A Computer Study of Nuclear Characteristics of the University of Maryland Reactor

Abstract

An analytical study of the University of Maryland Reactor was performed using an IBM 7094 computer. The nuclear analysis codes employed in the study were GAM-I, for fast group cross-section evaluation, Tempest-II, for thermal cross-sections, WANDA, for one-dimensional neutron transport calculations and PDQ, for two-dimensional transport calculations. A computer program, XS-1, for the calculation of effective control rod cross-sections was written as part of this study. The method employed in the determination of fast energy broad group cross-sections for non-fuel regions from that normally used. The neutron leakage from the fuel was used as a source of neutrons in lieu of the fission spectrum. This method was felt to give a better estimate of the neutron flux distribution in energy. The analytical results compared favorably with experimental data. The predicted K- effective was 0.993 versus 1.006 actual. The thermal neutron flux calculated and measured at the center of fuel elements generally agreed within 10%. The calculated neutron flux above 3 Mev, which is more difficult to match, was found to be low by approximately 30%. Control rod worth was predicted to be 5.6% Δρ for the center shim rod and 2.7% Δρ for the outside shim rod. Measured values are reported as 3.6% Δρ and 2.0% Δρ respectively.