[en] The objective of radiation protecting workers in the Russian Federation are achieved due to implementation of the strategy of keeping control by source as a base for safety assurances for protecting workers against risks arising from occupational exposure to ionizing radiation. Overall hierarchy structure of regulatory documents for protecting workers of the nuclear facilities against exposure to ionizing radiation is under creation in the Russian Federation for last five years. These documents implements the International Basic Safety Standards for Protection against Ionizing Radiation and for Safety of Radiation Sources and 1990 Recommendations of the ICRP in to the national radiation protection practice. The overall structure and the main regulatory documents from it are discussed. (author)

[en] The objective of radiation protecting workers against risks arising from exposure to ionizing radiation in the Russian Federation is: 1. To limit the harmfulness of routine exposure - to prevent deterministic effects by keeping the exposure doses below their thresholds and undertake every reasonable action in order to lower the probability of occurrence of stochastic effects to an acceptable level. 2. To limit the risk of potential exposure - to undertake every reasonable action in order to reduce the probability of occurrence of stochastic and deterministic effects to an acceptable level by improving the reliability of the radiation protection and safety system. These objectives are achieving due to implementation of the strategy of keeping control by source as a base for safety assurances for protecting workers against risks arising from exposure to ionizing radiation. The details of implementation of the Strategy of control by source as a base for protecting workers are discussed. (author)

[en] Epidemiological studies of nuclear industry personnel contain the significant abilities to assess the prolonged radiation exposure effects in the human health. The clarification of these assessments and following improvements of the scientific justification of radiation regulation require the expansion of factual basis of the research currently, Branch Medical Dosimetric Registry (BMDR) of atomic industry and nuclear power employees is under the development in Russian to compose a number of regional registries. This work is coordinated by the State Research Center- Institute of Biophysics (Moscow). The first phase of this project was devoted to the forming of the regional registry of Mayak PA employees (Ozersk, South Uranl region). the employee registries of Siberian Chemical Plant (SCP, Seversk, Tomsk region) and Mountain Chemical Plant (MCP, Zheleznogorsk, Krasnoyarsk region) are at the finalization. At later phases, BMDR will be added by the information on other enterprises and on operating NPP too. The paper describes the structure, general issues of the forming and current status of BMDR. The comparison of major BMDR features versus LSS registry (which is the one of basic components for international radiation protection recommendations and current radiation protection standards) demonstrates that BMDR information can be more preferable to assess the significance of the man made radiation at high and intermediate dose ranges. Particularly, the number of employees (20-40 year age range) exposed to doses specific to detectable radiation health effects (above 2000 mSv) is almost ten times more than that for LSS cohort. Besides, the health monitoring was elaborated since the employment start point (Whereas, since year 5 for LSS cohort). BMDR dose records were measured (against LSS reconstructed doses) and the employee exposure duration was equal to years and decade (alternatively to momentary exposure recorded in LSS). BMDR data quantity and quality correspond to basic requirements of international cooperative studies and can be easily integrated to these studies upon the registry finalization. The unique information accumulated by BMDR can predispose the assessment of the man made irradiation significance as well as to be the source of new fundamental considerations. (Author) 7 refs

[en] The site of temporary storage of spent nuclear fuel and radioactive waste, situated at Andreeva Bay in Northwest Russia, was developed in the 1960s, and it has carried out receipt and storage of fresh and spent nuclear fuel, and solid and liquid radioactive waste generated during the operation of nuclear submarines and nuclear-powered icebreakers. The site is now operated as the western branch of the Federal State Unitary Enterprise, SevRAO. In the course of operation over several decades, the containment barriers in the Spent Nuclear Fuel and Radioactive Waste storage facilities partially lost their containment effectiveness, so workshop facilities and parts of the site became contaminated with radioactive substances. This paper describes work being undertaken to provide an updated regulatory basis for the protection of workers during especially hazardous remediation activities, necessary because of the unusual radiation conditions at the site. It describes the results of recent survey work carried out by the Burnasyan Federal Medical Biophysical Centre, within a programme of regulatory cooperation between the Norwegian Radiation Protection Authority and the Federal Medical-Biological Agency of Russia. The survey work and subsequent analyses have contributed to the development of special regulations setting out radiological protection requirements for operations planned at the site. Within these requirements, and taking account of a variety of other factors, a continuing need arises for the implementation of optimisation of remediation at Andreeva Bay.

[en] Full text: The result of dosimetric expertise after internal contamination is usually presented as a 'best estimate' calculated from the available monitoring data and information about the conditions of exposure. Reference assumptions are applied to make up for unknown parameter values. However, previous intercomparisons demonstrated that the evaluation of internal dose is subject to a large uncertainty coming from both the technical limits of measuring apparatus and the assumptions made by the expert. In this work we propose an alternative approach to the assessment of internal dose that enables to report the confidence interval associated with the evaluated dose value. So far, four sources of uncertainties on input data have been considered: the physicochemical characteristics of the radioactive material (absorption type and diameter of aerosols), the time pattern of intake, the counting error and the stochastic variability of excretion or bioassay sampling. Three successive steps of approximation are suggested, depending on the expected level of dose, for which increasingly realistic parameter values should be sought and applied. The propagation of errors from input data to dose result is performed through Monte Carlo calculation. Finally, the results of dose calculation are presented in the form of mean possible dose value, medium possible dose value and upper 95^th percentile. This approach has been implemented in internal dose calculation softwares and applied to two example cases of routine monitoring of uranium and special monitoring of caesium. It provides a useful tool for both reporting the results of expertise and designing realistic monitoring programs, and it will be improved in the future by further investigation of the sources of uncertainty in internal dosimetry. (author)

[en] Measurements of actinide body burden using whole body counting spectrometry is hampered due to intensive absorption of γ-rays inside the patient's body, which depends on the anatomy of a patient. To establish the correspondence between pulse-height-spectra intensity and radionuclide activity, Monte Carlo calculations are widely used. For such calculations, the radiation transport geometry is usually described in terms of small rectangular boxes (voxels) retrieved from computed tomography or magnetic resonance images. The software for Monte Carlo-assisted calibration of whole body counting, which performs automatic creation of individual MCNP voxel phantoms, was checked in a quasi-in vivo experiment on large animals. During the experiment, pigs of 35-40 kg body mass were used as phantoms for measurement of actinides body burden. ²⁴¹Am was administered (via injection of a radioactive solution or via implantation of plastic capsules containing the radioactive material) into the lungs of pigs. The pigs were measured using the pure germanium low-energy γ-spectrometers. The images of animals were obtained using the computed tomography machine. On the base of these tomograms, MCNP4c2 calculations were done to obtain the pulse-height-spectra of the whole body counters. The experimental results were reproduced in calculations with error of less than 30% for ²⁴¹Am administered via injection and less than 10% for ²⁴¹Am administered inside the capsules. (authors)

[en] In the 1960s two technical bases for the Northern Fleet were created in the Russian northwest at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, receiving and storing radioactive waste and spent nuclear fuel. No further waste was received after 1985, and the technical bases have since been re-categorised as temporary storage sites. The handling of these materials to put them into a safe condition is especially hazardous because of their degraded state. This paper describes regulatory activities which have been carried out to support the supervision of radiological protection during recovery of waste and spent fuel, and to support regulatory decisions on overall site remediation. The work described includes: an assessment of the radiation situation on-site; the development of necessary additional regulatory rules and standards for radiation protection assurance for workers and the public during remediation; and the completion of an initial threat assessment to identify regulatory priorities. Detailed consideration of measures for the control of radiation exposure of workers and radiation exposure of the public during and after operations and emergency preparedness and response are complete and provided in sister papers. The continuing requirements for regulatory activities relevant to the development and implementation of on-going and future remediation activities are also outlined. The Norwegian Radiation Protection Authority supports the work, as part of the Norwegian Government's plan of action to promote improvements in radiation protection and nuclear safety in northwest Russia.

[en] We performed a comparative study of the formation of γН2АХ foci (a marker of DNA doublestrand breaks) in human bone marrow mesenchymal stem cells after 24-h incubation with ³Н-thimidin and tritium oxide with low specific activities (50-800 MBq/liter). The dependence of the number of γH2AX foci on specific activity of 3H-thymidine was described by a linear equation y=2.21+43.45x (R²=0.96), where y is the number of γH2AX foci per nucleus and x is specific activity in 1000 MBq/liter. For tritium oxide, the relationship was described by a linear equation y=2.52+6.70x (R²=0.97). Thus, the yield of DNA double-strand breaks after exposure to ³H-thymidine was 6.5-fold higher than after exposure to tritium oxide. Comparison of the effects of tritium oxide and X-ray radiation on the yield of DNA double-strand breaks showed that the relative biological efficiency of tritium oxide in a dose range of 3.78-60.26 mGy was 1.6-fold higher than that of X-ray radiation. Improvement of the methods of analysis of DNA double-strand breaks repair foci is highly promising in the context of creation of highly sensitive biodosimetry technologies for tritium compounds in humans.

[en] The assessment of internal dose is subject to a large uncertainty due to the limits of measuring technique and to the assumptions made by the expert. Here, we propose an approach to report the confidence interval associated with the evaluated dose. The sources of uncertainties considered so far include the date of intake, the physico-chemical characteristics of the radioactive material, the counting error and the stochastic variability of excretion. Three successive levels of approximation are suggested, depending on the expected dose, for which increasingly realistic parameter values should be sought and applied. Finally, the results of a Monte Carlo dose calculation are presented in the form of a statistical distribution of possible dose values. This approach has been applied to two cases of uranium and caesium exposure. (authors)