Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

Journal Article · Tue Aug 01 04:00:00 EDT 1978 · Nucl. Technol.; (United States)

OSTI ID:6385689

Albright, D C; Bari, R A

Analyses of primary pipe rupture accidents in the Clinch River Breeder Reactor were carried out with Brookhaven National Laboratory-modified versions of the DEMO code. The thermal transient responses of the core and radial blanket were calculated for a large number of initial conditions and plant configurations. These include studies of variations of pipe break size and operating power. Pipe ruptures commencing from two-loop initial operating conditions are presented. The sensitivity of the thermal response to variations in particular parameters within the estimated ranges of their uncertainties has been studied. Conditions under which sodium boiling in the core occurs are delineated.

• Research Organization: Brookhaven National Lab., Upton, NY
• OSTI ID: 6385689
• Journal Information: Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 39:3; ISSN NUTYB
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BREEDER REACTORS
CLINCH RIVER BREEDER REACTOR
COMPUTER CALCULATIONS
COMPUTER CODES
COOLING SYSTEMS
D CODES
EPITHERMAL REACTORS
FAILURES
FAST REACTORS
FBR TYPE REACTORS
LIQUID METAL COOLED REACTORS
LOSS OF COOLANT
PIPES
POWER REACTORS
PRIMARY COOLANT CIRCUITS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
RUPTURES
SODIUM COOLED REACTORS
TEMPERATURE GRADIENTS