Assessment of subchannel code ASSERT-PV for supercritical applications

The Canadian subchannel code ASSERT-PV has been modified and used in the development of a fuel assembly concept for the Canadian Supercritical Water-cooled Reactor (SCWR). This paper describes the modification to ASSERT-PV for the SCWR applications, and presents an assessment of the code against available experimental data. Full-bundle tests for the SCWR fuel assembly are not yet available; therefore, partial-bundle tests with a reduced number of rods and a shortened channel length were considered in the assessment. Three heat transfer experiments at supercritical pressures were selected for the assessment: (i) a Japanese 7-rod bundle water experiment; (ii) a Chinese 2 × 2 rod bare-bundle water experiment; and (iii) a Russian 7-rod bare-bundle Freon experiment. The rod surface temperature was taken as the key parameter in the assessment since the maximum fuel cladding temperature is a key criterion used in developing the SCWR fuel assembly concept. The ASSERT-PV predictions of rod surface temperature were compared against experimental data of measured wall temperatures. Six widely-known heat transfer correlations were assessed for their suitability in predicting the wall temperatures in rod bundles under conditions relevant to the Canadian SCWR. Overall, the Jackson correlation was found to be the most suitable for predicting the wall temperatures in the range of conditions covered by the selected experiments.