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Assessment of Potential Negative Impacts of the Main Gate Valve Closure Under Loss of Coolant Accident at VVER-440 Model 213 Nuclear Power Plant

[+] Author Affiliations
Chester Everline, Steve Meyer

Scientech, Inc., Concord Twp., OH

Gregory Gromov, Igor Lola, Stanislav Sholomitsky, Victor Mukoid, Alexander Sevbo

Energorisk, Ltd., Kiev, Ukraine

Paper No. ICONE12-49342, pp. 177-184; 8 pages
  • 12th International Conference on Nuclear Engineering
  • 12th International Conference on Nuclear Engineering, Volume 3
  • Arlington, Virginia, USA, April 25–29, 2004
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 0-7918-4689-X | eISBN: 0-7918-3735-1
  • Copyright © 2004 by ASME


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.

Copyright © 2004 by ASME



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