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 LaboratoryPlace of Publication: Idaho Falls, IdahoAdditional Info: Idaho National Engineering Lab., Idaho Falls (USA)
- Creation: 1980-01-01
- 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 ReportsCode: OSTI
Name: UNT Libraries Government Documents DepartmentCode: UNTGD
- 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.