Tem Characterization of Irradiated U3SI2/AL Dispersion Fuel (original) (raw)
Journal of Nuclear Materials, 2009
Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127 I beam up to an ion fluence of 2 Â 10 17 cm À2. Microscopy and mainly Xray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl 3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.
From arc-melted ingot to MTR fuel plate: A SEM/EBSD microstructural study of U3Si2
Journal of Nuclear Materials
U 3 Si 2 /Al fuel plates are widely used in Material Testing Reactors (MTRs) as Low Enriched Uranium driver fuel. In this paper, a reinvestigation of the microstructure of U 3 Si 2 particles is proposed to take full advantage of the new capabilities offered by Electron BackScattered Diffraction (EBSD) techniques. Using EBSD we demonstrate that most particles are single crystalline in as-fabricated plates. To understand this characteristic, linked to the microstructure of the starting material, an in-depth study of U 3 Si 2 ingots produced by arc-melting was performed at a laboratory scale; they were extensively characterized by EBSD, scanning electron microscope, energy dispersive spectroscopy and X-ray diffraction. It is shown that U 3 Si 2 grain may be large (up to several thousands of micrometers) and that they exhibit a strong preferential orientation, linked to the axial thermal gradient created in the arcmelting chamber. A significant impact of the cooling rate after arc-melting on the ingot microstructure is noticed: grains are smaller and more columnar when the cooling rate is high. A deviation from 3U/2Si stoichiometry caused for example by impurities induces the formation of a secondary phase, which exhibits a square spiral morphology. We then demonstrate that the cooling rate of U 3 Si 2 ingots has a direct influence on the characteristics of the powders obtained by crushing these ingots. Indeed the powder obtained from "slow" cooled ingots is found very close to the powder used for industrial MTR plates. On the contrary particles obtained from "fast" cooled ingots, are polycrystalline and more resistant to crushing. Thus, this work provides significant advances both in the characterisation of technological products like U 3 Si 2 /Al MTR plates and in more basic knowledge about the U 3 Si 2 phase formation.
Heavy ion irradiation of U–Mo/Al dispersion fuel
2006
The usage of high-density U-Mo/Al dispersion fuel for high burn up in research and test reactors seems to be limited by the unfavourable interdiffusion layer between the fuel and the Al-matrix, which develops during irradiation. This interdiffusion layer was observed up to now only after costly and time consuming in-pile irradiation and could not be created in out-of-pile experiments. This paper presents a new approach of creating such an interdiffusion layer out-of-pile by irradiation with heavy ions. An appropriate choice of heavy-ion irradiation simulates irradiation damage and deposition of fission fragments as it happens during in-pile irradiation and induces a diffusion process between the fuel and the Al matrix. An irradiation experiment and post-irradiation examinations are presented.
STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3
Urania (Tangerang), 2022
STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm 3. Uranium-silicide compound fuel dispersed in aluminium matrix (U3Si2-Al) have been used in a large number of research reactors around the world because of its excellent behavior under irradiation. This fuel also provides high uranium density with typical fuel loading up to 4.8 gU/cm 3 to compensate for the reduced fissile amount in LEU. To improve the density of current U3Si2-Al (2.96 gU/cm 3) used in Indonesian GA Siwabessy Multipurpose Research Reactor, U3Si2-Al dispersion fuel plate with density of 4.8 gU/cm 3 (U 235 ∼19.75%) had been irradiated in RSG GAS for 175 days at 15 MW power to burnup level of approximately 40%. The characterization was performed using SEM-EDS and optical microscope to study microstructure of the irradiatted fuel, largely the fission gas bubbles and the interaction layer between U3Si2 fuel and Al matrix. The average diameter of the bubbles with diameter from 0.06 to 0.55 µm was 0.21 µm. The interaction layer was identified as U(Al,Si)2,3 with thickness of approximately 1.5 µm. The relatively small fission gas bubbles and the interaction layer didn't cause swelling on the fuel and the overall performance of the fuel plate was very good.
Neutron Irradiation of U3Si5 and Al43Mo4U6 Compounds. First Results
Procedia Materials Science, 2015
In the framework of U(Mo) alloys qualification used as nuclear fuel for research reactors, post-irradiation experiments have shown that the porosity and swelling phenomena observed due to irradiation on dispersed fuel elements are associated to a poor behavior of the interaction layer grown by interdiffusion during fabrication and/or irradiation between U(Mo) particles and the matrix. Fission induced amorphization of some of the phases that form the interaction layer, has been proposed as one explanation to understand the failure. Out of pile diffusion couples experiments have shown that the phases UAl 3 , UAl 4 , Al 20 Mo 2 U and Al 43 Mo 4 U 6 form the interaction layer when pure Al is used as matrix while U(Al,Si) 3 , U 3 Si 5 , USi 2 , USi 2-x , Al 20 Mo 2 U and Al 43 Mo 4 U 6 are identified when Al(Si) is used instead of Al. The objective of this investigation is to study, independently, the behavior of each phase under neutron irradiation focusing on a possible amorphization. In this work, powders of U 3 Si 5 and Al 43 Mo 4 U 6 compounds were irradiated at RA1 reactor (Argentina). Samples were analyzed by optical microscopy, scanning electron microscopy, energy dispersive spectroscopy, wavelength dispersive spectroscopy and X-ray diffraction.
Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel
Journal of Nuclear Materials, 2014
U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding. 2. Examinations before thermal treatments 2.1. IRIS 4 fuel plate manufacturing and fresh fuel plate characterisation U-7 wt.% Mo particles (enrichment <20 wt.% 235 U) produced by centrifugal atomisation were selected [20,21]. They were
U-Si and U-Si-Al dispersion fuel alloy development for research and test reactors
As part of the National Reduced Enrichment Research and Test Reactor Program, Argonne National Laboratory (ANL) is engaged in a fuel alloy development project. The reduction of the /sup 235/U enrichment from above 90% to below 20% for such fuels would lessen the risk of diversion of the fuel for nonpeaceful uses. Fuel alloy powder prepared with low-enrichment uranium (
Journal of Nuclear Materials, 2011
Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525°C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475°C.
JOM, 2017
A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.