Experimental Studies on Development of the Method for Determining the Amount of Impuritated Gases in Nuclear Fuel

Abstract

The article is devoted to an issue of estimating the impurity gas amount in nuclear fuel in the aspect of the distracting contribution from released gases to the total pressure inside ampoule of the device in the simulating a severe accident with core melting. The paper presents a method based on measuring the pressure and temperature of gas in a closed values of the fuel elements during the fuel melting. The correctness of the developed methodology is confirmed by the results of experiments on the melting of fuel in a pulsed graphite reactor IGR with the implementation of a controlled neutron pulse.

Feasibility study mixed oxide fuel tests in the impulse graphite reactor

“…The calculation of the neutron-physical characteristics [24] of the "IGR reactor -experimental device" system was carried out using the combined computational model [25], which fully corresponds to the real design and takes into account the true distribution of the uranium concentration in the graphite blocks of the core. The calculation results showed that in experiments with a given configuration of the experimental device, the required height profile of the power in the fuel element [26] can be provided (Figure 3). In a fuel element with MOX fuel, a significant irregularity of radial energy release will be observed.…”