Thermal hydraulic design aspects of heat exchangers in PFBR

Development and induction of fast breeder reactors form the second stage of India's nuclear power programme. Towards this end, a test reactor FBTR (fast breeder test reactor) has already been built. The nuclear heat generated in the core is removed by primary sodium circulated by primary pumps. For the design of IHX and SG, the thermal loadings on the tubes and shell under steady state and transient conditions are essential. The thermal loadings under steady state can occur due to various reasons. The crossflow and bypass flow effects result in the peripheral tubes running hotter than the other tubes. In addition to these, the once through type SGs are prone to static and dynamic flow instabilities which induce cyclic thermal stress. Computer codes have been developed for analysing these instabilities. These instabilities will be taken care of by providing orifices in the tubes. Estimation of thermal loading, through large scale experiments in sodium is prohibitively expensive and time consuming. On the other hand analytical estimation would greatly reduce the need of scale model testing and aid in optimization of design. This paper presents a brief outline of the computer codes used and results of multidimensional analysis carried out for IHX and SG of PFBR. 4 refs., 6 figs., 2 tabs