Nuclear Power Plant Equipment Research Articles

Первичное регулирование частоты тока в энергосистеме атомными электростанциями на основе водородно-теплового аккумулирования

According to the Technical Requirements for Generating Equipment of Participants in the Wholesale Market of the Unified Energy System (UES) of Russia, 2016 to participate in the general primary frequency regulation (PRCF), the maneuverable characteristics of generating equipment of nuclear power plants with VVER reactors put into operation before 2009 should ensure frequency deviations guaranteed realization of the required primary power for loading up to 2% of the nominal electric power. For this, the current capacity of the reactor installation should be maintained at a level of not more than 98% of the nominal thermal power. The fulfillment of this requirement signifi- cantly reduces the installed capacity utilization factor (CUF) of reactor plant. In addition, at present in the UES of the Russian Federation there is a tendency towards an increase in the deficit of peak and half-peak capacities. The majority of fossil fuel-fired thermal stations are switched to the half-peak mode, which negatively affects their efficiency and reliability. In addition, the rise in price of natural gas makes it more profitable to sell it abroad instead of burning at power plants. On the other hand, an increase in the share of nuclear power plants is observed in the UES, which exacerbates the problems associated with the passage of minima and maxima of the daily load in the power system, due to the economically and technically justified need to load NPPs with maximum CUF. The authors developed an approach to solving this problem by combining NPPs with an environmentally friendly energy source - an autonomous hydrogen complex (AHC), which includes thermal batteries and an additional multi- functional steam turbine unit. The developed energy complex will allow energy to be accumulated during hours of minima load in the power system due to the electrolysis of water to produce hydrogen and oxygen, as well as the accumulation of hot water in the storage tanks. The accumulated energy can be used to generate super-nominal electrici- ty to cover the half-peak load zone in the power system. In addition, the presence of a low-power steam turbine installation will ensure uninterrupted power supply to consumers of their own needs at the NPP by using the energy of the residual heat from the reactor when the station is completely de-energized. Based on the proposed energy complex, a method has been developed to ensure the participation of NPPs in the PRCF in an energy system with a constant CUF. To assess the effectiveness of the proposed solution, a methodology for thermodynamic analysis of the energy complex based on the combination of NPPs with AHC was developed. The dependence of the required hydrogen fuel consumption and the efficiency of using off-peak electricity on the temper- ature of the feed water supplied to the hydrogen-oxygen steam generator from the hot water tanks is constructed. Based on the results obtained, the technical and economic efficiency of the developed energy complex is considered. The accumulated net present value was determined depending on off-peak electricity tariffs with three variants of the forecast dynamics of the half-peak electricity tariff, taking into account natural gas savings, reduced investment in NPP safety systems and the economic effect of ensuring the participation of NPPs in the PRCF with the plant load at 100%.

SOME ASPECTS OF NON-DESTRUCTIVE TESTING OF EQUIPMENT AND PIPELINES OF NUCLEAR POWER PLANTS

Topical aspects of non-destructive testing (NDT) of equipment and pipelines of power units of nuclear power plants (NPP) are considered. Non-destructive testing is an integral part of the entire life cycle of a nuclear power plant. Briefly analyzed documentation and types of non-destructive testing at the stage of manufacturing equipment and pipelines. Power of ultrasonic testing of welded joints of pipelines with a thickness of 2 to 6 mm. The main points of drawing up technological maps of non-destructive testing are noted. A brief analysis of the application of non-destructive testing in accordance with the requirements of the regulatory documents of the Russian Federation and the safety standards of the International Atomic Energy Agency is provided. It is shown that in recent years there has been a tendency to expand the role of NDTs from the classical approach of detecting and assessing defects to the use of NDTs in the study of metal degradation and ageing processes during the operation of NPP equipment and pipelines.

Distributed plasticity approach for nonlinear analysis of nuclear power plant equipment: Experimental and numerical studies

Numerical modeling for the safety-related equipment used in a nuclear power plant (i.e., cabinet facilities) plays an essential role in seismic risk assessment. A full finite element model is often time-consuming for nonlinear time history analysis due to its computational modeling complexity. Thus, this study aims to generate a simplified model that can capture the nonlinear behavior of the electrical cabinet. Accordingly, the distributed plasticity approach was utilized to examine the stiffness-degradation effect caused by the local buckling of the structure. The inherent dynamic characteristics of the numerical model were validated against the experimental test. The outcomes indicate that the proposed model can adequately represent the significant behavior of the structure, and it is preferred in practice to perform the nonlinear analysis of the cabinet.Further investigations were carried out to evaluate the seismic behavior of the cabinet under the influence of the constitutive law of material models. Three available models in OpenSees (i.e., linear, bilinear, and Giuffre-Menegotto-Pinto (GMP) model) were considered to provide an enhanced understating of the seismic responses of the cabinet. It was found that the material nonlinearity, which is the function of its smoothness, is the most effective parameter for the structural analysis of the cabinet. Also, it showed that implementing nonlinear models reduces the seismic response of the cabinet considerably in comparison with the linear model.

Effect of TiC particles on ductility dip cracking susceptibility of Ni-base superalloy

Ni-base superalloy FM-52M is mainly used in welded structures of key equipment of nuclear power plants, such as pressure vessels and heat dissipation evaporators. It is prone to form coarse columnar microstructures during welding and has significant ductility dip cracking (DDC) susceptibility. This work attempted to introduce micrometre-sized TiC particles into FM-52M welds to refine the microstructure and alleviate sensitivity to DDC. A Ni–TiC coating with good dispersion and uniformity of TiC particles was first prepared on the substrate by composite electroplating. The TiC particles were then directly introduced into the molten pool during tungsten–inert gas welding. Scanning electron microscopy and electron backscatter diffraction were used to analyse the microstructure of the weld and strain-to-fracture testing was performed to evaluate its DDC sensitivity. The experimental results showed that solidification grains of the weld were obviously refined and its high-temperature mechanical properties were reinforced. The minimum critical strain for DDC increased to approximately 3.5%, and the crack morphology was small and tortuous. Compared with the original material, the microstructure and DDC resistance of the weld were improved by introducing TiC particles.

Analysis and Optimization of Three-Beam Traverse Structure Elements for Reactor Cavity Reinforced Concrete Block Installation

Increased demands are placed on the reliability and strength of structures used in the installation of equipment for nuclear power plants. At the same time, an excessive safety margin of the equipment used leads to an increase in its dimensions, weight and a significant rise in cost. The issue of reducing the metal consumption of structures while maintaining the required performance criteria for this equipment is very relevant. The paper presents a model for optimizing the design of a three-beam traverse with a carrying capacity of 100 tons for reactor cavity reinforced concrete block installation. A verification calculation is carried out and recommendations on the optimization of design parameters are proposed on its basis.

Influence of the composition of α-titanium alloys on thermal conductivity

Among titanium alloys, modern α- and pseudo-α-alloys occupy a special place due to the unique combination of their mechanical properties, corrosion resistance, low density and high specific strength, which determines their effectiveness in various industries. Analysis of structural materials used for heat exchange equipment of nuclear power plants showed that the increase in the efficiency and compactness of tube systems made of a-titanium alloys is constrained by their thermal conductivity characteristic, which does not exceed 89 W/(m·K) at a temperature of 20°C. An exception is the VT1-0 grade alloy, the scope of which is limited to a maximum operating temperature of no more than 250°C. The paper considers the results of studies of a new titanium alloy of the Ti-Zr-Al-O composition with increased thermal conductivity for pipe systems of power equipment.

Evaluation of through-thickness residual stresses and microstructure in SA516 Gr. 70 steel welds

The critical working condition of nuclear power plant equipment necessitates meticulous determination of the welding process and parameters. In this work, some major influential factors of welding were investigated to observe their effects on the through-thickness residual stress distribution in multipass pressure vessel steel welds. In this regard, experiments were conducted to find the characteristics of residual stresses dispersed in SA516 Grade 70 steel welds of different groove geometries with distinct welding conditions. Three-dimensional finite element models of the weldments were developed considering a moving heat source with temperature-dependent material properties to simulate the welding thermal cycles and corresponding residual stress fields. Effects of weld groove geometry, number of weld passes, external constraints, and preheating on the through-thickness residual stress fields were studied. Additional attention was given to the evaluation of the heterogeneous microstructure and microhardness across the weld cross-section associated with their weld thermal history. Finally, the evolution of the through-thickness residual stresses attributed to subsequent weld passes was elaborated.

Study on Vibration Control of Double Motor Refrigeration Unit in Nuclear Power Plant

In order to solve the problem of excessive vibration during the operation of the refrigeration unit in nuclear power plant, taking the double motor refrigeration unit in operation as the research object, we studied the effective measures to reduce the vibration and improve the stiffness of the double motors in the near resonance region. On the basis of theoretical calculation, the weaknesses of the equipment stiffness were reinforced and reconstructed, and the secondary analysis and calculation of the stiffness and seismic performance of the reformed refrigeration unit were carried out to determine the feasibility of the optimization scheme. Based on the analysis, we optimized the equipment. After that, there was a significant decrease in the vibration of the refrigeration unit. The vibration measurement results indicated that the vibration value of the unit has dropped to the qualified range, effectively avoiding the equipment damage caused by excessive vibration. We provided a complete set of analysis ideas for the research and solution of vibration problems of rotating equipment in nuclear power plant.

Research on Maintenance Method of Safety-level Switching Power Supply Module in Nuclear Power Plant

With the technological development in the domestic electronic industry and the urgent demand of cost reduction and efficiency improvement in nuclear power plant, the electrical equipment in nuclear power plant has changed from replacing with brand-new equipment regularly or after failure to reusing after aging mitigation. At present, it is very urgent to research the maintenance methods of safety-level electrical equipment. Through the research on the failure mode and influence analysis, vulnerable components and parts, post-repair test and post-repair service life calculation of the safety-level switching power supply module, the maintenance methods are systematically sorted out. Through specific case analysis, the aging of power supply module can be mitigated and the reliability of it can be improved. At the same time, the method can be extended to other safety-level electrical equipment maintenance with one of the characteristics of unavailability of spare parts, high price, and large quantity or reverse demand.

Study on the Development of I&C Equipment Performance Monitoring Program for Nuclear Power Plants

The equipment performance monitoring program is designed to provide guidance for monitoring the performance of equipment, with an aim to enable engineers to identify hidden hazards as early as possible and take corresponding corrective actions to achieve the purpose of equipment reliability management of nuclear power plant. The nuclear power plant realizes monitoring, control and protection of the whole plant through I&C equipment, by which to ensure the safe, reliable and economic operation of the nuclear power plant. In this paper, a set of performance monitoring program process special for I&C equipment of nuclear power plant is developed. And the RGL system (rod control system and rod position system) of a nuclear power plant is taken as an example to illustrate, analyze and summarize, which lays a foundation for nuclear power plant to carry out equipment reliability management.

Brief Discussion on Evaluation & Demonstration of Operating Reliability of Safety-level I&C System Equipment in Nuclear Power Plant

Safety-level system equipment is essential for the nuclear power plant either to complete the reactor emergency shutdown, containment isolation, reactor core cooling and heat emission from containment and reactor, either to prevent large amounts of radioactive substances from being released into the environment, and is also the focus of system equipment reliability management. As the safety-level system equipment has seismic requirements, in this paper, the operating reliability of safety-level I&C system equipment is evaluated and demonstrated, and the general process of reliability evaluation & demonstration is worked out.

Analysis and Performance Evaluation of Casting Inclusions in the Reactor Coolant Pump Casing of Nuclear Power Plant

The casting quality of the coolant pump casing of the nuclear power plant reactor is directly related to the operational reliability and safety of the nuclear main pump, and plays a key role in the integrity of the pressure-bearing boundary of the reactor primary loop. In this paper, aiming at the low impact performance of the sample during the casting process of the main pump casing of a nuclear power plant, through using failure analysis tools like fishbone diagram from multiple dimensions such as material selection, design and technology, melting analysis, pouring process, riser design, and heat treatment process, and combining with metal macro-fracture analysis and micro-electron microscopy scanning methods for cause analysis, finally, it was found that the basic reason for the low impact performance of the pump shell is that the secondary inclusions appear on the fracture of the sample during the solidification of the molten steel. Using test-retest inspection and finite element mechanics simulation analysis, the comprehensive evaluation of the impact performance of the sample was obtain, which provides an effective solution for the analysis and evaluation of casting inclusions in water pumps of nuclear power plants, and also provides an important reference for the structural optimization and equipment research and development of water pump equipment of nuclear power plants.

Исследование процессов усталостного и коррозионно-усталостного разрушения псевдо- титанового сплава

The paper describes the results of fatigue and corrosion-fatigue tests of the pseudo α titanium alloy PT-3V, which is actively used in nuclear engineering for the manufacture of heat exchange equipment for modern nuclear power plants. Alloy PT-3V has an inhomogeneous coarse-crystalline structure with precipitates of β-phase particles along the grain boundaries of the lamellar shape. It is shown that smooth specimens tested according to the bending-with-rotation loading scheme do not show a noticeable decrease in the cyclic fatigue life when exposed to a neutral corrosive environment (3 % aqueous NaCl solution). However, specimens with a notch (stress concentrator) tested according to the cantilever bending loading scheme demonstrate sensitivity to the action of a corrosive environment at the stages of initiation and growth of fatigue cracks, as evidenced by a significant decrease in the number of cycles before crack initiation, as well as before specimen failure, in comparison with tests in air. Fractographic analysis of fractures of smooth specimens and specimens with a concentrator after fatigue and corrosion-fatigue tests has been carried out. The main stages of the initiation and growth of fatigue cracks are revealed. It has been established that a decrease in the resistance to the initiation and propagation of corrosion-fatigue cracks during testing of notched specimens may be due to the effect of hydrogen embrittlement, accelerated by stress concentration.

Comparative analysis of the assessment of material costs on an ARMS system and the elimination of the consequences of a radiation accident at an atomic energy facility

The work provides a comparative analysis of the material and financial costs for the development and operation of an Automated Radiation Monitoring System (ARMS) for monitoring of the environmental radiation situation and the damage associated with the elimination of the consequences of a radiation accident at an Atomic Energy Facility. Furthermore, we consider a brief scenario of the sequence of operations of the ARMS system related to the assessment of radioactive pollution of the environment, in the framework of a hypothetical accident of level 7, on the INES scale, and that of the Atomic Energy Facility. The work presents the main methods and means of the ARMS system. The system helps to carry out modern methods of assessing dose loads on the personnel and the population of the region exposed to a radioactive release torch. It also helps in the elimination of the effects of radioactive pollution of the environment and its damage to the population, agricultural land, parks and reserves. The maintenance work and the evacuation of the population from the contaminated area is considered. It shows that the material and financial costs of nuclear power plant equipment, which allows for radiation monitoring, maintenance, as well as work-related to prognostic assessments of radioactive pollution of the environment, are much less than the costs of eliminating the consequences of severe radiation accidents at nuclear power plants.

Plastic Strain Correction in Fatigue Analysis of Nuclear Power Plant Equipment

Plastic Strain Correction in Fatigue Analysis of Nuclear Power Plant Equipment

Optimization of Systems Repair Plans and Assessment of the Useful Life of Nuclear Power Plant Equipment

This paper presents a computer-aided method of planning the volumes of repairs of systems of nuclear power units and a method for calculating their gamma-percentile life. This planning is carried out on the basis of predicting the reliability indicator, the probability of no-failure operation for a certain time period, with the gamma-percentile life of the equipment being determined by solving the corresponding equations. The tasks considered are related to an important energy problem of extending the operation of nuclear power units. Its importance is determined mainly by economic feasibility: it is cheaper to assess the useful life of a nuclear power unit and, on this research basis, extend its operation, than create a new unit. It is also shown that the calculation of the probability of a radiation accident at a nuclear power unit is associated with the results of planning the repairs of its systems, with assessment of its useful life. An optimization problem is formulated: it is required to find such a plan for the volumes of repair of a system that, with limited repair costs, its reliability indicator for a given duration deviates least from the maximum permissible value. The solution to the problem is based on calculating the structural reliability of the system. A graphological image of the system is built in the form of a composition of graphological images of typical structures. After the reliability indicator of typical structures has been calculated, the structures are replaced with individual structural elements, which makes it possible to simplify the initial graphological image of the system in a computational scenario and calculate its reliability indicator. The determination of the repair volume is carried out by applying a version of the coordinate-wise optimization method. To assess the gamma-percentile life, a model is adopted, in which the recoverable equipment components have an unlimited life, although, of course, they "age", and the non-recoverable components spend their life up to the level when their replacement becomes conditioned by the violation of the requirement for the maximum permissible value of the system reliability indicator. Estimates of the gamma-percentile life of the equipment are calculated by planning system repairs on a sequence of intervals of annual energy production by a nuclear power unit.

Operability Study of Overhung Pump Being Under Special Dynamic Loads

One important part during designing of pumping equipment for nuclear power plants is to ensure pumping equipment operability and stiffness under extreme dynamic loads of natural and anthropogenic nature. One of the most dangerous and intensive natural sources is seismic load, and the most dangerous anthropogenic impact are considered to be air shock wave (ASW) and aircraft crush (AC) on the enclosures of the reactor building. At the present time mathematic models and methods for calculation of NPP equipment operability under special dynamic loads (SDL) are applied to define stress-strain state of the design in whole, to assess local deformation in zones having the highest level of equivalent stresses and for studying the design state in other extreme conditions affecting the safety of NPP. Calculation estimation of the design dynamic response to external loads of natural and anthropogenic nature is necessary to avoid wrong decision solutions at the design stage and manufacturing of new equipment.

ТЕСТОВЫЕ РАСЧЕТЫ ГИДРОДИНАМИКИ РАЗДАЮЩИХ КОЛЛЕКТОРНЫХ СИСТЕМ ТЕПЛООБМЕННИКОВ И РЕАКТОРОВ ЯЭУ

The paper presents the results of test calculations of hydrodynamics for distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The header system is a typical element of the flow path in the specified NPP equipment, their efficient and safe operation is a priority area for Rosatom State Corporation. The calculations have been performed with the use of computational fluid dynamics software systems known as CFD codes. The results of calculations are presented for the coolant turbulent flow in the flow paths of the flat and cylindrical DHS with different conditions for the coolant supply and removal. A comparative analysis of the calculation results and experimental data on hydrodynamics of DHS typical of the nuclear power plants has been carried out, their satisfactory agreement has been shown. An assumption is made of the legitimacy of using the CFD code for the calculation of hydrodynamics of various DHS types and their specific typical elements.

Stress Strain State of Steam Turbine Components under Plastic Deformation

Reliable operation of secondary equipment of PWR nuclear power plants is an integral part of nuclear and radiation safety of the entire NPP unit. The problem of stress strain state of steam turbine structural components under plastic deformation is considered. The theory of elastic-plastic deformations is used to solve the problem along with the finite element method. The paper presents the results of computer assessment of stress strain state of locking joint of working blades of the first stage of a intermediate-pressure cylinder (IPC) and high-pressure cylinder (HPC) body of a steam turbine, which makes it possible to characterize the degree of relaxation and stress redistribution in the structure in comparison with obtained earlier results. It is provided that the use of the presented calculation method in designing new structures of steam turbine components operating in the area of high thermal and power loads, taking into account the contact interaction of components, as well as different mechanical and physical properties of materials, and their changes depending on operating temperature, at this stage of software development allows one to identify problem areas in the design and prevent further breakdowns in the turbine. Based on the comparison of operational data of the developed design solutions and calculation assessment, it is proved that the chosen calculation method can significantly increase the operational reliability not only of the turbine unit but also the nuclear unit as a whole, as well as reduce economic costs caused by turbine unit downtime during maintenance.

Distributed parameter modeling for the steam generator in the nuclear power plant

Steam generator is an important equipment in nuclear power plants. This paper investigates the dynamics and heat transfer process in the vertical U-tube natural circulation steam generator. The steam generator under study is divided into four regions, i.e., the hot leg, the cold leg, the steam separator and the saturated steam chamber. Correspondingly, the mass, momentum and energy conservation equations are derived with distributed parameter modeling approach. The model system is validated with the history operation data from a Chinese nuclear power plant. The calculated thermodynamic properties of the saturated steam at the outlet of the steam generator are found to agree well with the measured values, which indirectly validates accuracy of the model system. Model-based dynamic simulation software for the steam generator is developed with which the heat transfer coefficient distribution, the pressure distribution, the temperature distribution, and the speed distribution of the working medium are obtained in real time.