QUARTERLY TECHNICAL REPORT ON SPERT PROJECT, OCTOBER, NOVEMBER, DECEMBER 1961

Technical Report · Thu Mar 15 04:00:00 EST 1962

DOI:https://doi.org/10.2172/4781091· OSTI ID:4781091

Schroeder, F. ed.

Spert I. An analysis of the transient temperature distribution in the fuel rod employed in the water-moderated, low-enrichment UO/sub 2/ core was performed. The calculated central and surface temperatures at Peak power for reactor transients of different reactor periods were found to be in good agreement with earlier experimental data. Spert II. The fuel-plate temperature distribution, heat transfer rate, and energy patition between fuel plate and water were calculated from the reactor power and surface temperature data for a D/ sub 2/O-moderated core. The reduced prompt neutron lifetime for the D/sub 2/O- moderated core was calculated from step-transient test data and compared with the statistical value. Spert III. Tests were performed to study the effects of coolant flow on the kinetic behavior of the reactor at elevated temperatures. Forced coolant flow was round to reduce the initial power peak for a 20-msec period excursion at 93 deg C, 2500 psig and 116 deg C, 200 psig. The system temperature response to the startup of a stagnant loop at 430 and 440 deg F, with the primary circulating loop at 450 deg F, was investigated. Engineering. The material of two failed Spert II core structure bolts was found to be of 17-4 PH, not 17-7 PH as specified. Review of the failure in the Spert III pressurizer vessel indicates that it was due to overheating and high-temperature creep. (D.L.C.)

• Research Organization: Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
• DOE Contract Number: AT(10-1)-205
• NSA Number: NSA-16-017235
• OSTI ID: 4781091
• Report Number(s): IDO-16750
• Country of Publication: United States
• Language: English

Related Subjects

COOLANT LOOPS
CREEP
DISTRIBUTION
EQUATIONS
EXCURSIONS
FAILURES
FUEL ELEMENTS
FUELS
HEAT TRANSFER
HEATING
HEAVY WATER MODERATOR
HIGH TEMPERATURE
LIFETIME
LIQUID FLOW
MECHANICAL STRUCTURES
PLATES
POWER
PRESSURE
PRESSURE VESSELS
PROMPT NEUTRONS
REACTIVITY
REACTOR CORE
REACTOR SAFETY
REACTOR TECHNOLOGY
REACTORS
SPERT-1
SPERT-2
SPERT-3
STARTUP
STATISTICS
SURFACES
TEMPERATURE
TESTING
TRANSIENTS
URANIUM DIOXIDE
VARIATIONS
WATER MODERATOR