Experiment data report for Semiscale Mod-1 Test S-02-4 (blowdown heat transfer test)
Recorded test data are presented for Test S-02-4 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-4 is one of several Semiscale Mod- 1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of- Fluid Test (LOFT) design basis. Test S-02-4 was conducted from an initial cold leg fluid temperature of 544$sup 0$F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full design core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66$sup 0$F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The uninterpreted data from Test S-02-4 are presented for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent. (auth)
- Research Organization:
- Aerojet Nuclear Co., Idaho Falls, ID (United States)
- DOE Contract Number:
- E(10-1)-1375
- NSA Number:
- NSA-33-016409
- OSTI ID:
- 4101003
- Report Number(s):
- ANCR-1234
- Resource Relation:
- Other Information: Orig. Receipt Date: 30-JUN-76
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
N77200 -Reactors-Power Reactors
Non-breeding
Light-water Moderated
Non-boiling Water-cooled
220900* -Nuclear Reactor Technology-Reactor Safety
*BLOWDOWN- HEAT TRANSFER
*PWR TYPE REACTORS- BLOWDOWN
DATA
FLOW RATE
FLUID FLOW
LOFT REACTOR
LOSS OF COOLANT
MOCKUP
PRESSURE GRADIENTS
SIMULATION
TRANSIENTS