Experiment data report for semiscale Mod-1 Tests S-03-1, S-03-2, S-03-3, and S-03-4 (reflood heat transfer tests) [electronic resource].
- Published:
- Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy, 1976.
- Physical Description:
- Pages: 136 : digital, PDF file
- Additional Creators:
- United States. Department of Energy. Office of Scientific and Technical Information
Access Online
- Restrictions on Access:
- Free-to-read Unrestricted online access
- Summary:
- Recorded test data are presented for Tests S-03-1, S-03-2, S-03-3, and S-03-4 of the Semiscale Mod-1 reflood heat transfer series (Test Series 3). The tests conducted in this series are separate effects core reflood tests performed to determine the reflood heat transfer characteristics of the 5.5-foot-long Mod-1 rod bundle. Tests S-03-1 through S-03-4 were forced feed reflood tests in which the reflood rate was held constant during each test, although the reflood rate varied from test to test. Tests S-03-1 through S-03-4 were conducted from an initial system temperature of about 290/sup 0/F at a pressure of 60 psia. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. In all four tests, reflood coolant at 153/sup 0/F was injected directly into the core barrel by means of a specially designed core inlet manifold. Test S-03-4 was performed with a peaked radial power profile, in contrast to the flat profile used in the first three tests. During reflood, core power was reduced from the initial level of 0.097 MW, according to the American Nuclear Society decay heat curve for pressurized water reactor (PWR) core decay heat, +20 percent. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS system was controlled to maintain constant pressure during the tests.
- Report Numbers:
- E 1.99:ancr-nureg-1306
ancr-nureg-1306 - Subject(s):
- Other Subject(s):
- Note:
- Published through SciTech Connect.
05/01/1976.
"ancr-nureg-1306"
Crapo, H.S.; Jensen, M.F.; Sackett, K.E.
Idaho National Engineering Lab., Idaho Falls (USA) - Funding Information:
- E(10-1)-1375
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