Main Circulation Pipeline Research Articles

Assessment of Power Equipment Operational Safety in the Sustainable Management of Residual Lifespan

Introduction. Ensuring the safe operation of energy facilities is a critical issue at all life cycle stages. It is associated with the adoption of management decisions when rescheduling and assessing the remaining resources, which are made based on a comprehensive assessment of the technical condition of the equipment, taking into account the economic efficiency and environmental requirements for further operation, which is the basis for technical and economic assessments. Aim and tasks. This study aims to develop a technical and economic assessment of the safe operation of power equipment to ensure sustainable management of the remaining service lifespan. Results. An algorithm for assessing pipeline system elements by the coolant environment is proposed, which allows for the pipeline load and changes in erosion and corrosion caused by coolant movement. It was proposed that the equipment be systematised by safety class and coolant. When conducting a technical and economic assessment, these parameters and the costs of extending the service life and upgrading the existing equipment were considered in the standard calculations of a certain type of equipment. For this purpose, mathematical models for calculating the residual life for each type, obtained and confirmed based on empirical studies, are proposed. Research findings into ensuring the reliability and safety of operation with various heat carriers have shown that it is necessary to consider low-cycle and high-cycle loads and erosion processes. It was found that the erosive wear rate is 0.3-0.4 mm/year for bending sections of the main circulation pipeline, where maximum loads occur relative to low-cycle and vibration loads, and the stress is 212.5 MPa when assessing the residual life of 7 years. Conclusions. The cost-effectiveness of reassigning the service life is reasonable, as it does not require additional investment in upgrading or replacing equipment subject to annual technical diagnostics. The systematisation of pipelines, improved codifiers, lifespan assessment models, and algorithms for evaluating the remaining life under various loadings enhance the regulatory framework for safe nuclear power plant operations and support technical and economic evaluations for extending their lifespan.

Use of NEWS and MMM technologies for diagnostics of NPP components

The paper provides information on the results of the R&D projects of the Research Centre Rez carried out between 2019 and 2022 for the Technological Agency of the Czech Republic as part of the "Perspective Diagnostic Methods" projects. In the first part of this paper, the development of the use of NEWS technology for the detection of cracks in pipelines is described. Results from development on half-pipe and pipe test pieces are presented, as well as results from measurements on the main circulation pipeline and connection pipes up to the first shut-off fitting at NPP Temelín. The method of certification and verification of the methodology is also described. The second part of this paper describes the development of the use of MMM technology to detect the state of tightening of bolts of flange joints. The development focused on determining the dependence of the tension in the bolt and the values obtained by the MMM method, determining the influence of the measurement location for the MMM method on the assessment of the condition of the flange joint, and determining the sensitivity of the method to unsatisfactory flange conditions such as undertightened bolts, overtightened bolts etc. is described. At the end, implementation plans are presented, either for specific applications at the NPP or in a subsequent R&D project.

RESEARCH BY DIRECT METHODS OF THE BASE METAL OF MAIN CIRCULATION PIPELINE OF THE POWER UNIT 3 OF THE SOUTH UKRAINE NPP AFTER 200 THOUSAND HOURS OF OPERATION

The base metal (BM) of the main circulation pipeline (MCP) of the unit 3 of the South Ukraine NPP, which has been in operation for 200 thousand hours, was investigated by direct methods. It is shown that the mechanical properties and structure of BM of the MCP after 200 thousand hours of operation satisfy the requirements of regulatory documentation for metal. A comparison of the test results of the BM in the initial state and after 100 and 200 thousand hours of operation was carried out. It is shown that in the process of operation there is a decrease in the rate of changes in the mechanical properties of BM, in particular, during the last 100 thousand hours of operation in comparison with the previous ones.

Modeling the process of mixing the coolant in the lower section and pressure mixing chamber of VVER-1200 (V-491)

The most representative accident scenario is determined and a physical and mathematical model is developed for studying the mixing of non-isothermal coolant flows in the structural elements of the V-491 reactor facility (VVER-1200), in which the motion of the medium is described in a three-dimensional non-stationary formulation. Based on the analytical estimates of the list of initiating events, a scenario was chosen with the connection of an idle loop of the main circulation pipeline to three operating ones without a preliminary power reduction. A computational algorithm and a numerical method have been developed for the computational analysis of the selected accident scenario and the justification of the safety of operation of the V-491 reactor facility (VVER-1200). During the numerical simulation, the RANS method was used, which consists in solving the Reynolds-averaged Navier–Stokes equations, the continuity equation and the energy equation. The SST k–ω model of turbulence by Florian Menter is used to close the equations. The verification of the developed physical and mathematical model and calculation procedure was carried out by modeling thermohydraulic processes in models with both a relatively simple geometric design (tee connection) and in a scale model of the reactor vessel (ROCOM experiment), including a lowering section and a pressure mixing chamber. The qualitative agreement of the numerical simulation results with the available data of physical experiments is shown. The results of numerical simulation of the mixing process of non-isothermal coolant flows in the section from the branch pipe of the “cold” thread of the main circulation pipeline to the lower boundary of the fuel of the VVER-1200 core (V-491) are presented. It is shown that the heterogeneity in the temperature distribution at the entrance to the core manifests itself up to 15.5 s of the calculated accident scenario. For calculations the following code coupling were used: Ansys Fluent/Rainbow-TPP.

THE SAFETY FACTORS OF PROBABILITY REACHING LIMITING STATES OF NPP EQUIPMENTS AND PIPELINES DURING DIFFERENT OPERATING MODES

Reliability theory makes it possible to determine the probabilities of destruction, reaching limit states, equipment and pipeline failures (E&P). Normative documents, as a rule, define requirements for integral indicators of reliability or safety of objects, but do not establish requirements for admissible values of probabilities for individual systems and data elements of objects. This article proposes the approach for determining the permissible values of the probabilities of reaching the yield stress or ultimate strength by operating stresses on the basis of design data on loads and data from certificates on the mechanical properties of metal. During operation, the achievement of the working stresses in metal of E&P value of the permissible stress values is a probabilistic event, in this fact, the probabilistic approach has been developed to obtain the safety factor of probability reaching limiting states. The developed approach is based on the methods of strength science, statistical analysis and probability theory. The object of application of the developed approach is thermal mechanical equipment used in different branches of technology. Using the example of the main circulation pipeline of the NPP with WWER-440, the presence of a reserve in the probability of reaching the limit state is shown and a relationship is established between the stress variation coefficient and the calculated probability for normal operation and abnormal operation. This approach determines the reserve for the probability of destruction, which will justify the extension of the resource or optimize the operating parameters of the objects under consideration.

Analysis of accidents of the WWER-1000 reactor in which emergency cooling heat exchangers operate

The object of research is the emergency operation modes of the WWER-1000 nuclear installation elements, which also use emergency cooling heat exchangers. The analytical studies carried out are based on an analysis of the operating documentation of the above-mentioned nuclear installation. An analysis of the elements of the emergency core cooling system, which includes emergency cooling heat exchangers, has been carried out. This analysis has shown that in order to localize the accidents of the WWER-1000 reactor unit, the coolant that comes out of the leak is collected in the tank of the plant and from there is pumped by pumps through the above-mentioned heat exchangers. This ensures that the heat of the coolant escaping from the leak is removed and the reactor core is effectively cooled using already cooled water.As a result of a comparative analysis of design accidents at the WWER-1000 reactor unit, it has been established that the emergency core cooling system is involved in accidents related to a rupture of the first and second reactor circuit pipelines within the containment. Such accidents include small, medium and large leaks in the primary circuit as well as ruptures in the steam line or feed water pipeline of the steam generator (within the containment). A detailed review of the parameters of the coolant that flows out of the leak and into the tip-tank shows that the most conservative accident involving emergency cooling heat exchangers is a «Large leak mode. Bilateral rupture of the main circulation pipeline DN 850 mm». In this accident, the temperature of the coolant in the tip-tanks reaches 110 °С and subsequently enters the inter-tube space of the heat exchangers.Currently, there is an urgent task to justify the safe operation of nuclear power plant equipment and emergency cooling heat exchangers in particular. The results of the research can serve as baseline data for determining the thermal stress state of the above-mentioned heat exchangers in accidents of the WWER-1000 nuclear installation.

Computational modeling of boric acid droplet entrainment during water-water energetic reactor operation in emergency boiling mode

The results of a computational modeling of the processes of droplet entrainment of boric acid by a vapor stream in the event of an emergency at a water-water energetic reactor (WWER) nuclear power plant associated with a guillotine rupture of the main circulation pipeline and the loss of all alternating current sources are presented in the article. The dependence received in the course of computational modeling makes it possible to estimate the maximum size of the droplets carried away by the steam stream during the emergency process in the WWER reactor unit during the operation of the complex of passive safety systems in the pressure range 0.2-0.5 MPa. It was established that to the end of 72 hours of an accident the maximum size of droplets carried away from reactor is 0.11 mm.

Estimation of the resource of pipes of the main circulation pipeline of a NPP by the criterion of crack resistance with account of operational degradation of properties

The results of an experimental determination of the characteristics of cyclic crack growth resistance of specimens made of stainless steel 08Kh18N12T are given. Compact samples 20 mm thick cut from Du 550–30 pipes of the primary circuit of the main circulation pipeline (MCР) were studied under eccentric tension in the initial state and after 100 thousand hours of operation in the conditions of the Novo-Voronezh NPP. Using a new engineering approach based on measuring the specimen tightening and constructing a cyclic elastoplastic fracture diagram for a specimen with a crack in the coordinates “tightening – the stress intensity factor, calculated taking into account the plasticity correction”, a calculated life estimation is given for the pipe of Du 550–30 MCР NPP, made of 08Kh18N12T steel, according to the criterion of cyclic crack growth resistance, taking into account operational degradation of properties caused by the running time of 100 thousand hours. The life (survivability), expressed in a number of cycles of change in internal pressure, for a pipe with the most dangerous defects (surface semi-elliptical crack and combined defect “pitting + crack”) is determined. An analysis of the influence of the level of initial damage and the type of hazardous defect on the survivability of the pipes is made. The results can be used to analyze damage and predict the life of pipes made of ductile metals and alloys, in cooling systems of reactors, as well as main pipelines of hydrocarbon transport.

Modeling the emergency processes proceeding in the containment of NPP with water-water energy reactor

The article presents the results of computational modeling of the development of accidents in the containment of NPP with water-water energy reactor (WWER) using the COCOSYS software. The containment of the Balakovo NPP with WWER-1000/V-320 was selected as the object of study. The results of the calculation of environmental parameters in the containment during a beyond design basis accident with a “large” leak from the primary circuit at the following initial events: break of the main circulation pipeline DN850 with a full cross section at the reactor inlet, with simultaneous failure of all AC sources including diesel generators for a long period, without operating personnel intervention. A comparison of the results of calculations with the results obtained using the ANGAR software certified by Rostekhnadzor is performed. The calculation results are in good agreement: emergency scenarios are similar; the maximum deviation from absolute pressure values is less than 10 %. Thus, it was concluded that the analysis of processes occurring in the NPP containment during an accident can be performed using the COCOSYS code. This software can be used to independent safety assessment of NPP required by the IAEA standards.

НЕОПРЕДЕЛЕННОСТЬ В РАСЧЕТАХ ИЗ-ЗА «МГНОВЕННЫХ» АВАРИЙНЫХ СИТУАЦИЙ НА ВВЭР-1000

The paper shows the occurrence of pressure waves and their propagation in the equipment of the first circuit of the WWER-1000 reactor plant in emergency situations associated with an instantaneous stop of the Main circulation pump or a two-way flow with an instantaneous two-way break in the cold thread of the main circulation pipeline. The influence of the time of initialization of the accident (pump stop, pipeline rupture) on the intensity of the process-amplitude, frequency of pressure changes is investigated. Pressure drops during an emergency on the main elements of the circuit are considered. It is shown that the maximum changes in the amplitude and frequency of both the pressure and the pressure drops on the circuit elements belong to the initial stage of the accident. The main attention is focused on the pressure drops on the equipment, because this parameter determines the dynamic loads on the equipment, which can lead to its failure.

Electroslag deposition with tape under a ceramic flux of a homogeneous single-layer anti-corrosion coating of the inner surface of pipes with a diameter of 850 and 990 mm for nuclear power plants

ABSTRACT Electroslag strip overlaying process with agglomerated flux using external magnetic field and without it is examined. Minimal strip size to use external magnetic field for improvement bead shaping quality is determined. The impact of external magnetic field’s parameters at bead shaping features is researched. Also as the optimal magnetic field’s parameters, the optimal mode parameters of the process (e.g. weld current, weld voltage, weld speed) are determined. Influence of displacement from zenith at bead shaping quality is examined. Proposed ‘slalom’ and ‘spiral’ realization of electroslag strip overlaying process provides incessancy of overlaying on pipe’s inner surface. The technology of electroslag strip homogeneous single-layer corrosion resistant overlaying with agglomerated flux on pipe’s inner surface with a diameter of 850 and 990 mm is developed. These pipes are applied for main circulation pipeline’s manufacturing of nuclear power plant type V-1000 and V-1200.

A Setup for Complex Investigation of Mechanical Characteristics of Structural Materials for NPP Equipment

The complex experimental equipment has been developed and manufactured in compliance with the existing standards for measuring hardness and strength characteristics via the method of instrumented indentation, as well as for short-term static uniaxial tensile tests of structural materials and small punch testing of miniature disk specimens in the macrorange of loads of 2–10,000 N. This allows one to obtain the necessary characteristics of hardness and strength in relation to the current state of materials of the main equipment at NPP and thereby minimize the volume of material used for the production of specimens on the same setup employing various normative techniques of both destructive and non-destructive inspections. The equipment is designed in the form of replaceable units: indentation, tension, and punching of miniature disk specimens, which are installed depending on the test method. As a result, there is a unified assembly that makes it possible to reduce the effect of instrument errors in the validity of the data obtained using different test methods and facilitate its modification according to the other tests. The developed experimental equipment has been tested using the segment of metal of the double ball pipe wall ∅990 × 70 mm (the basic metal is 10GN2MFA steel and the deposited material is 08Kh19N10G2B steel) of the Du850 main circulation pipeline of NPP from WWER-1000 reactor pressure vessel. The mechanical characteristics obtained by the methods of indirect control were compared with those obtained from standard tensile tests. At the same time, the deviation of the values of yield strength and ultimate strength obtained employing the indirect control methods from the values obtained in the standard tensile testing is within the limits recommended by the normative documents.

Static strength calculation of DN 350 branch pipe cutting-in in the main circulation pipeline of reactor plant WWER-1000

Nowadays the total electricity generated in nuclear power plants in Russian Federation is approximately 18% of all power generation. There are 36 reactors in operation in Russia, 19 of which are reactor plants WWER-1000. A planned operation lifetime of these plants is 30 years and most of them have served for this period. To extend reactor working lifetime it is necessary to carry out the calculation for possibility justification of continued operation of reactor equipment and pipelines. The paper presents the static strength calculation of DN 350 branch pipe cutting-in in the main circulation pipeline of reactor plant WWER-1000. According to calculation results a following conclusion is made: if the static strength of DN 350 branch pipe cutting-in meets the requirements of technical standard, the continued operation of this pipeline section will be possible.

Calculation analysis of the processes of boric acid mass transfer in the WWER core in boiling emergency mode

In this paper the processes of boric acid mass transfer in a WWER-TOI nuclear reactor core in case of the accidents with main circulation pipeline rupture and loss of all AC power supply are considered. The heat removal from the WWER-TOI reactor in case of an emergency process is determined by the use of the passive heat removal and passive core cooling systems. Passive core cooling system of WWER-TOI consists of three stages of hydro accumulators. These systems provide reactor cooling by feed of boric acid solution with concentration of 16 g H3BO3/kg H2O. In view of length of accident process, the boiling of the coolant with high content of boron and considering low concentration of boric acid in the steam leaving the core, conditions may arise for the possible accumulation and subsequent crystallization of boric acid in the reactor that can lead to a deterioration of the process of core heat removal. To assess the chance of accumulation and subsequent crystallization of H3BO3 in the core of the WWER reactor, a hand calculation analysis of the change of concentration of boric acid in the reactor in case of emergency mode was carried out. In accordance with the calculation results, if a boric acid concentration in hydro accumulators of passive core cooling systems is 16 g / kg H2O, than a large excess of the final boric acid concentration in the core is observed in accidents after approximately 43 hours of an emergency process. The options of reducing the concentration of H3BO3 to 8, 4, 2 and 1 g/kg H2O are considered in the calculation process. The obtained results allow us to conclude that a decrease in the concentration of boric acid in the system of hydro accumulators of the third stage would allow to avoid the crystallization of boric acid in the core in the event of an accident.

Effect of boric acid mass transfer on the accumulation thereof in a fuel core under emergency modes at NPPs with WMR

Boric acid mass transfer processes in the reactor facilities with WMR are considered for the case of an emergency with breaking of the main circulation pipeline (MCP) and the operation of the passive safety systems, such as first-, second-, and third-stage accumulator tank systems, and a passive heat removal system (PHRS). Calculation results are presented for a change in the boric acid concentration in the fuel core (FC) of a water-moderated reactor (WMR) in the case of an emergency process. The calculations have been performed for different values of drop entrainment of boric acid from the reactor (0, 0.2, 2%). A substantial excess of the maximum concentration of boric acid has been found to occur 24 hours after an emergency event with a break of MCP. It is shown that increasing the droplet entrainment of boric acid causes the crystallization and accumulation thereof in the FC to become slower. The mass of boric acid deposits on the elements of internals is determined depending on the values of drop entrainment. These results allow one to draw a conclusion concerning the possibility of accumulation and crystallization of boric acid in the FC, because the latter event could lead to a blocking of the flow cross section and disturbance in the heat removal from fuel elements. A review of available literature data concerning the thermal properties of boric acid solution (density, viscosity, thermal conductivity) is presented. It is found that the available data are of quite a general character, but it does not cover the entire range of parameters (temperature, pressure, acid concentrations) inherent in a possible emergency situation at nuclear power plants with WMR. It is demonstrated that experimental study of boric acid drop entrainment at the parameters inherent in the emergency mode of WMR operation, as well as the studies of boric acid thermal properties in a wide range of concentrations, are required.

Evaluation of the Effect of Thermal Stratification on the Residual Life of the Surge Line of the Pressure Compensator in the No. 5 Unit of the Novovoronezh NPP

This article is concerned with investigations of temperature stratifi cation of the fi rst-loop coolant in the surge line of the pressure compensator in the No. 5 unit of the Novovoronezh NPP. The actual stress-strain state and accumulated structural damage in the surge line in different operating regimes was analyzed by means of multiparametric computational and experimental monitoring. The experimental data demonstrate thermal stratifi cation, giving rise to a temperature differential over the cross section ranging from 30°C under normal operating conditions to 150°C, as a result of disruption of the normal operating conditions, and high thermal stresses. The analysis shows that the cyclic strength of the weld connection of the surge line can be exhausted by the end of the 60-yr life of the No. 5 unit. It is concluded that the actual residual stresses must be measured and calculations of the cyclic strength and the residual life of the surge line must be performed. It is suggested that a technology be developed to strengthen the weld-joint metal in order to reduce the formation of operational surface defects. Since 1980, tens of operating defects and even through cracks formed in pipelines because of stratifi cation were discovered in NPP with PWR and VVER [1, 2]. Defects in the form of shallow surface cracks in transitional surfacing were found in the piping of the pressure compensation system of the No. 5 unit of the Novovoronezh NPP during operational monitoring of the weld connections. According to the recommendations of the International Insurance Inspectorate the infl uence of thermal stratifi cation on the damage rate for the surge line of the pressure compensator should be determined. The function of the pressure compensation system is to create and maintain the coolant pressure in the fi rst loop in order to prevent the coolant from boiling in stationary regimes and to limit the deviation in transitional and emergency regimes. The pressure compensator communicates with the coolant in the hot string of the fourth loop of the main circulation pipeline (MSP) via the 426 × 40 mm connecting surge line (Fig. 1). While the reactor is operating, the surge line is heated owing to the closed circulation of the coolant. When the average temperature of the fi rst-loop coolant changes in transitional regimes associated with the disruption of equipment operation and accompanying a change in the load, part of the coolant fl ows from the pressurizer into the loop or from the loop into the pressurizer along the surge line. In addition, a limitation on the deviation of the pressure from the nominal value is attained by compression or expansion of the steam blanket in the pressurizer. Thus, the integrity of the surge line is directly related with the reliability of fi rst-loop operation. In the course of operation the following loading factors act on the pipeline: mass of the structure, internal pressure, temperature change, and temperature stratifi cation in transient regimes. Thermal stratifi cation is associated with the separation of the coolant fl ow over the section of the pipeline into a cold layer at the bottom and hot layer at the top. The temperature gradients in the coolant result in beyond design-basis stresses in

Investigation and development of technology for automatic argon-arc narrow gap welding of Du-850 pipes for reactor cooling plant to water–water reactors

The technology of automatic narrow-gap argon-arc welding of pipes of the main circulation pipeline Du-850 of water–water nuclear reactors is investigated and developed. Welding is carried out using automatic welding systems with Sv-10G1SN1MA welding wire. The welded joints were investigated by metallographic methods, and the mechanical properties were determined.

Estimate of the Effect of the Time Interval between Technical Inspections on the Strength Reliability of Certain First-Loop Components of a VVER-1000 Reactor

The results of a computational investigation of the possibility of and the safety conditions for switching from a 4- to an 8-yr time interval between technical inspections of the main circulation pipeline and the pressure compensator vessel in VVER-1000 reactors are presented. To this end, calculations of the critical and admissable sizes of defects in the main metal and weld seams in the operating regime, in accident situations, and during earthquakes have been performed. Calculations of the time for a through defect to reach a critical size for different operating periods have been performed. The influence of the hydrotesting pressure and the time interval between such tests on the operational safety and the effect of the time between technical inspections on the reliability of the indicated first-loop components of a VVER-1000 reactor with respect to the criterion of fracture resistance, taking account of the probabilistic nature of the initial data, is analyzed. The calculations were performed using normative and certified procedures geared toward the typical characteristics of steel and structures and the conditions of fabrication, assembly, and operation.

Possibility of reducing the maximum design pressure in the containment vessel of the VV�V-1000

This paper demonstrates up to what extent the pressure can be reduced in the case of an accident with rupture of the main circulation pipeline in the containment vessel of a nuclear power station with a VVER-1000, by the use of a bubbling plant.

Shock loads on units within the containment in the VV�R-1000 in the initial stage of an emergency involving failure in the main circulation pipeline

Shock loads on units within the containment in the VV�R-1000 in the initial stage of an emergency involving failure in the main circulation pipeline