Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series. [PWR]
Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident (LOCA).
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- E(10-1)-1375
- OSTI ID:
- 7157489
- Report Number(s):
- ANCR-NUREG-1287
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BLOWDOWN
COMPUTER CODES
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
MOCKUP
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTORS
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS