Development of the subcooled boiling model in the SPACE code for a thermal-hydraulic safety analysis of a small modular reactor

Abstract

After Fukushima disaster occurrence, the development of a small modular reactor (SMR) based on innovative safety systems of a passive type is recently brought to light again to inherently eliminate progressing the core melting and hydrogen explosion situations during any postulated or severe accidents. To appropriate cope with changed situation of the global nuclear market and warming, Korean government has determined to develop an innovative SMR, a so-called i-SMR, featured in inherent safety concept and the various passive systems since 2022. According to a phenomena identification and ranking table (PIRT) of legacy SMR design such as Westinghouse IRIS (International Reactor Innovative and Secure) and NuScale SMRs (Mario et al, 2004 and Kent et al., 2010), the collapsed water level in long adiabatic section at above active core is one of the figure of merits (FOMs), which means that it is a high important phenomenon with respect to the safety analysis. In addition, a passive containment cooling system (PCCS) of i-SMR with a tall adiabatic section installed at above a heat exchanger transferring the heat between containment vessel of a reactor module and atmosphere is easy to be occurring a flashing-induced instability (FII) in this system because of operating at low-pressure conditions. The PCCS has to supply a stable cooling capacity to the heat transfer boundary of the reactor module during any accident conditions. These phenomena are highly correlated with a subcooled boiling (SCB). Previous study has revealed that the SPACE code is highly evaluated for the prediction error of the void fraction in the vertical channel under high-pressure SCB conditions. In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical heated channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The first necessary refinement is a new net vapor generation (NVG) empirical correlation which is developed to improve the prediction of an incipient point of NVG using an artificial neural network (ANN) technique. The second modifications are related to the pumping factor based on the Končar model modified as departure bubble diameter and frequency correlations for high pressures and Zeitoun model for low pressures. Final refinement is a modified Bestion drift-flux model to appropriate predict an interfacial velocity of the SUBO experiments for low-pressure SCB conditions. To perform an effect of a new SCB model on the predictive void fraction, the simulations for 1) the SCB experimental tests of various TH conditions, 2) the representative integral effect tests of the legacy verification and validation (V&V) matrix for the SPACE code, and 3) CIRCUS experiment conducted for a flashing induced instability phenomenon were conducted. Based on the result of the simulations, the modified SPACE code was confirmed to the best predictions of SCB void fraction across the entire relevant pressure range as well as derived the reasonable predictions for the FII of important phenomena to design the PCCS of i-SMR. In conclusion, the refined SPACE code can be used as a safety analysis tool for advanced SMRs.