by Akaho, E.H.K.; Anim-Sampong, S. (Ghana Atomic Energy Commission, Accra (Ghana))
A two-group macroscopic cross-section data base has been generated using a simplified model for the Miniature Neutron Source Reactor (MNSR). The correction term to the basic macroscopic cross-sections due to varying effects of temperature along the axis of the fuel and burnup were expressed in the form of poly-nomials. The error analysis that the database is accurate enough to be used for neutronic calculations for the reactor. (author).
Source/Report Journal of the Kumasi University of Science and Technology; v. 14(1); ISSN 0855-0395; CODEN JKUTDP; Feb 1994; p. 40-47; 9 refs.; 2 tabs.; 19 figs
Publ. Year 1994