Investigations of postulated accident sequences for the Fort St. Vrain HTGR

Conference · Sat Dec 31 23:00:00 EST 1977

OSTI ID:6386052

Ball, S. J.; Cleveland, J. C.; Conklin, J. C.; Hatta, M.; Sanders, J. P.

The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 6386052
• Report Number(s): CONF-781144-3
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210300 -- Power Reactors
Nonbreeding
Graphite Moderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CALCULATIONS
ENRICHED URANIUM REACTORS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
VRAIN REACTOR