In support of the analyses for the Rivne Nuclear Power Plant (RNPP) VVER-440/213 (Ukraine) Safety Analysis Report (SAR), detailed MELCOR and CONTAIN models of the containment were developed. The RNPP containment features a bubble condenser tower with air locks and active and passive spray systems. Code input models were developed to accurately represent the containment volumes, room interconnections, structural masses, and the engineering safety features. Although MELCOR 1.8.3 [1] was the primary tool for the SAR containment analysis, comparison calculations were performed using CONTAIN Version 1.12 [2]. Consequently, both the response of the VVER-440 containment to limiting design conditions as well as a comparison of the two codes is presented. In the context of SAR requirements, the present application was performed for design basis accidents with conservative assumptions to compare the containment temperature and pressure with design criteria. The peak containment pressure and temperature were evaluated using the most intensive release of the primary and secondary coolant into the hermetic compartments, in particular, for the large break loss of coolant accident and main steam line break. Conservative coolant release data were evaluated using the RELAP5/Mod3.2 SAR model. The selection of the accident scenario, initial and boundary conditions, and the major results are presented. The results of the analyses will be included in the design basis accident analysis chapter of the RNPP SAR.

The paper presents the results of the special screening assessment, conducted under the Level 1 Probabilistic Risk Assessment (PRA) study for the Rivne Nuclear Power Plant (RNPP) Unit 1, Ukraine. This analysis has been intended to investigate potential for negative impacts of the closure of the Main Gate Valves (MGV) during loss of coolant accident (LOCA). RNPP Unit 1 is a VVER-440 Model 213 plant featuring six operating primary loops with MGVs installed at the hot and cold legs of each loop. Emergency operating instructions (EOIs) direct operators to isolate a leaking primary loop using the MGVs during a LOCA. From the experience of accident analyses for some Westinghouse plants it is known that closing loop stop valves (LSV, which are analogous to the VVER-440 MGVs) during accidents may result in more severe consequences. In particular, closing a LSV (or MGV) during a LOCA could inhibit steam venting through the break, and thus depress core levels. This would hasten the onset of core damage. Considering this experience, there has been a need to address such potential impacts in the RNPP Unit 1 PRA, taking into account design specifics of the VVER-440/213 plants. The analysis, which has combined qualitative considerations and thermal-hydraulic calculations with RELAP5/Mod3.2 model of the RNPP Unit 1 nuclear steam supply system, has revealed that such impacts are unimportant for this plant. The paper presents the approach and results of the assessment, and describes the supporting reactor thermal hydraulic evaluations.