Bruce Nuclear Generating Station B rapid cooldown test and validation of simulation model

Journal Article · Sun Sep 01 00:00:00 EDT 1985 · Nucl. Technol.; (United States)

OSTI ID:5912782

Chang, Y F; Langan, M D; Sermer, P; Watson, P C

The SOPHT code was assessed against Bruce Nuclear Generating Station B commissioning data from a heat transport system rapid cooldown. It was found that (a) under a rapid upstream depressurization, the steam relief valves, like orifices, had a lower discharge coefficient than the corresponding steadystate value and (b) the flashing of water in the steam generators during depressurization causes the at-power boiling heat transfer correlations to overpredict the steam generator heat transfer.

• Research Organization: Ontario Hydro, Central Nuclear Services, Toronto, Ontario
• OSTI ID: 5912782
• Journal Information: Nucl. Technol.; (United States), Vol. 70:3
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BRUCE-5 REACTOR
COOLING
REACTOR ACCIDENTS
BRUCE-6 REACTOR
BRUCE-7 REACTOR
BRUCE-8 REACTOR
COMPUTER CODES
S CODES
REACTOR COOLING SYSTEMS
COMPUTERIZED SIMULATION
DEPRESSURIZATION
HEAT TRANSFER
HYDRAULICS
PERFORMANCE TESTING
RELIEF VALVES
STEAM GENERATORS
VALIDATION
ACCIDENTS
BOILERS
CANDU TYPE REACTORS
CONTROL EQUIPMENT
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
EQUIPMENT
FLOW REGULATORS
FLUID MECHANICS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
MECHANICS
NATURAL URANIUM REACTORS
PHWR TYPE REACTORS
PRESSURE TUBE REACTORS
REACTOR COMPONENTS
REACTORS
SIMULATION
TESTING
THERMAL REACTORS
VALVES
VAPOR GENERATORS
220900* - Nuclear Reactor Technology- Reactor Safety
210400 - Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated