Simulation of a loss of coolant accident: Results of a standard problem exercise of the International Atomic Energy Agency
- Texas A M Univ., College Station (USA)
The purpose of this study was to compare the results generated from the IBM version of RELAP5/MOD2 to the experimental data of an International Atomic Energy Agency (IAEA) standard problem exercise. The standard problem exercise data were that of a 7.4% break loss-of-coolant accident conducted at a test facility in Hungary. The United States did not formally participate in this exercise whose aim was to assess the capabilities of computer codes and modeling techniques and in which a total of 17 organizations from 12 countries participated. The results obtained using the IBM version of RELAP5/MOD2 compared favorably with the experimental data. The experimental facility, PMK-NVH (Paks Model Circuit), is a scaled-down model of a Hungarian reactor, the VVER-440 Paks nuclear power plant. A volume and power scaling ratio of 1:2070 is used. The six loops of the actual reactor are modeled by one active loop called the PMK. The secondary loop in the experimental facility is the NVH loop. The coolant in the facility is water, and the operating conditions are the same as in the actual reactor. The orientation of the steam generator is horizontal, as opposed to the vertical design of once-through and U-tube steam generators. The parameters of the accident are that it starts at full power, a 3-mm cold-side break occurs at the upper head of the downcomer, there is no injection from hydroaccumulators, the high-pressure injection system corresponds to the case in which one-third of the pumps are available, and isolation of the secondary occurs immediately after transient initiation.
- OSTI ID:
- 6851637
- Report Number(s):
- CONF-890604--
- Journal Information:
- Transactions of the American Nuclear Society; (USA), Journal Name: Transactions of the American Nuclear Society; (USA) Vol. 59; ISSN TANSA; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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ACCIDENTS
AVAILABILITY
BOILERS
COMPUTERIZED SIMULATION
COOLING SYSTEMS
ECCS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
HIGH PRESSURE COOLANT INJECTION
LOSS OF COOLANT
PAKS-1 REACTOR
PAKS-2 REACTOR
PAKS-3 REACTOR
PAKS-4 REACTOR
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
STEAM GENERATORS
THERMAL REACTORS
TIME DEPENDENCE
TRANSIENTS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WWER TYPE REACTORS