Steam Line Research Articles

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.

Investigation into the Effect of Manufacturing Technology on the Fracture Toughness and Durability of Steam Pipes Made of 10Kh9MFB-Sh Steel

Notched sample modeling was used to demonstrate effect of technology used for manufacturing pipes from steel 10Kh9MFB-Sh (piercing-broaching or electroslag remelting) on metal fracture toughness under creep conditions of 600°C at subcritical and avalanche stages of destruction. According to the subcritical crack growth and metal fracture toughness resistance determined on the samples, the lifespan of pipes having initial defects was evaluated. Calculations were carried out for pipes of two types including a hot reheat (HRH) steam line operating under a pressure of 4.1 MPa and a main steam (MS) line operating under a pressure of 14.0 MPa. Taking into account the obtained characteristics of fracture toughness, the use of electroslag remelting technology was shown to provide a durability of about 3 times that of piercing-broaching approaches. Differences in the pressure level and wall thickness shown to result in a tenfold increase in the durability of the HRH and MS steam pipelines at 600°C justify a different cracking risk classification, which should be taken into account in conducting operational monitoring.

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

Recently, efforts have been made to apply the leak-before-break (LBB) design concept to the main steam line (MSL) piping in secondary systems of Korea standard nuclear power plants (KSNPs). In this study, the microstructures, mechanical properties, and consequential LBB margins of five candidate materials (SA335 Gr.P11/P22, SA508 Gr.3 Cl.2, SA508 Gr.3 Cl.1, and SA508 Gr.1A steels) were evaluated, and strong/weak points of candidate materials for applying MSL piping were analyzed through the comprehensive analysis including microstructure, mechanical properties, LBB margin, usage history, economics, and so on. SA508 Gr.1A steels, which has been used extensively as a piping material in NPPs and has a low fabrication cost, had a good combination of yield strength and fracture toughness. They had LBB margin of 1.11–1.21. SA508 Gr.3 Cl.1 steel had an excellent combination of yield strength and fracture toughness with an LBB margin of 1.64. SA508 Gr.3 Cl.2 steel had the highest LBB margin (1.84) because it had the highest strength; however, it had low fracture toughness and low upper shelf energy (USE). The effect of yield strength on the LBB margin was considerably higher than the effect of fracture toughness. SA508 Gr.3 Steels have high LBB margin owing to the high yield strength, but they have never been used for piping materials in NPPs before. SA335 Gr.P11 and P22 had the LBB margin of 1.1 and 1.23, respectively. However, the elongation and J-R fracture toughness of SA335 Gr.P11 and P22 decreased significantly at 286 °C owing to dynamic strain aging (DSA).

Global sensitivity analysis of the main steam line break accident by using sampling methods and surrogate models

Global sensitivity analysis (GSA) aims at quantifying the individual effects of input variables on the model outputs. The global sensitivity analysis is expressed by the Sobol’ indices. The Sobol’ indices can be obtained by different algorithms. These algorithms can be divided into two main categories: the sampling methods and the surrogate models. This work employs these methods to perform the global sensitivity analysis of Main Steam Line Break Accident (MSLB) of the AP1000 nuclear power plant. Specifically, the DAKOTA is coupled with an in-house containment analysis code to perform the sensitivity analysis. This study focuses on the effects of five key input parameters on safety parameters. The results show that atmospheric pressure has the most important impact on the peak pressure and peak temperature of the containment during the MSLB accident. Besides, compared with the sampling methods, the surrogate model methods need fewer samples and less time to obtain the Sobol’ indices.

Research of a two-phase steam system during its flow in the pipe and a device for measuring dryness steam

Water steam is widely applied in various fields of industry and heat power engineering. The article shows a wide application of water steam in heat-and-power engineering. In this regard, the issue of preserving and controlling the dryness of steam during its transportation via steam lines remains relevant. Quite detailed theoretical studies of steam parameters during its flow in the pipe have been performed. A comparative analysis of steam velocity and temperature plots depending on certain flow parameters and accepted approximations is performed. The analysis presented that gas-dynamic and thermodynamic processes can be considered approximately. This does not play a fundamental role in the general patterns of dropping liquid. As a result of numerical studies of the steam condensation kinetics, the specific volume and density of drop condensate along the pipe section are determined. The obtained results and the proposed methodological approach for determining the parameters of steam were used in the development and further improvement of the device for measuring the dryness of steam, the original construction of which was proposed earlier by one of the authors. The article presents the functional diagram and layout of the measuring unit of the gauge.

Coupled 3D thermal–hydraulic and neutron kinetic Relap5 model for transient analysis of a 1200-MWe nuclear PWR plant

This paper deals with the assessment of the fully integrated three-dimensional thermal–hydraulic and neutron kinetic modeling capability of the Relap5-3D system code. The interest in integrated system codes is growing in the field of safety analyses, where coupled three-dimensional simulations can significantly improve the level of understanding of spatial core phenomena in accident conditions. For this purpose, an asymmetric design basis accident, namely a main steam line break, has been simulated for a commercial 1200-MWe pressurized water reactor. The simulation has been performed with a one-dimensional plant model, including all the major primary, secondary and safety systems included in advanced PWR designs, upgraded with a three-dimensional thermal–hydraulic representation of the reactor vessel and a three-dimensional neutron kinetic core model. The analysis demonstrated the code capability to capture 3D core phenomena, not achievable with 1D plant model based on point kinetic.

Abrasive water jet machining process parameters optimisation for machining of complex profiled component from Inconel 718 super alloys

The demands for machining complex profile from difficult to machine materials find wide applications in various industries, that has been triggered the neck for development of advanced machining technologies, specifically non-conventional processes and associated optimisation of machining parameter. Keeping in view the applications of Inconel-718 material for aerospace applications, it has been selected as the specimen material for machining complex profiled component on KMT steam line SL-V50 water jet five axes CNC abrasive water jet machine (AWJM) at the research organisation. In the present work AWJM process parameters are optimised by Taguchi-Grey relational method on dimensions, geometries and surface roughness of complex curved contours. Additionally, the Pareto analysis of variance (ANOVA) of Grey relational grade is also applied to find the most significant process parameter. Traverse speed is found to be the most significant process parameter for profile cutting in AWJM, also confirmation experiments were conducted to validate the results. The experiment was done for a specified profile for cross sectional thickness of 6.6 mm and 18 sets of experiments were conducted based on stark's L18 orthogonal array.

Optimizing the Design of the Steam-Receiving Perforated Sheet Used in the PGV-1000MKO Steam Generator

The VVER-TOI reactor plant has for the first time been furnished with a new layout of steam generators in the NPP reactor building, according to which the steam produced in the steam generator is removed from it through one steam nozzle. The shift for using one steam-offtake nozzle results in that a nonuniform steam velocity field is produced in the separation volume between the evaporation surface and steam-receiving perforated sheets (SRPSs), which is caused by steam motion along the path with the lowest friction, and this can entail excessive carryover of moisture from the steam generator (SG). To obtain the necessary moisture-separation performance of a horizontal steam generator equipped with one steam nozzle, it is necessary to set up a uniform field of the steam velocity vector’s vertical component in the separation volume with limiting the pressure loss in the SG steam path down to the design value. Therefore, it is proposed to arrange nonuniform flow friction along the steam motion path from the evaporation surface to the steam line by using a variable SRPS perforation ratio. In this study, a set of optimization numerical analyses of the SRPS design for the PGV-1000MKO steam generator performed using the CFD methods was carried out for determining the perforation ratio. In carrying out the optimization numerical analyzes, the performance characteristics of the steam-receiving device equipped with one steam nozzle located in the steam generator’s upper part and consisting of more than 50 perforated sheets had to be made maximally close to the performance characteristics of the steam-receiving device equipped with ten steam nozzles. By using an elaborated computer model, a set of optimization numerical analyses of the SRPS design was carried out, and an optimized design version of the steam-receiving device with a variable perforation ratio for the PGV-1000MKO steam generator has been obtained.

KAJIAN TERHADAP KARAKTERISTIK PERPINDAHAN PANAS PADA LOOP THERMOSYPHON SEBAGAI ALAT RECOVERY PANAS TEMPERATUR RENDAH

The purpose of this study was to examine the effectiveness of the Loop Thermosyphon Heat Exchanger (LTHE) fluidized R134a refrigerant at variations in working fluid pressure of 0.8MPa and 1.2MPa. Thermosyphon is made of three parts, namely, evaporator, condenser and adiabatic part. Evaporators are made of 3/4 inch diameter copper pipe with a length of 3.5 cm. The condenser is made of 3/4 inch diameter copper pipe with a length of 15 cm. The adiabatic part of the steam line is made of copper pipes with a diameter of 3/4 inches and a length of 25 cm and a liquid line with a diameter of 3/8 inches with a length of 27 cm. Things that were investigated included the effect of air flow velocity (heated fluid) of 0.8 m/s, 1m/s, and 1.2m/s on the effectiveness of LTHE and the effect of an increase in evaporator temperature by 400C, 500C, and 600C on the effectiveness of LTHE. The study was conducted experimentally where the LTHE evaporator was heated with a heater. The results of this study show that the effectiveness of LTHE heat transfer work pressure of 1.2 MPa pressure is about 30-68% and higher than the effectiveness of LTHE heat transfer work pressure of 0.8 MPa pressure about 22-35%. Effectiveness decreases with increasing airflow velocity. The LTHE thermal resistance of 0.8 MPa pressure is around 0.3-0.450C/ W and is greater than the LTHE thermal resistance of 1.2 MPa pressure around 0.08-0.40C /W.

Partially heated lid-driven flow in a hexagonal cavity with inner circular obstacle via FEM

In the present article, numerical analysis is performed over the lid-driven hexagon cavity filled with water. The cavity is partially heated at the upper wall, that is moving with uniform velocity. Moreover, cylindrical obstacle is placed inside of hexagonal cavity and different constraint are defined at the surface. Governing partial diifferential equations are nondimensionlized before being solved numerically using Finite Element Method (FEM). The present numerical method is also applied to the previous reported models to validate the results reported in this article. The simulation is performed in order to determine the heat transfer rate, steam lines and isotherms behavior due to the variation of contained physical and geometrical parameters, such as: Reynolds number (200 ≤ Re ≤ 550), Richards number (10−6 ≤ Ri ≤ 1), Hartman number (0 ≤ Ha ≤ 20), heated length (0.1 ≤ HL ≤ 0.3) and cylindrical obstacle (cold, adiabatic and heated). The circular obstacle played an important role in the formation of isotherms by adjusting the various constraint at the surface. The outcome reveals that the heat length, Richardson number and Reynold number are strongly associated with heat transfer.

Simulation and uncertainty analysis of main steam line break accident on an integral test facility

In order to investigate the performance of a Generation III BWR safety system, an Integral test facility was designed and conducted, and several integral tests were performed. The best estimate safety analysis of Main Steam Line Break (MSLB) accident using RELAP5/MOD3.3 and RELAP/SCDAPSIM/MOD3.4 (from now on referred as RELAP/SCDAP3.4) was performed. By comparing the simulation results with the test results, the prediction capability of the RELAP5/MOD3.3 and the RELAP/SCDAP3.4 was compared and analyzed. The RELAP/SCDAP3.4 code was selected to perform the uncertainty study. According to the Phenomenon Identification and Ranking Table, 12 input treatable parameters and 29 source code coefficients were selected with imposed uncertainty distribution to conduct uncertainty analysis. The 95/95 uncertainty band of the minimum core collapsed level were obtained and analyzed. In addition, the contribution weight of input treatable parameters and source code coefficients for the minimum core collapsed level were identified with the Spearman rank correlation coefficient.

Non-stationary extreme value analysis applied to seismic fragility assessment for nuclear safety analysis

Abstract. Fragility curves (FCs) are key tools for seismic probabilistic safety assessments that are performed at the level of the nuclear power plant (NPP). These statistical methods relate the probabilistic seismic hazard loading at the given site to the required performance of the NPP safety functions. In the present study, we investigate how the tools of non-stationary extreme value analysis can be used to model in a flexible manner the tail behaviour of the engineering demand parameter as a function of the considered intensity measure. We focus the analysis on the dynamic response of an anchored steam line and of a supporting structure under seismic solicitations. The failure criterion is linked to the exceedance of the maximum equivalent stress at a given location of the steam line. A series of three-component ground-motion records (∼300) were applied at the base of the model to perform non-linear time history analyses. The set of numerical results was then used to derive a FC, which relates the failure probability to the variation in peak ground acceleration (PGA). The probabilistic model of the FC is selected via information criteria completed by diagnostics on the residuals, which support the choice of the generalised extreme value (GEV) distribution (instead of the widely used log-normal model). The GEV distribution is here non-stationary, and the relationships of the GEV parameters (location, scale and shape) are established with respect to PGA using smooth non-linear models. The procedure is data-driven, which avoids the introduction of any a priori assumption on the shape or form of these relationships. To account for the uncertainties in the mechanical and geometrical parameters of the structures (elastic stiffness, damping, pipeline thicknesses, etc.), the FC is further constructed by integrating these uncertain parameters. A penalisation procedure is proposed to set to zero the variables of little influence in the smooth non-linear models. This enables us to outline which of these parametric uncertainties have negligible influence on the failure probability as well as the nature of the influence (linear, non-linear, decreasing, increasing, etc.) with respect to each of the GEV parameters.

Guilan combined power plant in Iran: As case study for feasibility investigation of converting the combined power plant into water and power unit

This study aims to find a cost effective solution for converting an existing combined power plant into a CWP system including the membrane and thermal desalination units. Without applying major structural change in power plant, to supply fresh water, the steam lines extracted from heat recovery steam generator are used in the thermal desalination unit and membrane desalination system is connected to the gas turbine. An investigation is performed on the produced water and performance of power plant associated with different conditions of electricity and water demands. Also, the influence of steam corresponding to the low and high enthalpy conditions on the performance of thermal desalination unit and power plant is explored. Furthermore, in view of the production capacity of electricity associated with the gas units and consumption demands of various seasons, the influence of membrane desalination unit on the water production is illustrated. An economic investigation is performed to estimate the selling price of water in each unit. It is concluded that in the thermal unit, using HP line produces about 25,000 m3/day, while the LP line causes up to 7000 m3/day water.

Uncertainty and regression analysis of the MSLB accident in PWR based on unscented transformation and low rank approximation

The present studies focus on the quantification of uncertainty during the main steam line break accident scenario (MSLB) in PWR, assuming that there is a failure on the feed-water regulating valve of the broken steam generator. The scenario is characterized by the associated positive Doppler and coolant density reactivities which bring the core back to critical (return-to-power). Accordingly, the input uncertainty parameters are the Doppler and coolant density reactivities taking into account the correlation matrix among the input parameters, which is calculated by SCALE 6.2 code. The main safety parameters are the maximum cladding surface temperatures and average core power during the accident which are computed by ATHLET thermal-hydraulic code. The sampling-based uncertainty technique is considered to be the most dependable technique which can be applicable to any code, however it is computationally expensive. Therefore, it is important to develop efficient techniques which are capable of reducing the calculation time. The first approach is the SVD-UT where the Unscented Transform (UT) algorithm and singular value decomposition (SVD) are combined to generate a minimal sample points. In addition, due to the strong correlation between the input reactivities, the computational time can be further reduced by implementing the Low Rank Approximation (LRA) and revealing the active subspace.

A Multiscale and Multiphysics PWR Safety Analysis at a Subchannel Scale

A high-fidelity safety analysis method for pressurized water reactors (PWRs) is presented using a multiscale and multiphysics coupled code. Computational resolution of the conventional safety analysis can be greatly improved using this method in which the whole reactor vessel is modeled at a subchannel scale with around 5 million calculation meshes. Three-dimensional thermal hydraulics inside the reactor vessel is simulated using CUPID-RV with subchannel-scale thermal-hydraulic models for the reactor core. The subchannel models were validated using the legacy rod bundle experiments including single- and two-phase flow tests that were used in the validation of other subchannel analysis codes. The three-dimensional mesh was generated for the reactor vessel. Structured meshes were used in the core region for the subchannel model, and body-fitted unstructured meshes were applied for the downcomer, lower and upper plenums, and hot and cold legs. The number of meshes was optimized for a practical calculation. A three-dimensional core kinetics code (MASTER) and a one-dimensional system analysis code (MARS) were coupled with CUPID-RV for an accident analysis of PWRs. Subchannel-scale full-core steam line break accident analysis of the OPR1000 PWR was realized using the coupled code (MASTER/CUPID-RV/MARS) with a reasonable computation time, and thus, the present method can be used as a practical tool for three-dimensional safety analysis of PWRs.

Boilers and Steam Lines Chemical Cleaning

Boilers and Steam Lines Chemical Cleaning

An efficient method of key parameter screening for PCCS under SLB accident in AP1000

Variance decomposition is an effective sensitivity analysis method to screen the key parameters influencing passive safety system operation based on the uncertainties of thermal-hydraulic (T-H) model inputs. However, such method needs a large number of samples gained from T-H model running with the inputs sampled randomly from their probabilistic distributions, and the T-H model always takes quite long time to run once, then it will be a heavy calculation burden to do the analysis. In this paper, we propose a method to improve the analyzing efficient: based on the system T-H characteristics, the system behavior in a short time after an accident happening can represent the system T-H performance and be used to do the sensitivity analysis.Passive Containment Cooling System (PCCS) in AP1000 is used as a case study in our analysis, by which the heat produced in the containment can be transferred to the atmosphere through natural circulations. After steam line break (SLB) accident, the peak value of pressure in the containment appears within 1000s, we do the sensitivity analysis to screen key parameters in two ways: firstly, we screen the key inputs with variance decomposition method directly, 100 samples are gained from T-H model simulating 1000s, the inputs are sampled based on their probabilistic distributions, and the results show that air pressure is the most important parameter and the others don't have enough differentiation degrees. Then we get more 100 samples under the condition that air pressure is supposed as 0.1 MPa, here air temperature and steam mass flow are important ones besides air pressure. In another way, we analyze the correlation between pressure in the containment in a short time after SLB and the peak value according to the system T-H behavior, and get 600 samples from T-H model simulating 50s, the results are in accordance with that from T-H model simulating 1000s, air pressure, air temperature and steam mass flow are important parameters, and it just needs 1.55 h to calculate the important factor for one input.

Mechanising migration: Transnational relationships, business structure and diffusing steam on the Atlantic

The diffusion of steam into the transatlantic migration system of the nineteenth century, one of the most important developments in the history of human demography, is often explained by the technical progress of ships, which made the carrying of migrants under steam profitable. Existing historiography posits that early, basic paddle steamers were sustainable only with government mail contracts, whilst iron screw steamers later facilitated the emergence of a mass migrant trade. Data on steam company formation, durability and accounting for the mid-nineteenth century show that technical thresholds are not sufficient to explain the transfer to steam shipping of migrants. Determining factors were inter-regional relationships connecting engineering and demographic change, and, critically, the abandonment of capitalising novel steam lines. This article demonstrates that steam became usable because of endogenous transfer within well-established sailing services, a pivotal strategy adopted by those connected to centres of both innovation and migration.

Análisis de la Eficiencia Energética de una Caldera Industrial Alimentada por Leña

El objetivo de este trabajo fue caracterizar el rendimiento térmico de un generador de vapor con la intención de reducir el desperdicio de combustible y los costos involucrados. El potencial económico y social de un país tiene gran relación con su reserva energética y la manera como es explotada. El desarrollo provoca impactos ambientales significativos, siendo que la mayoría se debe a la generación y al uso de energía, cuya demanda se ha tornado cada vez mayor. La metodología empleada fue el análisis de una caldera industrial del tipo pirotubular modelo ATA 14 H 3N, de la marca ATA, con capacidad de 33,3 kg/s de vapor y una presión de hasta 1034 kPa, alimentada por combustible sólido (leña de eucalipto en troncos). Los datos necesarios para el análisis fueron obtenidos por medio de mediciones en campo y de informes de análisis químicos elaborados por empresas especializadas. El resultado de la eficiencia térmica de la caldera fue obtenido, por el método indirecto, en torno al 74%. Se concluye que para mejorar la eficiencia de la caldera será necesario optimizar el exceso de aire, así como el aislamiento de más tramos de la línea de vapor y la instalación de purgadores de condensado en los puntos distantes del centro de generación de vapor.

Specific Features of the Damageability of the Welded Joints of Steam Pipelines that Have Been in Service for a Long Time

The aim of the work is to identify the features of damageability of welded joints of heat-resistant steels (12Х1МФ, 15Х1М1Ф) for long-term creep exploitation. To achieve the goal in the study of the structure and damage, metallographic methods and methods for determining mechanical properties were used. Studying the peculiarities of using heat-resistant steels for steam pipelines and elements of their systems as applied to the production of power-generating blocks under creep conditions over 270.000 hours allowed: to establish the reasons for their damageability and destructibility; mechanisms for the development of creep cracks and fatigue cracks; connection structure with damage; structural changes. Established the relationship of damage with the presence of stress concentrators. The generalization of features of damage allowed to specify typical patterns of damage, which is advisable to reduce it. It was revealed that when welded joints are produced for more than 270.000 hours under the conditions of frequent start-stops of power units, their damage by the fatigue mechanism increases markedly. Revealed that damage by the mechanism of fatigue significantly depends on the structural state of the steam lines and their welded joints.