Engineering & Microstructure Scale PIE Report on EBR-II X441A Metallic Fuel Pins for the MORPH Experiment (original) (raw)
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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.
Low-Burnup Irradiation Behavior of Fast Reactor Metal Fuels Containing Minor Actinides
Nuclear Technology, 2009
Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and/or rare earths (REs) have been irradiated in the fast reactor PHÉNIX to examine the effects of adding those elements on metal fuel irradiation behavior. In this experiment, two MA-containing metal fuel pins, in which the test alloys U-19Pu-10Zr-2MA-2RE and U-19Pu-10Zr-5MA/U-19Pu-10Zr-5MA-5RE (wt%) were loaded into part of a standard U-19Pu-10Zr alloy fuel stack, and a reference fuel pin of U-19Pu-10Zr alloy without MAs or REs was set in an irradiation capsule. Two other capsules with this same configuration are also irradiated. Postirradiation examinations are conducted at ;2.5, ;7, and ;11 at.% burnup. For the low-burnup fuel pins, nonde-structive tests after irradiation have been performed, and the integrity of the pins was confirmed. The irradiation behavior of MA-containing metal fuels up to 2.5 at.% burnup was analyzed using the ALFUS code. The calculation results, such as the axial swelling distribution of a fuel slug or the extrusion behavior of bond sodium to the gas plenum, are consistent with the measurement data regardless of the addition of MAs and REs to the U-Pu-Zr alloy fuels. This observation result indicates that the macroscopic irradiation behavior of U-Pu-Zr fuels containing MAs and REs of 5 wt% or less is similar to that of U-Pu-Zr fuels up to ;2.5 at.% burnup.
Prototypic irradiation testing of high-density U-Mo alloy dispersion fuels
An irradiation test vehicle has been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of this experiment began in October 1999. This irradiation test is designed to obtain irradiation performance data on a series of high-density U-Mo alloy dispersion fuels at prototypic reactor conditions. The irradiation experiment contains 47 miniature fuel plates. The U-Mo alloy fuel compositions include: U-10Mo, U-8Mo, U-7Mo, U-6Mo, U-6Mo-1.7Os and U-6.1Mo-0.9Ru, all fabricated at densities >8 g-U/cm 3. U 3 Si 2 fuel plates have also been included in the test matrix at a density of 6 g-U/cm 3. A variety of fabrication techniques were used to produce the fuel powders, including machining, atomization, arc-sputtering and comminution. Fuel alloys are being tested in both as-fabricated and heat treated conditions. The U-6Mo alloy is being tested in both aluminum and magnesium matrices, while all others are being tested in an aluminum matrix only. The peak fuel temperatures are >200°C. The experiment will be discharged at peak fuel burnups of greater than 30 at.% U 235. Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics under conditions of fissile uranium densities and fuel temperatures prototypic of today's highest enrichment, highest power research reactors. This paper presents the design of the irradiation test and the irradiation conditions.
Fuel clad chemical interaction of U-Mo fast reactor fuel
Journal of Nuclear Materials, 2019
U-33 at.% Mo (16.8 wt.%) metallic fuel is a candidate material for metallic fuel Fast Reactor. One of the lifelimiting factors of fast breeder reactor clad is the fuel clad chemical interaction due to formation of low melting eutectics between U and Fe. This chemical interaction should be avoided or minimized to increase the fuel burn-up. Fuel clad chemical interaction between U-33 at.% Mo metallic fuel with T91 (9 Cr-1Mo) clad has been studied at 650 C, 675 C and 700 C, for different soaking time, through diffusion couple experiments. Development of microstructures, phase constituents and compositions during thermal treatment were examined by scanning electron microscopy and X-ray energy dispersive spectroscopy. In the fuel side, U 6 Fe phase is formed along with bcc-Mo in lamellar morphology through cellular precipitation. Due to slower diffusion of Mo compared to uranium from U-33 at.% Mo fuel, a Mo rich layer is formed at the slug surface. This Mo rich layer subsequently acts as a diffusion barrier layer and minimizes further growth of clad-wastage zone and slug-interaction zone. Multi-phase layer growth constants and activation energies have been calculated and compared with interactions reported in literature of other U-based fuels and cladding elements. The study indicates that U-33 at.% Mo/T91 fuel reduces fuel clad chemical interaction significantly compared with U-23Zr/Fe and U-23Zr/Fe-Cr.
Behavior of unirradiated Zr based uranium metal fuel under reactivity initiated accident conditions
Nuclear Engineering and Design, 2008
Reactivity initiated accident (RIA) tests with 4 unirradiated Zr based uranium metal fuel rods were performed to establish a criterion which should be observed under RIA conditions. Of the four tests, fuel failures were observed in the two tests that experienced the maximum energy depositions of 188 and 212 cal/g, respectively. However, the fuel failures were not observed at the place of a maximum energy deposition but at the position where the thermocouples were installed; one failed at the position whose local energy deposition was 150 cal/g, and the other one at the place with energy deposition of 170 cal/g. The fuel failures seem to have occurred because excessive pressure, which was caused by the partial melting of the fuel meat, was applied to the cladding with a reduced thickness. However, other parts of the fuel rods including the place of a maximum energy deposition maintained their integrity and a big change in the temperature and pressure in the internal capsule, which would be an indication of the fragmentation and dispersion of the fuel meat into the internal capsule, was not observed. Visual inspection also showed that, except for the thermocouple positions, there was no trace of clad failure such as the formation of brittle cracks in the cladding or melting of the cladding. Therefore, for the Zr based uranium metal fuel rods, it can be concluded that the threshold energy deposition above which fragmentation and dispersion of fuel meat into the primary coolant system is expected to occur could be higher than 212 cal/g.
Nuclear Engineering and Design, 2015
An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin Finite Element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U-Pu-Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all available data covering 0 to 100% zirconium concentration and a zirconium concentration dependent linear heat rate solution derived from Monte Carlo N-Particle (MCNP) simulations were developed. An iterative calibration process was applied to obtain optimized diffusion coefficients for U-Pu-Zr metallic fuels. Optimized diffusion coefficients suggest relative improvements in comparison to previous reported values. The most influential or uncertain phase is found to be the gamma phase, followed by alpha phase, and thirdly the beta phase; indicating separate effect testing should concentrate on these phases.
Microstructural characterization of U-Zr alloy fuel slugs for sodium-cooled fast reactor
Surface and Interface Analysis, 2012
In order to investigate a fabrication process of metallic fuel for sodium-cooled fast reactor, U-Zr fuel slugs were fabricated with gravity casting and injection casting. In order to reduce the formation of radioactive wastes, and to have a fine microstructure, spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. The microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscope. U-10wt.%Zr fuel slugs had some dispersion phases of Zr precipitates and Zr compounds of about 5 mm in size. The matrix was composed of fine laminar structure of about 0.15 mm in thickness. U grains became finer from about 30 mm in conventionally cast fuel slug to about 2 mm in atomized powder. Laminar structure became so fine from 0.2 mm in conventionally cast fuel slug to 0.1 mm in atomized powder. It is possible for atomized metallic fuel to have a fine microstructure, resulting in a higher fission gas release rate during irradiation.
Frontiers in Nuclear Engineering
Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swe...