Circuit Of Nuclear Power Plant Research Articles

Effect of Hydrazine on Redistribution of Corrosion Product Deposits in the Primary Circuit of a Shutdown Reactor in Propulsion-Type Nuclear Power Plants

A redistribution process of corrosion product (CP) deposits in the primary circuit of nuclear power plants (NPP) with propulsion service is examined. It is demonstrated that, after a shutdown of the NPP, the activity of activated corrosion products (ACP) and the concentration of stable corrosion products in the coolant increase considerably due to their entering the coolant from the deposits on surfaces of the primary circuit equipment. The corrosion products enter the coolant principally in the form of insoluble compounds, and the suspended solid particles above 0.4 μm in size account for a minimum of 90% of the total content of CPs in the coolant. This increases the CP concentration in the coolant by a factor of approximately 10, and the total activity of ACPs becomes more than 300 times the initial value determined during reactor operation. It was established that injection of hydrazine facilitated redistribution of the corrosion products and radionuclides associated with them between the surface of construction materials and the coolant: the corrosion products went into the coolant and the rate of their redeposition decreased by a factor of ten. This enables us to increase the fraction of highly active ACPs removed from the circuit by the normal cleaning system. In addition, a shift in the equilibrium on injection of hydrazine gives a higher activity of ACPs in the coolant after completion of the deposition process and removal of CPs that entered the circuit in the normal treatment system. The equilibrium activity of ACPs is approximately ten times the value specific for the installation that is cooled down without introduction of hydrazine. A decrease in this characteristic to stable low values is observed only after the reactor is put into operation.

Large eddy simulation of thermal turbulent mixing and reduction of temperature fluctuations with swirl in T-junction

Turbulent mixing of two fluids having a temperature difference can lead to temperature fluctuations in the T-junctions, particularly in the primary cooling circuit of nuclear power plants. Temperature changes produce thermal stresses causing thermal fatigue. This is a key topic related to nuclear security. This article presents the numerical results of two thermal turbulent mixing simulations, applied to a T-junction (OECD-NEA-Vattenfall T-junction). These simulations apply, respectively, to the case of a flow with and without rotational movements into the cold side. The numerical predictions were made using the large eddy simulation with the commercial code Fluent. The numerical results obtained in the first case were compared with the available experimental data and were found in an agreement satisfactory to the profiles of average variables and the spectrum of the temperature near the wall. The comparison with RMS profiles was less satisfactory in the center of the main tube, but was of good quality in the near-wall region. In the second simulation, the main movement of the fluid is affected by the spiral paths created by the swirl velocity caused at the cold inlet, which improves the mixing of hot and cold fluids downstream of the T-junction. From the analysis of these results, it clearly appears that the temperature fluctuations decrease near the wall in this case which leads to less high thermal loads on the tube, and hence a lesser thermal fatigue.

Structure and Mechanical Properties of Welded Joints for Nuclear Power Plants of Type MIR 1200

The paper deals with research, development and verification of production technology of selected welded joints for pressure vessels of primary circuits of nuclear power plants of type MIR 1200. Effect of various welding technology including simulation heat treatment on mechanical and fracture properties have been studied. Four type of homogenous 10GN2MFA – 10GN2MFA type of welded joints have been prepared for experimental programme. Conventional mechanical properties (tensile and impact test) as well as unconventional mechanical properties (fracture mechanics, low-cycle fatigue and stress corrosion cracking in water environment) have been studied. Effect of elevated working temperature on structure and material properties has been evaluated. Temperature dependencies of shear fracture have been plotted and effect of welding procedure on transition temperature shift has been evaluated. Experimental data have been compared with numerical simulation using FEM.

Application of Metallographic Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP

During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.

Impact of Ag and NOx compounds on the transport of ruthenium in the primary circuit of nuclear power plant in a severe accident

Ruthenium is a semi-volatile element originating as a fission product in nuclear reactors that can be released in case of a severe nuclear accident. In this work, the impact of atmosphere composition on the transport of ruthenium through the primary circuit was examined. The effects of silver nanoparticles representing aerosols and NO2 gas as a product of air radiolysis were studied. Quantification of ruthenium transported both as gas and aerosol was performed. Chemical composition of ruthenium species was evaluated. The transport of gaseous ruthenium through the facility increased significantly when NO2 gas was fed into the atmosphere. When both silver aerosols and NO2 were fed into the atmosphere, the transport of ruthenium in gaseous and aerosol forms was promoted. It was concluded that the composition of atmosphere in the primary circuit will have a notable effect on the speciation of ruthenium transported into the containment building during a severe accident and thus on the potential radioactive release to the environment.

A Review of Oxidation Phenomena in Ni Alloys Exposed to Hydrogenated Steam Below 500°C

The following is a review of the accumulated literature over the past 20 years for oxidation in Ni-Fe-Cr alloys exposed to 400°C to 500°C hydrogenated steam. Hydrogenated steam at temperatures of 500°C and below is considered to simulate 300°C to 360°C deaerated water often used in the primary circuit of nuclear power plants. Alloy 600 (Ni-16Cr-9Fe) is highly susceptible to stress corrosion cracking (SCC) in primary water and the peak crack growth rate is observed near the Ni/NiO equilibrium potential. It is believed that the oxidation tendency, internal and/or external, is directly related to the mechanism of SCC in Alloy 600. An initial review will be conducted exploring inter- and intragranular oxidation in Alloy 600 and the relationship to embrittlement and SCC. Following this, investigations into oxidation phenomena, in particular the possibility of internal oxidation, in alloys with high reactive element contents, such as Alloy 690 and Alloy 800, will be analyzed. Finally, the practical applicabilit...

Estimation of the Drop Size in Dispersed Flow

The formulas for calculating the characteristic drop size for the mean Sauter diameter have been compared. The question on various forms of the size distribution of drops has been considered. To substantiate the applicability of the compared formulas for calculating the thermohydrodynamics in the circuits of nuclear power plants, experimental data on the wall temperature in a dispersed flow have been used. It has been shown that the Sauter diameter values calculated using the wall temperature in the supercritical region are in good agreement with sparse direct measurements of the drop size in steam–water flows. The drop sizes calculated using the tested formulas obtained for two-component gas–liquid flows or for single-component flows of coolants (various kinds of freons) and liquefied nitrogen turned out to be much lower. It has been shown that it is necessary to recalculate the numerical coefficients in the considered formulas in using them for steam–water flows.

Modeling of corrosion product migration in the secondary circuit of nuclear power plants with WWER-1200

Models of corrosion and mass transfer of corrosion products in the pipes of the condensate-feeding and steam paths of the secondary circuit of NPPs with WWER-1200 are presented. The mass transfer and distribution of corrosion products over the currents of the working medium of the secondary circuit were calculated using the physicochemical model of mass transfer of corrosion products in which the secondary circuit is regarded as a cyclic system consisting of a number of interrelated elements. The circuit was divided into calculated regions in which the change in the parameters (flow rate, temperature, and pressure) was traced and the rates of corrosion and corrosion products entrainment, high-temperature рН, and iron concentration were calculated. The models were verified according to the results of chemical analyses at Kalinin NPP and iron corrosion product concentrations in the feed water at different NPPs depending on рН at 25°С (рН25) for service times τ ≥ 5000 h. The calculated рН values at a coolant temperature t (рНt) in the secondary circuit of NPPs with WWER-1200 were presented. The calculation of the distribution of рНt and ethanolamine and ammonia concentrations over the condensate feed (CFC) and steam circuits is given. The models are designed for developing the calculation codes. The project solutions of ATOMPROEKT satisfy the safety and reliability requirements for power plants with WWER-1200. The calculated corrosion and corrosion product mass transfer parameters showed that the model allows the designer to choose between the increase of the correcting reagent concentration, the use of steel with higher chromium contents, and intermittent washing of the steam generator from sediments as the best solution for definite regions of the circuit.

Micro-mechanical State at Tip of Environmentally Assisted Cracking in Nickel-based Alloy

Nickel-based alloys and austenitic stainless steels are widely used in the structures of primary circuit of nuclear power plants. Environmentally assisted cracking (EAC) of these materials is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. Researches show that EAC in nickel-based alloy is a process of oxide film rupture and reform at the tip of EAC in the high temperature water environment of nuclear power plants. To understand the micro-mechanical state at the tip of EAC, the stress-strain in the oxide film and the base metal at the EAC tip was simulated and discussed using a commercial finite element analysis code, which provides a foundation to improve the quantitative predication accuracy of EAC growth rate of nickel-based alloys and austenitic stainless steels in the important structures of nuclear power plants.

МАТЕМАТИЧЕСКАЯ МОДЕЛЬ НЕСТАЦИОНАРНОГО ТЕПЛОМАССОПЕРЕНОСА В КАНАЛЕ С ЖИДКОМЕТАЛЛИЧЕСКИМ ТЕПЛОНОСИТЕЛЕМ С УЧЕТОМ ПРОЦЕССА КРИСТАЛЛИЗАЦИИ-ПЛАВЛЕНИЯ

The mathematical model has been developed describing heat and mass transfer in a channel with liquid metal coolant when there is a crystallization-melting process. The model is based on a one-dimensional description of heat and mass transfer, taking into account the qualitative differences of the process depending on the condition of the contact surfaces. The thermodynamic equilibrium of the phases at the boundary of phase transfer is postulated. It is shown that a two-speed heterogeneous flow model describing the formation of solid layer on the surface of heat exchange is reduced to the solution of two more simple interrelated initial-boundary value problems: the Stefan problem of phase transfer and the single-phase flow model with a source and a variable cross-sectional area. The process of crystallization with the formation of micro fine liquid-metal suspension can be described in terms of the homogeneous model of heat and mass transfer. Represented mathematical model can be used for developing numerical algorithms and programs for calculating of heat-hydro-dynamics in steam generators and circuits of nuclear power plants with heavy liquid metal coolants. Keywords: homogeneous model, quasi-homogeneous model, heterogeneous model, the liquidmetal coolant, Stefan problem, interfacial exchange, steam generator, wet ability of surface, phase transfer. &nbsp

Modelling of Phased Array Ultrasonic Inspection of a Steam Generator Dissimilar Metal Weld

Phased array ultrasonic techniques (PAUT) are more and more used for inspection of the primary circuit of nuclear power plants in order to detect defects such as stress corrosion cracking (SCC) at the dissimilar metal welds (DMWs) between the main pipe lines and large components such as the pressure vessel, steam generators and pressurizers. Nevertheless, disturbances such as beam splitting and skewing may occur due to the anisotropic and inhomogeneous properties of the welding material. These disturbances affect the detection, localization and sizing of possible weld discontinuities. Simulation tools can help to understand these physical phenomena, optimize ultrasonic testing. Since strict safety requirements are leading to increased demands for the inspection of such components, a research task called MAPAID has been dedicated under the umbrella of “NUGENIA+” FP7 project to model and validate PAUT techniques for inspection of DMWs of nuclear power plants. As a part of the task, this study presents the phased array ultrasonic inspections and ultrasonic simulations of steam generator DMW. The testing has been recently developed in the CIVA in order to determine the index distances regarding the given orientation artificial defect analysis and the incident angles for longitudinal and transversal tests. By comparing the experimental and simulation tests results, conclusions can be drawn to quantitatively assess the contribution of phased array techniques to improved NDE performances of such parts, as well as the ability of simulation to help for design, optimization and interpretation of such inspections.

Characterization of Damage During Low Cycle Fatigue of a 304L Austenitic Stainless Steel as a Function of Environment (Air, PWR Environment) and Surface Finish (Polished, Ground)

304L austenitic stainless steel is notably used to manufacture pipes in primary cooling circuits of nuclear power plants operating at a temperature close to 300°C in a specific aqueous environment. Customarily, pipes are roughly polished, however, after some interventions, a few zones might be ground. Start-ups, shutdowns and operating transients of the reactor produce Low Cycle Fatigue (LCF) loadings with different strain rates. These solicitations and the modified surface finish create specific conditions whose effects must be taken into account. In order to understand if there is a joint influence of the PWR environment and surface roughness on the successive stages of fatigue damage leading to failure, fatigue tests were conducted at 300°C under a total strain amplitude of ± 0.6%. Two environments were used, laboratory air and simulated PWR primary aqueous environment. Two surface finishes were investigated, namely fine mechanical polishing and grinding (Rt of about 40μm). The analysis of the damage process is based on data concerning crack density, crack length and striation spacing. The integration of the relation between the striation spacing and the strain intensity factor ΔKɛ, provides an estimation of crack propagation. It is confirmed that in PWR environment one striation can be considered as the crack advance during one fatigue cycle. Moreover, the specific crack front geometry observed in ground specimens and resulting from successive microcrack coalescences explains the reduction in fatigue life induced by grinding.

원자로 압력 및 체적제어계통의 다단 오리피스 설계

다공 및 다단오리피스열을 해석할 수 컴퓨터코드를 사용하여 원자력발전소 내 공급 및 배출장치의 수축오리피스를 설계하였다. 원래 단공 오리피스가 설치되었으나 다공 오리피스 설치가 필요하여 설계 변경되었다. 오리피스 사이 거리는 충분한 압력회복거리로 유지되었다. 최소 정압은 공동현상 방지를 위하여 그 온도에 상응하는 증기압 이하로 유지되도록 설계하였다. 오리피스 설치 배관의 직경은 운전시 강성을 유지할 수 있도록 원래 크기보다 증가시켰다. Restriction orifices in the feed and bleed circuit of nuclear power plant are designed using computer program capable of handling multiple hole cascade orifice assembly. Single hole stages of orifice assembly are alternated with multihole stages where necessary. The distance between stages is such that it allows full pressure recovery. The minimum static pressure is higher than vapor pressure at the operating temperature so that cavitation does not occur. Piping sizes are reviewed and increased if necessary to improve rigidity.

Leakage Detection System for the Main Circulation Circuit of Nuclear Power Plants

The problem of leakage detection in the main circulation circuit of nuclear power plants is addressed. A self-calibrating system for reliable detection of incipient leakages is proposed.

Synthetic Aperture Focusing Technique for Detecting Transverse Cracks in Austenitic and Dissimilar Welds

The inspection of austenitic and dissimilar welds using ultrasound demands for sophisticated testing techniques. The application of reconstruction methods like the Synthetic Aperture Focusing Technique (SAFT) on the measurement results provides an appropriate approach for defect characterization and sizing. Nevertheless, the reconstruction algorithm has to consider the aniso-tropic wave propagation inside the inhomogeneous weld material. In recent years the detection of transverse cracks has become increasingly important for ensuring the structural integrity of pipes in the primary circuit of nuclear power plants or longitudinally welded, cladded pipes. However, relia-ble inspection techniques are hardly available. In this particular case, it is expected that the compar-atively long propagation path of the ultrasonic wave field inside the inhomogeneous weld material enhances the effect of anisotropy and influences the accuracy and the signal-to-noise-ratio of the reconstruction result. In this contribution we suggest an advanced ultrasonic testing technique for detecting and sizing of transversal cracks in austenitic and dissimilar welds. The method applies a SAFT reconstruction algorithm considering the anisotropy and the inhomogeneity. A V-arrangement of the transducers in pitch-catch technique is chosen to avoid a direct coupling on the weld face. The reconstruction algo-rithm is based on an extended 3-dimensional weld model and uses a ray-tracing approach for de-termining the wave propagation paths. Along with the reconstruction algorithm the transducer set-up and experimental results of different specimens with artificial transverse flaws are presented. The availability of the proposed method for crack sizing is assessed in comparison to conventional testing techniques.

CFD modeling of thermal mixing in a T-junction geometry using LES model

Turbulent mixing of fluids at different temperatures can lead to temperature fluctuations at the pipe material. These fluctuations, or thermal striping, inducing cyclical thermal stresses and resulting thermal fatigue, may cause unexpected failure of pipe material. Therefore, an accurate characterization of temperature fluctuations is important in order to estimate the lifetime of pipe material. Thermal fatigue of the coolant circuits of nuclear power plants is one of the major issues in nuclear safety. To investigate thermal fatigue damage, the OECD/NEA has recently organized a blind benchmark study including some of results of present work for prediction of temperature and velocity fluctuations performing a thermal mixing experiment in a T-junction. This paper aims to estimate the frequency of velocity and temperature fluctuations in the mixing region using Computational Fluid Dynamics (CFD). Reynolds Averaged Navier-Stokes and Large Eddy Simulation (LES) models were used to simulate turbulence. CFD results were compared with the available experimental results. Predicted LES results, even in coarse mesh, were found to be in well-agreement with the experimental results in terms of amplitude and frequency of temperature and velocity fluctuations. Analysis of the temperature fluctuations and the power spectrum densities (PSD) at the locations having the strongest temperature fluctuations in the tee junction shows that the frequency range of 2–5Hz contains most of the energy. This range of frequency is characteristic for thermal fatigue.

Case study of the applicability of cyclic hardening material descriptions in finite element simulation of cyclic thermal shocks

Finite element simulation of cyclic thermal loading induced fatigue can be used to predict the service time of components in industrial environments. Various damage criteria for low cycle fatigue have been developed, which require a temperature-dependent description of the cyclic plastic hardening of the material. In these descriptions the shape of the stress-strain loop is coupled to its cyclic evolution, which leads to problems under specific loading conditions and/or for materials not exhibiting a quick cyclic stabilization. The applicability of a basic, nonlinear kinetic-isotropic cyclic hardening description combined with an approach in which its parameters depend on the accumulated plastic strain and temperature is evaluated in a case study mimicking boundary conditions that occur in the primary cooling circuit of nuclear power plants. The model is calibrated with experimental data for a material that exhibits cyclic softening, requiring a modification of the hardening description.

Investigation of friction units of nuclear plants operating in heavy high-temperature liquid-metal heat carriers. Part 2. Experimental results

The tribological problems posed by heavy liquid-metal heat carriers used in the circuits of nuclear power plants are discussed. The results of experimental study of the vibrational wear of heat-exchanging tubes in the interfaces with the spacing grates are presented. These results are applicable for testing the vibrational strength of the steam generator of a fast neutron reactor with lead heat carrier. Experimental investigation and service-life tests of involute-profile cylindrical tooth gears and sliding bearings working in fast neutron reactors containing the lead heat carrier are conducted. Experimental and theoretical study is performed on the wear of steel and cast-iron sliding bearings in the recycling pumps of reactors operating in lead, eutectic lead-bismuth and lead-lithium alloys. It is shown experimentally that the tooth gears from 40X13 steel and SCh20 grey iron are capable of operating in the lead heat carrier for 500 and 400 hours, respectively, at the temperature 450°C when the concentration of thermodynamically active oxygen in the lead is close to saturation, the contact stress in the gearing is 758 MPa, the gear rotational frequency is 500 min−1 or less, and the gear ratio is 1.5 (the peripheral velocity of the gear and the wheel over the pitchline is 2.1 m/s). The likely causes of destruction of the gear teeth are adhesive wear and insufficient heat resistance (refractorability) of the tooth material at 450–500°C.

Diffuse reflection spectroscopy as a tool to measure low surface coverage of metal substrates by metallic oxide particles

Deposition of particles from a circulating fluid is a common industrial problem causing fouling or radioactive contamination in circuits of nuclear power plants. A few experimental methods have been developed to study the adhesion process of particles in order to determine the deposition kinetics. This work describes a setup based on diffusion reflectometry of visible light to perform real-time in-situ measurements. The deposition of magnetite and hematite particles onto stainless steel was studied by recording spectra of reflected visible light for several surface-coverage values. Scanning electron microscopy has shown that magnetite particles do not form a uniform layer. Simulations of the evolution of the reflected intensity as a function of the deposited amount of both oxides were performed on the basis of their absorption coefficients and particle sizes. The agreement between theoretical and experimental results shows that diffuse reflection spectroscopy could be a helpful in-situ method of quantifying fouling kinetics of metallic surfaces.

Modelling the attenuation in the ATHENA finite elements code for the ultrasonic testing of austenitic stainless steel welds

Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and makes ultrasonic non-destructive testing difficult. The ATHENA 2D finite element simulation code was developed to help understand the various physical phenomena at play. In this paper, we shall describe the attenuation model implemented in this code to give an account of wave scattering phenomenon through polycrystalline materials. This model is in particular based on the optimization of two tensors that characterize this material on the basis of experimental values of ultrasonic velocities attenuation coefficients. Three experimental configurations, two of which are representative of the industrial welds assessment case, are studied in view of validating the model through comparison with the simulation results. We shall thus provide a quantitative proof that taking into account the attenuation in the ATHENA code dramatically improves the results in terms of the amplitude of the echoes. The association of the code and detailed characterization of a weld’s structure constitutes a remarkable breakthrough in the interpretation of the ultrasonic testing on this type of component.