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A Conceptual Design of a Thorium-Uranium (233) Power Breeder Reactor

Description: From abstract: A conceptual design study has been performed for a sodium cooled, graphite moderated, thermal power-breeder reactor utilizing the Thorium-Uranium 233 breeding cycle. Several aspects of the design of the system are considered but no attempt has been made to supply all the details. It appears that the design presented is feasible and will allow the production of economic power as well as full utilization of thorium resources.
Date: February 1, 1954
Creator: Henrie, J. O. & Weisner, E. F.
Partner: UNT Libraries Government Documents Department

A Reversing Logarithmic DC Amplifier

Description: Purpose: Automatic recording equipment was designed for use with a high temperature Sykes experiment in which calorimetric measurements were to be made to temperatures approaching 2000* C. At such high temperatures, radiation becomes the dominant mechanism for heat transfer. The temperature differences which are used to determine the magnitude of this transfer no longer are directly proportional to it, but must be related by the Stefan-Boltzman law of radiation.
Date: January 1, 1954
Creator: Carter, R. L.
Partner: UNT Libraries Government Documents Department

A Remotely Controlled Welding Device for Joining Stainless Steel Tubes

Description: Abstract: The design and testing of experimental equipment for remotely joining stainless steel tubing by heliarc welding is described. This apparatus consists of a modified heliarc welding torch which is hydraulically controlled to maintain constant arc voltage. A suitable arc voltage sensing and control amplifier circuit was developed for this application.
Date: November 15, 1954
Creator: Mueller, Martin & Hecker, Eugene
Partner: UNT Libraries Government Documents Department

Sodium Graphite Reactor Quarterly Progress Report for July-September, 1954

Description: Reactivity calculations have been performed for the steady-state Pu feedback technique outlined in the previous progress report. A full-scale power plant study was initiated, based on sodium-graprite technology. A twin-core power plant is now considered to be the most promising configuration. Several design drawings are given of such a reactor, using slightly enriched U to produce Pu amd electrical power. The thermal neutron flux distribution in a cluster of 6 U rods was measured, and the results are compared with previous measurements for 7 rod cluster. The average thermal cycling of hollow U slug elements was begun. Results are given for 500 cycles between 100 and 500 deg C. A series of powder- compacted U alloys were thermal cycled between 200 and 700 deg C. Data on the transfur of radioactivity from Zr by Na has been obtained from a capsule of the first series of three miniharps. Fe, Al, and Cu were immersed in toluene end irradiated at 150 deg F in the MTR-Gamma canal. Toluene is being considered as a shield coolant for the SRE. The effect of 1-Mev electron irradlation on terphenyls was also studied. A venting tube arrangement has been designed for the Zr-canned graphite moderator. A number of thermal insulating brick amd fiber materials were sublected to liquid Na to study deterioration effects. The materials tested were JohnsManville Brick C-16 (Sil-O-Cel mortar), Superex Paste, and Eagle-Pitcher Mineral Wool. Encouraging results were obtained in an efiort to evaluate the effectiveness of Na decontamination by liquid ammonia. Pressure drop and flow characteristics of the latest design SRE fuel element have been completed. Design details of the 2-speed control rod drive assembly are given. Other aspects of the reactor control system, including design and component fabrication, are discussed. Gamma dose rates at the surface of the top shield ...
Date: December 1, 1954
Creator: Siegel, S. & Inman, G. M.
Partner: UNT Libraries Government Documents Department

A Sodium-Graphite Reactor Steam-Electric Station for 75 Megawatts Net Generation

Description: The major design features, nuclear characteristics and performance data for a nuclear fueled central station power plant of 75,000 kw net capacity are presented. The heat source is a Na cooled graphite moderated reactor. The design of the reactor takes full advantage of the experience gained to date on the Sodium Reactor Experiment (SRE); the plant described here is a straightforward extension of the smaller experimental SRE, which is now under construction. The fuel elements are made up of rod clusters and the moderator is in the form of Zr canned graphite elements. The performance of the reactor has been based on conservative temperatures and coolant flow velocities which result in a plant with "built-in reserve." Thus, as experience is gained and anticipated improvements in reactor fuel elements and construction materials are proven, the performance of the plant can be increased accordingly. Two reactor designs are described, one for operation with slightly enriched U fuel elements and the other for operation with Th--U fuel elements. The associated heat exchangers, pumps, steam, and electrical generating equipment are identical for either reactor design. An analysis of turbine cycles describes the particular cycle chosen for initial operation and discusses a method by which modern central station performance can be initially obtained. The design and performance data which are required to enable reliable estimates of the plant construction and operating costs to be made are established. (auth)
Date: March 22, 1955
Creator: Weisner, E. F. & Sybert, W. M.
Partner: UNT Libraries Government Documents Department

Transient Pressure in the Water Boiler Kinetic Experiment

Description: The results of theoretical studies on the effect of inertial pressures in water boiler transients are summarized as follows: (1) derivation of the equation for transient pressures in a partially enclosed container of water, (2) development of the equation of state for a mixture of water and radiolytic decomposition gas, (3) calculation of peak power as a function of reactor period, and (4) discussion of the effect of limited void space above the solution. It is intended that these results should serve as an indication of the magnitude of the inertial effects as a guide for planning the experimental program for KEWB. It is concluded that the attainable pressures represent no hazard to equipment, personnel or environs. (auth)
Date: March 18, 1955
Creator: Hetrick, D. L. & Remley, M. E.
Partner: UNT Libraries Government Documents Department

Use of a Water Boiler Reactor as a Production Test Facility

Description: The feasibility of the use of a water-boiler type reactor as a production test facility for making reactivity tests on various forms of uranium and uranium compounds was investigated. It is concluded that the reactor should be very useful for production testing of materials with the danger coefficient techniqus. A suggested production procedure is outlined. (auth)
Date: August 25, 1953
Creator: Remley, M. E.
Partner: UNT Libraries Government Documents Department