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Uncertainty Analysis of LB-LOCA in VVER-1000 Geometry

Mousavian, S. K ; Sharif University of Technology | 2003

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  1. Type of Document: Article
  2. Publisher: 2003
  3. Abstract:
  4. Evaluation of nuclear power plant performances during accidents is important in view of safety analysis. Several complex thermal-hydraulic system codes are developed for simulation of accidents and transients. Until 1989, only highly conservative methods were approved for evaluating the response of a nuclear power plant to a Large Break Loss of Coolant Accident (LB-LOCA). The conservative evaluation models did not simulate the actual phenomena occurring during a postulated LB-LOCA accurately. Therefore, many countries such as US (CIAU), France (IPSN), Germany (GRS), England (AEA), Spain (ENUSA) and Italy (CIAU) developed methods for calculating the uncertainty in the predictions of advanced thermal-hydraulic codes. The LB-LOCA in a cold leg of VVER-1000 geometry is studied by using the RELAPS system code. The effect of active and passive Emergency Core Cooling Systems (ECCS) and Emergency and Main Feed Waters (E/MFW), as well as the effect of coupling primary side and containment are considered. The Russian results are obtained by the DINAMIKA-97 code in Final Safety Analysis Report (FSAR). We evaluate the uncertainty bands of some important quantities i.e., surface cladding temperature, upper plenum pressure and primary mass inventory by CIAU method. The CIAU method has been developed by the University of Pisa for calculating the uncertainty analysis
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  7. Source: International Conference - Nuclear Energy for New Europe 2003, Proceedings, Portoroz, 8 September 2003 through 11 September 2003 ; 2003 , Pages 361-371
  8. URL: https://inis.iaea.org/search/search.aspx?orig_q=RN:36115770