Power-Cooling-Mismatch test series: PCM-1 experiment predictions. [PWR]

Technical Report · Thu Dec 01 04:00:00 EST 1977

OSTI ID:5241170

Stanley, C J; Domenico, W F

The objective of this single rod (0.914 m long) test is to extend the current data base on fuel rod behavior to include rod failure at power, under stable film boiling conditions. The operating conditions of the experiment have been specified such that a significant amount of fuel is expected to be molten when the rod fails. This large amount of molten fuel is expected to provide the possibility of molten fuel-coolant interaction (MFCI). The design analysis conducted for the PCM-1 experiment involved the determination of the rod power and coolant conditions that would result in significant fuel melting with cladding temperatures in the high ..beta..-phase, but below the zircaloy melting point. The results of this design analysis indicated that the following conditions should exist in the experiment: peak fuel rod power) 78.7 kW/m; rod internal gas mixture: 77.7% helium, 22.3% argon; coolant mass flux: 1356 kg/s . m/sup 2/; and coolant inlet temperature 600 K. For these conditions FRAP-T3 calculations predicted a maximum fuel centerline temperature of 3668 K at the axial hot-spot elevation (0.41 m) with approximately 60% of the fuel pellet radius being in the molten state. The maximum cladding temperature at the same elevation was predicted to be 1673 K. Cladding collapse was predicted for the fuel rod above 0.31 m with a high probability of clad failure during the stabilized film boiling portion of the test. Melting of the cladding was not indicated; and therefore, rod failure is expected to be initiated by oxygen embrittlement of the cladding in the vicinity of the axial hot spot.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 5241170
• Report Number(s): TFBP-TR-242
• Country of Publication: United States
• Language: English

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220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
FUEL ELEMENT FAILURE
MELTDOWN
POWER-COOLING-MISMATCH ACCIDENTS
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REACTOR ACCIDENTS
REACTORS
SIMULATION
TEMPERATURE DISTRIBUTION
WATER COOLED REACTORS
WATER MODERATED REACTORS