Open AccessStructural thermal tests on Advanced Neutron Source reactor fuel plates
Author(s) -
W. F. Swinson,
R.L. Battiste,
G.T. Yahr
Publication year - 1995
Publication title -
osti oai (u.s. department of energy office of scientific and technical information)
Language(s) - English
Resource type - Reports
DOI - 10.2172/212588
Subject(s) - coolant , materials science , nuclear engineering , stress (linguistics) , thermal , yield (engineering) , neutron temperature , aluminium , nuclear reactor , neutron , reliability (semiconductor) , composite material , nuclear physics , thermodynamics , engineering , physics , philosophy , linguistics , power (physics)
The thin aluminum-clad fuel plates proposed for the Advanced Neutron Source reactor are stressed by the high-velocity coolant flowing on each side of the plates and by the thermal gradients in the plates. The total stress, composed of the sum of the flow stress and the thermal stress at a point, could be reduced if the thermal loads tend to relax when the stress magnitude approaches the yield stress of the material. The potential of this occurring would be very significant in assessing the structural reliability of the fuel plates and has been investigated through experiment. The results of this investigation are given in this report
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