Results of TRACG analysis of SBWR transients, including model uncertainty

Conference · Thu Dec 31 23:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6960050

Marquino, W; Andersen, J G.M. ^[1]

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1. General Electric Co., San Jose, CA (United States)

Computer codes for boiling water reactor (BWR) transient analysis have relied on full-scale tests in a jet-pump reactor for validation and qualification, and also for determination of model uncertainty. The simplified BWR (SBWR) is a natural-circulation reactor design with significant differences from the plant for which full-scale tests are available. The objective of this work is to determine the uncertainty in applying the TRACG computer code through a procedure similar to the code scaling, applicability, and uncertainty (CSAU) methodology.

• OSTI ID: 6960050
• Report Number(s): CONF-931160--
• Journal Information: Transactions of the American Nuclear Society; (United States), Journal Name: Transactions of the American Nuclear Society; (United States) Vol. 69; ISSN TANSAO; ISSN 0003-018X
• Country of Publication: United States
• Language: English

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CONVECTION
DATA COVARIANCES
ENERGY TRANSFER
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TESTING
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TRANSIENTS
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