Steven L. Hayes

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

Abstract Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium–molybdenum (U–Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4–10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235 U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel–matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U–10Mo composition. Both of the U–10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Temperature gradient driven constituent redistribution in UZr alloys

Abstract Although the phenomenon of constituent redistribution is common in Uue5f8Puue5f8Zr alloys irradiated under a wide range of conditions, it has been observed in Uue5f8Zr alloys only at elevated temperatures. Redistribution is relatively rapid and is essentially complete by 5 at% burnup. Experimental observations of constituent redistribution in Uue5f8Zr fuel elements are presented and analyzed. A model based on a thermal diffusion mechanism is proposed, and its computer implementation is described. The model calculations, supported by experimental observation, indicate that the excess enthalpy of solution of the bcc γ-phase controls the redistribution process as an additional driving force. A heat of transport of −50 to −100 kJ/mol in this phase results in the best match between calculation and experimental observations. The model predicts that constituent redistribution will be observed only when a region of the fuel operates at temperatures above 935 K.

Analysis of constituent redistribution in the γ (bcc) U-Pu-Zr alloys under gradients of temperature and concentrations

Abstract Rods of a ternary alloy (71U–19Pu–10Zr by weight percent) were annealed under a temperature gradient of 220°C/cm for 41 days and examined for micro-structural development and compositional redistribution. An enrichment of Zr with concurrent depletion of U was observed within the γ (bcc) phase region on the hot-end side (T≅740°C). The experimental redistribution of the elements in the γ (bcc) phase was analyzed in the framework of multicomponent mass transport with due consideration of thermotransport and ternary diffusional interactions. Based on a new analysis involving an integration of interdiffusion fluxes in the diffusion zone, kinetic parameters related to the thermotransport and ternary interdiffusion were calculated for each component i over selected ranges of composition. The thermotransport coefficients of U, Pu, and Zr were in the approximate ratio of 1:2:−4.5 in the hot-end region. In addition, the interdiffusion flux contributions arising from the gradients of temperature and concentrations of U and Zr were estimated.

Performance of HT9 clad metallic fuel at high temperature

Abstract Irradiation testing and post-irradiation examination of HT9 clad U-10Zr metallic fuel has been carried out in the EBR-II reactor. Peak cladding temperatures in the range of 630 to 660°C were obtained in order to study high temperature breach behavior of metallic fuel. Of the 15 experimental fuel elements tested to 10 at% burnup, two breached by stress rupture. Breach behavior was found to be totally benign for reactor operations. Significant fuel/cladding chemical interactions involving lanthanide-series fission products and carbide depletion were observed in the hottest cladding region.

Irradiation behavior of U-Nb-Zr alloy dispersed in aluminum

Abstract Three U–Nb–Zr alloys (U–5Nb–3Zr, U–6Nb–4Zr, and U–9Nb–3Zr) were included in a screening irradiation test of low-enrichment aluminum matrix dispersion fuels. Fuel particles made from these alloys reacted readily with aluminum during fuel fabrication and post-fabrication annealing, resulting in large fuel plate thickness increases. Under irradiation, the behavior of U–5Nb–3Zr (wt%) alloy based fuel was poor at 41 at.% 235U burnup, showing indications of incipient breakaway swelling. The post-irradiation microstructural characteristics of U–6Nb–4Zr based fuel were somewhat better than those of U–5Nb–3Zr, but is marginal at 70 at.% burnup. U–Mo based fuels generally show less reaction on fabrication and better fuel performance characteristics during irradiation.

Irradiation behavior of U6Mn–Al dispersion fuel elements

Irradiation testing of U6Mn–Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U6Mn in an unrestrained plate configuration performs similarly to U6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×1027m−3. Fuel plate failure occurs by fission gas pressure driven `pillowing’ on continued irradiation.

Characterization of corroded metallic uranium fuel plates

Abstract A brief background on the history of the corrosion of uranium metal fuel plates from the Zero Power Physics Reactor (ZPPR) and the findings of a recent characterization of the corrosion are presented. The characterization encompassed visual examination, metallography, scanning electron microscopy, and X-ray diffraction. Corrosion of the plates has been observed essentially since their manufacture. The corrosion was found to have both general and localized forms. A black powder corrosion product associated with areas of localized attack was determined to be UH 3 , while the remainder of the corrosion product was UO 2+ x .

FASTER test reactor preconceptual design report summary

The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.