High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, July 1-September 30, 1981

Technical Report · Thu Dec 31 23:00:00 EST 1981

OSTI ID:5330337

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

Development work continued on the accident dynamics simulation codes ORTAP, BLAST, and ORECA for the Fort St. Vrain (FSV) reactor. New steam line and main steam bypass system models were developed and incorporated into ORTAP. An initial simulation of the FSV prestressed concrete reactor vessel and liner cooling system was developed and tested for use in the severe accident sequence analysis task.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5330337
• Report Number(s): NUREG/CR-2221-Vol.3; ORNL/TM-8128; ON: DE82005654
• 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
BUILDING MATERIALS
COMPUTER CALCULATIONS
CONCRETES
CONTAINERS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HTGR TYPE REACTORS
MATERIALS
MATHEMATICAL MODELS
PRESSURE VESSELS
PRESTRESSED CONCRETE
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
THERMAL STRESSES