THEORETICAL GROUNDS FOR THE EQUIPMENT'S OPERABILITY AND SAFETY ESTIMATIONS OF NUCLEAR POWER PLANTS WITH EMERGENT PROPERTIES TAKING INTO ACCOUNT

O.V. Yefimov, Yu.V. Romashov
2017 Trudy Odesskogo Politehničeskogo Universiteta  
Ю.В. Ромашов. Теоретичні основи оцінки працездатності обладнання атомних електростанцій та його безпечної роботи з урахуванням синергетичних ефектів. Обговорюються підходи до оцінки працездатності та безпеки обладнання АЕС. Метою даної статті є концептуальна розробка підходів, моделей і методів, які роблять можливим отримання оцінки працездатності і безпеки АЕС з урахуванням можливого виникають властивостей цих систем обладнання. Наукова і практична цінність результатів є розробка підходів,
more » ... робка підходів, моделей методів, які дозволяють отримати оцінки працездатності експлуатації обладнання та безпеки АЕС, беручи до уваги можливі емерджентні властивості, які нехтували в підходах, моделей і методів, добре знаних сьогодні. Розглягуто загальні підходи до побудови математичних моделей відповідального обладнання АЕС, яка з урахуванням взаємодії різних зовнішніх факторів, які можуть призвести до синергетичних ефектів. Основний результат є встановлена можливість математичного моделювання обладнання АЕС у вигляді систем, пов'язані крайової і початково-крайових задач, що дозволяє брати до уваги можливі емерджентні властивості. Головний висновок перспективних застосувань методидискретизація для чисельного аналізу елементів АЕС. Значення отриманих результатів полягає в розробці теоретичних основ для оцінки працездатності і безпеки АЕС обладнання шляхом з урахуванням можливих емерджентних властивостей. Ключові слова: елементи обладнання та технологічні процеси АЕС, зовнішні впливові фактори, синергетичні ефекти, працездатність, міцність, руйнування, надійність, математичні моделі, чисельне інтегрування A.V. Yefimov, Yu.V. Romashov. Theoretical grounds for the equipment's operability and safety estimations of nuclear power plants with emergent properties taking into account. The approaches to the evaluation of equipment's operability and safety of NPP are discussed. The purpose of this article is the conceptual development of approaches, models and methods that make possible to obtain assessments of the equipment's operability and safety of NPP taking into account possible emergent properties of the systems. The scientific and practical value of the results of this article is to develop approaches, models and methods that allow us to obtain estimates of the equipment's operability and NPP safety, taking into account the possible emergent properties that are neglected in approaches, models and methods well-known today. General approaches to the construction of mathematical models of the responsible equipment of NPP, which taking into account the interaction of various external factors that can lead to synergistic effects, are considered. The main result is the established possibility of mathematical modelling of NPP equipment in the form of systems of connected boundary and initial boundary value problems, which allows us to take into account possible emergent properties. The main conclusion is promising applications semi-discretization methods for numerical analysis of NPP elements. The value of obtained results consists in developing of theoretical foundations for estimating the equipment's operability and safety of NPP by taking into account possible emergent properties. Evidence of emergent properties presence based on the operation experience of NPP units. At the present time, the problem of the operability estimating of the critical elements of the nuclear power plant equipment, including the in-house elements of nuclear reactors and steam generators, has not been fully resolved, as evidenced by damages of different equipment's elements that occur periodically during the operation of nuclear power units, and the causes of many such damages are still accurate not understanding. Cases of failure in the operation of the control and protection system due to the delay in the fall of the regulating rods due to the curvature of the guide channels of the fuel assemblies since 1992 have been recorded on most VVER-1000 units [5] . The causes of the guide channels curvature consist in the loss of stability due to excessive compression [5] , which agrees with the approximate estimates of the critical compressive forces of the guide channels [6]. Although the measures taken to reduce the compressive forces have led to a positive effect [5], the reasons for the excessive "clamping" of the guide channels have not been precisely established and, apparently, are due to the emergent properties of the complex of external factors. There are all known cases of depressurization of fuel rod cladding [5] , which can be dangerous because of the increase in radioactivity of the coolant because the cladding is the first protective barrier separating nuclear fuel from the surrounding environment; exceeding the permissible number of destroyed fuel cladding requires the shutdown of the nuclear reactor and the reloading of nuclear fuel. Depressurization is a consequence of the destruction of the material of the fuel cladding, which occurs with the simultaneous occurrence of many processes, including thermo elastic deformation, corrosion damage, creep, embrittlement under mechanical contact between fuel and cladding, or high gas overpressure, which are determined by thermal conductivity and heat transfer, emission of gaseous products fission and diffusion [7]. Thus, the destruction of the fuel cladding should be considered as the emergent properties effect, i.e. as an unobvious result of the interaction of a multitude of factors. At the same time, there is an uncertainty of the initial data necessary for estimating the durability of the fuel cladding [8], which leads to the possibility of destruction due to the unevenness of the neutron flux and the probabilistic nature of nuclear reactions and other factors. The damages of the output collectors were discovered in 1986 after several years of operation on 25 PGV-1000 steam generators and led to repairing of two and replacing the 23 steam generators at operating NPP [9]; these damages are due to the emergent properties presence and, as noted in [9] , are caused by the complex multifactor effect of different nature factors. It was possible to develop measures that reduce the residual stresses and allow excluding such damage of the output collectors in the future [9], despite the lack of a deep understanding of the processes that caused damage. Nevertheless, the study of the damaging process of the output collectors of PGV-1000 steam generators is of considerable interest, especially due to the fact that the damages was observed only at the output collectors having a lower temperature than the input collectors. Damages of the heat exchanging pipes of steam generators of NPPs with VVER are due to the corrosion cracking of stainless steels [9] . The stress corrosion cracking occurs only under the combined action of tensile stresses and corrosive environments [10] and must be considered as the emergent properties effect. It is required the development of a model for stress corrosion cracking of stainless steels [10], taking into account the mechanical stress factors and chemical composition of the corrosive medium, to estimate the quantitative data about the effect of mechanical stress factors and chemical properties of the corrosive medium on cracking time and reliability parameters of heat exchanging pipes of steam generators of nuclear power plants with VVER which was carried out in [11]. The results of the researches [10, 11] showed that relatively small changes in the tensile stress and concentration of the corrosive component lead to a significant change in the time of stress-corrosion cracking of the heat exchanging pipes of steam generators of NPP with VVER, which explains the possibility of cracking of individual heat exchanging pipes and the normal operation of other heat exchanging pipes in one steam generator at the same time. More reasonable estimates of cracking time require researches of regularities in the concentration of corrosive active components on heat exchanging piped of steam generators of nuclear power plants.
doi:10.15276/opu.2.52.2017.07 fatcat:6oedgyg6jffupf6tngpbd4zgqy