Multirod burst test program. Quarterly progress report, October--December 1975. [BWR; PWR]
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.
- Research Organization:
- Oak Ridge National Lab., Tenn. (USA)
- Sponsoring Organization:
- ERDA
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 7359468
- Report Number(s):
- ORNL/NUREG/TM-10
- Country of Publication:
- United States
- Language:
- English
Similar Records
Multirod burst test program quarterly progress report, July--September 1976. [BWR; PWR]
Preliminary Multirod Burst Test Program results and implications of interest to reactor safety evaluations
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Technical Report
·
Mon Jan 17 23:00:00 EST 1977
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OSTI ID:7311013
Preliminary Multirod Burst Test Program results and implications of interest to reactor safety evaluations
Conference
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Sat Dec 31 23:00:00 EST 1977
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OSTI ID:6147077
Boundary effects on Zircaloy-4 cladding deformation in LOCA simulation tests. [PWR; BWR]
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Thu Dec 31 23:00:00 EST 1981
·
OSTI ID:5275056
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
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210200 -- Power Reactors
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LOSS OF COOLANT
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WATER MODERATED REACTORS
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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