Experiment data report for Semiscale MOD-2A Natural-Circulation Test Series (tests S-NC-8B and S-NC-9)
This report presents test data recorded for Tests S-NC-8B and S-NC-9 of the Semiscale Mod-2A Natural Circulation Test Series. These tests are part of a series of Semiscale tests that investigate the thermal-hydraulic phenomena resulting from operational transients involving loss of mechanical primary coolant circulation in a pressurized water reactor. Both tests also simulated a loss-of-coolant accident resulting from a 0.4% communicative cold-leg break. These tests provide experimental data to develop and assess the analytical capability of computer models used to predict the results of small-break loss-of-coolant accidents or operational transients involving the loss of primary pumping ability. The primary objective of Tests S-NC-8B and S-NC-9 was to experimentally characterize the thermal-hydraulic behavior of a system during single-phase, two-phase, and reflux natural circulation conditions experienced in the course of an integral small break with and without the presence of emergency core cooling water. Of special interest were the effects on single-phase natural circulation flow caused by changes in core power, primary pressure, and external heater power. This report presents the uninterpreted data from Tests S-NC-8B and S-NC-9 for future analysis. The data, presented as graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5337508
- Report Number(s):
- NUREG/CR-2648; EGG-2184; ON: DE82013828; TRN: 82-013473
- Resource Relation:
- Other Information: Includes 2 sheets of 48x reduction microfiche. Portions of document are illegible
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
SIMULATION
PWR TYPE REACTORS
ECCS
EXPERIMENTAL DATA
NATURAL CONVECTION
ACCIDENTS
CONVECTION
DATA
ENGINEERED SAFETY SYSTEMS
INFORMATION
NUMERICAL DATA
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled