85%. The swelling behavior of the fuel was similar to that observed in previous U3Si(subscript x) irradiation experiments. The resulting fuel plate thickness increases and associated changes in coolant channel cross-sections were well within acceptable limits.

Post-irradiation Examination Techniques for Research Reactor Fuels
Post-irradiation Examination Techniques for Research Reactor Fuels

Author: IAEA

Publisher: International Atomic Energy Agency

Published: 2023-04-18

Total Pages: 149

ISBN-13: 920102021X

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Post-irradiation examination (PIE) is an indispensable step in the selection of new or improved research reactor fuel, and in the characterization and understanding of its in-core behaviour. This publication provides an introduction to PIE techniques. It describes a typical PIE process from intercycle inspections in the reactor pool or channel, to hot cell PIE, which is subdivided into non-destructive and destructive testing techniques with their typical output, advantages and drawbacks, and their applicability to understanding fuel irradiation behaviour. Much of the work presented in this publication originated from the research and development of new low enriched uranium research reactor fuels. Intended readers include research reactor operators, regulators and their technical support organizations, fuel developers and manufacturers, laboratory staff, and policy makers.

Postirradiation Examination of a Low Enriched U3Si2-Al Fuel Element Manufactured and Irradiated at Batan, Indonesia
Postirradiation Examination of a Low Enriched U3Si2-Al Fuel Element Manufactured and Irradiated at Batan, Indonesia

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Published: 1994

Total Pages: 9

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The first low-enriched U3Si2-Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U3Si2-Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 [mu]m (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 [mu]m) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat.

Postirradiation Analysis of Experimental Uranium-silicide Dispersion Fuel Plates
Postirradiation Analysis of Experimental Uranium-silicide Dispersion Fuel Plates

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Published: 1985

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Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235U. Fuel plates containing 33 v/o U3Si and U3Si2 behaved very well up to this burnup. Plates containing 33 v/o U3Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U3Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U3Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs.

Postirradiation Examination of High-density Uranium Alloy Dispersion Fuels
Postirradiation Examination of High-density Uranium Alloy Dispersion Fuels

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Published: 1998

Total Pages: 11

ISBN-13:

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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.

Fuel Plate Failure Experiments and Analyses in Irradiated U-10Mo Alloy
Fuel Plate Failure Experiments and Analyses in Irradiated U-10Mo Alloy

Author: Francine Joyce Rice

Publisher:

Published: 2017

Total Pages: 0

ISBN-13:

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The Materials Management and Minimization (M3) Program intends to qualify a new high-density low-enriched-uranium (LEU) U--Mo monolithic fuel to enable conversion of six US high-performance research reactors (USHPRRs). This thesis presents the preliminary results and discussions related to post-irradiation blister anneal studies and fission product release scoping studies performed on U--Mo monolithic fuel plates. Blister anneal testing on irradiated fuel plates is a temperature-resolved failure-threshold measurement technique historically used to assess fuel plate stability under off-normal operating conditions. The effects of fuel composition, geometry, fission density, and irradiation conditions are presented herein as parameters that were investigated for their impact on blister-threshold temperatures. The fission-product-transport scoping study successfully characterized the release, transport and temperature-resolved deposition behavior of iodine and cesium. Two failure temperatures were evaluated: 600 and 1250°C. Testing was performed in the main hot cell at the Materials and Fuels Complex located at Idaho National Laboratory.

The Use of U3Si2 Dispersed in Aluminum in Plate-type Fuel Elements for Research and Test Reactors
The Use of U3Si2 Dispersed in Aluminum in Plate-type Fuel Elements for Research and Test Reactors

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Publisher:

Published: 1987

Total Pages:

ISBN-13:

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A high-density fuel based on U3Si2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U3Si2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U3Si2 particle swelling rate is approximately the same as that of the commonly used UAl(subscript x) fuel particle. The presence of minor amounts of U3Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U3Si2-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

Comprehensive Nuclear Materials
Comprehensive Nuclear Materials

Author: Todd R Allen

Publisher: Elsevier

Published: 2011-05-12

Total Pages: 3552

ISBN-13: 0080560334

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Comprehensive Nuclear Materials, Five Volume Set discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerators, and for diverse functions in fuels, cladding, moderator and control materials, structural, functional, and waste materials. The work addresses the full panorama of contemporary international research in nuclear materials, from Actinides to Zirconium alloys, from the worlds' leading scientists and engineers. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environment Fully integrated with F-elements.net, a proprietary database containing useful cross-referenced property data on the lanthanides and actinides Details contemporary developments in numerical simulation, modelling, experimentation, and computational analysis, for effective implementation in labs and plants