Evaluation of thermal response in Fort St. Vrain reactor primary system to a design basis depressurization accident followed by cooling with two Pelton wheel drives operating at 7000 rpm

Technical Report · Mon Dec 01 00:00:00 EST 1975

DOI:https://doi.org/10.2172/4131039· OSTI ID:4131039

Sanders, J. P.; Turner, W. D.; Ball, S. J.; DeVault, R. M.; Giles, G. E.; Paul, D. D.

Independent computations have been performed at ORNL to determine the thermal and hydraulic response of the primary system of the Fort St. Vrain reactor to a specified design basis depressurization accident (DBDA). For this accident it is stated that two of the four auxiliary drives on the circulators start operating at 7000 rpm after a 5-min delay. The calculations at ORNL supported the values given by the General Atomic Co. for primary system flow and pressure loss, and they confirmed the heat removal capacity of the flooded steam generators. The fuel and outlet helium temperatures calculated by ORNL generally reached higher values during the transient due to the fact that the ORNL calculations were performed using a single-channel calculation that could not include the effect of interregional heat conduction. An ORNL code that is developmental, but which includes interregional conduction, agrees well with the GAC values. The results of these calculations are presented graphically. 18 references. (auth)

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-33-014127
• OSTI ID: 4131039
• Report Number(s): ORNL-TM-5140
• Resource Relation: Other Information: Orig. Receipt Date: 30-JUN-76
• Country of Publication: United States
• Language: English

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220900* -Nuclear Reactor Technology-Reactor Safety
*VRAIN REACTOR- DESIGN BASIS ACCIDENTS
COMPARATIVE EVALUATIONS
COMPUTER CODES
HEAT TRANSFER
HYDRAULICS
PRIMARY COOLANT CIRCUITS
REACTOR SAFETY
STEAM SYSTEMS
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS