Повышение экономической эффективности эксплуатации энергоблоков атомных станций с учетом обеспечения безопасности

The particulars of the design of new-generation ASCTP and experience in implementing the system in the No. 3 unit of the Kalininskaya nuclear power plant are examined. The basic requirements and the main approaches for developing ASCTP for the power-generating units of nuclear power plants with newgeneration VVER, including protection from failures for general reasons in the safety control systems, expansion of the functions of the system of the top block level, intellectually signaling, automation level of technological processes, and servicing of the means of automation are presented. It is noted that the principle of optimizing the application of modern information technologies taking account of experience in design and operation of power-generating units with mandatory maintenance of the updated base must play an important role in determining the requirements for and making technical decisions concerning ASCTP for new-generation VVER.

We describe the methods and present the results of an examination of removable elements of fittings of two power units of the Rovno Nuclear Power Plant after 10 and 13 years of operation. It was discovered that, after the indicated periods of operation, the technical state of these elements was satisfactory and the material of the rods had no serious defects or other types of damage that could restrict the possibility of their subsequent operation. According to the regulations accepted in nuclear power engineering, the operating modes of fittings realized at nuclear power plants are light-duty as far as the number of loading cycles is concerned. Therefore, the degree of local fatigue damage to the metal of the rods does not reach its critical value. The outlined procedure of examination of removable elements of fittings is comprehensive. It is completed by low-cycle fatigue testing of the damaged metal and enables one to perform an adequate analysis of the possibility of subsequent trouble-free operation of the corresponding equipment.

The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy.

Data on the modern state and development trends for coal and nuclear power engineering in Russia up to 2030 are generalized. It is emphasized that from the viewpoint of strategy, coal and uranium fuel will be the main energy carriers. The forecast of energy consumption is made; the “roadmap” of new power-generating units of heat and nuclear power plants on the territory of Russia is presented.

The full-scale simulator (FSS) of nuclear power plant (NPP) unit is designed for operating personnel professional joint training of the power unit's main control room using its real full-scale model. Testing of all FSS systems has an important place in the software part of the development of FSS. Currently, the development of software testing for FSS systems is required. The article provides an overview ozf modern testing systems. In addition, the article presents the algorithms for the operation of tests designed by using Pytest – the framework for writing software tests. The paper presents the following tests: overpower test and check for the presence of chugging; test of the correctness of the decay heat model in terms of amplitude in the event of scram and stable attenuation of decay heat after scram.

The principles for constructing a problem-oriented system for designing a technology for disassembling the equipment in the power-generating units of nuclear power plants which are being decommissioned, starting from the rules of organization and use of the design space on the basis of informational and expert methods and taking account of the influence factors of the environment where the technology functions, are presented. 2 figures, 10 references.

The study is devoted to the selection and justification of circuit solutions for the composition of an intermediate steam super heater using the example of the thermal circuit of the steam turbine unit of the South-Ukraine Nuclear Power Plant (SU NPP).Steam turbine units with water-water power reactors, which are currently in operation and planned for installation in the future, operate on wet steam, therefore measures are taken to dehumidify the steam during its expansion in the turbine unit -intermediate separation and steam superheating according to the following schemes: -separation and two-stage steam superheating; -separation and single-stage steam superheating.Most plants use separation and two-stage intermediate steam superheating.There is no justification for the use of such a scheme in domestic literature.Methodology.A thermal calculation of the steam turbine unit was performed using the method of successive approximations: the productivity of the steam generator was calculated using two-stage and single-stage intermediate steam superheating; a comparative assessment of the mass and size indicators of the separator-steam super heater and the thermal efficiency of the unit was performed.The initial and final parameters of the working fluid and the parameters of the working fluids in the regeneration system correspond to the thermal diagram of the steam turbine unit of a nuclear power plant.Results.A comparison of the technical indicators of two options for using steam intermediate superheating for a steam turbine unit at a nuclear power plant allows conclusions to be drawn about the questionable feasibility of using twostage intermediate steam superheating.Two-stage steam superheating (compared to single-stage) complicates the thermal scheme of the steam turbine unit, the design of the separator-steam superheater, and reduces the productivity of the steam generator (increases the thermal efficiency of the unit) by only 0.28 %.Given the insignificant share of fuel costs for a nuclear power plant, a reduction in total operating costs in the range of hundredths of a percent can be expected.With single-stage intermediate steam superheating, the mass, size, and cost indicators of the steam super heater are improved by reducing the heat transfer surface (by about a third) as a result of increasing the mean logarithmic temperature head by 34 %.Practical significance.The results of the study can be taken into account when selecting and justifying design decisions regarding the composition of steam turbine installations for promising nuclear-fueled steam turbine installations.

With rapid development of informatization and digitalization technologies, digital transformation is pushing ahead in industry around the world, especially in nuclear power operation and maintenance. In the full life-cycle digital transformation plan of nuclear power enterprises, the design institute has accumulated a large number of digital design results, such as the three-dimensional model data of nuclear power plants. These 3D models are handed over to the operation organizations to build digital, visual and intelligent operation and maintenance solutions. Without digital models, field staffs have to check the paper procedures and technical data before or even during the on-site inspection and maintenance preparation, however, this is adverse for field work. The application of 3D digital auxiliary tools and platforms can effectively improve management efficiency and reduce potential risks in nuclear power operation and maintenance. This paper proposed a digitalization and visualization method to improve the digital management level, and developed a 3D visual operation management platform for Daya Bay nuclear power plant. With the platform, nuclear power operation organizations could handle the daily maintenance work in an efficient way, so that to improve the operation management level.

A nuclear power plant is a complex technological facility. The introduction of new standards and experience gained in the operation of domestic and foreign nuclear power plants requires the introduction of changes in the operating organization of existing nuclear power plants and improvements in their equipment. A general survey of the modernization of existing nuclear power plants with RBMK-1000 type reactors is presented. Measures for modernization of the power-generating units of nuclear power plants carried out following the accident at the Chernobyl nuclear power plant and at the present time are discussed. Because of modernization, the safety, reliability and economic efficiency of nuclear power plants have improved. The principal focus is on modernization of the reactor plant.

The WWER-1000 reactor is operated at 13 of the 15 operating power units of Ukraine's nuclear power plants (NPPs). Ensuring long-term and safe operation of such reactors is the basis for reliable operation of all 13 Ukrainian nuclear power plants units and the guarantor of Ukraine's energy security. The determining and leading factor influencing the safety and proper residual life of the WWER-1000 reactor vessel is the radiation embrittlement of the reactor steel. The consequences of radiation embrittlement of reactor steel are negatively manifested in emergencies with cooling of the core. This process itself – radiation embrittlement – accumulates constantly and gradually. Therefore, it is important to monitor it by periodically performing ongoing rapid assessments of the brittle strength of the WWER-1000 reactor vessel (along with other factors, including cyclic damage, as discussed in a previous publication). Therefore, it is important to use the calculated express methods of periodic assessment of the brittle strength of the WWER-1000 reactor vessel with guaranteed accuracy. The effectiveness of the approach is supported by low cost of resources – engineering staff, fast and relatively simplified use of computers and software. As an example and confirmation of the applicability of the proposed approach, an expert rapid assessment of the fragile strength and residual life of the reactor vessel of Unit № 1 of the South-Ukrainian Nuclear Power Plant was performed. This takes into account the actual, passport characteristics of its metal. The negative impact of the rigid regime with cooling of the WWER-1000 reactor of Unit № 1, not taken into account by the operating organization (South-Ukrainian Nuclear Power Plant) when extending its designated resource / service life, is shown. timely to clarify complex factors, technical aspects and parameters, as well as – their possible negative effects on the safe operation of systems and elements of nuclear power plants.

The comment by Walter Scheider that Three Mile Island “remains an icon of a profit-driven industry cutting corners” echoes one by Anatoly Alexandrov, president of the Soviet Academy of Sciences and a strong supporter of the “RBMK” reactor, a particular type built only in Russia and used in the Chernobyl nuclear power plant. Alexandrov said that “such an accident [as TMI] can only happen in America where they put profits ahead of safety.” Lecturing in Dubna, Moscow, and Gatchina just after TMI, I told listeners that if they believed Alexandrov, they were condemned to have a serious accident in the country within a decade. Alas, I was right. The centrally planned economy of the Soviet Union did far worse in ensuring safety than the US, and the Chernobyl accident occurred.The profit motive, if suitably guided by good analysis tools, can enhance safety. Fortunately, we now have “risk-informed regulation.”Much of the improvement in safety since TMI has been profit driven. It was the industry that set up the Institute of Nuclear Power Operations and the World Association of Nuclear Operators. Those organizations set safety targets and guidelines and put pressure on members to follow them. Analysis by the Nuclear Regulatory Commission and university groups strongly suggests that if the safety targets and procedures, guided by a rigorous analysis, are met, the reactor will be safer. Other than closing it, the safest way to operate a power plant is to have it running smoothly and continuously. That is also the most profitable, so the requirements of safety and profitability tend to coincide; the analysis spotlights those areas where they do not. Industry efforts have increased plant availability from 60% in the 1970s to 92% today.The cost of nuclear power to the consumer depends very much on public acceptance, and it is the antinuclear movement that has set up the expensive, often unscientific roadblocks to that acceptance. There are signs that improved plant performance in the past 20 years has increased public acceptance and therefore profitability. 1 1. R. Wilson, Int. J. Global Energy Issues (in press). However, few utility company presidents would order a new nuclear power plant today without some assurance that past roadblocks will not be reinstituted.Although disposal of the waste from nuclear fission is a problem, it is a much smaller one than the disposal of carbon dioxide from burning coal. That waste can produce the climate change that is Alan Robock’s professional concern. But his stated opinion that “the waste disposal problem is not solvable in the near future” can only apply to the political problem rather than to the scientific one. Independent committees agree that a technical solution is possible. Political maneuvers in late 2005 delayed or prevented the temporary storage of nuclear waste on the Goshute Indian Reservation in Utah. It is but one example of the political impasses that delay the storage of nuclear waste material. The funding cut in December 2006 for the US Department of Energy’s presentation of the case for the Yucca Mountain nuclear waste repository is another.Scheider and the authors of the references he quoted incorrectly blame the power company for the confusion at TMI. However, neither the Associated Press nor any major newspaper accurately quoted the press releases from the Nuclear Regulatory Commission. From noon on the day of the accident, I was involved in explaining events to the public, and I could say with assurance that the expressed fears were vastly overstated. I recommended that the press quote NRC press releases verbatim and then comment as they wished, but that recommendation was ignored.A study by the European Commission (http://externe.jrc.es/index.html) states unequivocally that if coal plants were forced to pay their full external costs, they would not be built. But Scheider is right in one important respect. The lay public is not stupid, and the details of nuclear power can be explained to them. For such an explanation I recommend David Bodansky’s excellent book. 2 2. D. Bodansky, Nuclear Energy: Principles, Practices, and Prospects, 2nd ed., Springer, New York (2004). REFERENCESSection:ChooseTop of pageREFERENCES <<1. R. Wilson, Int. J. Global Energy Issues (in press). Google Scholar2. D. Bodansky, Nuclear Energy: Principles, Practices, and Prospects, 2nd ed., Springer, New York (2004). Google Scholar© 2007 American Institute of Physics.

The most important results of the scientific and scientific-organizational activities of the Institute for Safety Problems of Nuclear Power Plants of the National Academy of Sciences of Ukraine in 2023 are presented. Even under the conditions of a large-scale invasion of Russian troops on the territory of Ukraine and the continuation of hostilities by the Russian aggressor, the employees of the Institute completed all the planned tasks, which primarily became possible due to the introduction of a number of organizational measures, the effectiveness of which is confirmed by a significant increase in the quality of scientific results, their implementation, as well as the publication activity of researchers. Over the years of eliminating the consequences of the Chornobyl nuclear power plant accident, performing scientific research at the Shelter object and in the Chornobyl Exclusion Zone, a unique experience has been accumulated that allows us to solve complex tasks of ensuring the appropriate level of nuclear, radiation and environmental safety of the energy industry of Ukraine and facilities, which use nuclear technologies. The results of the Institute’s research work are implemented in practical activities at power units of nuclear power plants of Ukraine, during the scientific and technical support of works on the Shelter object transformation into an ecologically safe system and the construction of the Centralized Spent Nuclear Fuel Storage Facility. The importance of the Institute in solving topical and strategically important scientific and technical tasks in the war and post-war periods has been approved by the management of the National Academy of Sciences of Ukraine, experts in evaluating the effectiveness of its activities, as well as international partners. All this indicates the possibility of further development and, if necessary, the expansion of the scope of complex tasks, which are solved by specialists of the Institute.

The simple graphic method proposed for assessment of the operational safety of operating power-generating units of nuclear power plants makes it possible to:

Examination of VVER-440 RPV steel re-irradiation behavior using materials from operating units

Four types of surveillance programs were (are) realized in Slovak NPP’s: • “Standard Surveillance Specimen Program” (SSSP) was finished in Jaslovske´ Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4; • “Extended Surveillance Specimen Program” (ESSP), was prepared for Jaslovske´ Bohunice NPP V-2 with aim to validate the SSSP results; • For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program “Modern Surveillance Specimen Program” (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life; • For the Bohunice V-1 NPP was finished “New Surveillance Specimen Program” (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years; • New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV’s austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: • increasing of neutron fluence measurement accuracy; • substantial improvement the irradiation temperature measurement; • fixed orientation of samples to the centre of the reactor core; • minimum differences of neutron dose for all the Charpy-V notch and COD specimens; • the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV’s unit-1 and 2, based on the fundamental postulate — to provide the irradiation embrittlement monitoring till the end of units operation. The “New Surveillance Specimen Program” (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The results showed us that the NPP V-1 RPV’s can be operated without any technical limitation for minimum next 20 years. Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1 and 2) was prepared as the part of research project dealing with the WWER-440 units ageing management.