Experiment data report for Semiscale Mod-1 Test S-02-8; blowdown heat transfer test. [PWR]
Recorded test data are presented for Test S-02-8 of the Semiscale Mod-1 blowdown heat transfer test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system, and to provide a data base for a regulatory standard problem. Test S-02-8 was conducted from an initial cold leg fluid temperature of 542/sup 0/F and an initial pressure of 2,262 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full 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 to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The purpose of the report is to make available the uninterpreted data from Test S-02-8 for future data analysis andtest 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.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- DOE Contract Number:
- E(10-1)-1375
- OSTI ID:
- 7249409
- Report Number(s):
- ANCR-NUREG-1238; TRN: 77-001154
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BLOWDOWN
HEAT TRANSFER
PWR TYPE REACTORS
DATA
LOSS OF COOLANT
MOCKUP
SIMULATION
ACCIDENTS
ENERGY TRANSFER
INFORMATION
REACTOR ACCIDENTS
REACTORS
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled