ISSN Online: 2688-7231
ISBN Online: 978-1-56700-524-0
Proceedings of the 26thNational and 4th International ISHMT-ASTFE Heat and Mass Transfer Conference December 17-20, 2021, IIT Madras, Chennai-600036, Tamil Nadu, India
Integrated steady and transient pool thermal hydraulic analysis of primary sodium circuit
FBR1&2 is a proposed two loop 600 MWe pool type Indian fast reactor design using liquid sodium as coolant. Towards optimization of design aspects, a comprehensive CFD model of reactor pool and immersed components is developed for detailed thermal hydraulic studies. The main focus of the present work is on resolving temperature distributions of important structural components during reactor transients. During such conditions, both inner and main vessels along with immersed components are subjected to rapid temperature changes, making thermal loads important. Primary coolant, viz., liquid sodium exacerbates the problem of thermal loads on immersed components due to its high thermal conductivity and resultant low boundary layer attenuation. The transient investigated as part of this work is reactor SCRAM. During reactor SCRAM reactor power and flow are brought down from high temperature-high flow normal operation state to low temperature-low flow shutdown state. This leads to fast evolution of reactor pool flow and temperatures. Evolution of temperatures of all important reactor components as a result of this transient is estimated which is necessary for thermo-mechanical analysis of these components.