Thermal hydraulic analysis of a steam generator tube rupture accident with total loss of high pressure safety injection

A Steam Generator Tube Rupture (SGTR) accident with total loss of High-Pressure Safety Injection (HPSI) is one of the beyond Design Bases Accidents (DBAs). In most Probabilistic Safety Assessments (PSAs), this accident was considered as a core damage sequence without additional operations actions to prevent core damage. Primary background of this estimation was based on the simple fact that this sequence was a beyond DBA. From the major view-point of DBA, a main purpose of SGTR accident analysis estimates whether the amount of radioactive materials to be released to the environment satisfies a design limitation. On the other hand, from the viewpoint of PSA, especially level 1 PSA, a main purpose of it estimates whether this sequence results in core damage. Because PSA should consider all risk factors including severe accident risk, the release amount of radioactive materials suggested in the design limitation of DBA is considerably less than that treated in PSA. This accident regarded as a core damage sequence because of a beyond DBA resulted in conservative estimation. Therefore, we need to identify whether this accident sequence is spontaneously reached in core damage or not. The present study primarily focused on the understanding of thermal hydraulic behavior of this sequence in order to establish realistic bases of PSA. A thermal hydraulic analysis of a SGTR without HPSI sequence has been performed using MARS2.1 thermal hydraulic code. The analysis was applied to Ulchin units 3 and 4 that was one of the Korea Standard Nuclear Power plants (KSNPs). Single-tube guillotine break at hot-leg side in a steam generator during full power operation was assumed. We also assumed that Reactor Coolant System (RCS)'s coolant charging system and Auxiliary Feedwater system (AFW) were available and no operators' actions were considered. The result of the analysis showed that no core damage was occurred during 24 hours and RCS maintained a high-pressure state. This result means that operators should perform additional actions in order to achieve a safe plant state. We identified that the previous estimation of the accident sequence was more conservative than result from the present study. However, to obtain a precise estimation of the sequence, it is necessary that additional thermal hydraulic analyses have to be performed. We expect that our tentative effort assists an improvement of the current PSA models