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Design and analysis of a thermal hydraulic integral test facility for bushehr nuclear power plant

Khoshnevis, T ; Sharif University of Technology | 2009

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  1. Type of Document: Article
  2. DOI: 10.1016/j.pnucene.2008.10.004
  3. Publisher: 2009
  4. Abstract:
  5. In this paper, design and analysis of a thermal hydraulic integral test facility for Bushehr Nuclear Power Plant (NPP) is presented. The Bushehr Integral Test Facility (BITF) is a test facility designed to model the thermal-hydraulic behaviours of the Bushehr NPP (VVER-1000) pressurized water reactors currently in use in IRAN. These reactors have unique features that differ from other PWR designs. The BITF simulates the major components and systems of the reference NPP, making it possible to examine postulated small and medium break a loss of coolant accidents (LOCAs) and operational transients. The BITF is a volume-scaled model (1:1375). To ensure that gravitational forces remain equal to those in the reference reactor, the major components and systems in the BITF preserve 1:1 elevation equivalence to the reference reactor. The facility has four loops (each one consists of a hot leg, a steam generator, a loop seal, a main circulation pump and a cold leg), a pressurizer connected via two surge line to the hot leg of the loops 2, 4, the emergency-core-cooling system (ECCS) which is provided by an active pump simulating high and low pressure injection systems, and four hydro-accumulators. The report also contains a comparison between experimental data of PSB test facility and RELAP5 calculations of BITF facility under steady state condition of the reactor power 15% from the nominal. © 2008
  6. Keywords:
  7. Accidents ; Cooling systems ; Gravitation ; Hydraulics ; Loss of coolant accidents ; Nuclear energy ; Nuclear industry ; Nuclear power plants ; Nuclear reactor accidents ; Nuclear reactors ; Pressurized water reactors ; Pumps ; Steam accumulators ; Steam power plants ; Testing ; Water cooled reactors ; BITF test facility ; PSB test facility ; RELAP5/MOD3.2 ; Volume scaling ; VVER-1000 reactor ; Test facilities
  8. Source: Progress in Nuclear Energy ; Volume 51, Issue 3 , 2009 , Pages 443-450 ; 01491970 (ISSN)
  9. URL: https://www.sciencedirect.com/science/article/pii/S0149197008001303