Experiment data report for semiscale Mod-1 tests S-02-9 and S-02-9A (blowdown heat transfer tests)
Recorded test data are presented for Tests S-02-9 and S-02-9A of the Semiscale Mod-1 blowdown heat transfer test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Tests S-02-9 and S-02-9A were conducted from initial cold leg fluid temperatures of 542 and 544$sup 0$F, respectively, and initial pressures of 2,253 and 2,263 psia, respectively. 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 full core temperature differential (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 in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs. Both tests were conducted with a flat radial power profile. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. However, pressure suppression system coolant spray was not used. (auth)
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- DOE Contract Number:
- E(10-1)-1375
- NSA Number:
- NSA-33-022303
- OSTI ID:
- 4069249
- Report Number(s):
- ANCR-1236
- Resource Relation:
- Other Information: Orig. Receipt Date: 30-JUN-76
- Country of Publication:
- United States
- Language:
- English
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