Experiment data report for Semiscale Mod-1 Test S-04-4 (Baseline ECC test). [PWR]

Technical Report · Fri Oct 01 00:00:00 EDT 1976

OSTI ID:7125425

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

Recorded test data are presented for Test S-04-4 of the Semiscale Mod-1 Baseline ECC Test Series. This test is among 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 (PWR) system. Test S-04-4 was conducted from an initial cold leg fluid temperature of 541/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 and reflood transient using downcomer volume scaled coolant injection parameters, modified to account for additional coolant injected directly into the lower plenum. In addition, the volume of the lower plenum was reduced from that used in previous tests to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 66/sup 0/F at a core power level of 1.52 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.52 MW in such a manner as to simulate the predicted surface heat flux response of nuclear fuel rods during a loss-of-coolant accident.

• 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: 7125425
• Report Number(s): TREE-NUREG-1003; TRN: 77-006254
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
CORE FLOODING SYSTEMS
DEPRESSURIZATION
ECCS
MOCKUP
SIMULATION
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
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