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SMART is a small sized integral type pressurized water reactor and its design minimizes the potential for an occurrence of transients and accidents. To achieve these purposes, engineered safety features enhance the accident resistance. A large volume of semi-passive pressurizer can accommodate a wide range of pressure transients. Low core power density increases the thermal margin, and negative fuel and moderator temperature coefficients yield beneficial effects on a core self-stabilization, and limit the reactor power during accidents. Considering the enhanced engineered safety features and design characteristics, anticipated transient without scram scenarios are adopted to evaluate the strength of the SMART design. The peak pressure does not exceed 116% of the nominal pressure and the fuel can maintain its integrity for the transient. The intrinsic SMART design characteristics having passive safety features and a large primary coolant inventory can maintain a stable condition safely although the reactor protection and diverse protection systems are not available for the anticipated transient without scram transients.

CUPID-SG (CUPID code for Steam Generators) is a computer code for analyzing two-phase flows and heat transfer in PWR steam generators. It was derived from CUPID (Component Unstructured Program for Interfacial Dynamics), which is a component-scale thermo-hydraulic analysis code developed at KAERI (Korea Atomic Energy Research Institute). CUPID-SG shares the code structure with CUPID, which adopts two-fluid, three-field conservation equations, and semi-implicit numerical algorithms. CUPID-SG supports unstructured meshes to handle complex geometries, and adopts constitutive models of two-fluid model based system analysis codes for two-phase boiling flows. FRIGG tests are the benchmark in confirming the correct implementation of the transport models and the applicability of CUPID-SG to two-phase flows over heated tube bundles. A comparison with the FRIGG experiments showed that the constitutive models are properly implemented, and that CUPID-SG is capable of analyzing vertical two-phase boiling flows over tube bundles typically appearing in steam generators.

A heat transfer test facility, SPHINX, which uses carbon dioxide as a medium at supercritical pressures, has been built at KAERI. A series of experiments are under way for various geometries including tubes of several diameters and narrow annulus passages of a concentric and eccentric layout. The experiments aim to collect heat transfer data and to provide an empirical heat transfer correlation required for a SCWR design. In this paper the test results for tubes of 4.4 mm and 9.0 mm IDs, and a concentric annular passage (ϕ8 mm × ϕ10 mm × L1800 mm) are presented for certain combinations of the heat fluxes and mass fluxes. The heat transfer coefficients produced in the tests were compared with those from the existing heat transfer correlations with different media. A new correlation was introduced for the experiment data presented in this paper.