Post-test analysis of the BTF-104 severe-fuel-damage experiment using the CATHENA thermal-hydraulics code

Book · Mon Jul 01 04:00:00 EDT 1996

OSTI ID:250989

Popov, N K; Walters, L C; Tran, T V ^[1]; DeVaal, J W; Dickson, L W; Irish, J D ^[2]

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1. Atomic Energy of Canada Ltd., Pinawa, Manitoba (Canada)
2. Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada). Chalk River Labs.

In the BTF-104 (Blowdown Test Facility) experiment a single fully-instrumented, CANDU{reg_sign}-type fuel element was subjected to conditions representative of a Loss-Of-Coolant Accident (LOCA) with additional Loss-Of-Emergency-Coolant Injection (LOECI). Depressurization was followed by a period of degraded steam cooling achieving fuel temperatures of the order of 1800-1900 C. Material and coolant temperatures, and fission-product releases were monitored. The principal objective of the BTF-104 experiment was to determine the timing and amount of short-lived fission product release from the fuel element at the target fuel temperature. Information generated from BTF experiments is used by the CANDU Owners Group (a consortium of AECL and Canadian nuclear utilities) for validating computer codes used in licensing CANDU reactors. Simulations of the BTF-104 experiment were conducted using the CATHENA code (Canadian Algorithm for THErmalhydraulic Network Analysis) to provide a better understanding of the thermal-hydraulic phenomena observed in the experiment. This paper summarizes the major events and conditions of the BTF-104 experiment, describes the CATHENA idealization of the BTF-104 fuel stringer and BTF loop, and discusses important results obtained by these simulations.

• OSTI ID: 250989
• Report Number(s): CONF-960306--; ISBN 0-7918-1226-X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
CANDU TYPE REACTORS
COMPUTERIZED SIMULATION
DAMAGE
ECCS
EXPERIMENTAL DATA
FISSION PRODUCT RELEASE
FUEL ELEMENTS
HEAT TRANSFER
LOSS OF COOLANT
LOSSES
PRIMARY COOLANT CIRCUITS
REACTOR LICENSING
REACTOR SAFETY EXPERIMENTS
TEST FACILITIES
THEORETICAL DATA