Tem Characterization of Irradiated U3SI2/AL Dispersion Fuel (original) (raw)
Related papers
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.
Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment
2009
This paper outlines the thermal analysis for the irradiation of high density uranium-silicide (U3Si2 dispersed in an aluminum matrix and clad in aluminum) booster fuel for a Boosted Fast Flux Loop designed to provide fast neutron flux test capability in the ATR. The purpose of this experiment (designated as Gas Test Loop-1 [GTL-1]) is two-fold: (1) to assess the adequacy of the U3Si2/Al dispersion fuel and the aluminum alloy 6061 cladding, and (2) to verify stability of the fuel cladding boehmite pre-treatment at nominal power levels in the 430 to 615 W/cm2 (2.63 to 3.76 Btu/s•in2) range. The GTL-1 experiment relies on a difficult balance between achieving a high heat flux, yet keeping fuel centerline temperature below a specified maximum value throughout an entire operating cycle of the reactor. A detailed finite element model was constructed to calculate temperatures and heat flux levels and to reveal which experiment parameters place constraints on reactor operations. Analyses we...
Effect of Si-content on U3Si2 Fuel Microstructure
2020
The development of U3Si2 as an accident tolerant nuclear fuel has gained research interest due to its promising high uranium density and improved thermal properties. In the present study, three samples of U3Si2 fuel with varying silicon content have been fabricated by a conventional powder metallurgical route. Microstructural characterization via scanning and transmission electron microscopy reveals the presence of other stoichiometry of uranium silicide such as USi and UO2 in both samples. The detailed phase analysis by x-ray diffraction shows the presence of secondary phases, such as USi, U3Si, and UO2. The samples with higher concentrations of silicon content of 7.5 wt.% displays additional elemental Si. These samples also possess an increased amount of the USi phase as compared to that in the conventional sample with 7.3 wt.% silicon. The optimization of U3Si2 fuel performance through the understanding of the role of Si content on its microstructure has been discussed.
Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix
Journal of Nuclear Materials, 2011
In the framework of the IRIS-TUM irradiation program, several full size, flat dispersion fuel plates containing ground U(Mo) fuel kernels in an aluminum matrix, with and without addition of silicon (2.1 wt.%), have been irradiated in the OSIRIS reactor. The highest irradiated fuel plate (with an Al-Si matrix) reached a local maximum burnup of 88.3% 235 U LEU-equivalent and showed a maximum thickness increase of 323 lm (66%) but remained intact. This paper reports the post irradiation examination results obtained on four IRIS-TUM plates. The evolution of the fission gas behavior in this fuel type from homogeneously dispersed nanobubbles to the eventual formation of large but apparently stable fission gas bubbles at the interface of the interaction layer and the fuel kernel is illustrated. It is also shown that the observed moderate, but positive effect of Si as inhibitor for the U(Mo)-Al interaction is related to the dispersion of this element in the interaction layer, although its concentration is very inhomogeneous and appears to be too low to fully inhibit interaction layer growth.
Non-destructive Pre-irradiation Assessment of UN / U-Si “LANL1” ATF formulation
2016
The goal of the Advanced Non-destructive Fuel Examination (ANDE) work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels, ultimately also to irradiated fuels. The results of these characterizations provide complete pre-and post-irradiation on length scales ranging from mm to nm, guide destructive examination, and inform modelling efforts. Besides technique development and application to samples to be irradiated, the ANDE work package also examines possible technologies to provide these characterization techniques pool-side, e.g. at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) using laser-driven intense pulsed neutron and gamma sources. Neutron tomography and neutron diffraction characterizations were performed on nine pellets; four UN/ U-Si composite formulations (two enrichment levels), three pure U3Si5 reference formulations (two enrichment levels), and two reject pellets with visible flaws (to qualify the technique). The 235 U enrichments ranged from 0.2 to 8.8 wt. %. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U3Si5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. We have also proposed a data format to build a database for characterization results of individual pellets. Neutron data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. This report provides the results for the characterized samples and discussion in the context of ANDE and APIE. We quantified the gamma spectra of several samples in their received state as well as after neutron irradiation to ensure that the neutron irradiation does not add significant activation that would complicate shipment and handling. We demonstrated synchrotron-based 3D X-ray microscopy on the composite fuel materials, providing unparalleled level of detail on the 3D microstructure. Furthermore, we initiated development of shielding containers allowing the characterizations presented herein while allowing handling of irradiated samples.