H.G. Johnson
Abstract
The CORA code is a steady state/transient, core thermal hydraulics code for the FFTF Reactor. A brief overview of the code development and use is presented.
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The CORA code is a steady state/transient, core thermal hydraulics code for the FFTF Reactor. A brief overview of the code development and use is presented.
Key concepts: Nuclear engineering, Transient (computer programming), Cluster (spacecraft), Nuclear reactor core, Core (optical fiber), Transient analysis, Materials science, Nuclear physics