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Mass Transfer Coefficients and Interfacial Area in a One Stage Pulse Column

Description: Over-all mass transfer coefficients were determined independently of the interfacial area by considering the operation of the pulse column to be a stage- wise process. The calculated coefficients describe only the operation of the pulse column for the free rise of organic droplets through a stage, since in the downstroke the aqueous phase passes through the organic phase in the form of rivulets preferentially wetting the plates. The driving force for transfer was based on the aqueous phase conc… more
Date: November 1, 1960
Creator: Konopik, A. E. & Burkhart, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A CORROSION STUDY OF WELDED STAINLESS STEEL FUEL ELEMENTS

Description: Fuel element corrosion studies designed to and in selecting and evaluating SM-2 fuel element welding techniques are discussed. Tests on type 347 ss plate type fuel elements welded by the selected tungsten inert gas technique showed good corrosion integrity of specimens under a variety of conditions including a 500-hr test under simulated SM-2 conditions of flow and coolant chemistry. (auth)
Date: December 22, 1960
Creator: Bergen, C. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Civilian Power Reactor Program. Part I. Summary of Technical and Economic Status as of 1960. Heavy Water-Moderated Power Reactors

Description: A revision of section 8 of TID-8516, Part I, is presented. The reactor concept which is presented is a change from the former design, a pressurized- pressure vessel-indirect cycle plant to a boiling-pressure tubedirect cycle plant. A description of the plant and a summary of characteristics for the 110- and 325- Mwe systems are given. (J.R.D.)
Date: August 19, 1960
Creator: Hutton, J. H.; Davis, S. A.; Graves, C. C. & Duffy, J. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Mixing of a Cold Liquid Jet With a Boiling Liquid Stream

Description: The turbulent mixing of a cold liquid jet with a stream of vapor bubbles and saturated liquid is analyzed for a single axial liquid jet. The cases where the issuing jet velocity is very much greater than and nearly equal to the free stream velocity are considered. Transport of axial momentum and of scalar quantities in the stream are calculated and the results applied to the transport of enthalpy in a bubbly mixture. (D.L.C.)
Date: October 21, 1960
Creator: Bankoff, S. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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An Investigation of the Structural Integrity of Selected Components of the Oak Ridge Research Reactor

Description: An investigation was made to determine the structural behavior of selected components of the Oak Ridge Research Reactor for increased power level conditions. It was found that a reactor cooling water outlet temperature of 150 deg F will cause severe plastic strain cycling in the aluminum housings for the large test facilities. Increasing the reactor cooling water flow rate of 21,000 gpm will cause plastic deformations in certain reaons of the core box. These latter deformations can be tolerated… more
Date: July 22, 1960
Creator: Corum, J M; Greenstreet, B L; Maxwell, R L & Rosenthal, M W
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PERIODIC RADIATION SURVEY OF REACTOR PLANT CONTAINER AND COMPONENTS AFTER SHUTDOWN. CORE I, SEED I. SECTION 3. Test Results (T-612076)

Description: The effect of isolation and forced draining on the radiation level of the 1AC coolant purification hairpin loop was investigated. The activity in the horizontal and vertical legs (U-bend) of the loop piping was determined by a radiochemical analysis of the crud deposits. (C.J.G.)
Date: May 22, 1960
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Valve Stem Freeze Seal for High-Temperature Sodium

Description: Valve stem freeze seals for high-temperature service in advanced sodium- cooled reactor systems were studied. An experimental model, suitable for use with a 6-in. size valve, operated satisfactorily under a variety of conditions. The freeze seal region was cooled by natural convection to ambient atmosphere; cooling by both circumferential and longitudinal finned sections was experimentally studied. The operating conditions included sodium bulk temperatures up to 1300 tained F, sodium pressures … more
Date: July 30, 1960
Creator: McDonald, J.S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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ANNUAL SUMMARY RESEARCH REPORT IN ENGINEERING FOR JULY 1, 1959-JUNE 30, 1960

Description: Work was continued on the determination of size and distribution of dispersed phase droplets in a pulse column. The droplet behavior and dispersed phase hold-up in Yorkmesh packing was studied with an equilibrated system of methyl isobutyl ketone and water with the ketone phase dispersed. An investigation is being made of the recovery of copper and EDTA from rare-earth ion exchange wastes. The effect of vapor properties on entrainment from bubble cap trays is being investigated. The design, con… more
Date: August 1, 1960
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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LAMPRE Project Quarterly Report: May 12960

Description: With the initiation of the LAMPRE Program Quarterly Report the opportunity is taken to review the objectives of the program as well as to summarize its present status. The program can be conveniently divided into two separate but related phases. The first is concerned with the design and construction of the first reactor experiment, called LAMPRE I. The second involves problems of specific interest for LAMPRE I as well as to the general class of molten plutonium fueled reactors. A general descr… more
Date: June 1960
Creator: Glasstone, Samuel
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SOME STEADY-STATE THERMAL CHARACTERISTICS OF A THREE-LOOP REACTOR POWER SYSTEM

Description: The three-loop power system which is to be used with Experimental Breeder Reactor No. II (EBR-II) was analyzed to determine the coolant flow rate requirements at various power levels, coolant temperatures at various power levels, effects of heat exchanger sizes (system optimization), and effects of control errors. An intermediate heat exchanger, preheater, evaporator, and superheater are included in the EBR-II power system. Constant thermal resistances and physical properties, and perfect insul… more
Date: March 1, 1960
Creator: Bump, T.R. & Monson, H.O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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ON THE PROBLEM OF LIQUID ENTRAINMENT

Description: THe problem of liquid entrainmment by a gas phase is considered with the purpose of gathering, presentinf, and correlating available information which may be useful for predicting the liquid carry-over in a boiling water reactor at high pressure. Correlations which predict water carryover as a function of pressure, of the height of the vapor dome, and of the vapor mass flow rate are presented and discussed. Because entrainment depends upon the hydrodynamic conditions and upon the expansion of a… more
Date: October 1, 1960
Creator: Yeh, G. C.K. & Zuber, N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SHIELDING OF DEMINERALIZERS AND FILTERS IN THE HFIR PRIMARY COOLANT SYSTEM

Description: S> Thicknesses of ordinary concrete required to shield the demineralizers and filters in the HFIR primary water system were computed for normal operating conditions and for abnormal conditions such as a meltdown of the fuel within the reactor. About 4 1/2 ft, 3 1/4 ft, and 4 1/4 ft of concrete are required to shield the cation exchange unit, the anion exchange unit, and the filter unit, respectively, to the most stringent of the following radiation levels: (a) 0.75 mr/hr for normal reactor oper… more
Date: July 25, 1960
Creator: McLain, H A & Haack, L A
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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LAMPRE Project Quarterly Report: August 1960

Description: The LAMPRE-I project is summarized in terms of capsule development, sodium system, LCX-III, gas system, and fuel transfer and storage. The 2-Mw Sodium Test Facility was initially placed in operation on Mar. 15, 1960. Since startup, operation has been continuous except for 12 forced shutdowns required for maintenance of auxiliary steam-system equipment and repair of refractory setting in the gas-fired sodium heater. Essentially no maintenance or adjustment hns been required on sodium components … more
Date: September 1960
Creator: Glasstone, Samuel
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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High Flux Isotope Reactor--a General Description

Description: The High Flax lsotope Reactor (HFIR) is being planned for construction at Oak Ridge National Laboratory as a supporting facility in the program of investigation of the properties of the transplutonium elements. The reactor will be a flux-trap reactor consisting of a berylliumrefiected, light-water-cooled annular fuel region surroundin g a light-water island. An irradiation sample of 200 to 300 g of Pu/sup 242/ will be placed in the island where a thermalneutron flux of approximately 3 x 10/sup … more
Date: March 1, 1960
Creator: Cole, T E
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Shielding of Pipes in the HFIR Primary Coolant System

Description: The thicknesses of ordinary concrete required to shield pipes in the ORNL High Flux Isotope Reactor primary water system were computed for normal operating conditions and for abnormal conditions such as a defective fuel plate or a meltdown of the fuel within the reactor. About 6 ft of concrete is required for the pipes at the outlet of the reactor, and 2 ft of concrete is required for the pipes located about 1 1/2 min downstream from the reactor vessel-. These thicknesses of concrete reduce the… more
Date: June 30, 1960
Creator: McLain, H A & Haack, L A
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Mechanisms and Kinetics of Uranium Corrosion and Uranium Core Fuel Element Ruptures in Water and Steam

Description: The mechanisms and kinetics of uranium corrosion and fuel element ruptures were investigated in water and steam at 170 to 500 deg C and at 100 to 2800 psig. The fuel element samples were coextruded Zircaloy-clad uranium-core rods and tubes which were defected prior to exposure. Uranium corrosion was found to be the sum of two processes; direct oxidation by water, and oxidation of uranium hydride intermediate. Fuel element ruptures occur in two stages; an initial induction period followed by an … more
Date: July 21, 1960
Creator: Troutner, V.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SM-2 REACTOR CORE AND VESSEL REVIEW REPORT FOR PERIOD DECEMBER 15, 1959 TO MARCH 18, 1960

Description: The SM-2 core lifetime was calculated and stuck rod criteria formulated. The control system could shutdown the core under the most adverse conditions with any single rod stuck out. The one-half-inch Eu/sub 2/O/sub 3/ flux supressors adequately supressed the power spike at the bottom of the core. The nuclear effects of increasing the dimensione of the fuel matrix were calculated. A new analysis of various reactivity measuremerts performed durirg the SM-2 critical experiments was made to test the… more
Date: March 30, 1960
Creator: Hoover, H.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Sodium Graphite Reactor: Tommorrow's Power Plant

Description: A description is given of the Advanced Sodium Graphite Reactor Power Plant, including the reactor, heat transfer systems, generatirg plant, control systems, and the economics of producing 256 Mw(e). The safety of this design is due to its unusually low operating pressure, absence of chemically incompatible materials in the core, and excellent stability under atatic and dynamic conditions. The reactor is being constructed at Hallam, Nebraska, at a probable cost of 0 to 0/kw, exclusive of the fir… more
Date: April 25, 1960
Creator: Beeley, R. J.; Lowell, E. G.; Polak, H. & Renard, J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Vibration and Collapse Testing of SM-2 Fuel Elements. Period Covered : January 1959 to January 1960

Description: Experiments were conducted on SM-2 fuel elements to determine the presence of flow-induced fuel plate vibrations, their frequency and amplitude, and to verify the ability of welded fuel elements to resist any tendency towards a fiow-induced channel collapse. Test facilities, instrumentation, and procedures are described in detail. Results of tests showed that vibrations are practically non-existent in SM-2 welded fuel eleinents, welded fuel elements will effectively oppose channel collapse, and… more
Date: July 1, 1960
Creator: Morehouse, D. J. & Christenson, J. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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CONTROL AND DYNAMICS PERFORMANCE OF A SODIUM COOLED REACTOR POWER SYSTEM

Description: In an attempt to prove the feasibility of liquid sodium as a reactor coolant, a design stady of three vital system components was undertaken: the intermediate exchanger, the boiler, and the superheater. Investigation shows that, during transient changes in load, the steam and sodium systems' temperature changes will not significantly exceed the small steady-state temperature changes, provided that the control system mairtains the design steady-state ratio between the flow rates and power level … more
Date: January 1, 1960
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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STUDY OF FUEL TEMPERATURE AND FLOW EFFECTS OF PLUGGING IN SRE FUEL CHANNELS

Description: The fuel surface temperatures and flow rate effects which might be produced by plugs in SRE fuel channels were investigated. It was found that plugs which locally insulate more than 20 to 30% of the surface of a fuel rod are dangerous. Heat transfer through the mcderator would not significantly affect the sensitivity of the outlet thermocouple to change in mass flow through a fuel channel, provided the reactor power is above 2 Mw and the channel is not completely plugged. If a 5% or greater inc… more
Date: March 15, 1960
Creator: Noyes, R.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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RADIOACTIVE WASTE DISPOSAL PROJECTS. IDAHO CHEMICAL PROCESSING PLANT TECHNICAL PROGRESS REPORT FOR JULY-SEPTEMBER 1959

Description: The progress of extensive laboratory and pilot plant investigations on the fluidized bed process for the con version of radioactive liquid wastes to solids is reported. These studies are directed toward obtaining information on the dynamics of fluidized bed operation, the removal of volatile fission products and solids particles from gases, the development of equipment and operating techniques, and the various long term disposal aspects of all radioactive wastes. Laboratory studies on the remov… more
Date: July 1, 1960
Creator: Slansky, C. M.; Warzel, F. M. & Stevens, J. I.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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