ROSA-AP600 experiment simulating a steam generator tube rupture transient

Conference · Mon Dec 01 04:00:00 EST 1997

OSTI ID:552162

Nakamura, Hideo ^[1]; Kukita, Yutaka ^[2]; Ghan, L S; Schultz, R R ^[3]

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1. Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)
2. Nagoya Univ., Aichi-ken (Japan)
3. Lockheed Idaho Technologies Company, Idaho Falls, ID (United States)

Thermal hydraulic response of the Westinghouse AP600 reactor to a postulated steam generator tube rupture (SGTR) transient was investigated experimentally using the ROSAN Large Scale Test Facility (LSTF) which had been modified to simulate the AP600 reactor with a volume ratio of 1/30.5. The experiment simulated a simultaneous double-ended rupture of multiple U-tubes in one of two SGs. The AP600 passive safety features, passive residual heat removal (PRHR) and core makeup tanks (CMTs), were found to have an adequate heat removal capacity to keep the core filled with subcooled coolant without any operator actions. The primary-to-secondary heat transfer became reversed in the broken SG because of a rapid decrease in hot leg temperature due to the passive heat removal. The SG relief valves did not open after isolation of the SGs, for either broken or intact SG. Significant thermal-stratification, however, was present in the cold leg that includes the low-temperature PRHR return flow. 7 refs., 12 figs.

• Research Organization: American Nuclear Society, La Grange Park, IL (United States)
• OSTI ID: 552162
• Report Number(s): CONF-970607--Vol.2
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
LOSS OF COOLANT
PWR TYPE REACTORS
REACTOR SAFETY EXPERIMENTS
SECONDARY COOLANT CIRCUITS