Experiment data report for Multirod Burst Test (MRBT) Bundle B-5. [PWR]

Technical Report · Wed Aug 01 04:00:00 EDT 1984

DOI:https://doi.org/10.2172/6517908· OSTI ID:6517908

Chapman, R H; Crowley, J L; Longest, A W

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., TN (USA)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6517908
• Report Number(s): NUREG/CR-3459; ORNL/TM-8889; ON: TI84015735
• Country of Publication: United States
• Language: English

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ACCIDENTS
ALLOYS
DATA
DEFORMATION
EXPERIMENTAL DATA
FUEL ASSEMBLIES
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
INFORMATION
LOSS OF COOLANT
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRAINS
STRESSES
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS
TEST FACILITIES
THERMAL STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS