U-10Mo Alloy Research Articles

Kinetics of cellular transformation and competing precipitation mechanisms during sub-eutectoid annealing of U10Mo alloys

Transformation kinetics of metastable body-centered cubic γ-UMo phase in U-10 wt.% Mo alloy during annealing at sub-eutectoid temperatures of 500 °C and 400 °C has been determined as a function of time using detailed microstructural characterization by scanning electron microscopy, X-ray diffraction analysis, scanning transmission electron microscopy, and atom probe tomography. Based on the results, we found that the phase transformation is initiated by cellular transformation at both the temperatures, which results in formation of a lamellar microstructure along prior γ-UMo grain boundaries. At 500 °C, cellular transformation is the main mode of transformation. However, at 400 °C, cellular transformation and precipitation of ordered γ′ U2Mo phase within γ-UMo grain interior creates a competitive situation. The kinetics of the cellular transformation was found to be very sluggish at 400 °C compared to the kinetics observed at 500 °C as measured by the transformed area fraction. The hardness of γ-UMo grain interior also was measured and correlated with the observed microstructural evolution. The implication of these results on the low-enriched U-10 wt.% Mo alloy fuel fabrication process development is discussed.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 °C at a rate of 1 °C/min (∼125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.

Microstructural anomalies in hot-isostatic pressed U–10 wt.% Mo fuel plates with Zr diffusion barrier

Microstructural anomalies in the co-rolled-and-HIP'ed U–10wt.% Mo (U10Mo) metallic fuel plate with Zr diffusion barrier assembly were examined as a function of HIP temperature (from 520 to 580°C) and duration (45, 60, 90, 180 and 345min) by scanning and transmission electron microscopy. The anomalies observed in this study are carbide/oxide inclusions within the U10Mo fuel alloy, and regions of limited interaction between the U10Mo alloy and Zr barrier, frequently associated with carbide/oxide inclusions. In the U10Mo alloy, the cF8, Fm3m (225) UC phase (a=4.955Å) and cF12, Fm3m (225) UO2 phase (a=5.467Å) were observed throughout the U10Mo alloy with an approximate volume percent of 0.5 to 1.8. The volume percent of the UC–UO2 inclusions within the U10Mo alloy did not change as functions of HIP temperature and time. These inclusion phases, located near the surface of the U10Mo alloy, were frequently observed to impede the development of interdiffusion and reaction between the U10Mo alloy and Zr diffusion barrier. The regions of limited interaction between the U10Mo and Zr barrier decreased with an increase in HIP temperature, however no noticeable trend was observed with an increase in HIP duration at constant temperature of 560°C.

Fabrication of Powder from Ductile U-Mo Alloys for Use as Nuclear Dispersion

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. Thats why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-dehydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and γ-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed.

Assessment of residual stresses on U10Mo alloy based monolithic mini-plates during Hot Isostatic Pressing

This article presents an assessment of the residual stresses in U–10 wt.% Mo (U10Mo) alloy based monolithic fuel plates and the elasto-plastic response to thermo-mechanical processing. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Understanding of the three-dimensional residual stress field is important for understanding the in-reactor performance of these plate-type fuels. To define fuel-cladding stress–strain characteristics, a thermo-mechanical finite element model was developed. During fuel plate fabrication, the hot pressing temperature approaches the melting temperature of the cladding, so that temperature dependent material properties were incorporated to improve the accuracy of the model. By using elasto-thermo-plastic material models, it was determined that the cladding material (Al6061-O) is subjected to tensile stresses that exceed its proportional limits. The fuel foil is subject to compressive stresses and remains below yield. The residual stresses in the plates are significant, and therefore, should not be neglected. In particular, the simulations indicate the presence of high stress gradients at the fuel/cladding interface, thus emphasizing the need for a high quality bond.