A method is developed for calculating the temperature coefficient of ηf for heterogeneous reactor lattice cells on a fairly rigorous basis, using only microscopic material constants as input data. The method is based on the integral transport equation, and involves flux and adjoint weighting the temperatures derivatives of the kernels of the integral operators. Temperature coefficients are obtained for a localized temperature increase, as well as for a uniform increase in cell temperature. The coefficients are separated, on physical grounds, into ‘spectrum’ and ‘transport’ effects. The numerical accuracy of the method is found to be limited, at the present time, by the uncertainties in fuel reaction cross sections. The method is used in a brief survey of temperature effects in natural-uranium/graphite lattices. The transport temperature coefficients are shown to yield the dependence of the thermal multiplication factor on a velocity-averaged diffusion coefficient. The spectrum temperature coefficients give the dependence of the thermal multiplication factor on average neutron velocity and disadvantage factor. Non-diffusion effects are noticed when the region near the fuel is heated. The results of the method are compared with published experimental results for natural-uranium/graphite lattices. Good agreement between theory and experiment is obtained. The influence of reactor operating conditions on temperature coefficients is reproduced by the theory.