A Data Development Project and a Coordinated Research Project of the International Atomic Energy Agency (IAEA) aimed at updating the library of dosimetry cross sections International Reactor Dosimetry File: IRDFF are underway. To validate the new evaluations the thermal neutron capture cross-sections were first compared against the Keyzero and Mughabghab values as well as against the previous IRDFF-2002 and -90 data. Furthermore, a series of shielding benchmarks available from the SINBAD database were used to check and validate the new IRDFF dosimetry file, version v1-04 (updated 6Li(n,t) data were taken from v1.05). Several benchmark experiments performed at the Frascati Neutron Generator (FNG), ENEA Frascati, and in ASPIS, AEA Technology, Winfrith were analysed. The main purpose of repeating the calculations with the new dosimetry cross sections was to check for any improvement between measured and calculated reaction rates (compared to IRDF-2002 as well as IRDF-90) and removal of some inconsistent trends in the results for different monitors. Since the dosimetry data represent a relatively small part of the overall uncertainty, the major part coming from the transport cross section and model approximations, these results can be considered as an indirect validation of the new IRDFF dosimetry library. The compensation of errors between the transport cross sections and dosimetry data is likely. To obtain additional information potentially useful to conclude on the impact of transport and dosimetry cross-section uncertainties and their compensation, as well as on the computer code modelling uncertainties, the results using different transport cross-sections and computer codes (DOORS and MCNP) are presented for several benchmark analyses.