Large Sample Neutron Activation Analysis (LSNAA) was applied to perform non-destructive elemental analysis of a ceramic vase. Appropriate neutron self-shielding and gamma ray detection efficiency calibration factors were derived using Monte Carlo code MCNP5. The results of LSNAA were compared against Instrumental Neutron Activation Analysis (INAA) results and a satisfactory agreement between th...

Radon, Rn, is radioactive noble gas which decays by alpha emission with half-life of 3.825 d. Its short-lived progeny, Po, Pb and Bi (Po), are alpha and beta radioactive and they emit gamma radiation as well. Radon progeny can be inhaled by humans where they deposit on the inner layers of bronchi, and bronchioles. Particles (alpha, beta and gamma) emitted in radioactive decay damage surrounding...

Criticality safety reanalyzes of the NPP Krško spent fuel pool has been performed. Limiting criticality cases at optimal moderation conditions are calculated using different SCALE code versions (SCALE-4.4a, SCALE-5.1 and SCALE-6.1) with 44-group ENDF/B-V 238-group ENDF/B-VII.0 derived cross section libraries. Relative small discrepancies observed. Additional verification is provided SERPENT MCN...

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by Generation IV International Forum (GIF) for next generation innovative nuclear energy systems. It was among a group more than 100 prototypes and his commercial availability expected 2030. GFR has common goals as rest GIF advanced types: economy, safety, proliferation resistance, sustainability. Several fuel design such...

The deuterium-tritium neutron generator should be fully shielded for the safety of operators participating in experiments since D-T is commonly used activation experiments. In this study, MCNP5 code was to simulate shielding effect thermalization device previously designed by our group with Pb and boron-containing polyethylene as material. dose rate outside previous can not meet requirement, so...

in this research, shielding and criticality safety calculations carried out for interim storage and transportation cask in the tehran research reactor. such dual purpose cask is being designed to the spent fuel elements of research reactors. the monte carlo mcnp5 code calculation was utilized for the criticality safety analysis and origen2.1code was used for shielding calculation. according to ...