ABSTRACT
As part of an ongoing dose reconstruction project, equivalent organ dose rates from photons and neutrons were estimated using the energy spectra measured in the central hall above the graphite reactor core located in the Russian Mayak Production Association facility. Reconstruction of the work environment was necessary due to the lack of personal dosimeter data for neutrons in the time period prior to 1987. A typical worker scenario for the central hall was developed for the Monte Carlo Neutron Photon-4B (MCNP) code. The resultant equivalent dose rates for neutrons and photons were compared with the equivalent dose rates derived from calculations using the conversion coefficients in the International Commission on Radiological Protection Publications 51 and 74 in order to validate the model scenario for this Russian facility. The MCNP results were in good agreement with the results of the ICRP publications indicating the modeling scenario was consistent with actual work conditions given the spectra provided. The MCNP code will allow for additional orientations to accurately reflect source locations.
Subject(s)
Models, Biological , Nuclear Reactors , Occupational Exposure/analysis , Occupational Exposure/standards , Organ Specificity/radiation effects , Radiometry/methods , Radiometry/standards , Calibration , Computer Simulation , Dose-Response Relationship, Radiation , International Cooperation , Models, Statistical , Monte Carlo Method , Neutrons , Photons , Radiation Dosage , Radiation Protection/standards , Reference Standards , Reproducibility of Results , Russia , Sensitivity and SpecificityABSTRACT
One of the objectives of the Joint Coordinating Committee for Radiation Effects Research Project 2.4 is to document the methodology used to determine the radiation doses in workers from the Mayak Production Association who were exposed to plutonium. The doses have been employed in numerous dose response studies measuring both stochastic and deterministic effects. This article documents both the historical (pre-1999) and current ("Doses 1999") methods used by the FIB-1 scientists to determine the doses. Both methods are based on a three-chamber lung model developed by the FIB-1 scientists. This method was developed in partial isolation from the West and has unique characteristics from the more familiar ICRP biokinetic models. Some of these characteristics are the use of empirically based transportability classifications and the parameter modification for chelation-therapy-enhanced excretion data. An example dose calculation is provided and compared to the dose that would be obtained if the ICRP models were used. The comparison demonstrates that the models are not interchangeable and produce different results.
