Setting radiation safety standards from a modern perspective

The basic ideas published more than 50 years ago in A. D. Sakharov’s article “Radioactive Carbon from Nuclear Explosions and Zero-Threshold Biological Effects” are examined from the standpoint of the scientific knowledge accumulated up to 1958 in the field of radiobiology and the effect of radiation on the human body at ultralow doses. The present state of radiological risk regulation is examined in the light of the problems which Sakharov touched upon. The need for improving the normative-legal foundation of radiological safety and protection in connection with the publication of new ICRP recommendations and the development of new safety standards started under IAEA aegis is discussed.     

Quantitative criteria for rehabilitation of the territory of shore servicing bases of the naval fleet

A variant of the application of the norms and regulations currently operating in our country which ensure radiation safety for workers, the general public, and the environment is presented to valdiate the criteria for rehabilitation of the territory of the shore servicing bases of the naval fleet. The main normative-legal documentation on the rehabilitation of the radiation dangerous objects is analyzed and the international and domestic experience in performing such work, including radiation accidents, is examined. The quantitative criteria used in practice for the residual radioactive contamination of industrial objects and housing developments and the environment are singled out, and an attempt is made, on the basis of a generalization of the information available, to adapt individual tenets in the interests of rehabilitation of radiation dangerous shore-based objects of the fleet. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 35–49, July, 2006.     

Analysis of risk in taking measures for radiation and social protection of the public

The necessity of risk analysis for support of radiation and public protection measures and substantiated application of risk analysis for the development of a scientific-methodological and normative-legal base is examined. Recommendations are given for the development and use of risk analysis. For radiation accidents, a priori and postfactum risk assessment and harm must be distinguished (before and after an accident has occurred). In the latter case, detailed risk assessment is needed, especially to establish a link between sickness and the accident as well as for compensation of harm to the health of victims. A collection of risk indicators for different applications is given. A new additional risk indicator is proposed. This indicator is the relative loss of duration of life (lost years of life, referred to the year during which the source of risk acted). It is intended for establishing a universal limit on risk, suitable for any source of risk. The value of this basic limit is proposed and values of dose and risk limits derived from it for ionizing radiation and other sources of risk are presented. Requirements for methods of dose and risk assessment are formulated. 3 figures, 1 table, 16 references. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 384–395, November, 1999.     

Control of the safe operating period of equipment and pipelines in reactor facilities which are in operation or under construction

The basic assumptions for ensuring safe operation of the components of nuclear facilities, based on controlling service lifetime characteristics, are presented. It is shown up for the Du300 RBMK-1000 pipes, which were made of corrosion-resistant austenitic steel, that this technology can be used effectively in operating power generating units. The complex of measures, developed and validated in the last few years, for monitoring and assessing the technical state of weld seams in pipelines, using the safety concepts “leak before rupture” and “prevention of failure” as well as methods for suppressing the proneness of weld seams to form cracks by the corrosion cracking method under stress, has made possible not only safe operation of Du300 pipelines but it is also a basis for optimizing the volumes and periodicity of operational nondestructive monitoring. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 61–65, July, 2007.     

Safety regulation — 20 years after the Chernobyl accident

The main evolutionary stages and the successes, achievements, and drawbacks of the work performed by an inspection agency after the Chernobyl accident are analyzed. An assessment of the changes which have occurred in the functions of the state inspection agency is made on the basis of different points of view concerning the experience gained in safety regulation. The improvements made over a period of 20 yr in this agency are summarized, and the measures taken to increase the effectiveness of its work are indicated. __________ Translated from Atomnaya énergiya, Vol. 100, No. 4, pp. 282–297, April, 2006.     

Validation of the periodicity of checks of nuclear power plant safety systems using the bars system of programs

The application of the BARS program system for choosing the optimal parameters for checking the serviceability of the safety systems of nuclear power plants is described. The types of checks as well as the mathematical models determining the reliability of the components of the system for each type of check are described. Calculations of the failure probability of the safety systems of nuclear power plants which incorporate components with different levels of reliability are presented. The values of the optimal period between serviceability checks for each type of system are determined. An evaluation of the effect of failures of a general type on the reliability of safety systems is presented for two forms of the α-factor model.     

Estimate of the parameters of feedback on power and the stability of IBR-2 for different average power levels

Experimental and model studies of the parameters of fast feedback on power as a function of the average power of IBR-2 have been performed. Transient power processes caused by square fluctuations of reactivity have been investigated. The changes in the parameters are estimated for average power ranging from 0.5 to 1.5 MW. The results obtained are compared with data from previous experiments performed in 1984–1996. It is noted that the influence of feedback on power decreases as the reactor operating time increases. The model of a reactor with parameters of feedback on power which correspond to one series of experiments is investigated for stability by the frequency method. It is shown that at the regular average power level 1.5 MW a reactor in a self-regulating regime (i.e., without an automatic regulator) possesses an adequate margin of stability. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 89–93, August, 2007.     

Analysis of the possibilities of the neutron-resonance method of determining the elemental composition of matter

Mathematical modeling of the neutron-resonance method analyzing the elemental and isotopic of matter is performed. The method is based on investigation of γ cascades, emitted by nuclei as a result of the radiative capture of neutrons. Gamma cascades of the determined element (effect) and elements of the matrix (background) are modeled and these cascades are detected with a γ-ray multisectional 4π detector. The response functions of the detector are presented for the case of an analysis of the gold content in a natural medium. It is shown that the introduction, performing together with neutron spectrometry, the spectrometry of γ cascades makes it possible to increase the effect/background ratio by a factor of 100. It is noted that this ratio can be further increased by improving the method used to detect γ cascades and analyzing the results of the measurements.     

Regional system for registration and monitoring radioactive waste

A regional structure is proposed for the State registration and monitoring of radioactive waste for the Moscow region. Consideration is given to the problems, the information communications, the role of an interregional center for information analysis, and to solving the problem of the registration of radioactive waste as part of this system. An approach is proposed to monitoring the formation of waste which is based on ensuring conditions for not exceeding the temporary storage norms and for the correct time for removing radioactive waste. An algorithm is given for the efficient planning of its removal taking account of the requirements for radiation safety. Scientific-Production Association MosNPO “Radon”, Moscow. Translated from Atomnaya énergiya, Vol. 86, No. 6, pp. 465–470, June, 1999.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

Effect of fission products yield library on the calculation of the nuclide composition of the fuel in sodium-cooled fast reactors

The calculation of the composition of irradiated fuel for different degrees of burnup is a basic problem in the analysis of nuclear-radiological safety of objects holding spent fuel assemblies. The yield of fission products is one of the important initial indicators in burnup calculations. Methods for compiling libraries of fission products yield on the basis of the ENDF/B up-to-date evaluated nuclear data files are described. The nuclide composition of uranium oxide and uranium-plutonium-zirconium metal fuel in sodium-cooled fast reactors is analyzed by means of high-precision calculations performed with different fission product yields libraries using different computer codes MONTEBURNS–MCNP5–ORIGEN2 and the results are presented.     

Taking account of the effect of the curvature of VVÉR-1000 fuel assemblies on the power of the fuel elements

A comparative analysis is made of the deterministic and statistical methods of taking into account the effect of the curvature of VVéR-1000 fuel assemblies on the power of fuel elements. The fuel-element distribution of the energy release in the core for any random distribution of the gaps between the fuel assemblies is simulated, using the MEX code, on the basis of precise calculations (MCU code) and design calculations (BIPR-7 and PERMAK codes). The Monte Carlo method (Zazor code) was used to model the nominal density distribution of gaps in the core for different degrees of curvature of the fuel assemblies. It is shown that the power gain, obtained for the fuel elements by the probabilistic-statistical method, due to the curvature of the fuel assemblies is smaller and makes it possible to substantiate core safety with large perturbations, in contrast to the deterministic “maximum gaps near the most-energy stressed fuel element” method. 5 figures, 1 table, 3 references. Special Design Office “Gidropress.” Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 210–213, September, 1999.     

Shock-wave deformation and fracture of zirconium dioxide ceramic and concrete

It is noted that interest has increased in refractory materials such as zirconium dioxide ceramic and concrete for use as linings of traps for the melt of a reactor core. The shock-wave profiles are investigated under dynamic conditions, the dynamic strength is determined, and comparative experiments are performed to investigate the spallation resistance of these materials. It is noted that damping properties of ceramic are manifested: this is the reason why ceramic is more resistant to spallation than concrete. 5 figures, 3 tables, and 14 references. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 113–119, February, 2000.     

The integrity of the pressure containment of the high-temperature reactor

The safety concept for ensuring the integrity of the pressure retaining containment is determined by the structural and system-specific inherent safety characteristics and features of the high-temperature reactor. The integrity of the pressure retaining containment, i.e. the elimination of a major failure, is achieved by a system of measures ensuring a high standard of quality and safety. The fundamental cornerstones of this safety concepts are the stringent requirements in the design and manufacture in view of an optimized production technology as well as specific structural solutions such as, e.g., the prestressed concrete reactor vessel. Additional safety measures such as the quality control performed independently of the manufacturer's works and the in-service inspection, have to be considered as redundant safety measures. The in-service inspection can be limited to the confirmation of safety-relevant data and analysis of deviations from these data. Recurrent non-destructive tests within the PCRV are not required, however, possible to a certain extent.     

Simulation of the first-loop water chemistry regime of a nuclear power system with a two-phase water coolant

An approach to simulating the water-chemistry regime for transient and stationary operating states of a nuclear power system is proposed. The approach is based on the designer’s philosophy. The first loop of the nuclear power system is divided into standard units. The special features of the reactor design and the crux of the problem being solved determine the number of units and the interaction between them. The units can be structural components of the loop as well as coolant components. The thermochemical and radiation-chemical transformations of the components of the water coolant are simulated at each unit. The interaction between the physicochemical processes at the units is expressed in the flow of the coolant components between them. The approach developed is illustrated on the analysis of the water-chemistry regime in implementing a passive safety system — a promising solution geared toward increasing the operational safety of the first loop of a nuclear power system. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 19–25, January, 2006.     

Validation of nuclear and radiation safety of a container for spent amb reactor fuel assemblies at the Beloyarskaya nuclear power plant

An approach is proposed for validating the nuclear and radiation safety of a container for spent fuel assemblies from AMB-100 and-200 reactors at the Beloyarskaya nuclear power plant. To validate the radiation safety, the characteristics of fuel assemblies and their classification according to the average fuel burnup in the casing, and the intensities of n and γ radiation in the casing are analyzed. Nuclear safety is validated on the basis of the concept of a “model” casing. This model makes it possible to obtain an upper estimate of the effective coefficient of neutron multiplication for all real casings with fuel assemblies. Calculations are used to determine the minimum necessary thickness of the vessel, bottom, and cover for 17-and 35-place casings. It is shown that no special neutron protection is needed. The container design to be developed meets the IAEA and OPBZ-83 safety standards. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 423–428, June, 2006.     

Real-time measurement of hydrogen generation level during beryllium dust oxidation by steam depending on the dust arrangement geometry

The paper concentrates on the safety issues in the International Thermonuclear Experimental Reactor (ITER) and describes the experiment on the measurement of hydrogen generation rate in case of Ingress of Coolant Event (ICE)—leak inside the vacuum vessel during interaction between water and beryllium (Be) dust. The ICE situation in ITER was simulated in a facility; the active spectroscopy was used to define the hydrogen content by the dynamics of oxidant concentration at a sampling frequency up to 10 Hz. Hydrogen release in time at temperatures of 500-900 °C is investigated, and different versions of dust arrangement are considered, i.e. on the surface and in a slot between armoring tiles at different initial density. The obtained results are compared with the known experiments.     

核电厂安全管理工作探讨

搞好核电厂安全管理是核电厂安全生产的重要保证.核电厂安全管理包括建立健全安全责任制、安全生产的监督和管理及信息反馈、掌握安全防护规程、用科学态度处理放射性安全工作、保证工业安全及做好安全分析工作。我国核电刚刚起步,学习国外先进的安全规程,吸取他人的经验教训,探讨适合我国国情的安全管理规律是十分必要的.     

The concept of "safety"

Conclusions The systematic view of safety presented here is of a general nature, in contrast to the majority of other studies (for example [1–3]). Many authors define safety as a state of protectedness. However, if one understands protectedness in the sense adopted in the present paper, the concept of safety has a wider significance. In [3] safety is also considered from a systems point of view and three attitudes in society are distinguished characterizing the defined position of some subject-agent in the situation of guaranteeing safety. However, the whole course of the discussions and the concept adopted for safety differ considerably from those given here. Special attention has been devoted in the present paper to the relative nature of safety and the importance has been mentioned of distinguishing and comprehensively studying the different aspects of safety when adopting solutions. By virtue of the systems approach and the relative nature of safety, its definition as properties of an object is imprecise, although the concept of a safety object can be used if one takes account of its relative nature. The safety threshold defines the standard of safety, Although there is no scale of safety in terms of this concept, standards of safety can be compared in different fields or different countries. The lowest level of acceptable risk, i.e., the most rigorous establishment of the safety threshold, provides a basis for speaking of a higher standard of safety. Kurchatov Institute, Russian Scientific Center. Translated from Atomnaya énergiya, Vol. 80, No. 6, pp. 415–420, June, 1996.     

The provision of radiation safety of personnel in the extraction of uranium ores

The radiation hazard to the respiratory organs of miners in uranium mines is assessed and a set of measures is worked out for ensuring radiation safety in the prospecting and extraction of uranium ores. Measures are described and justified which, as a result of achieving them, enable a standard to be set for a mining atmosphere and also to prevent external radiation overexposure and radioactive contamination of the integumenta of people working in the uranium mines.Translated from Atomnaya Énergiya, Vol. 19, No. 2, pp. 161–168, August, 1965