Assessment Study on the PMK-2 Total Loss of Feedwater Experiment Using RELAP5 Code (NUREG/IA-0200)
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Date Published: March 2001
Lappeenranta University of Technology
Department of Energy Technology
P.O. Box 20
FIN-53851 Lappeenranta, Finland
Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Code Application and Maintenance Program (CAMP)
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
The main objective of the present report is to evaluate the predictability of RELAP5/MOD3.2.2 Beta computer code for the thermal-hydraulic system behavior during a total loss of feed water (LOFW) transient. The relevant experiment was conducted in the PMK-2 facility at the KFKI Atomic Energy Research Institute in Budapest, Hungary. The test simulated a beyond design-basis accident scenario with unavailability of the hydro-accumulators. For prevention of core damage, accident management strategies were applied, including a primary side bleed-and-feed procedure with intentional depressurization of the secondary side. After a brief description of the facility, the profile of the experiment is presented. Modeling aspects are discussed in the RELAP analysis of the test. Emphasis is placed on the ability of the code to reproduce the system response to opening of the pressurizer safety valve and feeding of coolant by the high pressure injection system. The calculations revealed that the code was capable of predicting the parameters with a sufficiently good overall agreement.