Subchannel thermal-hydraulic analysis at AP600 low-flow steam-line-break conditions

Journal Article · Fri Dec 01 04:00:00 EST 1995 · Nuclear Technology

OSTI ID:170261

Morita, T; Olson, C A; Sung, Y X; Connelley, Jr, J F; Novendstern, E H; Kapil, S; Rosenthal, P W ^[1]

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1. Westinghouse Energy Systems, Pittsburgh, PA (United States)

The AP600 reactor core approaches buoyancy-dominated flow at the departure from nucleate boiling (DNB)-limiting period of a postulated steam-line--break accident. The reactor core has a highly skewed power distribution at this time due to the conservative assumption of a withdrawn rod cluster control assembly (stuck rod). Under such conditions, strong buoyancy-induced core cross flow occurs, and coupled nuclear and thermal-hydraulic interactions become important. To analyze the transient, Westinghouse Electric Corporation has coupled THINC-IV with a neutronic code (ANC). Applicability of the THINC-IV subchannel code to the low-flow conditions with a steep radial power gradient is verified with existing rod bundle test results. The code predictions are in excellent agreement with the test data. The coupled codes provide a realistic three-dimensional simulation of core power by considering core flow distributions and the resultant enthalpy distributions in neutronic feedback. The safety analysis using the coupled code demonstrates that the DNB design basis is met during the postulated steam-line-break accident.

• OSTI ID: 170261
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 3 Vol. 112; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
CROSSFLOW SYSTEMS
FUEL CHANNELS
FUEL ELEMENT CLUSTERS
HEAT TRANSFER
HYDRAULICS
POWER DISTRIBUTION
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR CORES
SAFETY ANALYSIS
T CODES