Comparisons of RELAP4/MOD6 with Semiscale blowdown data. [PWR]

Technical Report · Sat Jul 01 00:00:00 EDT 1978

OSTI ID:6541702

Sahota, M S

The ability of the RELAP4/MOD6-Update 3 computer program to predict the phenomena occurring in a pressurized water reactor during the blowdown-refill phase of a loss-of-coolant accident is independently assessed. Calculations of RELAP4 system models are compared with data from several Semiscale experiments. The sensitivity of the calculated results to core phase slip, break flow, critical heat flux correlation, system nodalization and the initial conditions has been investigated. Areas of RELAP4 deficiencies have been noted. Semiscale Mod-1 modeling techniques for future analyses have been recommended.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 6541702
• Report Number(s): CVAP-TR-78-023
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CALCULATIONS
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS