Abstract. This paper deals with Post-CHF (critical heat flux) heat transfer with the focus on different regimes of film boiling. The new thermal-hydraulic code TUBE 2.0 is presented. This code uses the equation of energy conservation and predefined correlations to establish wall temperature, the departure of nucleate boiling ratio as well as other parameters of cooling in a simple geometry -an isolated channel. With experimental data of inverted annular film boiling from Stewart, the best-performing correlation for calculation of post-CHF heat transfer in the channel was determined. Finally, the new presented code TUBE 2.0 and subchannel code SUBCAL owned by Chemcomex a.s. are compared using results of various experiments conducted by Becker. Data from Stewart could not be used because of inability to predict the onset of boiling crisis with several correlations.

The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

No abstract

This paper presents the results of an analysis of flow distribution in VVER-1000 mixed core consisting of fuel assemblies with non-identical spacing grids. The calculation was carried out using the modified subchannel code SUBCAL-AZ which allows to calculate 3D thermal-hydraulic characteristics of the coolant flow in the full core subchannel model coupled with the neutron-physical code ANDREA. This full core subchannel model was created in three variants depending on the ANDREA calculations. The first variant (homogeneous core) consisted of 163 hydraulically identical fuel assemblies TVSA-T mod.2, whereas the other variants (mixed cores) consisted of fuel assemblies TVSA-T mod.0, mod.1 and mod.2. These fuel assemblies mainly differ in types, number and axial coordinate of spacing grids and also in diameter of guide tubes. The influence of mixed core to flow distribution was obtained by comparing these variants.

Abstract. This paper presents the results of an analysis of lateral coolant flow between adjacent fuel assemblies with non-identical spacing grids in a mixed core consisting of TVSA-T mod.1 and TVSA-T mod.2 fuel assemblies. The calculation was carried out using modified subchannel code SUBCAL which allows to calculate 3D thermo-hydraulic characteristics of the coolant flow in the full three fuel assemblies model. This full three fuel assemblies model was created in two variants. The first variant consisted of three hydraulically identical fuel assemblies TVSA-T mod.1, whereas the second variant consisted of two fuel assemblies TVSA-T mod.1 and one fuel assembly TVSA-T mod.2 which mainly differ in types, number and axial coordinate of spacing grids and also in diameter of guide tubes. The influence of mixed core to lateral coolant flow and hence coolant temperature was obtained by comparing these two variants. The power distribution was taken from presumed mixed core fuel reload calculated by macro-code ANDREA. Finally there were also provided a comparison of results achieved by subchannel analysis approach with calculation of similar problem using CFD code ANSYS CFX by TVEL, the fuel supplier.