Modern BWR core designs are characterized by an increased usage of Partial Length Rod fuel assemblies, which have been designed principally to enhance the shutdown margins and the stability performance. The increased moderator/fuel ratio induces less negative isothermal temperature coefficients (ITC) that can become positive at cold conditions. This can become a matter of concern during nuclear heating if criticality is reached late in the rod withdrawal sequence due to some non-negligible residual xenon poisoning or at end-of-cycle (EOC). This was recently experienced in a Swiss BWR when startup was made at EOC following a short shutdown period for maintenance. A small uncontrolled power transient occurred just after that criticality had been reached due to a positive ITC combined with a deactivation of one subsystem of the plant residual heat removal system. Because the validation basis of modern three-dimensional (3-D) core neutronic/thermal-hydraulic transient codes at such conditions is rather scarce, it was considered as an appropriate case to analyze with the SIMULATE-3K code that is currently being established as principal 3-D kinetic solver at PSI. The results presented in this paper show that the qualitative behavior of the transient can be reasonably captured while the quantitative agreement with plant data is found to be highly sensitive upon the ITC magnitude. Small uncertainties in that coefficient are sufficient to affect completely the simulation quality and in that context, an adequate upstream steady-state 2-D lattice / 3-D reactor analysis methodology is shown to play a major role. During the transient, the thermal-hydraulic modeling becomes in addition very important in order to adequately capture the formation of void, recalling the low pressure conditions, as this strongly affects the subsequent evolution of the reactivity coefficients and thereby, the core dynamical response.