Experiment data report for semiscale Mod-1 test S-01-6

Analysis of the thermal-hydraulic behavior resulting in early critical heat flux and evaluation of CHF correlations for the semiscale core. [PWR; LOFT]

“…5The excellent calculation capability of the General Electric Company correlation and the good capability of the MacBeth, Biasi, and B&W-2 correlations for the Semiscale core indicates that these steady state correlations are adequate for blowdown analysis in Semiscale, and that no new CHF calculation methods are needed for the Semiscalc core. (6) The steady state correlations which needed modification in order to be applicable to blowdown with reverse flow were not able to provide an adequate calculation of CHF in the Semiscale core. 7The observed radial and azimuthal temperature variations at the same elevation for different heater rods indicate that local subchannel behavior existed.…”

Heat transfer correlation development and assessment: a summary and assessment of return to nucleate boiling phenomena during blowdown tests conducted at the Idaho National Engineering Laboratory (INEL). [PWR; BWR]

“…The cold leg break blowdown simulations of the Semiscale testing program were found to have a definite RNB axial distribution. For example, the rewet distribution for Test S-02-7 [ 33 ] (similar to Test S-02-9) is shown in Figure 7. From this figure, it can be seen that rewets occurred predominantly in the upper part of the core.…”

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1976

“…1 Intact loop pump differential pressure (Test S-05-2) · ; 4 3. Intact loop pump differential pressure (Test S-05-2A) 5 4. Intact loop pump differential pressure (Test S-05-2B) 5 '· .…”