Academic literature on the topic 'Thermohydraulic parameters'

Modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor were simulated using the TRAC-PD2 and Open FOAM thermohydraulic codes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009) based on energy and mass conservation equations for the three-dimensional unsteady flow of a two-phase mixture. Coupled simulation of the dynamics of neutronic and thermohydraulic processes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009, Bolshagin et al. 2009, Galanin 1971, Weinberg and Wigner 1961, Ovchinnikov and Semenov 1988, Report LA-UR-03-1987) aims to improve the qualitative understanding and the quantitative presentation of their effects on safety. Investigating these modes using the above thermohydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using the industrial testing method, this providing the basis for solving the problems of ensuring the reliability, operational safety and efficiency of nuclear power plants. A modern nuclear reactor is a complex system studying and calculating which requires more than the use of simple theoretical models. Thermohydraulic calculations are an essential part of most engineering and technological development works in nuclear power. Since, in conditions of an NPP, no technologically conventional way can be used to verify and update the results and findings of an a priori analysis on the basis of commercial tests, investigations based on codes are used in some cases as the tools to study and predict the parameters of thermohydraulic processes in the reactor’s circulation circuit. The main purpose of the study is to calculate and investigate, based on codes, modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor in order to determine if calculated parameters conform to the acceptance criteria established by regulatory documents.

There is a significant lack of reliable information on the physical characteristics of thermohydraulic processes in emergency heat transfer modes when cooling the surface of fuel rods with light water coolant with supercritical thermodynamic parameters, in particular, on the physics of heat transfer processes and hydromechanics in the critical area. It is shown that in these conditions there is physical uncertainty about the causes of deteriorating heat transfer, which limits the possibility of creating effective calculation techniques for reliable determination of the upper limit of safe forcing of the heat transfer process in the core. At present, the vast majority of theoretical and experimental studies of thermohydraulic processes in the near-critical area have been performed only for the socalled “normal” heat transfer, which corresponds to the heat removal conditions with mixed turbulent convection of superheated to “gas” state of light water coolant in its inertial mode. Attention is paid to the possible appearance of macromolecular ensembles on this surface in the form of pseudo-vapor formations, which are capable of causing an emergency mode of pseudo-film boiling. On the basis of the given experimental data of various authors existence of rather deep physical analogy between processes of heat exchange in supercritical thermodynamic system and unheated boiling at subcritical parameters of the heat carrier is proved. Existence of the pseudo-boiling process in the conditions of supercritical thermodynamic parameters makes it impossible to use in the thermohydraulic calculation the empirical dependences for “hot” gas for the range of active zones operational parameters.

One of the milestones leading to the unit transition to long-term operation is the assessment of the current technical state of equipment and pipelines during the periodic safety review. An integral part of this assessment are the strength calculations within the necessity to take into account rapid thermohydraulic parameters changing during non-stationary transients. Such calculations are characterized by the need for taking into account the relationship between equipment units and pipelines of the reactor coolant system. This task requires additional development of a complex three-dimensional FE model of reactor coolant system. The submodeling approach was used to perform the strength calculation of primary equipment. This approach involves the step-by-step process of strength calculations including the development of a thermohydraulic model with thermohydraulic analysis, the creation of a simplified complex finite element model of the reactor unit containing basic equipment and pipelines, validation, detailed transfer of boundary condition datasets for finite element model and determination of stress strain state for all calculation modes, and performance of the specified calculations of the equipment by a method of submodeling at the end. In general, the implementation of the whole complex of thermohydraulic and strength calculations within one program code is a quite complicated task and requires creating a separate approach. This paper presents the results of developing a comprehensive approach using the coupled calculations in the thermohydraulic and strength codes.

The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level) was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.

A computer code TRUCHAM developed to model the coupled heat and moisture flow through a porous medium is used to assess the thermohydraulic performance of the clay-based engineered barrier encapsulating nuclear waste containers in a deep geological disposal vault in the borehole emplacement concept. This paper contains an overview of the development of the numerical model and its application to the buffer–container experiment at the Underground Research Laboratory of Atomic Energy of Canada Limited Research. The thermohydraulic transport parameters required for the analysis were determined by specially designed laboratory experiments. The needs for further development of the model and the material properties are identified. Key words : buffer, clay barrier, coupled heat and moisture, heater experiment, numerical model, nuclear waste disposal, thermal diffusivity, unsaturated soil.

Heat exchanger performance can be improved via the introduction of vortex generators to the airside surface, based on the mechanism that the generated longitudinal vortices can disrupt the boundary layer growth, increase the turbulence intensity and produce secondary fluid flows over the heat transfer surfaces. The key objective of this paper is to provide a critical overview of published works relevant to such heat transfer surfaces. Different types of vortex generator are presented, and key experimental techniques and numerical methodologies are summarized. Flow phenomena associated with vortex generators embedded, attached, punched or mounted on heat transfer surfaces are investigated, and the thermohydraulic performance (heat transfer and pressure drop) of four different heat exchangers (flat plate, finned circular-tube, finned flat-tube and finned oval-tube) with various vortex-generator geometries, is discussed for different operating conditions. Furthermore, the thermohydraulic performance of heat transfer surfaces with recently proposed vortex generators is outlined and suggestions on using vortex generators for airside heat transfer augmentation are presented. In general, the airside heat transfer surface performance can be substantially enhanced by vortex generators, but their impact can also be significantly influenced by many parameters, such as Reynolds number, tube geometry (shape, diameter, pitch, inline/staggered configuration), fin type (plane/wavy/composite, with or without punched holes), and vortex-generator geometry (shape, length, height, pitch, attack angle, aspect ratio, and configuration). The finned flat-tube and finned oval-tube heat exchangers with recently proposed vortex generators usually show better thermohydraulic performance than finned circular tube heat exchangers. Current heat exchanger optimization approaches are usually based on the thermohydraulic performance alone. However, to ensure quick returns on investment, heat exchangers with complex geometries and surface vortex generators, should be optimized using cost-based objective functions that consider the thermohydraulic performance alongside capital cost, running cost of the system as well as safety and compliance with relevant international standards for different applications.

The fundamental thermophysical features of the heat exchange process between the heated wall of a vertical channel and the light-water coolant of supercritical parameters concerning the conditions of heat-generating assemblies channels and cores of perspective energy nuclear reactors are considered. The available methods and recommendations for determining the limits of thermal load are analyzed. It is a guarantee the absence of the characteristic dangerous mode possibility of deteriorated heat exchange in these conditions and corresponding sharp rise in the channels wall temperature, which threatens their destruction. The physical nature of the occurrence of degraded heat transfer regimes remains unclear, and the existing approaches to the implementation of thermohydraulic calculation in such conditions are not sufficiently justified. The complex nature of intercellular heat and mass transfer in the fuel assembly and the presence of individual thermohydraulic cells with reduced levels of heat transfer intensity indicate that the existing method of determining the area of degraded heat transfer in the reactor core channels with supercritical parameters of the coolant is significantly simplified. Insufficient data and research results have been revealed to create adequate methods of heat-hydraulic calculation, suitable for taking into account the peculiarities of the heat transfer process complex flow under conditions of supercritical parameters of the coolant. The application of such methods should be the basis for ensuring the safe operation of prospective reactors and minimizing potential losses of a different nature from accidents caused by the destruction of cores through unacceptable heat transfer modes. To this end, the main direction of further research is identified.

This PhD study, developed at Universitat Politècnica de València (UPV), aims to cover the first phase of the benchmark released by the expert group on Uncertainty Analysis in Modeling (UAM-LWR). The main contribution to the benchmark, made by the thesis' author, is the development of a MATLAB program requested by the benchmark organizers. This is used to generate neutronic libraries to distribute among the benchmark participants. The UAM benchmark pretends to determine the uncertainty introduced by coupled multi-physics and multi-scale LWR analysis codes. The benchmark is subdivided into three phases: 1. Neutronic phase: obtain collapsed and homogenized problem-dependent cross sections and criticality analyses. 2. Core phase: standalone thermohydraulic and neutronic codes. 3. System phase: coupled thermohydraulic and neutronic code. In this thesis the objectives of the first phase are covered. Specifically, a methodology is developed to propagate the uncertainty of cross sections and other neutronic parameters through a lattice physics code and core simulator. An Uncertainty and Sensitivity (U&S) analysis is performed over the cross sections contained in the ENDF/B-VII nuclear library. Their uncertainty is propagated through the lattice physics code SCALE6.2.1, including the collapse and homogenization phase, up to the generation of problem-dependent neutronic libraries. Afterward, the uncertainty contained in these libraries can be further propagated through a core simulator, in this study PARCSv3.2. The module SAMPLER -available in the latest release of SCALE- and DAKOTA 6.3 statistical tool are used for the U&S analysis. As a part of this process, a methodology to obtain neutronic libraries in NEMTAB format -to be used in a core simulator- is also developed. A code-to-code comparison with CASMO-4 is used as a verification. The whole methodology is tested using a Boiling Water Reactor (BWR) reactor type. Nevertheless, there is not any concern or limitation regarding its use in any other type of nuclear reactor. The Gesellschaft für Anlagen und Reaktorsicherheit (GRS) stochastic methodology for uncertainty quantification is used. This methodology makes use of the high-fidelity model and nonparametric sampling to propagate the uncertainty. As a result, the number of samples (determined using the revised Wilks' formula) does not depend on the number of input parameters but only on the desired confidence and uncertainty of output parameters. Moreover, the output Probability Distribution Functions (PDFs) are not subject to normality. The main disadvantage is that each input parameter must have a pre-defined PDF. If possible, input PDFs are defined using information found in the related literature. Otherwise, the uncertainty definition is based on expert judgment. A second scenario is used to propagate the uncertainty of different thermohydraulic parameters through the coupled code TRACE5.0p3/PARCSv3.0. In this case, a PWR reactor type is used and a transient control rod drop occurrence is simulated. As a new feature, the core is modeled chan-by-chan following a fully 3D discretization. No other study is found using a detailed 3D core. This U&S analysis also makes use of the GRS methodology and DAKOTA 6.3.<br>Este trabajo de doctorado, desarrollado en la Universitat Politècnica de València (UPV), tiene como objetivo cubrir la primera fase del benchmark presentado por el grupo de expertos Uncertainty Analysis in Modeling (UAM-LWR). La principal contribución al benchmark, por parte del autor de esta tesis, es el desarrollo de un programa de MATLAB solicitado por los organizadores del benchmark, el cual se usa para generar librerías neutrónicas a distribuir entre los participantes del benchmark. El benchmark del UAM pretende determinar la incertidumbre introducida por los códigos multifísicos y multiescala acoplados de análisis de reactores de agua ligera. El citado benchmark se divide en tres fases: 1. Fase neutrónica: obtener los parámetros neutrónicos y secciones eficaces del problema específico colapsados y homogenizados, además del análisis de criticidad. 2. Fase de núcleo: análisis termo-hidráulico y neutrónico por separado. 3. Fase de sistema: análisis termo-hidráulico y neutrónico acoplados. En esta tesis se completan los principales objetivos de la primera fase. Concretamente, se desarrolla una metodología para propagar la incertidumbre de secciones eficaces y otros parámetros neutrónicos a través de un código lattice y un simulador de núcleo. Se lleva a cabo un análisis de incertidumbre y sensibilidad para las secciones eficaces contenidas en la librería neutrónica ENDF/B-VII. Su incertidumbre se propaga a través del código lattice SCALE6.2.1, incluyendo las fases de colapsación y homogenización, hasta llegar a la generación de una librería neutrónica específica del problema. Luego, la incertidumbre contenida en dicha librería puede continuar propagándose a través de un simulador de núcleo, para este estudio PARCSv3.2. Para el análisis de incertidumbre y sensibilidad se ha usado el módulo SAMPLER -disponible en la última versión de SCALE- y la herramienta estadística DAKOTA 6.3. Como parte de este proceso, también se ha desarrollado una metodología para obtener librerías neutrónicas en formato NEMTAB para ser usadas en simuladores de núcleo. Se ha realizado una comparación con el código CASMO-4 para obtener una verificación de la metodología completa. Esta se ha probado usando un reactor de agua en ebullición del tipo BWR. Sin embargo, no hay ninguna preocupación o limitación respecto a su uso con otro tipo de reactor nuclear. Para la cuantificación de la incertidumbre se usa la metodología estocástica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Esta metodología hace uso del modelo de alta fidelidad y un muestreo no paramétrico para propagar la incertidumbre. Como resultado, el número de muestras (determinado con la fórmula revisada de Wilks) no depende del número de parámetros de entrada, sólo depende del nivel de confianza e incertidumbre deseados de los parámetros de salida. Además, las funciones de distribución de probabilidad no están limitadas a normalidad. El principal inconveniente es que se ha de disponer de las distribuciones de probabilidad de cada parámetro de entrada. Si es posible, las distribuciones de probabilidad de entrada se definen usando información encontrada en la literatura relacionada. En caso contrario, la incertidumbre se define en base a la opinión de un experto. Se usa un segundo escenario para propagar la incertidumbre de diferentes parámetros termo-hidráulicos a través del código acoplado TRACE5.0p3/PARCSv3.0. En este caso, se utiliza un reactor tipo PWR para simular un transitorio de una caída de barra. Como nueva característica, el núcleo se modela elemento a elemento siguiendo una discretización totalmente en 3D. No se ha encontrado ningún otro estudio que use un núcleo tan detallado en 3D. También se usa la metodología GRS y el DAKOTA 6.3 para este análisis de incertidumbre y sensibilidad.<br>Aquest treball de doctorat, desenvolupat a la Universitat Politècnica de València (UPV), té com a objectiu cobrir la primera fase del benchmark presentat pel grup d'experts Uncertainty Analysis in Modeling (UAM-LWR). La principal contribució al benchmark, per part de l'autor d'aquesta tesi, es el desenvolupament d'un programa de MATLAB sol¿licitat pels organitzadors del benchmark, el qual s'utilitza per a generar llibreries neutròniques a distribuir entre els participants del benchmark. El benchmark del UAM pretén determinar la incertesa introduïda pels codis multifísics i multiescala acoblats d'anàlisi de reactors d'aigua lleugera. El citat benchmark es divideix en tres fases: 1. Fase neutrònica: obtenir els paràmetres neutrònics i seccions eficaces del problema específic, col¿lapsats i homogeneïtzats, a més de la anàlisi de criticitat. 2. Fase de nucli: anàlisi termo-hidràulica i neutrònica per separat. 3. Fase de sistema: anàlisi termo-hidràulica i neutrònica acoblats. En aquesta tesi es completen els principals objectius de la primera fase. Concretament, es desenvolupa una metodologia per propagar la incertesa de les seccions eficaces i altres paràmetres neutrònics a través d'un codi lattice i un simulador de nucli. Es porta a terme una anàlisi d'incertesa i sensibilitat per a les seccions eficaces contingudes en la llibreria neutrònica ENDF/B-VII. La seua incertesa es propaga a través del codi lattice SCALE6.2.1, incloent les fases per col¿lapsar i homogeneïtzar, fins aplegar a la generació d'una llibreria neutrònica específica del problema. Després, la incertesa continguda en la esmentada llibreria pot continuar propagant-se a través d'un simulador de nucli, per a aquest estudi PARCSv3.2. Per a l'anàlisi d'incertesa i sensibilitat s'ha utilitzat el mòdul SAMPLER -disponible a l'última versió de SCALE- i la ferramenta estadística DAKOTA 6.3. Com a part d'aquest procés, també es desenvolupa una metodologia per a obtenir llibreries neutròniques en format NEMTAB per ser utilitzades en simuladors de nucli. S'ha realitzat una comparació amb el codi CASMO-4 per obtenir una verificació de la metodologia completa. Aquesta s'ha provat utilitzant un reactor d'aigua en ebullició del tipus BWR. Tanmateix, no hi ha cap preocupació o limitació respecte del seu ús amb un altre tipus de reactor nuclear. Per a la quantificació de la incertesa s'utilitza la metodologia estocàstica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Aquesta metodologia fa ús del model d'alta fidelitat i un mostreig no paramètric per propagar la incertesa. Com a resultat, el nombre de mostres (determinat amb la fórmula revisada de Wilks) no depèn del nombre de paràmetres d'entrada, sols depèn del nivell de confiança i incertesa desitjats dels paràmetres d'eixida. A més, las funcions de distribució de probabilitat no estan limitades a la normalitat. El principal inconvenient és que s'ha de disposar de les distribucions de probabilitat de cada paràmetre d'entrada. Si és possible, les distribucions de probabilitat d'entrada es defineixen utilitzant informació trobada a la literatura relacionada. En cas contrari, la incertesa es defineix en base a l'opinió d'un expert. S'utilitza un segon escenari per propagar la incertesa de diferents paràmetres termo-hidràulics a través del codi acoblat TRACE5.0p3/PARCSv3.0. En aquest cas, s'utilitza un reactor tipus PWR per simular un transitori d'una caiguda de barra. Com a nova característica, cal assenyalar que el nucli es modela element a element seguint una discretizació totalment 3D. No s'ha trobat cap altre estudi que utilitze un nucli tan detallat en 3D. També s'utilitza la metodologia GRS i el DAKOTA 6.3 per a aquesta anàlisi d'incertesa i sensibilitat.¿<br>Mesado Melia, C. (2017). Uncertainty Quantification and Sensitivity Analysis for Cross Sections and Thermohydraulic Parameters in Lattice and Core Physics Codes. Methodology for Cross Section Library Generation and Application to PWR and BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/86167<br>TESIS

The chapter is devoted to the design and performance of adsorptive chillers. Basic types of design and operating principle of adsorptive chillers were analyzed. Advantages and disadvantages performance of one-, two-, three-, and four-bed solar power adsorptive chillers are compared. Performance of adsorptive refrigerators based on composite adsorbents was studied. The correlation between the adsorbent composition and the coefficient of energy performance of the adsorptive chiler was revealed. An optimal composition of adsorbent 'silica gel – sodium sulphate' is stated to be of 20% silica gel and 80% sodium sulphate. The maximal values of the coefficient of performance of cycle of studied solar adsorptive chiller about of 1.14 are stated for composites containing about 20 wt. % silica gel and 80 wt% sodium sulphate. As a consequence of decreasing of adsorbent mass, the coefficient of performance is shown to increase when sodium sulphate content in the composite increased. Regeneration process parameters of the composite were shown to strongly affect on the coefficient of performance of the adsorptive chiller. The growth of the coefficient of performance is stated to result from decreasing the difference between adsorbent temperature and regeneration temperature from 85 to 55°C. The basic factors affecting the net coefficient of energy performance of the adsorptive solar refrigerator were stated daily solar radiant flux alongside with composition of the adsorbent and difference between adsorbent temperature and temperature regeneration. Net coefficients of performance of solar adsorptive refrigerator based on composite ‘silica gel – sodium sulphate' were stated to change from 0.25 to 0.34 during operating period. Utilization of the adsorption heat is suggested to warm the heat carrier which applied to heat adsorbent during regeneration. The ways to improve the design and performance of adsorptive solar chillers are suggested. The first one involves the introduction of solar collectors made of cellular polycarbonate plastics in the design of adsorptive solar chiller. Instantaneous efficiency coefficient were calculated as special thermal performance-solar radiant flux surface density ratio, optical efficiency factor is determined as special thermal performance-solar radiant flux surface density ratio at the equal temperatures of heat transfer medium and environment, reduced heat loss factor being calculated as the product of solar collector efficiency factor and net heat loss coefficient. The environmental test of developed collectors PSK-AV2-3, PSK-AV1-2, PSK-AV2-1, PSK-VS1-2, PSK-VS2-2, PSK-VS2-3, PSK-ST10-PW were conducted. The correlation of their results with laboratory tests when the thermohydraulic stand applied is shown. Relative accuracy of laboratory and environment tests was shown to be not exceeding 5 – 7%. The optical efficiency factor and the coefficient of thermal losses of polymeric solar collectors were determined. On the basis of the dependencies of the efficiency of the solar collectors vs. the reduced temperature, optimal designs of the polymeric solar collectors for the adsorption chilling solar systems are determined to be depended on the temperature of the regeneration temperature of the sorbents. As the temperatures of the regeneration of composite adsorbent ranged from 50ºС to 60ºС, appliance of the collectors PSK-AV2-1, PSK-CT10-PW occur to be expedient, and PSK-AB2-3, PSK-VS2-3, PSK-AB1-2, PSK-VS2-2, and PSK-VS1-2 are revealed to be more efficient when regeneration temperatures increased over 80 ºС. Thermotechnical characteristics of designed polymeric solar collectors are shown to surpass conventional metal and vacuum collectors. The perspectives of polymeric solar collectors in the design of adsorptive chilling solar plants were shown. Another way to improve the performance of adsorptive solar chillers concerns with equipping it with a photosensitive element and an electric drive, which will allow changing the angle of slope of the adsorber to the horizon depending on the intensity of the solar radiation. The chapter can be useful for design the efficient adsorptive chilling plants.

Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics analysis code, named STAF, is developed based on FLUENT. With STAF code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This code is developed based on the porous media theory. In this code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF code. The calculated results are compared with a general steam generator thermohydraulic analysis code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.

This paper deals with understanding the thermohydraulic characteristics of one type of zig-zag square microchannel. Studies are done to understand the individual effect of geometric parameters like hydraulic diameter, orientation angle and connector length for Reynolds number between 100 and 400. The hydraulic diameter is varied between 100 μm and 300 μm with increment of 100 μm. Studies are carried out for orientation angle of 10°, 20° and 30°. The connector length of the study conducted in this paper is varied from 200 μm to 400 μm to 600 μm. The results are presented in terms of enhancement of Nusselt number and Poiseuille number over that in straight square microchannels. With increase in hydraulic diameter the enhancement in Nusselt number and Poiseuille number increased at a specific Reynolds number. Similarly with increase in the orientation angle the enhancement in these two thermohydraulic characteristics is observed for a specific Reynolds number. With increase in connector length both Nusselt number and Poiseuille number increased over that in straight channels of the same hydraulic diameter for a particular Reynolds number. Enhancement in Nusselt number and Poiseuille number is observed with increase in Reynolds number for all zig-zag microchannels. For the three fold increase in each of the geometric parameters the smallest enhancement is observed with increase in connector length. The greatest enhancement in thermohydraulic characteristics is observed with change in orientation between the arm and the connector.

Abstract With the development of the nuclear energy industry, small modular reactors (SMRs) have become an important option in China’s energy development due to their advantages in terms of safety and economics. The helical-coil steam generator is an important part of SMR, and the structure parameters need to be optimized urgently to reduce costs and improve safety. In this paper, various thermohydraulic parameters were used as indicators for the design quality of the helical-coil steam generator, such as the volume of the equipment, velocity of the flow, maximum temperature of the tubes, and natural circulation capability. The optimization of these target parameters has important theoretical and practical values. The genetic algorithm method is widely used to processes data efficiently, and it is able to reach a optimum design rapidly, which makes it a good choice for the optimization of parameters. The results show that this method could efficiently reduce the volume of the helical coil steam generator.

The possibility of instabilities in future nuclear reactors cooled by supercritical water is a matter of concern due to sharp changes in thermodynamic properties of coolant within the core. In the present work, a lumped parameter dynamic model of supercritical water-cooled reactor has been developed for analysis of coupled neutronic-thermohydraulic instabilities. The coolant channel is divided into two nodes with a moving boundary between them. The heater wall dynamics is described by a lumped parameter energy balance. Point neutron kinetics with one group of delayed neutrons has been used to model the power dynamics. Simple non-dimensional equations of state have been obtained for evaluating thermodynamic properties. Stability analysis has been done for various values of parameters such as the reactor power, coolant mass flow rate and inlet temperature. Stability maps have been plotted in the parameter planes. Dynamic simulations have been performed in the time domain to study the nature of operating transients. The stability analysis with neutronics is found to be more conservative. Transient simulations without neutronics indicate a supercritical Hopf bifurcation and the existence of a stable limit cycles in the unstable region. However, simulations with coupled neutronics indicate a subcritical Hopf bifurcation and the existence of unstable limit cycles in the stable region. Therefore, the analysis with neutronics is more conservative and shows that the system can be unstable for large perturbations, even if it is stable for small perturbations.

Concerning the radioactive waste management, thermoaeraulic of a new horizontal and passive concept in a closed space is studied. Corrosion is avoided, as heat transfer use confined air in natural convection. An experimental real scale mock-up has been carried out to demonstrate feasibility of this new concept and to provide reliable data for the validation of numerical simulation methods. The design of such facilities implies thermal-aeraulic studies in order to predict air and wall temperatures. For that reason, numerical CFD tools have to be validate in order to model the cooling of spent nuclear fuel in a real interim storage and make sure that safety thresholds never exceed critical values. The experimental loop represents a slice of the real module at the same scale in height and length. It consists of four superposed horizontally annular spaces (length: 5m; air gap: 5cm) where inner cylinder is equipped with electrical wires, aiming to reach a uniform and controlled power density. A natural air circulation appears between the tubes and a special cooled roof. The whole system is inside an air-conditioned building. For each annular space, inner surfaces are painted in black with a known emissivity. Instrumentation includes thermocouples in the air flow channel on the cylinders and on the roof wall, and air velocity measurement (hot wire transducer and Laser Doppler Velocimetry). During experiments, the external temperature and the heating power have been controlled and adjusted to simulate various thermal conditions. Numerous configurations and geometrical parameters have been studied to optimize the cooling conditions. The main experimental results obtained with different values of the parameters like the kind of cooling (water or air), power density (500 to 750W/m2) are presented as well as the heterogeneous power distribution on the four channels. Steady-state local temperatures are presented, with air flow distribution within annular space. Using this database, a special CFD tool has been developed to model complex and coupled phenomena. Then, numerical simulations results have been compared on a reference test case, firstly with water cooling and secondly with natural air cooling. Experimental and numerical data are presented and analysed. From that numerical validation, thermohydraulic model’s extrapolation is now in progress to design a real interim storage with 49 heated sub-channels and propose power loading in order to respect safety rules, especially temperature criteria on the wall.

Rolls-Royce has embarked on a Small Modular Reactor (SMR) design programme, which is developing a safe, deployable and affordable Nuclear Power Plant. An element of the work undertaken is the determination of the performance of nuclear reactor plants through the application of thermal hydraulic analysis methods. These methods are used throughout the product lifecycle; beginning with exploring the performance potential of design options and ultimately providing analysis to support the safety justification. To provide performance analysis at the earliest stages of design it is imperative to have a flexible analysis method that evolves alongside the design process while being able to inform the system architects at each phase. To achieve this goal an analysis method has been developed that allows for varying plant configurations with adjustable plant component geometries. The process of updating each of the plant component subsystem models has been automated and linked with an overarching database that allows system designers to directly update geometrical parameters. This approach to early concept design has led to the practice of two-stage verification and validation. The first stage is the traditional understanding of V&amp;V, the verification of the models and supporting scripts through the use of a predefined rigorous test suite which encompasses unit tests through to full system integration tests. The second stage is data verification and validation, as the method has the ability to constantly update based on data input the verification of this input data is paramount. The design data is hierarchical, which in addition to a mean and an uncertainty distribution, requires knowledge of a transfer function. Each item of data is owned by the department that is the expert in that field, for example the steam generator dimensions are owned by the components team. It is the responsibility of each of the data owners to verify their data and ensure that any transfer functions are implemented correctly. As data is used in a cross discipline setting it requires a central storage location, this is the Master Parameter List (MPL). This repository defines the data, its format, its owner, and the last time it was updated among other things. This repository is the single point of truth for every data item used in analysis. The aim of this paper is to demonstrate the way in which a flexible analysis method is created along with the framework that provides confidence in the analysis results through a two phase data driven V&amp;V approach.

Passive safety systems represent one field of research concerning the safety-related enhancement of nuclear power plants. Passive safety systems can ensure the safe removal of decay heat without an input of electrical or mechanical energy for commissioning or operation. The heat removal chain is guaranteed on the basis of the physical principles condensation, heat conduction, boiling and natural circulation. The thermal hydraulic processes in passive safety systems disagree with the plant-specific thermal hydraulics because of different operating conditions. Since the established system codes are validated for the plant-specific conditions, the operational behavior of passive safety systems is currently not sufficiently predictable. On this account, the German Federal Ministry of Education and Research initiated the joint project PANAS to investigate the decay heat removal by passive safety systems on the basis of experimental analyses, modelling and validation. Object is the heat removal chain in advanced boiling water reactors consisting of emergency condensers (EC; heat transfer from reactor core to core flooding pools) and containment cooling condensers (CCC; heat transfer from the containment to the shielding/storage pool). At Technische Universität Dresden, the test facility GENEVA was constructed for the experimental investigation of the operational behavior of the CCC. GENEVA models the CCC concerning the original thermal hydraulic conditions of the heat source and heat sink as well as the tube geometry for the heat transfer. In this way, the comparability of the thermal hydraulic phenomena is given. Previous experiments focused on the stability analysis of the natural circulation in the test facility. The focus of PANAS is on the condensation process of saturated steam at the outside of the slightly inclined tubes and the convection respectively boiling of both a stable and an unstable two-phase flow inside these tubes. For a detailed analysis, condensation rates at the outside as well as the flow structure inside have to be investigated experimentally. Therefore, the instrumentation in the heat transfer section of GENEVA is considerably enhanced. This enhancement comprises an optical measuring system for the film thickness or droplet size of the condensate, a tipping scale for the condensate mass flow, void probes for the steam void fraction and more than 100 thermocouples outside and inside the tubes for temperature profiles in axial, radial and azimuthal direction. By reference to these parameters, it is possible to examine the thermal hydraulic models for the heat transfer. The paper outlines the available models in system codes regarding condensation and boiling concerning the operating conditions of the CCC. Since dropwise condensation could be observed in previous experiments and the condensation models in system codes focus on film condensation, the review is extended beyond native models. A sensitivity analysis of the reviewed models regarding condensation shows huge differences concerning the value of the heat transfer coefficient. Furthermore, the courses of the condensation models present different dependencies regarding the heat transfer coefficient and the wall temperature. Due to this, the necessity of the experimental investigation and later the revision of the condensation models in system codes is confirmed. The comparison of the reviewed models with first experimental results outlines the tendency for the numerical description of the condensation process. Based on the investigation and validation of models concerning the heat transfer processes in the CCC, the operational behavior will be accurately predictable by established system codes, which enhances the safety investigation and the licensing. Although the conception of this investigation is founded on the CCC, the adapted models will be able to characterize the heat transfer processes boiling and condensation for saturation conditions at a relatively low pressure (maximum 4 bar) and for natural convection in general.

A small scale water-cooled experimental facility was built in order to study the complex thermohydraulic phenomena taking place in the Reactor Cavity Cooling System (RCCS) during the normal operation (steady-state case) an during accident scenario when forced convection is lost. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers. The pipes are connected via cold and hot manifolds to a water tank located on top of the cavity. Due to the complexity of the expected thermal hydraulic phenomena, a RELAP5-3D input deck was prepared in order to predict the main thermohydraulic parameters, mainly coolant flow rate and temperatures. During the facility shakedown, the coolant flow was constantly monitored in order to observe the natural circulation startup phase and some interesting features of the coolant behavior were observed. The comparison of the preliminary experimental results from a test run with the prediction of the RELAP5-3D simulations helped validating the assumptions and simplifications adopted in the model for future simulations of steady-state and transients, and confirmed the potentiality of the system code for analysis of such systems. In the present paperwork, a detailed description of the experimental facility and the RELAP5-3D model are provided. Preliminary experimental results from different test runs are described and compared with the RELAP5-3D simulation results.

Steam generator is one of the key equipments in the pressurized water reactor, from the performance point of view, it is necessary to apply optimization techniques to the design of the steam generator. In this work, the optimal designs of a U-tube steam generator (UTSG), taking minimization of the total volume and net weight as objective respectively, are carried out by considering thermohydraulic and geometric constraints using a complex-genetic algorithm (CGA). And the sensitivities of some parameters which influence the total volume and net weight of UTSG are also analyzed. Under the condition of constant secondary thermalhydraulic parameters of the steam generator, the optimal design indicates an obvious effect taking either the overall volume or the total weight of the steam generator as the objective. The optimization results show that the proposed optimal method is feasible and effective. And the results of optimal designs and sensitivity analysis would provide guidance in the engineering design of UTSG.