Primary containment capacity of Prototype Fast Breeder Reactor against core disruptive accident loadings

Severe accident analysis methodologies of sodium cooled fast reactors are critically re-looked in recent years and several scenarios are postulated. Development of advanced numerical and experimental techniques is carried out with the objective of determining realistic energy release, which is still evolving internationally. In view of this, a parametric study on mechanical energy release values in the range 100–1000 MJ under a core disruptive accident condition, has been carried out for a 500 MWe pool type Prototype Fast Breeder Reactor (PFBR) and associated mechanical consequences are assessed. The investigation focuses on issues related to structural integrity of primary containment, reactor containment building and post accident cooling capability. The assumptions made on estimation of mechanical energy release have been brought out with appropriate justifications. Fluid and structural dynamics analysis results are presented and discussed in detail depicting the loading sequences and various physical phenomena involved. alysis carried out for PFBR indicates that primary containment has high potential to withstand the transient forces generated by energy release, corresponding to the work potential even more than 1000 MJ. The sodium release to the reactor containment building through top shield penetrations under sodium slug impact phenomenon is limited with higher energy releases, mainly due to large dimensional changes of vessel associated with shorter transient duration and fall of sodium free level in the main vessel. The sodium level fall, a maximum value of 0.88 m for 1000 MJ is found to be acceptable for the PFBR. Hence, high uncertainties in the energy release estimation are not of great concern in PFBR safety studies. However, the deformations of decay heat exchangers immersed in the sodium pool could limit the acceptable work potential to 500 MJ derived based on a separate experimental study.