[en] The use of Monte Carlo particle transport codes for reactor fluence/dosimetry calculations is nowadays a state-of-the-art approach. At Paul Scherrer Institut the MCNPX code is used for studies related to LWR vessel and internals neutron exposure. In order to enlarge the in-house validation database as well as assess ability for ex-vessel dosimetry studies, the 'H.B. ROBINSON-2' neutron dosimetry benchmark available from the OECD/NEA SINBAD Database has been analyzed recently. The first results obtained employing the JEFF-3.1 library, indicated a rather good performance of the Monte Carlo based solution. Especially, an improvement was observed for the fission dosimeters located in the reactor cavity, versus the originally reported results obtained with the DORT code. However, the statistical precision of the preliminary MCNPX-2.4.0 modeling, mostly for the (n,p) and (n,α) dosimeters, had to be improved. For that, optimization studies have been performed using some of the standard variance reduction options available with MCNPX code, and the given paper presents the obtained estimates for the dosimeters specific activities C/E values. The work is currently continued with the focus on refinement of some modeling approximations involved in the performed analyses. (authors)