INVESTIGATION OF NEUTRON-INDUCED REACTION CROSS SECTIONS FOR SEVERAL NUCLEAR REACTOR STRUCTURAL MATERIALS
Abstract
Selection of the appropriate materials as a structural components of nuclear reactor are of key importance to implement the highest efficiency and security. Zr, Fe, Cr, Sn and Nb are commonly used materials involved in structural alloy inside the reactor. The presented result of neutron-induced reaction cross section calculations for Cr50-52-53-54, Fe54-56-57, Nb93, Sn117-119-124, Zr90-91-92 have been computed using Constant Temperature Model, Back Shifted Fermi Gas Model, Generalised Super Fluid Model and Microscopic level densities presented in TALYS 1.9 nuclear code. The calculations have been repeated by changing the level density parameter a for each isotope and model to observe changes in the model assumption. The obtained results are compared with the experimental data taken from the literature.
Keywords
Level density models, Level density parameter, TALYS 1.9, Reactor structural components
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