Thermophysical property and material compatibility of accident-tolerant UO2-U3Si2 composite fuel = 사고저항성 UO2-U3Si2 복합핵연료의 열물성 및 재료양립성

다국어 초록 (Multilingual Abstract)

One of the important lessons learned from the Fukushima accident was the low thermal conductivity of UO2 is not something supposed to be overlooked, at least not any longer. The most widely used commercial nuclear fuel, UO2, in fact has the lowest thermal conductivity among all major types of nuclear fuel including U-Zr, U-Mo, UN, UC, and U3Si2. After the accident, many accident-tolerant fuel (ATF) concepts, usually with enhanced thermal conductivity, have been resuscitated and newly suggested; however, after over a decade of research, most of them are still in the testing phase.
Uranium sesquisilicide (U3Si2) is a strong candidate material for ATF since it can simultaneously provide 4-8 times higher thermal conductivity and ~16% higher fissile density than UO2 unlike other non-fissile higher thermal conductivity additives considered for nuclear fuel. Similar benefits could be expected by adopting uranium mononitride (UN) and uranium monocarbide (UC); however, their corrosion resistance in for water-cooled reactor is generally considered unacceptable. The corrosion resistance under water/steam environment is also an outstanding issue for U3Si2, but perhaps considerably manageable. For instance, shorter fuel rod design with reduced fuel pellet diameter has been suggested mainly by Westinghouse Electric Company (WEC) to mitigate the consequence of cladding rupture; however, it will certainly increase the fuel cost.
The in-reactor corrosion issue of U3Si2 could be more economically handled by adopting UO2-xU3Si2 (x < ~50 wt%) composite fuel. In this approach, the concerns are laid on the potential microstructural defect formation between UO2 and U3Si2, e.g., micro-crack and secondary phase formation, which may deteriorate the fuel thermal conductivity and the corrosion resistance, which are the main subject of this dissertation.
A comprehensive experimental investigation on UO2-xU3Si2 (x = 10, 30, and 50 wt%) composite has been conducted in particular order to examine its applicability as an accident-tolerant fuel for water-cooled reactors. Firstly, the fuel manufacturing process was developed, for both conventional sintering (CS) and spark plasma sintering (SPS), using a surrogate material (CeO2-xCeSi2) to minimize radioactive waste generation during the development phase. Then, the UO2-U3Si2 pellets using UO2.02 and UO2.15 powder have been prepared and characterized using X-ray diffractometry (XRD), scanning electron microscopy (SEM), energy dispersive X-ray spectrometry (EDS), and laser flash analysis (LFA). The stoichiometric UO2.02-U3Si2 was selected as as-fabricated or beginning-of-cycle (BOC) composite fuel since the interaction layer (IL) formation between UO2 and U3Si2 has to be suppressed during the fabrication. In contrast, hyper-stoichiometric UO2.15-U3Si2 was considered as an end-of-cycle (EOC) composite fuel assuming a significant discharge burnup that might be challenged for advanced reactor designs. Finally, with the fully characterized fuel pellets, the oxidation resistance of UO2-U3Si2 was evaluated under high temperature air and pressurized water environment using thermogravimetric analysis (TGA). Diffusion couple study of UO2-U3Si2 with a selected ATF cladding material, FeCrAl alloy, was also carried out to glimpse the material compatibility of the composite fuel.
From the surrogate tests (CeO2 instead of UO2, and Ce3Si2 for U3Si2), spark plasma sintering was selected over conventional sintering based on simultaneously obtained higher pellet density and homogeneous microstructure; thus, all UO2-U3Si2 composite pellets used for this study were spark plasma sintered. In the SPS-ed UO2-xU3Si2 (x = 10, 30, 50 wt%) pellets, the formation of U3Si inclusions and USi phase regions was confirmed, which was more significant in high temperature (> 1400 °C) SPS-ed pellets. With increasing U3Si2 composition and sintering temperature, the SPS-ed pellet density per theoretical density was monotonically increased up to ~93%TD. The thermal conductivity of UO2-U3Si2 composite was decreased with increasing temperature, while the amount of decrement was decreased with increasing U3Si2 composition. Hyper-stoichiometric UO2.15-50wt% U3Si2 pellets exhibited 5-7% lower thermal conductivity than that of UO2.02 due to significant secondary USi phase zone formation. In contrast, stoichiometric UO2-50 wt%U3Si2 exhibited up to 100% higher thermal conductivity than UO2, which may suggest progressively accelerated formation of interaction layer and secondary phase with increasing burnup.
Significantly enhanced oxidation and corrosion resistance of UO2-U3Si2, compared to U3Si2, was concluded from the following results: (1) ~130 ℃ higher onset temperature in dynamic oxidation in air, (2) ~50% less weight gain for isothermal oxidation at 300 ℃ in air, and (3) noticeably mitigated pellet degradation under high temperature pressurized water.
The diffusion couple study on UO2-U3Si2 composite with an FeCrAl alloy showed the formation of 5-10 μm thick U-Si-O layer after 100 h annealing at 600 ℃. With further elevated annealing temperature of 800 ℃, U and Si infiltration to FeCrAl matrix was also observed after 100 h annealing, which may potentially limit the reactor operating temperature or the use of the FeCrAl alloy with the composite fuel.
This series of metallurgical and thermophysical investigations partially demonstrated the feasibility of UO2-U3Si2 composite as an accident-tolerant fuel, which might be an only option to offer both higher thermal conductivity and higher fissile density at the same time. It also indicated that suppressing the formation of secondary phases (U3Si and USi) with lower melting temperature and/or thermal conductivity during the fuel fabrication and reactor operation is a key to the successful utilization of the type of inhomogeneous composite fuel. Thus, concerning irradiation-accelerated interaction layer formation between UO2 and U3Si2, further experimental demonstration of the safety and performance of the composite fuel needs to be carried out as a future study, which could utilize, if not in-pile test in a research reactor, in-situ heating ion-beam irradiation under various mechanical stress and temperature conditions matching to an advanced reactor design under consideration.