Abstract
A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.
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