Abstract
For the enhancement of the safety of sodium-cooled fast reactors (SFRs), the core cooling capability in the natural circulation (NC) condition using a dipped-type direct heat exchanger (D-DHX) installed in the upper plenum of the reactor vessel (RV) has been investigated in Japan Atomic Energy Agency. The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the coreplenum interaction (CPI) which is the phenomena to decrease the maximum core temperature by the coolant at low temperature from the D-DHX flowing into the core. To judge the adequacy of the numerical results for a validation study, the degree of difference (DoD) between the numerical and experimental results must be measured. The area validation metrics (AVM) and the modified AVM (MAVM) based on the probability box (p-box) method were examined to measure the DoD accounting for the contribution of fluctuation phenomena. The DoD value was measured as the degree of the closed area between the cumulative distribution functions evaluated from the transient data of the numerical and experimental results. Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
