Thermalhydraulics studies in support of the ASTRID reactor design

International audienceA large number of thermal hydraulics studies are conducted while designing a nuclear reactor.For normal operating conditions in a pool type Sodium Fast Reactor, some design options are chosen according to the thermal hydraulic behavior of the primary circuit in order to minimize the thermal loadings on the structures and the components. For example:- thermal hydraulics studies of the hot plenum aim to verify the thermal hydraulic stability of the jet coming from the core (for nominal and partial loads), evaluate the thermal fluctuations on the Above Core Structure grid, verify the position of the thermal stratification in the hot collector and the homogeneity of the intermediate heat exchangers flow supply,- thermal hydraulics studies of the cold plenum aim to verify the lack of thermal stress on pump-diagrid connecting pipes and strongback due to heterogeneities of temperatures at the intermediate heat exchangers outlet and determine temperatures stratification in the upper part of the cold plenum.For dimensioning transients, thermal hydraulics studies provide input data for damage evaluations.For accidental conditions, thermal hydraulics studies aim at checking the respect of safety criteria to ensure tightness or integrity. In particular, the study of transient leading to decay heat removal in natural convection contribute to the components layout design in the reactor block and to the secondary loops design. The study of unprotected transients aim at checking the effectiveness of design options of the core and the reactor.This paper gives an overview of thermal hydraulics studies realized in the framework of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project and the evolutions of the design as a result of these studies