PSI - Issue 54

Oleksii Ishchenko et al. / Procedia Structural Integrity 54 (2024) 241–249 Author name / Structural Integrity Procedia 00 (2019) 000 – 000

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transient, was also considered. Hydraulic resistance coefficients were also specified in the form of distributed values (similar to work Ishchenko et al. (2021)), according to the types of nuclear fuel in the cartogram. Using an optimized representative model, the initial dynamics of a 1 second transient decompression process was calculated for both 1-phase and non-equilibrium 2-phase formulations using SPHM. Two possible cases of LB LOCA were considered - rupture of the “cold loop” (CL#1, Fig. 1) and the “hot loop” (HL#1, Fig . 1). The minimum time integration step was 10 -5 seconds, and the maximum – 10 -1 (almost at the end of the calculations). The initial state was formed for a nominal thermal power level of 3000 MW, the coolant temperature at the inlet was assumed to be 564 K, and the mass flow rate was 4850 kg/s. The operation of the loops is considered symmetrical. Comparison of the main integral characteristics (pressure drop, coolant heating, bypass percentage) showed good agreement with the design data (deviations do not exceed 1 - 5%).

Fig. 2. Formation of energy release in the VVER-1000 reactor core.

The initial dynamics of shock wave propagation into the reactor core barrel is shown in Fig. 3. Based on the results obtained, it is clear that the fundamental difference between 1-phase and 2-phase formulations is that in the second case the pressure level on the outer surface of the barrel is stabilized due to the presence of a flow crisis both when the cold and hot loop ruptures. To be strict, this method of estimating dynamic forces during the initial dynamics of decompression is valid only within 1 period of the shock wave, which is about 30-50 ms. This time period means the shock wave's reflection from the reactor's upper part and its return to the initiating boundary of the rupture. After the shock wave reaches the hot/cold loop, the BC's should consider the presence of a system influence, including the operation of main circulation pumps, which will be accounted in the future. Nevertheless, the maximum amplitude of shock loads on such elements as the reactor core barrel, baffle and nuclear fuel is realized within the above-mentioned time, which is generally acceptable for analysis without considering the system influence.

Fig. 3. Dynamics of pressure changes on the outer surface of the reactor core barrel (a) cold leg rupture; (b) hot leg rupture.

Depending on the complexity of further strength analysis, the procedure for preparing the input data set also differs. The loads on all reactor elements can generally be divided depending on the surfaces (Fig. 4) on which CFD allows obtaining static pressure and velocity. In the case of applying an analysis that considers the wall thickness of an element (for example, a reactor core barrel), there is no data preparation problem since the absolute pressures on each surface are the boundary conditions for the loads assessment. If special models are used that are based on shell theory, then the load parameters are the pressure drops between the conditional internal and external surfaces.