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- Title
MCNP APPROACHES FOR DOSE RATES MODELING IN LABORATORY FOR NEUTRON ACTIVATION ANALYSIS AND GAMMA SPECTROMETRY AT OSTRAVA.
- Authors
Uhlář, Radim; Alexa, Petr
- Abstract
The Laboratory for Neutron Activation Analysis and Gamma Spectrometry at the VŠB-Technical University of Ostrava was equipped with the neutron generator MP320 operating on the principle of the deuterium–tritium fusion and producing 108 neutrons/s at maximum. To ensure radiation protection of radiation workers and public outside the laboratory, the concrete shielding was designed and its protection efficiency was validated by MCNP simulations. Three approaches to calculate the dose rates were compared. The dose rates were estimated for the ORNL MIRD phantom located at the relevant positions (Tally F6 and *F8) and using the MCNPX mesh tally feature with the new ICRP Publication 116 flux-to-dose conversion factors. It was proven that the Approach II in which the absorbed dose rates due to neutrons for all organs are computed using the cell tally F6 and the photon dose calculation is performed by the *F8 energy deposition tally is the most valuable one.
- Subjects
OSTRAVA (Czech Republic); NUCLEAR activation analysis; NEUTRON generators; TRITIUM; RADIATION protection; ABSORBED dose; SPECTROMETRY; DESIGN protection
- Publication
Radiation Protection Dosimetry, 2019, Vol 185, Issue 1, p116
- ISSN
0144-8420
- Publication type
Article
- DOI
10.1093/rpd/ncy209