Pressurized Water Small Modular Reactor (SMR), Design Basis Accident Analysis using the ASTEC code

According to classifications adopted by the IAEA, small reactors are characterized by an equivalent electric output of less than 300 MW while medium-sized reactors by an equivalent electric power between 300 and 700 MW. Pressurized water small and medium sized reactors (SMR) generally adopt an integral layout of the primary circuit with in-vessel location of steam generators and control rod drives; one compact modular loop-type design features reduced length piping, an integral reactor cooling system accommodating all main and auxiliary systems within a leak-tight pressure boundary, and leak restriction devices. In this paper, a description is given of the development of the modelling and noding of the primary loop, secondary loop passive core cooling system and containment for a SMR, based on the available data of the SPES3-IRIS integral test facility. SPES3-IRIS is under construction at SIET laboratories in Piacenza (Italy), simulating with 1:100 volume scale and 1:1 height scale, the primary, secondary, containment and safety systems typical of the IRIS small modular reactor. Three ASTEC code modules were adopted: the ICARE module to predict the in-vessel phenomena, the CESAR module to compute two-phase thermal–hydraulics in the Reactor Cooling System (RCS) and for the control and safety systems, and the CPA module to evaluate thermal–hydraulic and aerosol behaviour in the reactor containment. The SMRs as well as the advanced nuclear water-cooled reactors rely on containment behaviour to achieve some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts. Thus, to simulate correctly the main phenomena involved during an accident scenario, the coupling between primary circuit and containment has to be reproduced accurately. Furthermore, given that the containment plays a fundamental role during every accident scenario, it has to be taken into account just as a real safety system. The worst design basis event for the SMR was analysed, and the calculated results were compared with those obtained by the University of Zagreb in collaboration with Westinghouse using the coupled codes RELAP-GOTHIC. The aim of this work is to evaluate the applicability of ASTEC coupled modules in the safety analyses of the new reactor systems with strong interaction between primary system and containment.