Pretest analysis of Semiscale Mod-3 baseline test S-07-8 and S-07-9
This document contains a pretest analysis of the Semiscale Mod-3 system thermal-hydraulic response for the second and third integral tests in Test Series 7 (Tests S-07-8 and S-07-9). Test Series 7 is the first test series to be conducted with the Semiscale Mod-3 system. The design of the Mod-3 system includes an improved representation of certain portions of a pressurized water reactor (PWR) when compared to the previously operated Semiscale Mod-1 system. The improvements include a new vessel which contains a full length (3.66 m) core, a full length upper plenum and upper head, and an external downcomer. An active pump and active steam generator scaled to their pressurized water reactor (PWR) counterparts have been added to the broken loop. The upper head design includes the capability to simulate emergency core coolant (ECC) injection into this region. Test Series 7 is divided into three groups of tests that emphasize the evaluation of the Mod-3 system performance during different phases of the loss-of-coolant experiment (LOCE) transient. The last test group, which includes Tests S-07-8 and S-07-9, will be used to evaluate the integral behavior of the system. The previous two test groups were used to evaluate the blowdown behavior and the reflood behavior of the system. 3 refs., 35 figs., 12 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5496764
- Report Number(s):
- IDO-1570-T5; ON: TI85015483
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
REACTOR CORES
BLOWDOWN
COMPUTERIZED SIMULATION
CRITICAL HEAT FLUX
ECCS
FLOW RATE
MOCKUP
PIPES
REACTOR SAFETY
RESPONSE FUNCTIONS
RUPTURES
TEMPERATURE GRADIENTS
TIME DEPENDENCE
VERY HIGH TEMPERATURE
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FAILURES
FLUID MECHANICS
FUNCTIONS
HEAT FLUX
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
SIMULATION
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled