Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

Conference · Mon Jan 01 00:00:00 EST 1990 · Transactions of the American Nuclear Society; (USA)

OSTI ID:5874142

Kervinen, T; Riikonen, V; Ritonummi, T

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing the heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.

• OSTI ID: 5874142
• Report Number(s): CONF-901101-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (USA), Vol. 62; Conference: American Nuclear Society (ANS) winter meeting, Washington, DC (USA), 11-15 Nov 1990; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
WWER TYPE REACTORS
REACTOR SAFETY EXPERIMENTS
DATA ACQUISITION SYSTEMS
ECCS
FLOW RATE
FUEL CHANNELS
HEAT EXCHANGERS
NATURAL CONVECTION
PLANNING
PRIMARY COOLANT CIRCUITS
R CODES
REACTOR MONITORING SYSTEMS
REACTOR SAFETY
S CODES
SCALE MODELS
SECONDARY COOLANT CIRCUITS
SIMULATION
STEAM GENERATORS
TEST FACILITIES
TWO-PHASE FLOW
ACCIDENTS
BOILERS
COMPUTER CODES
CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLUID FLOW
FLUID MECHANICS
MASS TRANSFER
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
STRUCTURAL MODELS
THERMAL REACTORS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled