Most of the physical properties of oxide fuel such as lattice parameter, diffusion coefficient, and thermal conductivity are affected by the O/M ratio.

The lattice parameter is needed for calculation of the theoretical density (TD) ratio in the fuel fabrica­tion process. The thermal expansion coefficient, which is defined as the temperature dependency of the lattice parameter, is also an important thermophysical property in fuel design when the variation in heat transport between the fuel and the cladding tube by thermal expansion of the fuel pellets and the stress to the cladding tube by fuel pellets under irradiation are evaluated.

The lattice parameters and thermal expansion coefficients of actinide dioxides are summarized in Table 2 in Section 9.1.З.1. As mentioned in Section 9.1.3.1.2, the dependency of the lattice parameter of stoichiometric mixed oxides on their chemical composition usually obeys Vegard’s law. The lattice parameter of MOX fuel decreases with an increase in the plutonium content. In the hypostoi — chiometric region, the lattice parameter of MOX fuel increases with a decrease in O/M ratio. In addition, Leyva et al.14 showed that the lattice parameter of (U, Gd)O2 decreases with an increase in Gd content.

As mentioned in Section 9.1.3.1.2, Vegard’s law is applied to the evaluation of lattice parameters as a function of composition and temperature in many cases (refer to Figure 13 in Section 9.1.3.1.2). It means that the thermal expansion coefficient of MOX fuel is independent of plutonium content. Martin15 showed that the thermal expansion coeffi­cient of MOX fuel tends to increase with an increase in deviation from stoichiometry in the hypostoichio — metric region.

The melting point of oxide fuel is one of the most important thermophysical properties for fuel design and performance analyses. As the chemi­cal composition and the O/M ratio of the oxide fuel change the melting point of the fuel itself, fuel design and performance analysis should be done in consid­eration of not only the chemical composition at the time of fuel fabrication but also its variation subsequent to nuclear transmutation during reactor operation. In addition, the melting point is also used in the estimation of sintering temperature, as men­tioned before.

Section 9.1.2 shows that the melting point of uranium oxide has its largest value near the stoichio­metric region and the melting point decreases with an increase in deviation from stoichiometry (refer to Figure 1 in Section 9.1.2.1). Further, the melting point of stoichiometric MOX decreases with an increase in plutonium content (refer to Figure 7 in Section 9.1.2.7). In the hypostoichiometric MOX, the melting point of MOX fuel increases with a decrease in O/M ratio.16 Beals et al.17 studied the UO2-GdO15 system at high temperatures and showed that the melting point of Gd bearing UO2 decreases with an increase in Gd content.

During reactor operation, the heat generated in the oxide fuel pellets flows from the central high temperature region to the low temperature periphery of the pellets, and consequently thermal equilibrium is achieved in the pellets. To evaluate the tem­perature distribution when thermal equilibrium is reached, thermal conductivity is one of the most important thermophysical properties. As thermal conductivity is a function of O/M ratio, density, chemical composition, and so on, the variation in chemical composition that occurs during reactor operation should be noted, along with the evaluation of the melting point, as mentioned before.

As mentioned in Section 9.1.6.2, thermal con­ductivities of oxide fuel decrease with an increase in temperature up to 1600-1800 K but increase with an increase in temperature beyond this range

(refer to Figures 33 and 34 in Section 9.1.6.2). The factors which heavily influence the thermal conduc­tivity are O/M ratio and fuel density. Thermal con­ductivity decreases significantly with an increase in deviation from stoichiometry and with a decrease in density. In addition, the thermal conductivity of a gadolinium-bearing uranium oxide decreases signifi­cantly with an increase in Gd content.18,19

2.15.2.1.3.1 Solubility in nitric acid solution

When the nuclear fuel cycle is considered, the disso­lution of oxide fuel is the essential first step in

100

E о

O)

E,

ф

й

c

о

Й

о

(Л

<л

b

1

Figure 3 Dissolution rate of mixed oxide of uranium and plutonium with various Pu contents as a function of the nitric acid concentration. Reproduced from Oak Ridge National Laboratory. Dissolution of high-density UO2, PuO2, and UO2-PuO2 pellets in inorganic acids, ORNL-3695; Oak Ridge National Laboratory: Oak Ridge, TN, 1965.

aqueous reprocessing. The solubility and dissolution rate of oxide fuel in nitric acid solution are important parameters related to the capabilities of the reproces­sing process. Generally, it has been supposed that the dissolution of MOX fuel decreases with an increase in the plutonium content. The maximum plutonium content of MOX driver fuel for fast reactors has been limited to about 30%, from the viewpoint of solubility in nitric acid solution.

There have been many studies on the solubility of oxide fuel in nitric acid solution.2 — From the results of these studies, it has been supposed that the factors affecting the dissolution rate of MOX are the fuel fabrication conditions (homogeneity of the admixture of UO2 and PuO2, sintering conditions and plutonium content, etc.) and the fuel dissolution conditions (nitric acid concentration, solution tem­perature, dissolution time, etc.) (see Figure 3).