Sensitivity studies of loss-of-coolant accidents in the Savannah River production reactors
- Los Alamos National Lab., NM (USA)
Loss-of-coolant accident (LOCA) analyses were completed using the Transient Reactor Analysis Code (TRAC) to support the U.S. Department of Energy efforts to restart the production reactors located at the Savannah River Site. The break location and pump operation after the LOCA were the parameters varied for these sensitivity studies. Three location of double-ended guillotine break were studied: plenum inlet, pump suction, and pump discharge. Three pump operation scenarios were also studied: continued operation of both ac and dc pumps, tripping of the ac motor at 2 s after the LOCA, and tripping of the ac motor at 200 s after the LOCA. The production reactors use low pressure and temperature heavy water as the process fluid. The reactor has a moderator tank that contains the fuel channels. Above the moderator tank is an upper plenum that distributes the heavy water to each fuel assembly. The heavy water flows down through the fuel channels and into the moderator tank. From the tank, the water is pumped back to the upper plenum through six loops. Each loop contains a pump and two heat exchangers. Four of the loops have an emergency core coolant system (ECCS) connection. This TRAC model has been benchmarked extensively against data taken in the actual reactors or in prototypical models of the components of the reactors. The calculations were completed using a version of TRAC-PF1/MOD 2 that was updated to include heavy water properties and other changes that are specific to the production reactors.
- OSTI ID:
- 5821986
- 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
Similar Records
Savannah River Site reactor hardware design modification study
Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ECCS
PERFORMANCE
LOSS OF COOLANT
SENSITIVITY ANALYSIS
SPECIAL PRODUCTION REACTORS
REACTOR SAFETY
AC SYSTEMS
BENCHMARKS
COMPUTER CODES
COMPUTERIZED SIMULATION
DC SYSTEMS
DEPRESSURIZATION
FAILURES
FUEL ASSEMBLIES
HEAT TRANSFER
HYDRAULICS
LEVEL INDICATORS
MODERATORS
MOTORS
PIPES
PRIMARY COOLANT CIRCUITS
REACTOR ACCIDENTS
REACTOR OPERATION
REACTOR START-UP
RUPTURES
SAVANNAH RIVER PLANT
STEADY-STATE CONDITIONS
T CODES
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
US DOE
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MEASURING INSTRUMENTS
MECHANICS
NATIONAL ORGANIZATIONS
OPERATION
POWER SYSTEMS
PRODUCTION REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
SIMULATION
START-UP
US AEC
US ERDA
US ORGANIZATIONS
220900* - Nuclear Reactor Technology- Reactor Safety
220700 - Nuclear Reactor Technology- Plutonium & Isotope Production Reactors