Hypothetical accident scenario analyses for a 250-MW(T) modular high temperature gas-cooled reactor

Conference · Tue Jan 01 04:00:00 EST 1985

OSTI ID:5537814

Harrington, R M; Ball, S J; Cleveland, J C

This paper describes calculations performed at Oak Ridge National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission's HTGR Safety Research Program, to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e. upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced helium primary coolant circulation, loss of primary coolant pressurization, and loss of heat sink were studied and are discussed. Comparisons of calculated and measured response for a flow reduction test on the German reactor AVR are also presented.

• DOE Contract Number: AC05-84OR21400
• OSTI ID: 5537814
• Report Number(s): CONF-851125-
• Country of Publication: United States
• Language: English

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