Experimental evaluation of the heat transfer performance of sodium heated once through steam generator

Abstract Steam generator is a crucial component in a nuclear power plant because its availability is directly linked to the availability of heat transport system and thus the plant availability. In Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction in India, eight number of steam generators each with a heat transfer capacity of 156 MWt transfers 1250 MW of heat from secondary sodium to the conventional steam/water system. The sodium heated once through steam generator with 23 m long seamless straight tubes produces super heated steam at 17.2 MPa pressure and 493 °C temperature. A model steam generator of 5.5 MWt power was tested in steam generator test facility of Indira Gandhi Center for Atomic research for validating the thermal hydraulic and mechanical design of the steam generator. The testing revealed the adequacy of heat transfer capability of the steam generator to transfer the intended power. From the experimental data it is estimated that the steam generator has 8.3% more tube surface area than the required to produce steam at nominal conditions. This paper gives the details of the model steam generator, heat transfer experiments conducted to validate the thermal design and the method for estimating the additional heat transfer area in once through type steam generator.