Actions for Experiment data report for semiscale MOD-1 tests S-03-A, S-03-B, S-03-C, and S-03-D (reflood heat transfer tests). [PWR].

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Experiment data report for semiscale MOD-1 tests S-03-A, S-03-B, S-03-C, and S-03-D (reflood heat transfer tests). [PWR].

Published
United States : [publisher not identified], 1976.
[Oak Ridge, Tennessee] : [U.S. Atomic Energy Commission], 1976.
Physical Description
microfiche : negative ; 11 x 15 cm

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Summary
Recorded test data are presented for Tests S-03-A, S-03-B, S-03-C, and S-03-D 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 Mod-1 rod bundle. Tests S-03-A through S-03-D were forced-feed reflood tests in which the reflood rate was held constant during each test. The tests were conducted to investigate the effects on system response resulting from variations in operating conditions of pressure, temperature, core power, reflood coolant, subcooling, and peak heater rod thermocouple temperature at reflood initiation. Test S-03-A was conducted from an initial system temperature of about 230/sup 0/F at a pressure of 20 psia. Tests S-03-B through S-03-D were conducted from an initial system temperature of about 290/sup 0/F at a pressure of 60 psia. In all four tests, reflood coolant was injected directly into the core barrel by means of a specially designed core inlet manifold. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. All four tests were conducted with a flat radial power profile. During reflood, core power was reduced from the initial level according to the American Nuclear Society (ANS) decay heat curve plus 20 percent for pressurized water reactor (PWR) core decay heat. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS was controlled to maintain constant pressure during the tests.
Report Numbers
ANCR-NUREG-1307
Other Subject(s)
Collection
U.S. Atomic Energy Commission depository collection.
Note
DOE contract number: E(10-1)-1375
OSTI Identifier 7265929
Research organization: Idaho National Engineering Lab., Idaho Falls (USA).
View MARC record | catkey: 38041406