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

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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 ... continued below

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Pages: 101

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Siegel, S. & Inman, G. M. December 1, 1954.

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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 were measured, together with the heat generation in the top thermal and biological shields. Revised plans for a fuel element and moderator-cell handling coffin, fuel storage system, and waste disposal system are outlined. (For preceding period see NAA-SR1049.) (K.S.)

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Pages: 101

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NTIS

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  • Other Information: Decl. Mar. 4, 1957. Orig. Receipt Date: 31-DEC-58

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  • Report No.: NAA-SR-1109(Rev.)
  • Grant Number: AT-11-1-GEN-8
  • DOI: 10.2172/4315537 | External Link
  • Office of Scientific & Technical Information Report Number: 4315537
  • Archival Resource Key: ark:/67531/metadc1019383

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Office of Scientific & Technical Information Technical Reports

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Creation Date

  • December 1, 1954

Added to The UNT Digital Library

  • Oct. 15, 2017, 10:09 p.m.

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  • Jan. 22, 2018, 11:36 a.m.

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Siegel, S. & Inman, G. M. Sodium Graphite Reactor Quarterly Progress Report for July-September, 1954, report, December 1, 1954; Downey, California. (digital.library.unt.edu/ark:/67531/metadc1019383/: accessed May 26, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.