Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Abstract

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters -the temperature and burnup -on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements.The cladding deformation caused by swelling of uranium dioxide is one of the main factors which limit the service life of high-temperature fuel elements of space nuclear power systems [1]. The computational-theoretical and experimental data on swelling under high-temperature irradiation are limited and contradictory [2]. According to some data, the swelling increases linearly with burnup, while according to other data the swelling saturates. Some works predict the existence of a maximum temperature dependence of swelling, while other works predict a sharp increase of swelling with temperature. These discrepancies could be due to the characteristic behavior of uranium dioxide due to the migration of bubbles of gaseous fission products and structural changes under high-temperature irradiation. The existing computational-theoretical models do not fully take into account the combined effect of these processes. The BARS statistical model of swelling (Bubble Analysis of Radiation Swelling of Fuel), which takes account of the characteristic behavior of bubbles of gaseous fission products and the change in the size and shape of uranium dioxide grains at high temperature, is proposed for analyzing the effect of the operational parameters on high-temperature swelling of uranium dioxide and to search for ways to increase the radiation resistance of oxide fuel.Reactor tests of samples [3] have made it possible to obtain data on free (not restrained by cladding) swelling of various modifications of uranium dioxide at high temperature and to develop a model of the deformation behavior of fuel elements SDS (Stress-Deformation State of Fuel Element). Fuel samples tested for density, oxygen content, grain size and shape, phase composition, and impurity content were used to obtain data on unrestrained swelling. This made it possible to determine the specific nature of the high-temperature radiation behavior of uranium dioxide and its modifications and to provide representative input data for the SDS codes. This factor and the possibility of...

Uranium dioxide mass transfer in the central channel of the fuel kernel of a fuel element of a single-element energy-generating channel

“…As a response, local constrictions which appear in the initial channel increase the deformation rate of the cladding; the formation of a closed vacuum-sealed cavity results in unacceptable loads on the cladding by gaseous fission products leaving the fuel. Because of small axial temperature gradients, longitudinal mass transfer is a slow process, comparable in duration to the service life, and therefore requires detailed studies taking account of the following concomitant factors: cyclic participation of oxygen in transfer with superstoichiometric composition of the uranium dioxide, change of the oxygen coefficient of uranium dioxide as a result of oxygen outflow from the system into the gas phase and inflow of additional oxygen from the dioxide with uranium burnup, local leaks of heat from the kernel through the system of spacers, as well as a change of the fuel temperature during radial mass transfer.Since strengthened emitter cladding is used in a single-element EGC, effectively redistributing the volume change, during swelling of the uranium dioxide, in the direction of the central channel of the kernel and the working fuel temperature characteristic for such an EGC leads to outflow of radiation pores in the same direction, the channel in the kernel remains almost unchanged because of the mutually compensating effect of these processes [3]. For this reason, questions concerning…”

“…Since strengthened emitter cladding is used in a single-element EGC, effectively redistributing the volume change, during swelling of the uranium dioxide, in the direction of the central channel of the kernel and the working fuel temperature characteristic for such an EGC leads to outflow of radiation pores in the same direction, the channel in the kernel remains almost unchanged because of the mutually compensating effect of these processes [3]. For this reason, questions concerning the swelling of uranium dioxide, creep of the kernel, and kinetics of the outflow of radiation pores remain outside the scope of the present article.…”

“…For this case, the expression for the flow of free oxygen Q O 2 (t) outside the fuel element at time t follows from well-known relations for the conductivity of long tubes and channels in the molecular flow regime [9] and has the form where r 1 , l 1 , and T 1 are, respectively, the radius, length, and temperature of the surface of the central channel on the top end of the reflector (see Fig. 1), which is the main gas-dynamic resistance in the path of the flow of molecules leaving the fuel element; m O 2 is the mass of an oxygen molecule; and P O 2 is the oxygen pressure at the entrance into the reflector, calculated using relation (3). The average deviation of the fuel composition from the stoichiometric composition at the time t according to the computed values of Q UO 2 (0, t), Q UO 3 (0, t), and Q O 2 (0, t) is determined from the relation (5) where N U 0 is the initial number of uranium atoms in the fuel kernel, m i is the mass of the molecule of the ith component, and β(t) is the fuel burnup at the time t.…”