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Conference Article: Strength analysis code for graphite structural components in uranium-graphite nuclear reactors

Rodchenkov, B.S.; Platonov, P.A.; Manevskij, V.N.; Kashirin, B.A.; Chugunov, O.K. (Research and Development Inst. of Power Engineering (Russian Federation))

Abstract

The Strength Analysis Code for graphite components used in uranium-graphite channel-type reactors has been developed in the USSR on the basis of results of reactor graphite property investigations including its properties under neutron irradiation conditions as well as on the operating experience of graphite structural components in the uranium-graphite reactors. The graphite stack forms the basis for the uranium-graphite reactor design. The strength analysis of the graphite components are performed following the physics and thermal physics calculations needed to assess thermal and radiation stresses in the analyzed components as well as following the account of variations in physics and mechanical properties of graphite under neutron irradiation and temperature. Strength analysis principles in the code are based on limiting states determined for achievement of critical fluence, crack formation under static loading, accumulation of ultimate strain, and initiation of through crack under static and cyclic loading. Each of these is discussed in the paper.

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key words: crack propagation; lwgr type reactors; mathematical models; mechanical properties; physical radiation effects; reactor components; stress analysis; thermal stresses; graphite moderated reactors; radiation effects; reactors; stresses; water cooled reactors
Reference:
Specialists' meeting on the status of graphite development for gas cooled reactors. Tokai, Ibaraki (Japan). 9-12 Sep 1991
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--690, pp:225-230