The SCDAP/RELAP5/MOD3.1 computer code was used to determine both the reactor and containment system behavior in the course of the accident. Several modules of the code were modified for this study to simulate the process of dissolved nitrogen release from the primary water to the vapor phase within the reactor vessel.

The computer simulation results showed that the reactor coolant leak rate through the pipe break decreases rapidly during the initial 30 seconds of the accident. Its value reduces from 16.5 kg per s at the accident beginning to about 3 kg per s, when the reactor end of the pressurizer surge line becomes uncovered and the leak flow changes from liquid to vapour. The subsequent discharge rate of steam follows the primary pressure (Fig. 3).

The heat removal from the primary system through the steam generator has an essential influence on the reactor coolant discharge rate and the rate of reactor pressure reduction. Heat transferred to the steam generator secondary exceeds the energy transferred to the coolant from the core for almost the whole duration of the accident, causing a decrease in the primary system pressure and temperature. However, the process of nitrogen release from the primary water (Fig. 4) leads to a reduction in the steam condensation rate at the primary side of the uncovered steam generator tubes. The break flow reduces at a lower rate as the result of this process. A significant part (about 20 %) of the released nitrogen mass remains in the vapor phase inside the reactor pressure vessel and localizes mostly above the water level inside the steam generator.

The containment shell pressure continues to rise even after the first safety membrane rupture at 16 s and reaches its maximum value just before the second membrane ruptures at approximately 2650 s (Fig. 5). The values of containment atmosphere pressure and temperature do not exceed permissible design limits of 0.6 MPa and 150 C during the accident progression. Their maximum simulated values are 0.34 MPa and 130 C, respectively.

A steady pressure decrease in all containment system compartments begins at ~5000 s of the accident. At the same time, a positive pressure difference between the containment shell and the PSP compartment starts to reduce to zero and becomes negative at -7000 s. This result is explained by decreasing discharge rate of the primary coolant and continuing heat removal from the containment atmosphere to the containment vessel walls and interior equipment, including the metal-water shielding tank.

During the next phase of the accident, a gradual displacement of water from the PSP compartment to the containment shell is observed as the result of increasing pressure difference between these compartments. At ~ 10000 s, the rising water level inside the containment shell covers the pressurizer surge line break located at the elevation of ~3.5 m above the shielding tank bottom level. After this event, the reactor begins to refill with suppression pool water (Fig. 6). The mass of coolant inside the reactor pressure vessel increases up to the value of -6000 kg and stabilizes beginning at -13000 s. During the last 5000 s of the transient, the calculated leak flow rate has a small value. The direction of the flow alters irregularly, so that the coolant mass within the reactor remains almost invariable for the rest of the accident simulation.

The results of the accident simulation indicate that a sufficient inventory of the primary coolant remains within reactor vessel to maintain reliable core cooling. The critical heat flux is not reached in the core during the accident progression.

The further state of the reactor plant will depend on assumptions concerning the degree of leak-tightness of the containment system compartments, RHR system working conditions, and operating staff efforts to activate failed safety systems. The available grace period is long enough to undertake measures for accident control.

Conclusions

In the course of performing the simulations, it was found that:

1. Neither core uncovering nor the critical heat flux in the core was reached without any operator action during the accident progression.

2. The process of nitrogen release from liquid to vapour phase within the reactor pressure vessel led to increased total loss of the primary coolant, but the values of containment shell atmosphere pressure and temperature did not exceed their permissible design limits.

3. Reactor refilling with the water of pressure suppression pool driven by the pressure difference between containment system compartments occurred during the last phase of the accident.