Radiation Safety During the Construction of the «Sarcophagus» («Shelter» object) at the 4-th Unit of the Chernobyl Nuclear Power Plant in May-November 1986

This year marks the 35-th anniversary of the events related to the liquidation of the consequences of the accident at the Chernobyl nuclear power plant (Chernobyl NPP) and their most important stage - the closure of the destroyed 4-th power unit with a special structure, which was initially called the «Sarcophagus» (later the «Shelter» object). This article presents information about the period May-November 1986 in the aspect of radiation safety of the builders of this unique facility.

Correction Coefficients for Measuring the Ambient Dose Equivalent Rate of Neutrons

Calculations of the response for the most widely used neutron dosimeters at the Russian nuclear power plant (NPP) have been performed. It is shown that in some cases it is necessary to introduce a correction for the measured value of the ambient dose equivalent rate (AEDR). The experimentally tested values of the correction for measuring AEDR in the containment rooms of NPP with VVER-1200 are given.

Prospects for Range Extension of Large Area Sources with Alpha/Beta-Emitting Radionuclides, Precision Factors

The article deals with the problems of testing, calibration and application concerning foreign radiation protection instruments due to mismatch of regulatory requirements in Russian and international standards. In particular, there isn’t unified nomenclature of radiometric sources. Promising directions for range extension of large area sources are presented. Some precision factors are discussed.

Evaluation of Feasibility of Gamma-Spectrometer NaI PAK-01 and SAS Na M3 Software Application for Regular Radionuclide Analysis of NPP Permissible Wastes

The feasibility of correct NaI gamma-spectrometry activity measurement for each nuclide in 131I, 132I, 133I, 134I, 135I, 133Xe, 135mXe, 135Xe and 222Rn composition is presented. To get this result the special matrix method M3 and SAS Na M3 software were used for spectra processing. SAS Na M3 software was developed for complex NaI gamma-spectra processing. Special algorithms and auxiliary software are used to overcome the problems of the classic spectra processing matrix method. Being used for spectrum processing SAS Na M3 software determines the nuclide composition of the sample, activity of nuclides identified and activities uncertainties. The activity values estimation is made for nuclides not identified in the sample measured but included in SAS Na M3 software nuclides library. The values of minimal detectable activities for NaI ∅3''× 3'' gamma-spectrometer and 1 hour measuring time are ~ 0.6 Bq for 131I, 132I, 133I, 134I and ~ 2 Bq for 135I.

New Algorithm for Upgrading of Portal Radiation Monitors with Plastic Detectors. Part 3. Description of the Algorithm

New algorithm is described in the third part of the paper. It uses method of energy windows and is based on iterative scheme. Values of smoothing and thresholds are optimized in each energy window and time step during of movement of transport through control zone. Some parameters of monitor which are measured on the place of its installation or manufacturing are used as input information.

Adapted Method of Radiation Response Function Calculation for Development of High-Sensitivity Reference Meters Based on Scintillation Detection Units with NaI(Tl) Crystals for Dosimetry of Low-Intensity and Near-Background Levels

Metrological support of photon radiation fields of low-intensity and near-background levels (0.04–100 μSv/h (μGy/h)) by ambient equivalent dose rate or kerma rate in air using scintillation detectors with NaI(Tl) crystals looks promising and in demand in dosimetry, but nontrivial due to the complex dependence of efficiency registration of gamma quanta from energy. The solution of such problems with the use of these detectors can be based on the use of the radiation response functions, which are functionals of the energy distribution of the radiation field fluence. The paper proposes a method for calculating the radiation response function adapted for solving metrological support problems for creating high-precision dosimetric measuring instruments based on scintillation detection units with NaI(Tl) crystals.

About Some Questions of the Organization of Work of Workers in the Conditions of Exposure to Ionizing Radiation

The article discusses the need to make changes to the regulatory documents that classify harmful and (or) dangerous production factors that stimulate and regulate the work of workers under the influence of ionizing radiation based on the concept of a linear dependence of the risk of stochastic effects on the received dose.

On the Possibility of Reducing the Measurement Error of Spectrometers with Proportional Gas Detectors in the RC-CR Formation of Output Signals

When analyzing the elemental composition of the substance under study, the task is to determine the true intensity of the radiation lines. The error in determining the intensity is caused by errors in the spectrometer. Reducing the error can be achieved by taking into account (correcting) the errors, as well as reducing them. In this paper, we propose a method for reducing the measurement error by reducing the miscalculations with an acceptable deterioration in the energy resolution of the spectrometer. An algorithm for automatically setting the duration of the output signals of the spectrometric amplifier and the width of the differential discriminator window for amplitude analysis is presented.

Occupational Exposure to Neutrons at the Beloyarsk NPP

Studies of the energy distribution of neutron radiation at the workplaces of the Beloyarsk NPP were carried out. At 1 and 2 power units, occupational exposure of neutron irradiation occurs during operations for loading spent nuclear fuel into special railway carriage. At power units 3 and 4, operations accompanied by neutron irradiation can be divided into 3 groups: (1) work in rooms adjacent to the reactor core; (2) manipulation of radioisotope neutron sources; (3) work with fresh and spent nuclear fuel. Based on the data obtained on the energy distribution of the neutron radiation flux density, the ‘true’ values of the ambient dose equivalent rate H*(10), the individual dose equivalent rate Hp(10) and the integral neutron radiation flux density at individual workplaces were determined. For each group of workplaces, Fluence-toambient dose equivalent conversion coefficients are determined, which lie in the range from 12 to 295 pSv⋅cm2. Correction factors for individual thermoluminescent dosimeters, taking into.

Current Problems in the Development and Certification of Measurement Methods in the Field of Radiation Monitoring

The article discusses the features of the development and certification of measurement methods in the field of radiation monitoring, including measurements at nuclear power facilities. The results of the analysis of actual problems related to the development and certification of methods are presented.