Starting in the 1950s, significant efforts have been devoted to developing metallic nuclear fuels for use in sodium fast reactors. During irradiation, one performance limitation for this type of fuel ...is fuel cladding chemical interaction (FCCI), which occurs after the fuel swells and contacts the cladding. This can cause formation of interaction zones on the cladding inner surface that have low mechanical strength and that can contain relatively low-melting phases. Significant in-pile and out-of-pile research has been performed to improve understanding of how FCCI affects fuel performance and also to identify ways to mitigate and possibly eliminate this phenomenon. In celebration of the diamond anniversary for the Journal of Nuclear Materials, this paper reviews and highlights key aspects of the metallic fuel FCCI research performed over the last 60 years, and discusses some of the most current research being performed today.
Molybdenum (Mo) has been reported as a potential candidate for the diffusion barrier between U-10Mo fuel and AA6061 cladding in monolithic fuel plates. Interdiffusion and reaction between Al and Mo ...was investigated using solid-to-solid diffusion couples annealed in the temperature range from 425° to 625 °C. Thermo-kinetic parameters, i.e. growth rate and integrated diffusion coefficients, were determined using Wagner’s approach. At low temperatures (T ≤ 450 °C), negligible diffusion interaction was observed between Al and Mo. At intermediate temperatures (T = 475° - 600 °C), diffusion couples were observed to mainly develop a layer of Al8Mo3 intermetallic phase, along with a thin, discontinuous layer of Al22Mo5 intermetallic phase between the terminal Al and Al8Mo3 layer. At high temperature (T = 625 °C), continuous layers of both of Al8Mo3 and Al22Mo5 intermetallic phases developed in the diffusion couple. The Al8Mo3 phase grew with a needle-like, non-planar morphology at intermediate temperatures, contrary to the planar growth of the intermediate phase in binary diffusion couple as per Gibbs phase rule. In general, the degree of non-planarity of needle-like Al8Mo3 phase increased with an increase in annealing time at 500 °C, but decreased with an increase in annealing temperature. This non-planar growth was simulated based on three-dimensional anisotropic diffusion using a quadric representation of the diffusivity tensor.
Thermally induced phase reversion kinetics of decomposed α + γ′ phases to γ phase was investigated for the U – 10 wt% Mo (U–10Mo) hypoeutectoid alloy using X-ray diffraction (XRD), scanning electron ...microscopy (SEM), and transmission electron microscopy (TEM). The as-cast U–10Mo alloy was homogenized to the γ phase at 1073 K for 96 h, and decomposed at sub-eutectoid temperature of 773 K as a function of time from 240 to 1200 h. The decomposition of γ phase initiated via two mechanisms: (1) discontinuous precipitation (DP) or cellular precipitation along the grain boundaries and inclusions; and (2) continuous precipitation (CP) within the γ grains as γ′ needles. After 720 h at 773 K, the DP colonies completely consumed the continuous γ matrix, after which discontinuous coarsening (DC) was observed. The completely-decomposed U–10Mo alloy samples were then annealed at 843, 853, and 863 K as a function of time to document the kinetics of α+γ’ → γ phase reversion. The reference intensity ratio analysis was employed to XRD patterns to quantify the phase amount after each heat treatment. A combinatorial kinetic model, consisting of the classical JMAK nucleation and growth, and spherical particle dissolution, was employed to describe the α+γ’ → γ phase reversion as a function of time and temperature. This phenomenological model provided a good approximation for the phase reversion within the temperature and time ranges examined in this study. Furthermore, a diffusion enhancement approach was employed for a set of typical irradiation parameters to estimate the irradiation effect on the α+γ’ → γ phase reversion kinetics.
The microstructure of fuels evolves with irradiation and extensive microstructural characterization has been performed both on as-fabricated and irradiated U–Mo fuels for both dispersion and ...monolithic designs since initiation of the program to transition research reactors around the world from using high enriched uranium to low enriched uranium fuels. Developing mechanistic understanding of the U–Mo fuel microstructure and properties, which ultimately govern fuel performance during reactor operation, is an important part of the fuel qualification effort. With recent progress in implementation of modern characterization techniques in nuclear fuels research, a critical review of these efforts is warranted. In this review article, the authors provide a compilation and impartial assessment of the available experimental data involving microstructural evolution of U–Mo fuels, gaps in fundamental understanding of irradiation-induced phenomena, and outline path forward for fuels research. This article is intended as a resource for fuel designers and modelers that recaps recent research trends and a guide for prioritizing future experimental work on U–Mo fuels.
Burnable absorbers can benefit nuclear reactors of virtually any design byproviding reactivity control for extended fuel cycles,tritium production,burning of long-lived radionuclides, and reactor ...safety. When selecting the ideal burnable absorber type and its design, one must consider the resulting impact on the reactor’s fuel cycle design and cost, reactivity, thermal hydraulics, manufacturing, and radiation response. These selection criteria, as well as neutronic and thermophysical material property requirements, may be vastly different depending on whether the burnable absorber is intended for use in a commercial water-cooled reactor, a research reactor, or a next-generation advanced reactor system. A recent integration, synthesis, description of past and present technologies, and identification of existing gaps and areas of future research is lacking on these important topics. This paper includes a fundamental description of the use of burnable absorbers and their impacts on reactivity, absorber depletion, self-shielding, basic thermophysical properties, and the use of burnable absorbers in next-generation nuclear applications.
Hot isostatically pressed AA6061 cladding is an important structural component of the high performance, Zr-laminated U-10Mo monolithic fuel system for the application in research and test reactors. ...In this study, the mechanical behavior of two diffusion bonded aluminum alloy, AA6061, was examined using tensile testing. Solid-to-solid diffusion bonding between two pieces of AA6061 was performed by isothermal annealing at 560 °C for 1.5 h, and diffusion couples were subsequently cooled via three different cooling methods: furnace cooling, air cooling, and water quenching. Dog-bone shaped tensile specimens, with 10 mm in gauge length (with diffusion bonded interface in the middle), and 1.5 × 1.5 mm2 gauge cross-sections, were fabricated from the diffusion bonded AA6061 by electro-discharge machining. Yield strength (% EL at failure) of furnace cooled, air cooled and water quenched tensile specimens determined was 82–89 MPa (10–30%), 112–116 MPa (10–14%), and 149–164 MPa (10–17%), respectively. This variation in mechanical behavior was examined with cooling-rate dependent, concentrated precipitation of Mg2Si at the diffusion bonded interface, with due respect for mechanical properties of the AA6061 alloy that inherently vary as a function of cooling rate from 560 °C. Finite element analysis using ABAQUS was employed to augment experimental findings with the appropriate microstructural constituents and alloy properties. Results suggest that the strength is dominated by matrix/bulk properties of AA6061, while ductility is strongly influenced by the cooling method dependent presence of Mg2Si precipitates at the interface.
The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to ...be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.
Fuel-cladding chemical interaction (FCCI) is a phenomenon that occurs at the fuel-cladding interface during the irradiation of U-Zr and U-Pu-Zr metallic nuclear fuel and stainless steel cladding. The ...inter-diffusion zone that develops places both the fuel and cladding at risk through the reduction in cladding strength and the formation of a (U,Pu)/Fe eutectic in the fuel. Due to the impact FCCI has on limiting fuel pin burnup, there is a need for better understanding of the governing FCCI mechanisms in order to make accurate predictions using fuel-performance codes. By performing a critical review of previous work, the physics of FCCI can be separated into individual phenomena so that targeted models can be developed for each. Through examination of experiments conducted both in- and out-of-reactor, the behavior of lanthanides provides a natural separation of models by tracking their behavior through (1) production and transport in the fuel to the clad, (2) interaction with macroscopic changes in fuel topography including cracking and swelling, and finally (3) inter-diffusion at the fuel-cladding interface. Informed by past experience, phenomenological models can be built for each separate effect and subsequently combined in an integral fuel-performance simulation. Prototypical simulation approaches at each level have been included, as well as suggestions for several experiments to help bolster the understanding of irradiated fuel. A robust and predictive FCCI model will provide fuel-performance codes with the ability to predict clad failure and/or fuel eutectic melting. Armed with this information, advanced concepts such as palladium doped fuel, ODS steels, or mitigating reactor designs may be able to reduce FCCI enough to extend fuel burnup beyond its current limits, potentially boosting safety margins and reducing cost through higher fuel utilization.
Microstructural characterization of U-Mo dispersion fuel plates irradiated in the RERTR-7 experiment in the Advanced Test Reactor was performed to study the dependence of fuel matrix composition in ...irradiated U-Mo fuels. The evolution of the fission gas pores and fuel matrix interaction in U-7Mo fuel plates with pure Al and Al-2wt%Si matrix compositions were characterized using optical and scanning electron microscopy in conjunction with image processing techniques. In addition, the combined contribution of Mo composition, fission power, and temperature was studied. Image analysis allowed determination of the fission gas pore (FGP) size, morphology and distribution, as well as the growth of the fuel matrix interaction (FMI) layer. The isolated effect of the Al matrix on the microstructural evolution of U-7Mo fuels revealed that the growth evolution of FGPs as well as the FMI layer is influenced by the matrix. The mean FGP area and FMI growth are 31% and 25% higher in U-7Mo/Al compared to U-7Mo/Al-2Si dispersion fuel, suggesting that U-Mo dispersion fuel with a thicker FMI layer (low thermal conductivity) has the potential to grow larger FGPs in the U-7Mo, possibly because of the higher temperature in the fuel meat. Previous studies have shown that there is an inverse relationship between Mo composition and the FMI, as a result, the microstructural response of U-10Mo and U-7Mo was also investigated and compared. Because the FMI layer thickness is 40% higher in the U-10Mo specimen than the U-7Mo specimen, the combined effect of increased fission power, and temperature for the U-10Mo sample may counteract the propensity of U-Mo fuels with increased Mo content to have a decreased amount of FMI.
High burnup structure formation in U-Mo fuels Smith, Charlyne A.; Biswas, Sudipta; Miller, Brandon D. ...
Journal of nuclear materials,
20/May , Letnik:
563, Številka:
C
Journal Article
Recenzirano
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This study used electron microscopy techniques and phase field modeling to investigate the mechanisms responsible for developing the high burnup structure in U-Mo fuels. The results show that grain ...subdivision is initiated by polygonization; however, evidence of dynamic recrystallization is present with increasing fission densities. At 2.5 × 1021 fissions/cm3, LAGB misorientation of 4° was measured near the grain boundary of neighboring grains, which supports polygonization as the leading grain subdivision mechanism. However, with increasing fission densities, the formation of HAGB subgrain clusters are evident, marking the activation of another mechanism − dynamic recrystallization. Previously formed LAGBs transform into new HAGB grains, which is reflected in the increasing number frequency of HAGBs to LAGBs. At the same time, the HAGB subgrain clusters also undergo polygonization, suggesting that polygonization is a continuous mechanism. Based on the simulations performed in this study, polygonization may continue to occur until a mean grain diameter of ∼382 nm is achieved. Based on the results of this work, LAGB to HAGB transformation is associated with a strain-driven mechanism potentially caused by fission gas-induced plastic deformation and is consistent with recrystallization. Polygonization and dynamic recrystallization occurs in tandem with polygonization leading grain subdivision. Based on the fission densities evaluated in this study, authors postulate that the critical fission density at which dynamic recrystallization (DRX) becomes active is between 2.5 × 1021 fissions/cm3 and 3.5 × 1021 fissions/cm3.
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