Actions for Multirod burst test program. Quarterly progress report, October--December 1975. [BWR; PWR].
Multirod burst test program. Quarterly progress report, October--December 1975. [BWR; PWR].
- Author
- Chapman, R. H.
- Published
- United States : [publisher not identified], 1976.
[Oak Ridge, Tennessee] : [U.S. Atomic Energy Commission], 1976. - Physical Description
- microfiche : negative ; 11 x 15 cm
- Summary
- Internally pressurized unirradiated Zircaloy tubes containing tubular electric heaters (to simulate nuclear fuel pellet heating) will be tested to failure in a low-pressure superheated-steam environment. These assemblies will be heated over a 915-mm (approximately 36-in.) length at a constant rate of 28/sup 0/C/sec (50/sup 0/F); differential pressures will range from about 700 to 14,000 kPa (100 to 2000 psi), corresponding to approximate rupture temperatures from 1200 to 700/sup 0/C (2200 to 1300/sup 0/F). In addition to measurements of cladding surface temperature and internal pressure during the transient test, data will be obtained on pre- and post-test flow resistance (for the multirod arrays) and on deformation, rupture strain, and channel blockage (as measured by sectioning of tubes and tube bundles). Progress made on design and construction of components and systems, development tests and evaluations, and procurement of long delivery items are summarized. Preliminary results of single-rod tests are presented.
- Report Numbers
- ORNL/NUREG/TM-10
- Other Subject(s)
- 21 specific nuclear reactors and associated plants
- 210100 - power reactors, nonbreeding, light-water moderated, boiling water cooled
- 210200 - power reactors, nonbreeding, light-water moderated, nonboiling water cooled
- 22 general studies of nuclear reactors
- 220900 - nuclear reactor technology- reactor safety
- Accidents
- Alloys
- Bwr type reactors
- Deformation
- Fuel cans
- Fuel element failure
- Loss of coolant
- Pwr type reactors
- Reactor accidents
- Reactors
- Research programs
- Simulation
- Test facilities
- Tin alloys
- Transients
- Water cooled reactors
- Water moderated reactors
- Zircaloy
- Zirconium alloys
- Zirconium base alloys
- Collection
- U.S. Atomic Energy Commission depository collection.
- Note
- DOE contract number: W-7405-ENG-26
OSTI Identifier 7359468
Research organization: Oak Ridge National Lab., Tenn. (USA).
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