Annual Progress Report on Fuel Element Development for Fiscal Year 1961 Metadata

Metadata describes a digital item, providing (if known) such information as creator, publisher, contents, size, relationship to other resources, and more. Metadata may also contain "preservation" components that help us to maintain the integrity of digital files over time.

Title

  • Main Title Annual Progress Report on Fuel Element Development for Fiscal Year 1961
  • Series Title Fiscal Year 1961

Creator

  • Author: Gibson, G. W.
    Creator Type: Personal
  • Author: Shupe, O. K.
    Creator Type: Personal

Contributor

  • Sponsor: United States. Department of Energy.
    Contributor Type: Organization
    Contributor Info: USDOE

Publisher

  • Name: Phillips Petroleum Company. Atomic Energy Division.
    Place of Publication: Idaho Falls, Idaho

Date

  • Creation: 1962-03-01

Language

  • English

Description

  • Content Description: Progress in fuels and materials development is summarized. Major areas of investigation include a materials study by means of sample fuel plates containing uranium alloys or cermets, burnable poisons, non-uniform fuel and poison distributions and clad with various aluminum alloys; and an engineering study of fuel element geometries optimized in heat transfer, hydraulics, and materials strength. Up to 45 wt% U-Al alloys, 6 to 65 wt% UO/-Al and U3O6-Al dispersions, including enrichments ranging from 20% to 93%, were tested to 70% burnup in de-ionized water at 200 deg F in the MTR. Their performance at higher temperature is still being investigated. Test results for the MTR conditions indicate that all of the compositions investigated to date will successfully withstand even the longest irradiation at these conditions if properly fabricated. Some high strength aluminum alloy claddings, not yet fully tested, show some peculiar surface effects which may be related to corrosion. Metallographic studies of irradiated cermets reveal a reaction'' (diffusion) zone produced around UO/sub 2/ particles in contact with aluminum. These zones are being studied by means of x-ray diffraction, electron microscopy, and electron microprobe analysis. From engineering studies has come promise of improved heat removal and lower pumping requlrements for reactors through artificial roughening of fuel plates. Computer optimizatlon studies and hydraulic tests indicated 80% improvement in heat transfer or 60% less flow for the same heat load are obtainable for MTR conditions. Heat transfer test results from 0.110 x 2.624 ' electrically-heated channels using heat fluxes up to 2.88 x 10/sup 6/ Btu/hr-ft/ sup 2/, sgree better with correlations based on bulk temperatures than with the more widely used modified Colburn equation. In this range, a modifled Colburn equation with a 20% safety factor, as is presently used, seems adequate. However, an equation based on the bulk coolant temperature could be used employing a smaller safety factor because of its greater accuracy. ( auth)
  • Physical Description: Pages: 112

Subject

  • Keyword: Fuels
  • Keyword: Heat Transfer
  • Keyword: Water Coolant
  • Keyword: Uranium Alloys
  • Keyword: Hydraulics
  • Keyword: Dispersions
  • Keyword: Colburn Equation
  • Keyword: Configuration
  • Keyword: Fabrication
  • Keyword: Equations
  • Keyword: Corrosion
  • Keyword: Diffusion
  • Keyword: Irradiation
  • Keyword: Computers
  • Keyword: Coolant Loops
  • Keyword: High Temperature
  • Keyword: Variations
  • Keyword: Aluminum Alloys
  • Keyword: Mtr
  • Keyword: Canning
  • Keyword: Poisoning
  • Keyword: Distribution
  • Keyword: Enrichment
  • Keyword: Inspection
  • Keyword: Burnup
  • Keyword: Testing
  • Keyword: Electron Microscopy
  • Keyword: Uranium Oxides
  • Keyword: Fuel Elements
  • Keyword: Metallography
  • Keyword: Uranium Dioxide
  • Keyword: Fuel Cans
  • Keyword: Cermets
  • Keyword: Surfaces
  • Keyword: Reactor Technology
  • Keyword: Burnout
  • Keyword: Fluid Flow
  • Keyword: X Radiation
  • Keyword: Zones
  • Keyword: Plates
  • Keyword: Performance

Source

  • Other Information: Orig. Receipt Date: 31-DEC-62

Collection

  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report

Format

  • Text

Identifier

  • Report No.: IDO-16727
  • Grant Number: AT(10-1)-205
  • DOI: 10.2172/4782699
  • Office of Scientific & Technical Information Report Number: 4782699
  • Archival Resource Key: ark:/67531/metadc1053712