Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

Technical Report · Tue Feb 01 04:00:00 EST 1977

DOI:https://doi.org/10.2172/7318864· OSTI ID:7318864

Feldman, E. M.; Collins, B. L.; Sackett, K. E.

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection 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 (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545/sup 0/F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test 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 65/sup 0/F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux 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.
• Sponsoring Organization: US Energy Research and Development Administration (ERDA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 7318864
• Report Number(s): TREE-NUREG-1051
• Country of Publication: United States
• Language: English

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