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.