Today, Small Modular Light Water Reactors based on the use of natural circulation for removing core power in steady operational and transient condition, are one of the key design options for the deployment of nuclear reactor technology considering its advantage in term of inherent safety, design simplicity, simplified parallel construction, reduction of construction time, and reduction in finance and operation cost. In order to develop deterministic safety analyses, best estimate thermal-hydraulic system codes need to be validated against natural circulation phenomena typical of an integral design during steady-state and transient conditions. The goal of this paper is to summarize the results of the validation activity of the TRACE code against the experimental database developed in the OSU-MASLWR facility, designed to thermal hydraulically characterize the Multi-Application Small Light Water Reactor (MASLWR) basis for the NuScale design. The TRACE code qualitative and quantitative accuracy, by using the Fast Fourier Transform based Methods (FFTBM), will be presented and discussed. Particular attention will be focused on the capability of the code to predict the natural circulation in the integral design, the primary-to-secondary heat transfer in helical coil steam generator, and the passive primary-to-containment coupling. The tests selected for the analyses are the OSU-MASLWR-OOl, aiming at characterizing the thermal-hydraulic coupling of the passive primary-to-containment in design basis accident conditions, and the OSU-MASLWR-002 test aiming at characterizing the single-phase natural circulation and the primary-to-secondary heat transfer in a helical coil steam generator. © 2019 American Nuclear Society.
Validation of the trace code against small modular integral reactor natural circulation phenomena
Mascari, F;
2019-01-01
Abstract
Today, Small Modular Light Water Reactors based on the use of natural circulation for removing core power in steady operational and transient condition, are one of the key design options for the deployment of nuclear reactor technology considering its advantage in term of inherent safety, design simplicity, simplified parallel construction, reduction of construction time, and reduction in finance and operation cost. In order to develop deterministic safety analyses, best estimate thermal-hydraulic system codes need to be validated against natural circulation phenomena typical of an integral design during steady-state and transient conditions. The goal of this paper is to summarize the results of the validation activity of the TRACE code against the experimental database developed in the OSU-MASLWR facility, designed to thermal hydraulically characterize the Multi-Application Small Light Water Reactor (MASLWR) basis for the NuScale design. The TRACE code qualitative and quantitative accuracy, by using the Fast Fourier Transform based Methods (FFTBM), will be presented and discussed. Particular attention will be focused on the capability of the code to predict the natural circulation in the integral design, the primary-to-secondary heat transfer in helical coil steam generator, and the passive primary-to-containment coupling. The tests selected for the analyses are the OSU-MASLWR-OOl, aiming at characterizing the thermal-hydraulic coupling of the passive primary-to-containment in design basis accident conditions, and the OSU-MASLWR-002 test aiming at characterizing the single-phase natural circulation and the primary-to-secondary heat transfer in a helical coil steam generator. © 2019 American Nuclear Society.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
