SPERT PROJECT QUARTERLY TECHNICAL REPORT FOR OCTOBER, NOVEMBER, DECEMBER 1959

Technical Report · Wed Aug 24 00:00:00 EDT 1960

OSTI ID:4149910

Schroeder, F ed

SPERT I. The kinetic-test program using the Oak Ridge National Laboratory BSR-II core at Spert I was planned to yield information pertaining to the safety of the core with and without the ORNL safety system operative and, in addition, to supply detailed performance characteristics for the safety system during short-period power excursions. The data would also add to the store of information on the selflimiting behavior of cores with different physical and nuclear properties. The small stainless-steel plate type BSR-II core was installed in the Spert I reactor and neutronflux distributions, control-rod worths, and void and temperature coefficients have been measured. The neutron- flux peaking in a water-filled hole in this core was also determined as an aid to the planning of future capsule type experiments. SPERT III. The void coefficient of reactivity was measured in the Spert III reactor as a function of system temperature. The average of all measurements was about --2 x 10/sup -4/ dollars/-cm/sup 3/ or --40 cents/% void, which was not in significant disagreement with the density coefficient of 44 cents/% density change ( independent of temperature and pressure) obtained from the previous temperature and pressure coefficient data. The reactivity worth of voids in the interior of the Spert III control-rod-poison sections was investigated by the use of a mock- up control rod placed in various core positions. It was estimated that voiding the entire interiors of all eight control-rod-poison sections with the rods fully inserted would result in a reactivity increase of about 2 dollars. A severe blowout of the stuffing box on a motor-operated primary-flow control valve was investigated and the failure was attributed to scoring and galling of the valvc stems. Modifications have been made in all such valves to prevent future failures of this type. Failure of a pressure fitting on the reactor top head caused wide-spread wctting of the splash-protected electrical and instrument leads with resultant shorting, arcing, and spurious indications on the control panel. All control wiring and instrument leads in the reactor area were redesigned to provide reliable operation in a steam-saturated atmosphere. INSTRUMENTATlON. The response of chromelalumel thermocouples peened to the surface of aluminumclad fuel plates was investigated as a function of the depth of the junction. It was concluded that the variations in the depth at which a thermocouple is peened into the surface of the 0.020-in.-thick aluminum cladding introduce no significant differences in indicated temperature for reactor periods as short as 14 msec. THEORETICAL. In order to investigate the effect of longer thermal-time constants, prompt negative coefficients (as provided by Doppler broadening), and pin type fuel geometry on reactor kinetic behavior, it is planned to perform kinetic tests in the Spert I facility with a low-enrichment core utilizing UO/sub 2/ fuel rods of the type used in the N. S. Savannah. The coredesign parameters for use of these fuel rods in Spert I have been determined. A core loading of 570 fuel pins containing 32.3 kg of U/sup 235/ is estimated to provide an available cold excess rectivity of about 3 dollars. (auth)

• Research Organization: Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
• DOE Contract Number: AT(10-1)-205
• NSA Number: NSA-14-025012
• OSTI ID: 4149910
• Report Number(s): IDO-16616
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-60
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ALUMEL
ALUMINUM
ALUMINUM ALLOYS
CHROMEL
CHROMIUM ALLOYS
DIFFERENTIAL EQUATIONS
FUEL CANS
FUEL ELEMENTS
INSTRUMENTS
MANGANESE ALLOYS
NICKEL ALLOYS
PLANNING
REACTIVITY
REACTOR CORE
REACTOR FUELING
REACTORS
SAVANNAH
SILICIDES
SPECTRAL SHIFT
SURFACES
TEMPERATURE
THERMOCOUPLES
URANIUM DIOXIDE
VARIATIONS