TRACG prediction of gravity-driven cooling system response in the SBWR/GIST facility LOCA tests
- General Electric, San Jose, CA (USA)
General Electric (BE) Nuclear Energy has initiated work on technology programs in support of the advanced light water reactor (ALWR) plants under contract to the U.S. Department of Energy (DOE). Work has been performed under the advanced boiling water reactor (ABWT) design verification program and the simplified boiling water reactor (SBWR) program. The objective of the SBWR program is to develop the key features of a simplified reactor design. The gravity-driven cooling system (GDCS) is an important feature of the SBWR design. The main objectives of the GDCS test program at GE were to demonstrate the technical feasibility of the GDCS concept by performing a section-scaled integrated systems test of the SBWR design and to provide a data base to qualify the TRACG computer code for use in SBWR accident analysis. This paper describes the qualification of TRACG for GDCS applications. The calculational capability and analytical models of TRACG are tested by performing assessment analysis for five loss-of-coolant-accident (LOCA) tests in the GDCS Integrated Systems Test (GIST) facility. The results of the qualification comparisons are presented and TRACG application ranges are discussed.
- OSTI ID:
- 5874504
- Report Number(s):
- CONF-901101--
- Journal Information:
- Transactions of the American Nuclear Society; (USA), Journal Name: Transactions of the American Nuclear Society; (USA) Vol. 62; ISSN TANSA; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACCURACY
BLOWDOWN
BWR TYPE REACTORS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
CONTROL EQUIPMENT
CONVECTION
DATA BASE MANAGEMENT
DEPRESSURIZATION
DESIGN
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EQUIPMENT
FLOW REGULATORS
FLUID FLOW
FLUID MECHANICS
FUNCTIONS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MANAGEMENT
MASS TRANSFER
MECHANICS
NATIONAL ORGANIZATIONS
NATURAL CONVECTION
OPERATION
PERFORMANCE
PRESSURE SUPPRESSION
PRESSURE VESSELS
REACTOR ACCIDENTS
REACTOR OPERATION
REACTOR SAFETY EXPERIMENTS
REACTORS
RESPONSE FUNCTIONS
SCALE MODELS
SIMULATION
STRUCTURAL MODELS
T CODES
TEST FACILITIES
THREE-DIMENSIONAL CALCULATIONS
TWO-PHASE FLOW
US DOE
US ORGANIZATIONS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS