Global analysis of bundle behavior in pressurized water reactor specific CORA experiments

Journal Article · Thu Apr 01 04:00:00 EST 1993 · Nuclear Technology; (United States)

OSTI ID:6629638

Hering, W ^[1]; Minato, Kazuo; Nagase, Fumihisa ^[2]

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1. Inst. fuer Kernenergetik und Energiesysteme Pfaffenwaldring, Stuttgart (Germany)
2. Japan Atomic Energy Research Inst., Tokai-Mura (Japan)

At Kernforschungszentrum Karlsruhe, out-of-pile bundle experiments are performed in the CORA facility to investigate the behavior of light water reactor fuel elements during severe fuel damage accidents. To analyze the phenomena observed during the tests, such as claddin failure, oxidation, and deformation, as well as their influence on the post test bundle state, four pressurized water reactor specific tests are selected: CORA-2, CORA-3, CORA-5, and CORA-12. From each of these tests, a detailed global analysis using all the measured temperatures, pressures, and fluid compositions as well as videoscope information has been performed. To describe the post test bundle state quantitatively, axial profiles of the bundle cross-section area, the damage state of the rods, the average cladding oxidation, and the damage to the pellets are measured. The effects of CORA-specific components on the bundle melt progression and the measured axial profiles are identified and assessed. Most of the observations during the tests as well as the post test bundle state can be explained by the established common sequence of phenomena. For a better understanding of the melt progression, some physical phenomena, such as the energy release associated with the double-sided oxidation of the cladding, the melt release, or the melt relocation, must be analyzed in detail.

• OSTI ID: 6629638
• Journal Information: Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 102:1; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

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