The article presents the results of field studies in the area of the Belarusian NPP in the pre-operational period. The «background» contents of gamma-emitting radionuclides in individual components of the environment are determined. The main array of dose rate measurements in the area of the NPP construction site is in the range 0.048 ÷ 0.089 μSv/h. External radiation in the surveyed area is formed at 96% due to 40K, 226Ra and 232Th. The information obtained can be used to correctly interpret the data of future radiation monitoring during normal operation of nuclear power plants.

The determination of the physicochemical forms of radioiodine in the gas-air environment of an industrial nuclear reactor is necessary to solve related problems – technological control and radiation safety. In the technological context, the results obtained make it possible to adequately assess the efficiency of purification of emissions of radioactive iodine isotopes, the choice of instruments and methods for controlling emissions. In the context of radiation safety, research results make it possible to correctly assess the radiation effects on the environment and humans, substantiation of emission standards for the atmosphere and confirmation of the safety of operation of an industrial reactor installation. The research method is based on the difference in the deposition of radioiodine on a set of one AFA-RMP aerosol filter and six filters of the AFA-SI type, which makes it possible to separately determine the 131I aerosol, easily and hardly sorbed form. It has been shown that the non-purified gas-aerosol mixture mainly contains radioactive iodine in the form of gaseous hardly adsorbed compounds. For 131I, the most probable percentage in volumetric activity of hardly adsorbed, easily adsorbed compounds and iodine aerosols was obtained. Based on the data obtained, an assessment of dose loads was carried out taking into account the annual emissions of the reactor installation and weather conditions. A conservative approach to assessing the radiation exposure of 131I emissions is 47 times higher than the assessment taking into account its physicochemical forms.

A method for non-destructive monitoring of the content of natural radionuclides in building materials has been developed. Spectrum measurements of gamma radiation are carried out with a pre-calibrated field gamma spectrometer. The calculation of the average specific activity of natural radionuclides in building materials is carried out by comparing the calculated flux density of unscattered gamma quanta normalized to the specific activity, and the experimentally measured count rates in the photopeak. calculated for the geometry of the room under study and the location of the detector. Application of the developed method makes it possible to estimate the average activity of natural radionuclides in building materials without destruction.

An oral dosimeter of mixed gamma-neutron radiation for emergency exposure conditions has been developed. The energy dependence of the neutron radiation dosimeter sensitivity is close to the energy dependence of the specific effective dose per unit flux density. For neutron fields containing a significant contribution of fast neutrons, the uncertainty of the dosimeter readings is no more than 25% for the anteroposterior radiation geometry and no more than 35% for the rotation geometry. In neutron fields with a predominance of particles with thermal and intermediate energies, the dosimeter overestimates the effective radiation dose by 2.5 times for the anteroposterior geometry and 3.3 times for the rotation geometry. A staging experiment was carried out, which included placing individual dosimeters inside a canister simulating the torso of a standard adult in a neutron radiation field. The conditionally true values of the effective dose were obtained using the energy and angular distribution of the neutron radiation flux density. Differences in the dosimeter readings and the conditionally true value of the effective dose do not exceed 2.