A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis
Name
MIT-EL-79-018-06526448.pdf
Size
5.05 MB
Format
Adobe PDF
Checksum (MD5)
c20458f69853ae3d125db0136434db72
Author(s)
Kazimi, Mujid S.
Massoud, Mahmoud
Date Issued
February 1980
Publisher
MIT Energy Laboratory
Series/Report no.
MIT-EL
79-018
Abstract
A review is made of the computer codes developed in the
U.S. for thermal-hydraulic analysis of nuclear reactors. The
intention of this review is to compare these codes on the
basis of their numerical method and physical models with
particular attention to the two-phase flow and heat transfer
characteristics. A chronology of the most documented codes
such as COBRA and RELAP is given. The features of the recent
codes as RETRAN, TRAC and THERMIT are also reviewed. The
range of application as well as limitations of the various
codes are discussed.
Subjects
Nuclear fuel elements |x Computer programs.
Two-phase flow.
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