Published March 1982
| Version v1
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In-reactor fatigue crack propagation testing
Description
An in-reactor fatigue experiment was conducted in the Oak Ridge Research Reactor to investigate the effects of dynamic irradiation on fatigue crack propagation. Based on the desired test parameters and the reactor interface constraints, a pneumatically actuated fatigue machine was developed. The apparatus was designed to cycle a specimen train in the tension-tension mode between specific load limits at a frequency of 0.0167 s-1 (1 cycle/min). The test environment consisted of sodium at a temperature of 4250C
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS., PC A02/MF A01 as DE82013972.
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Additional details
Publishing Information
- Imprint Pagination
- 10 p.
- Report number
- HEDL-SA--2671-FP
Conference
- Title
- Topical meeting on fast, thermal and fusion reactor experiments.
- Dates
- 12 - 15 Apr 1982.
- Place
- Salt Lake City, UT (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 14725166
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CRACK PROPAGATION; FATIGUE; HIGH TEMPERATURE; ORR REACTOR; REACTOR EXPERIMENTAL FACILITIE; REACTOR INSTRUMENTATION; SODIUM
- Descriptors DEC
- ALKALI METALS; ELEMENTS; ENRICHED URANIUM REACTORS; MECHANICAL PROPERTIES; METALS; REACTOR COMPONENTS; REACTORS; TANK TYPE REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-820406--3.