A way of modeling the PMK-NVH integral test facility with RELAP5 thermal-hydraulic code is presented. Two code versions, MOD2/36.05 and MOD3 5m5, are compared and assessed. Modeling is demonstrated for the International Atomic Energy Agency standard problem exercise no. 2, a small-break loss-of-coolant acident, performed on the PMK-NVH integral test facility. Three parametric studies of the break vicinity modeling are outlined, testing different ways of connecting the cold leg and hydroaccumulator to the downcomer and determining proper energy loss discharge coefficients at the break. Further, the nodalization study compares four different RELAP5 models, varying from a detailed one with more than 100 nodes, down to the miniature one, with only ∼30 nodes. Modeling of some VVER-440 features, such as horizontal steam generators and hot-leg loop seal, is discussed.