An experimental and analytical study of the void coefficient of reactivity in the Ford Nuclear Reactor (a fully enriched, swimming pool type) has been completed. A stream of air bubbles was used to introduce voids. Out-of-pile calibration of the air flow system was necessary to account for variation in bubble rise velocity with average air concentration. This method is extremely simple except for the calibration procedure. With the results presented in this paper, the void coefficients of other reactors with similar fuel elements (18 plate, BSR type) can be measured without the necessity for recalibration. For the calculation of uniformly distributed void coefficients, relatively simple two-group diffusion theory is shown to be accurate provided the variation of leakage in all three dimensions is taken into account. This variation of leakage is computed by the use of a buckling iterative procedure. Second order effects, such as the variation of effective thermal neutron temperature and disadvantage factor, may be neglected. For the calculation of localized void effects, the buckling iteration method is inaccurate due to the nonseparability of axial and radial flux distribution in this small core. To improve the accuracy an extension of this method to several region iteration is suggested. The principal value of this type of calculation is the short computer time required.