At present, OpenMC cannot tally dose responses directly. To get dose, you’d have to come up with your own flux-to-dose conversion factor and then apply it to the tallied flux (possibly as a function of energy) as a post-processing step. Another thing to keep in mind is that OpenMC does not handle photon or electron transport yet, so you will only be able to determine dose due to neutrons.
In the future, please post questions directly to the user’s group email (email@example.com). I receive all the messages there as do many of our other developers/users, so if I can’t reply promptly, someone else may be able to assist you.