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SNAP II Power Conversion System Topical Report No. 16, 2500-Hour Endurance Test of Mercury Rankine Cycle Power System

Description: SNAP 1 is the designation for the 0.5-kw nuclear auxiliary power supply intended for application in a satellite. SNAP 1 was designed to convert thermal energy from the decay of a radioisotope into electrical energy using a Rankine engine with mercury as the working fluid. A successful 2500-hour endurance test is described of a complete developmental version of the SNAP 1 power conversion system utilizing a prototype turbomachinery package, an electrically heated boiler, and an air-cooled condenser. Indications from the data obtained during the test and from inspection of the system following the test were that many more hours of satisfactory operation could have been obtained on all major system components except the rotating unit pump. The mercury-lubricated bearings, the turbine, and the alternator, all demonstrated excellent endurance capability. Based on previous component tests, it is concluded that the pump performance deterioration was caused by air entrainment in the liquid Hg. (auth)
Date: January 1, 1961
Creator: Grevstad, P.E.
Partner: UNT Libraries Government Documents Department
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AEC FUELS AND MATERIALS DEVELOPMENT PROGRAM. Seventh Annual Report.

Description: This report is the seventh annual report of the unclassified portion of the Fuels and Materials Development Programs being conducted by the General Electric Company's Nuclear Materials and Propulsion Operation under Contract AT(40-1)-2847, issued by the Fuels and Materials Branch, Division of Reactor Development and Technology, of the Atomic Energy Commission. This report covers the period from January 31, 1967 to January 31, 1968, and thus also serves as the quarterly progress report for the final quarter of the year.
Date: January 1, 1968
Partner: UNT Libraries Government Documents Department
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Supporting Analysis for Thermal Suitability of Fuel Elements for SM-1A Core I Loading

Description: A recommended SM-1A Core I loading chart was derived from available, metallurgically acceptable elements at the SM-1A and SM-1 sites. The derivation was based on local thermal and hydraulic considerations of minimum elementto- element coolant channel clearances. These clearances were determined from field inspection measurements of outer fuel plate spacing, as modified by analytical calculations of plate ripple growth during exposure to reactor operating thermal stresses. (auth)
Date: January 10, 1962
Creator: Brondel, J. O.
Partner: UNT Libraries Government Documents Department
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Chemical Technology Division, Unit Operations Section Monthly Progress Report, June 1961

Description: An interfacial viseometer was built for use in an interfacial phenomena study. Installation of a 6-in.-ID foam separation column system was completed. The dispersiondrying-sintering characteristics of six low-nitrate batches of thoria sol material were studied. The average effective porosity of the CuO pellets used for reactor helium purification was determined to be 0.0545 for H/ sub 2/ transport and 0.0526 for CO transport. In continuous Zirflex dissolution studies, no H/sub 2/O/sub 2/ decomposition was observed when 10% H/sub 2/O/sup 2 was fed into boiling dissoivent through a water-cooled nozzle and the oxygen concentration in the scrubbed off-gas could be used to control the H/sub 2/O/sub 2/ concentration in the dissolver. The free fluoride in Zirflex solutions must be maintained above 1 molar in order to prevent uranium precipitation at low concentrations of uranium even though the F/sup -//U ratio exceeds 100. Chopped stainless steel-clad UO/sub 2/ sections were leached in a 4 stage pyrex leacher model using 6, 7, and 8 M nitric acid as the dissolvent. The temperature distribution expected within fuel elements consisting of square arrays of tubes was calculated for shipping conditions assuming heat to be transferred only by radiation. HETS values were calculated for uranium stripping under 5% TBP flowsheet conditions. Very fine particles were obtained by quenching fused salt droplets in water. A waste calcinatron run was made using TBP-25 waste and the close coupled evaporator-calciner system. (auth)
Date: January 23, 1962
Creator: Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Partner: UNT Libraries Government Documents Department
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