Experiment data report for Multirod Burst Test (MRBT) Bundle B-5. [PWR]
A reference source of MRBT bundle B-5 test data is presented with interpretation limited to that necessary to understand pertinent features of the test. Primary objectives of this 8 x 8 multirod burst test were to investigate the effects of array size and rod-to-rod interactions on cladding deformation in the high-alpha-Zircaloy temperature range under simulated light-water reactor loss-of-coolant accident (LOCA) conditions. B-5 test conditions, nominally the same as used in an earlier 4 x 4 (B-3) test, simulated the adiabatic heatup (reheat) phase of an LOCA and were conducive to large deformation. The fuel pin simulators were electrically heated (average linear power generation of 3.0 kW/m) and were slightly cooled with a very low flow (Re approx. 140) of low-pressure superheated steam. The cladding temperature increased from the initial temperature (335/sup 0/C) to the burst temperature at a rate of 9.8/sup 0/C/s. The simulators burst in a very narrow temperature range, with an average of 768/sup 0/C. Cladding burst strain ranged from 32% to 95%, with an average of 61%. Volumetric expansion over the heated length of the cladding ranged from 35% to 79%, with an average of 52%. The results clearly show deformation was greater in the bundle interior and suggest rod-to-rod mechanical interactions caused axial propagation of the deformation.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 6517908
- Report Number(s):
- NUREG/CR-3459; ORNL/TM-8889; ON: TI84015735
- Resource Relation:
- Other Information: Includes 6 sheets of 48x reduction microfiche
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL ASSEMBLIES
STRAINS
TEST FACILITIES
LOSS OF COOLANT
THERMAL STRESSES
PWR TYPE REACTORS
DEFORMATION
EXPERIMENTAL DATA
FUEL CANS
FUEL ELEMENT FAILURE
FUEL RODS
PRESSURE GRADIENTS
REACTOR SAFETY
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS
ZIRCALOY
ACCIDENTS
ALLOYS
DATA
FUEL ELEMENTS
INFORMATION
NUMERICAL DATA
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled