Investigations of postulated accident sequences for the Fort St. Vrain HTGR
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; TRN: 79-005133
- Resource Relation:
- Conference: 2. US-Japan seminar on HTGR safety technology, Fuji, Japan, 24 Nov 1978
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
VRAIN REACTOR
REACTOR ACCIDENTS
COMPUTER CALCULATIONS
REACTOR SAFETY
TEMPERATURE GRADIENTS
ACCIDENTS
ENRICHED URANIUM REACTORS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
POWER REACTORS
REACTORS
SAFETY
220900* - Nuclear Reactor Technology- Reactor Safety
210300 - Power Reactors
Nonbreeding
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