Testing and analysis of the Semiscale Mod-1 heater rod design Metadata

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  • Main Title Testing and analysis of the Semiscale Mod-1 heater rod design


  • Author: Larson, T.K.
    Creator Type: Personal


  • Name: Idaho National Engineering Laboratory
    Place of Publication: Idaho Falls, Idaho
    Additional Info: Idaho National Engineering Lab., Idaho Falls (USA)


  • Creation: 1980-01-01


  • English


  • Content Description: The use of electrically heated nuclear fuel rod simulators in the Semiscale Program is traced from a historical viewpoint. The design of the Semiscale Mod-1 electrical heater rod and core simulator is discussed. Heater rod thermal response during transient thermal-hydraulic depressurization experiments conducted in the Mod-1 system, and analysis techniques and tests conducted to help quantify heater rod characteristics and behavior are presented.
  • Physical Description: Pages: 32


  • Keyword: Heating 220900* -- Nuclear Reactor Technology-- Reactor Safety
  • Keyword: Simulation
  • STI Subject Categories: 22 General Studies Of Nuclear Reactors
  • Keyword: Heaters
  • Keyword: Electric Heating
  • Keyword: Reactor Safety Experiments


  • Conference: International symposium on fuel rod simulators-development and application, Gatlinburg, TN, USA, 22 Oct 1980


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


  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Article


  • Text


  • Other: TI85004761
  • Report No.: CONF-801091-3
  • Grant Number: AC07-76ID01570
  • Office of Scientific & Technical Information Report Number: 6029080
  • Archival Resource Key: ark:/67531/metadc1102877


  • Display Note: NTIS, PC A03/MF A01 - GPO.