FUEL CYCLE PROGRAM. A BOILING WATER REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Eleventh Quarterly Progress Report, January-March 1963

Technical Report · Mon Apr 01 00:00:00 EST 1963

DOI:https://doi.org/10.2172/4644665· OSTI ID:4644665

Howard, C.L. comp.

Even though VBWR shutdowns were required for location and removal of five failed fuel assemblies (HPD Program), the increase in fuel exposure was good. Fuel exposures wili pass the values at which cold worked stainless steel cladding was failing under the HPD Program. Failure of the 0.005-inch cold worked stainless steel clad fuel rods in assembly 8L was traced to strain cycling fatigue. A study of tapered fuel rods indicates a potential advantage for us of a variable water/fuel ratio along the flow channel. Natural circulation tests in the hydraulic stability loop were conducted over a range of conditions from stable, to oscillatory with exponential decay, to self-sustaining oscillation of constant amplitude, to unstable oscillations with divergent amplitude. The response to impulses in power input shows the effect of the time delay for transporting steam voids up through the riser. The data permit calculation of oscillation frequency, damping coefficient, time lags, and show the magnitude and character of pressure and velocity changes. The data, which have an experimental scatter of plus or minus 10% maximum, show that burnout heat fiux: decreases with increasing flow up to 2 x 10/sup 8/ lb/hr-ft/sup 2/; has a maximum for hydraulic diameter between 0.25 and 0.5 inch; and decreases for pressure increases between 600 to 1400 psi. A correlating equation for the data is given. The data are compared to results of others. Tests of special geometries show that the burnout heat flux: decreases 22 to 50% when the heated rod is within 0.033 inch of the channel wall; is unchanged upstream of a plate-type spacer; decreases 35 to 50% when the rod surface is roughened by sandblasting; is increased 20 to 40% by use of a rough liner. The four-rod test section is operating satisfactorily and 17 critical heat fiux data points are obtained at 1000 psia and flows of 0.5, 1.0, and 1.5 x 10/sup 6/ lb/hr-ft/sup 2/. In each case the critical heat flux occurred at the exit end and on the side of the rod facing the corner of the channel. The evaluation of film trippers (rough liner) on the unheated channel walls indicates considerable promise for increasing the burnout heat flux limit. The theory of operation is that the liquid film on the unheated wail is sheared off and dispersed, thus adding to the liquid film on the heated rod. Measurements with a heater rod bowed so that it is in contact with the channel wall show that the critical heat flux is decreased by a factor of two or more from values with normal clearance. Temperature measurements on the rod, when operating past the critical heat flux, were in the order of magnitude of 1000 deg F for heat fluxes of about 500,000--600,000 Btu/hr-ft/sup 2/. Chemical analyses for radial variations in isotopic composition within a fuel pellet are nearly completed and are compiled for interpretation. (N.W.R.)

• Research Organization: General Electric Co. Atomic Power Equipment Dept., San Jose, Calif.
• Sponsoring Organization: USDOE
• DOE Contract Number: AT(04-3)-189
• NSA Number: NSA-17-035140
• OSTI ID: 4644665
• Report Number(s): GEAP-4215
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-63
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
BOILING
BURNOUT
COLD WORKING
CONFIGURATION
COOLANT LOOPS
FAILURES
FATIGUE
FILMS
FLUID FLOW
FUEL CANS
FUEL CYCLE
FUEL ELEMENTS
HEAT TRANSFER
HEATING
HIGH TEMPERATURE
HYDRAULICS
IRRADIATION
ISOTOPES
LIQUIDS
MEASURED VALUES
OPERATION
OSCILLATIONS
PELLETS
PLATES
POWER
PRESSURE
PULSES
QUALITATIVE ANALYSIS
REACTORS
RODS
SHUTDOWN
SPACERS
STABILITY
STAINLESS STEELS
STEAM
TESTING
TRANSPORT
VARIATIONS
VBWR
VELOCITY
VOLUME
WATER MODERATOR