Experiment data report for Semiscale Mod-1 Test S-02-6 (blowdown heat transfer test)

Technical Report · Sat Jan 01 00:00:00 EST 1977

OSTI ID:7119257

Collins, B L; Crapo, H S; Sackett, K E

Recorded test data are provided for Test S-02-6 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. Test S-02-6 was conducted to investigate the response of the Semiscale Mod-1 system to a slow depressurization transient resulting from a simulated 6 percent single-ended cold leg break. During system depressurization, power to the Mod-1 electrically heated core was controlled to simulate the surface heat flux response of nuclear fuel rods prior to departure from nucleate boiling. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into the cold legs of both the intact and broken loops.

• Research Organization: SEE CODE- 9502158 EG and G Idaho, Inc., Idaho Falls (USA). Idaho National Engineering Lab.
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 7119257
• Report Number(s): TREE-NUREG-1037; TRN: 77-009163
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BLOWDOWN
MOCKUP
LOSS OF COOLANT
PWR TYPE REACTORS
HYDRAULICS
SIMULATION
ACCIDENTS
FLUID MECHANICS
MECHANICS
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