G.L. Hofman

Development of very-high-density low-enriched-uranium fuels

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun.

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.

Swelling behavior of U-Pu-Zr fuel

Recent fast reactor driver fuel tests in the Experimental Breeder Reactor II (EBR-II) have demonstrated good performance of U-Pu-Zr fuel alloys to burnups >15 at. pct. Postirradiation examination of these tests has yielded a large amount of fuel-swelling data and metallographic information. These data show that prior to making contact with the cladding tube, metallic alloy fuel swells rapidly due to its high fission-enhanced creep rate and irradiation growth. Measurements of macroscopic fuel slug growth during the entire fuel pin lifetime can now be understood by properly taking into account the irradiation environment and microstructural changes for the various alloy compositions. It has been determined that fission rate, temperature, and plutonium concentrations influence the observed macroscopic swelling rate.

Crystal structure stability and fission gas swelling in intermetallic uranium compounds

Abstract A correlation between fission-induced amorphization and the behavior of fission gasses in intermetallic uranium compounds is proposed. Changes in fission-gas mobility and in the plastic flow rate of the fuel, that result from such a transformation, are likely to be responsible for the large fission-gas-driven swelling-rate increases that have been observed in certain fuels. This study attempts to assess the propensity for amorphization of intermetallic compounds as potential test-reactor fuels by means of their thermodynamic properties. The compound U 6 Mn was thus identified as a probable candidate for a stable high-density intermetallic fuel to be used in dispersion-fuel elements.

Experimental studies of U-Pu-Zr fast reactor fuel pins in the experimental breeder reactor-ll

Argonne National Laboratory’s Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.

Temperature gradient driven constituent redistribution in UZr alloys

Abstract Although the phenomenon of constituent redistribution is common in UPuZr alloys irradiated under a wide range of conditions, it has been observed in UZr alloys only at elevated temperatures. Redistribution is relatively rapid and is essentially complete by 5 at% burnup. Experimental observations of constituent redistribution in UZr 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.

Dynamics of irradiation-induced grain subdivision and swelling in U3Si2 and UO2 fuels

Abstract Observations on low-temperature swelling of irradiated uranium silicide dispersion fuels have shown that the growth of fission-gas bubbles is strongly affected by fission rate. The plot of swelling versus fuel burnup exhibits a distinct “knee” that shifts to higher fission density with increased fission rate. State-of-the-art models of fission-gas behavior do not predict such a dependence. Below the knee, no gas bubbles can be detected by scanning electron microscopy. Just at the knee, gas bubbles are seen to form in a heterogeneous fashion. Above the knee, the bubble population rapidly multiplies and the bubble size increases with fission density. “Subdivision” of the original grains has been observed in high-burnup uranium dioxide. In addition, the peripheral region of LWR fuel pellets reveals an increasingly porous microstructure with burnup. Observations of this “rim effect” show that an extremely fine-grained structure formed by subdivision of the original fuel grains is associated with the porous microstructure. The singular observation that gas-bubble swelling is strongly dependent on fission rate has led us to speculate that a dense network of subgrain boundaries forms in UO2 and U3Si2 at high burnup. Fission-gas bubbles nucleate at the newly formed boundaries and then grow at an accelerated rate relative to that of fission-gas bubbles in the bulk material. A theoretical formulation is presented wherein the stored energy in the material is concentrated on a network of sink-like nuclei that diminish with dose due to interaction with radiation-produced defects. Grain subdivision is induced when the energy per nucleus is high enough that the creation of grain-boundary surfaces is offset by the creation of strain-free volumes, with a resultant net decrease in the free energy of the material. This formulation, applied within the context of a mechanistic treatment of fission-gas-bubble behavior, is shown to provide a plausible interpretation of the observed phenomenon.

Irradiation performance of U-Mo monolithic fuel

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

An alternative explanation for evidence that xenon depletion, pore formation, and grain subdivision begin at different local burnups☆

Abstract In order to interpret the recent observation that xenon depletion, pore formation, and grain subdivision occur successively at increasing local burnups, a rate-theory-based model is used to investigate the nucleation and growth of cavities during low-temperature irradiation of UO 2 in the presence of irradiation-induced interstitial-loop formation and growth. Consolidation of the dislocation structure takes into account the generation of forest dislocations and capture of interstitial dislocation loops. The loops accumulate and ultimately evolve into a low-energy cellular dislocation structure. The cell walls have been previously identified as recrystallization nuclei. The calculations indicate that nanometer-size bubbles are associated with this cellular dislocation structure while the observed micron-size bubbles are presumed to be either preexisting pores deformed by adjacent grains and/or new pores formed in the new recrystallized grain-boundary junctions. Subsequent to recrystallization, gas released from the recrystallized grains feeds the preexisting pores and the recrystallized grains may appear to form a preferential concentration of subdivided grains around the growing pores. This picture is illustrated in a sequence of photomicrographs of irradiated U 3 O 8.

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.