Benchmarking the RELAP5 L-Reactor model with Savannah River reactor test data
A quality-assured RELAP5 input model of the L-Reactor at the Savannah River Plant (SRP) was developed to support the analysis of a loss-of-coolant accident (LOCA). The RELAP5/MOD2.5 computer code and the L-Reactor model were benchmarked against SRP data to demonstrate their applicability for thermal-hydraulic analysis of SRP reactors. The code and model were benchmarked against data from several different reactor system tests including the 1985 AC Process Flow Tests, the 1983 Cavitation Tests, the 1987 AC Pump Trip Tests, and the 1970 Starved Pump Tests. Results of the benchmark calculations were favorable, yielding confidence in the capability of RELAP5 and the L-Reactor model to determine system response during normal and transient operation, including a LOCA. 17 refs., 30 figs., 11 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (USA)
- Sponsoring Organization:
- DOE/NE
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 7048376
- Report Number(s):
- EGG-EAST-8336; ON: DE90009789
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
BENCHMARKS
COMPARATIVE EVALUATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL EQUIPMENT
COOLING SYSTEMS
ECONOMICS
ENERGY SYSTEMS
ENERGY TRANSFER
EQUIPMENT
FAILURE MODE ANALYSIS
FLOW RATE
FLOW REGULATORS
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
L REACTOR
LOSS OF COOLANT
MECHANICS
OPERATION
PRESSURE EFFECTS
PRODUCTION REACTORS
PUMPS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY
REACTORS
RISK ASSESSMENT
SAFETY
SIMULATION
SPECIAL PRODUCTION REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
T CODES
TEMPERATURE EFFECTS
TRANSIENTS
VALVES
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
BENCHMARKS
COMPARATIVE EVALUATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL EQUIPMENT
COOLING SYSTEMS
ECONOMICS
ENERGY SYSTEMS
ENERGY TRANSFER
EQUIPMENT
FAILURE MODE ANALYSIS
FLOW RATE
FLOW REGULATORS
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
L REACTOR
LOSS OF COOLANT
MECHANICS
OPERATION
PRESSURE EFFECTS
PRODUCTION REACTORS
PUMPS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY
REACTORS
RISK ASSESSMENT
SAFETY
SIMULATION
SPECIAL PRODUCTION REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
T CODES
TEMPERATURE EFFECTS
TRANSIENTS
VALVES