The crystal structure of uranium dioxide is of fluorite type, as described in Section 2.1. Theoretical density (TD) of UO2 is 10.96gcm~3 at room temperature.

The melting point of UO2 is about 2850 °C (the literature gives a variety of values). However, it has appreciable vapor pressure at lower temperatures leading to weight losses during sintering. Fuel pellets of UO2 are generally produced using powder metallurgy techniques. The properties of UO2 often depend on the processing. The actual density (AD) of UO2 fuel may vary from 80% to 95% of the TD depending on the size/crystallite shape of the powder particles and the actual fuel fabrication pro­cess. However, there are more fabrication techniques that have come to improve fuel fabrication. High density of UO2 fuel fabricated has the following advantages: (a) high uranium density, (b) higher thermal conductivity, (c) high capability to con­tain and retain fission product gases in the fuel, and (d) large linear power rating of the fuel element. Figure 7.19 shows the bulk density of UO2 fuels with respect to oxygen/uranium ratio. In general, UO2 has a lower thermal conductivity compared to other uranium-based metallic and ceramic fuels, and its thermal expansion coefficient is relatively high. However, it got a smaller specific heat.

The thermal conductivity of UO2 has been measured repetitively since the late 1940s; however, modern measurement techniques have produced a significant insight into the transport mechanisms within the UO2 fuel. At temperatures rang­ing from room temperature to about 1800 K (1527 °C), the transport of energy within UO2 is controlled by lattice vibrations that cause a temperature-based decrease in the thermal conductivity trend. However, above 1800 K and up to the melting temperature, a small ambipolar polaron contribution reverses the trend and begins to increase the thermal conductivity. UO2 is a ceramic that is dominated by phonon-phonon interactions and as such the thermal conductivities are low at all temperatures. Figure 7.20 shows the thermal conductivity as a function of temperature.

Both UO2 and UN exhibit very similar temperature-dependent specific heats (Cp) that rise much more rapidly at elevated temperatures than typical ceramic materials (Figure 7.21). Both fuels exhibit a behavior that presents specific heat values nearly twice the Dulong-Petit value near their melting points. Significant research has shown that at low temperatures, the specific heat of both UO2 and UN is governed by lattice vibrations that can be predicted based on the Debye model. Over the range of 1000 (727 °C)-1500 K (1227 °C), the specific heat is governed by the har­monic lattice vibrations and above this temperature the specific heat is governed by crystal defects such as Frenkel pairs up to 2670 K (2397 °C). Above 2670 K, Schottky

TEMPERATURE C)

Figure 7.22 The variation of total thermal expansion and linear thermal expansion coefficients of UO2 as a function oftemperature Ref. [2].

defects become a dominant component in the specific heat for both ceramic materials.

The thermal expansion coefficients of both stoichiometric and nonstoichiometric uranium dioxide vary with temperature, as shown in Figure 7.22.

Mechanical Properties

Uranium dioxide has a tensile strength of just ~35 MPa and Young’s modulus of ~172 GPa. Strength increases with temperature most probably due to closure of pores (sintering effect). Above 1400 °C, it rapidly loses strength and can be subject to plastic deformation.

Irradiation Effects

UO2 is dimensionally stable to high-radiation exposure (>1020ncm~2) and rela­tively high burnup. Under neutron irradiation of sufficient flux, the fuel pellets may fragment by radial cracking. Furthermore, axial and circumferential cracking can also be observed. It is noted that the pellet cracks only in the initial period into few pieces and then stay that way for prolonged duration, implying that the cracking is related to thermal stresses and is not due to mechanical degradation. However, displacement damage caused by fission fragments may enhance the cracking effect.

Uranium dioxide also tends to release volatile fission products from free sur­faces. Fission gases could include Br, I, Te, Xe, Kr, and other related nuclides [17]. The amount of fission gas a depends on many factors, such as porosity, other microstructural characteristics, irradiation time, and irradiation temperature.

Figure 7.23 shows the radiation swelling as a function of fuel burnup at different irradiation temperatures.

The creep rates of a stoichiometric UO2 fuel as a function of stress in both uni­rradiated and irradiated conditions are shown in Figure 7.24. The in-reactor creep rate in the temperature regime of 800-900 °C does not depend much on tempera­ture, but it depends on neutron flux and stress. So, radiation-induced creep oper­ates in this temperature regime. However, in the higher temperature regime (>1200 °C), the creep rate becomes strongly temperature dependent.

7.3.2