Simulation and conceptual analyses of a stable natural core cooling system in an integrated small modular PWR

Abstract

Some of advanced Small Modular Reactors (SMRs) have much more passive and inherently safety features as well as other advantages such as rapid and more flexible installations than the conventional large nuclear reactors. In order to attain an optimum thermal-hydraulic design condition and analyze the transient phenomena in a small size reactor, a modified single-phase model is applied to simulate primary loop and estimate the main loop parameters of an advanced SMR especially for stability analyses. In this paper, the modified model is considering flash-out phenomena through the core and riser parts in addition of the single-phase model, besides using a four-region model for the integral pressurizer zone. Except some simplifying assumptions in governing correlations and fluid properties that do not make a major source of error, the results demonstrate acceptable accuracy in testing intervals. Particularly, stable states are studied during natural convection phenomenon. Firstly, the self-developing thermal-hydraulic code takes into account the parameters and analyses using the MATLAB software. Then, the RELAP5/MOD3.2 is also used to simulate and analyze the system. It uses a complete two-phase system of continuity equations. It is worth mentioning that, the RELAP5 code was especially proposed to be considered and initialized based on some main active safety system systems particularly pumps. A comparison of the results from both models for stability tests (maximum affordable core power) shows that the modified single-phase model makes suitable prediction during the steady state and also during small and quick deviations from the steady state conditions. On the other hand, the RELAP5 code can initialize and follow the transients of the natural cooling system without any pump. A parametric study has been performed to search the preliminary domain of the design. Then, a typical periodic oscillation is used to take system behavior and responses against possible perturbations or instabilities. Results are fairly promising to reach a stable core cooling system based on a complete natural convection. Finally, results show that including and using a 100 KW heater can stabilize the induced non-linear instabilities. It could also be noted that, the ability of a complete passive heat removal system could be mentioned as the most important passive safety feature of the future nuclear reactors. Academics still remember the Fukushima-Daiichi disaster. Accordingly, in this study, only a small heater has been added to the initial pattern. Then the developed model enhances the stability and safety performance of such an integral Pressurized Water Reactor (iPWR).