Dryout and Departure from Nucleate boiling (DNB) are utmost thermal-hydraulic concerns for the safety of LWRs. The behavior of two-phase flows at these conditions is still not fully understood. There is at least a need for a good local velocity and void fraction database at these conditions. This database can be exploited by CFD codes, thereby leading to understanding and predicting DNB and boiling crisis. Since these conditions occur in LWR at pressures greater than 70 bar and temperatures above 285 $^oC$, most instrumentations fail at these conditions. So there is a need for developing or optimizing new instruments for this specific objective. This study will look into the application of Hot Wire Anemometry (HWA) for this application. Previous experiments at near saturation conditions were studied, the hurdles of application of HWA in the HWAT loop at KTH were also investigated. Finally, the deposition of thin film on the HWA sensors for protection was studied. / Dryout och avvikelse från kärnkokning (DNB) är extrema termiska hydrauliska problem för säkerheten för LWR. Tvåfasflödets beteende under dessa förhållanden är fortfarande inte helt förstådd. Det finns ett behov av en god lokal hastighets- och tomrumsfraktionsdatabas under dessa förhållanden. Denna databas kan användas av CFD-koder, vilket leder till att förstå och förutsäga DNB och den kokande krisen. Eftersom dessa förhållanden förekommer i LWR vid tryck större än 70 bar och temperaturer över 285 oC, misslyckas de flesta instrument vid dessa förhållanden. Så det finns ett behov av att utveckla eller optimera nya instrument för detta specifika mål. Denna studie kommer att undersöka tillämpningen av Hot Wire Anemometry (HWA) för denna applikation. Tidigare experiment vid nästan mättnadsförhållanden dissekerades, hinder för tillämpningen av HWA i HWAT-slingan vid KTH undersöktes också. Slutligen undersöks avsättningen av tunn film på HWA-sensorerna för skydd.

OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg.The main objective of present thesis is to build a TRACE calculation model for simulating thermal hydraulic behaviors in LSTF and PV upper-head SBLOCA, so as to assess different modeling options and parameters of TRACE code. The results show that TRACE code well reproduce the complex physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimates the break mass flow from upper head which affects the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the flow area of leakage between downcomer and hot leg. Besides, the effect of the break location, low pressure injection system (LPIS) and accumulator setup are also studied.

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 were developed to validate the TRACE code (version 5.0 patch 2). The simulated transient thermal-hydraulic behaviors during the LOCA tests including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core are compared with experimental data. The simulation results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model, choked flow model , insulator in the steam dome, K-factor in the test section, and pump trip, on the results. The sensitivity analyses show that both the models and parameters have significant influence on the outcome of the model.

With the coming of Generation IV systems, there is a need for thermal-hydraulic codes to model such advanced reactors. Codes for High Temperature Gas-cooled Reactors (HTGRs) already exist, but often suffer from insufficient validation and little user experience. Therefore, some existing codes created for Light Water Reactors are updated with HTGR related features. In this study, the feasibility of providing the TRACE thermal-hydraulics code with those features is analysed. Two models were used, one of a pebble bed core, one of a prismatic block reactor. For this purpose the effective conductivity test of the High Temperature Test Unit was taken as a benchmark for the pebble bed core. For the prismatic block reactor a model of the GEMINI+ reactor was created. This would allow to simulate not only steady state, but also Depressurised Loss of Forced Cooling scenarios. For both models the effective conductivity is known to play a major role and, consequently, a model to incorporate such feature was developed and implemented within TRACE's control system module. Results show that TRACE has a good potential for HTGR simulation, but currently available models still provide unstable solutions. It is concluded that TRACE needs additional adjustments in order to be employed for HTGR safety analyses in the future. / Med fjärde generationens system på ingång, finns det ett behov av termisk-hydrauliska koder för att modellera sådana avancerade reaktorer. Koder för gaskylda högtemperaturreaktorer (HTGR) finns redan, men lider ofta av otillräcklig validering och liten användarupplevelse. Därför uppdateras vissa befintliga koder som skapats för lättvattenreaktorer med HTGR-relaterade funktioner. I denna studie analyseras möjligheten att tillhandahålla TRACE termisk-hydraulisk kod med dessa funktioner. Två modeller användes, den ena av en pebble-bed reaktor, den andra av en prismatisk blockreaktor. För detta ändamål togs det effektiva konduktivitetstestet för högtemperaturtestenheten som ett riktmärke för pebble-bedens härd. För den prismatiska blockreaktorn skapades en modell av GEMINI+-reaktorn. Detta skulle göra det möjligt att simulera inte bara steady state, utan även scenarier med trycklös förlust av forcerad kylning. För båda modellerna är den effektiva konduktiviteten känd för att spela en stor roll och följaktligen utvecklades och implementerades en modell för att införliva en sådan funktion inom TRACEs kontrollsystemmodul.. Resultaten visar att TRACE har en god potential för HTGR-simulering, men för närvarande tillgängliga modeller ger fortfarande instabila lösningar. Slutsatsen är att TRACE behöver ytterligare justeringar för att kunna användas för HTGR-säkerhetsanalyser i framtiden.

No description available.

The effects of appendages on flow characteristics and scalar mixing in gap-connected twin-subchannel geometries has been assessed. The assessment considers a symmetric, rectangular compound channel geometry connected by a single rectangular gap using computational fluid dynamics (CFD). Detailed numerical models (geometry and turbulence), characterizing the full test section from a reference experimental study, are generated and validated against measurements. Time varying details of the gap induced periodic structures and appendage induced vortices are captured through calculations in an unsteady Reynolds averaged Navier-Stokes (RANS) framework coupled with the Spalart-Allmaras (SA) turbulence model closing the RANS equations. Companion simulations are performed at each of two Reynolds numbers (2690 and 7500), one with and one without a gap-centered appendage. The appendage size modelled is representative of CANDU endplates. The appendage effects on flow characteristics and mixing are isolated through comparison of the associated simulations. In the absence of appendages, fluid exchange between subchannels is dominated by quasi-periodic flow pulsations through the gap formed due to flow instability in the near gap region. Without a gap-centered appendage, the magnitude, frequency, and structure length of the gap flow pulsations are well predicted by the model at both Reynolds numbers. The total tracer transfer between subchannels is reasonably well predicted for Re = 2690 (within approximately 17% of the experimental value). The model fails to capture the measured increase in scalar transfer through the gap with increased Reynolds number, underpredicting scalar mixing by 55% at Re = 7500. An argument is presented that the use of an isotropic turbulence model in the channel (SA), which precludes the development of channel secondary flows, is the source of the discrepancy between modelled and measured mixing at Re = 7500. Appendages, such as those introduced by end plates or bearing pads in CANDU fuel bundles, augment the exchange process between subchannels. With an appendage representative of a CANDU fuel bundle endplate introduced into the gap region, crossflow velocity and frequency are predicted to increase immediately downstream of the appendage due to flow diversion and vortex shedding. The higher local frequency is shown to be consistent with the vortex shedding frequency calculated for a stationary rectangular cylinder at the gap conditions. Further downstream, gap induced instabilities begin to re-establish as the dominant contributor to crossflow pulsations although they are not fully recovered by the test section exit. Mixing is augmented more by the appendage with increasing Reynolds number for the range examined. / Thesis / Master of Applied Science (MASc) / The fuel bundle and pressure tube assembly in the core of a CANDU reactor forms an intricate web of subchannels of varying geometries with interconnecting gaps. Heat generated within the fuel bundles is removed by coolant flowing through the pressure tube and within the bundle subchannels. Although flow is nominally axial along the length of the rod bundles, coolant is free to move between subchannels through the gaps by a variety of mechanisms. Detailed fluid flow in these rod bundle geometries is a complex 3D phenomenon, strongly affected by fluid turbulence and flow instabilities associated with the subchannel geometry. This flow is investigated in the current study and extended to include the effect of appendages, which hold the fuel rods in place, to determine their impact on mixing along the length of the bundle. Particular applications of the results of this study are in the areas of nuclear reactor performance and safety. The extent of coolant exchange between subchannels affects the local subchannel flow and temperature and, as a result, local cooling at the fuel element surfaces. Fuel element cooling is a principal component of reactor analysis under both normal operating conditions and postulated accident scenarios.

<p dir="ltr">Flow restrictions like wire-wrapped rod bundles and U-bend geometries are critical in fast reactor cores, Pressurized Water Reactor (PWR) steam generators, and helical-coil steam generators. This study investigates their effects on single-phase and two-phase flows to enhance the understanding of thermal-hydraulic phenomena in nuclear reactor systems. A hydrodynamic test facility for a 7-pin wire-wrapped rod bundle was designed and constructed using scaling analysis. The test section consists of seven wire-wrapped stainless-steel rods, each 19.05 mm in diameter, housed in an acrylic hexagonal channel measuring 2092 mm in height. Experimental pressure drop data was collected across a wide range of Reynolds numbers. The data validated the Upgraded Cheng and Todreas Detailed (UCTD) correlation within ±10% accuracy, and the <i>SST k-ω</i> turbulence model was identified as the most reliable for predicting pressure drops in 7-pin wire-wrapped rod bundle through Computational Fluid Dynamics (CFD) simulations. For the U-bend study, a new experimental database was established using the existing Purdue University separate-effects test facility, featuring a 25.4 mm inner diameter pipe and a U-bend with curvature to diameter ratio of 9. Detailed local data was collected under eight test conditions at ten measurement locations, which includes void fraction, gas velocity and bubble diameter, measured with miniaturized four-sensor conductivity probes, while pressure loss was obtained using pressure transducers. Mechanistic models were developed to characterize the U-bend effects, including pressure loss, variance of void fraction, U-bend dissipation length and bubble velocity. The Lockhart-Martinelli correlations can be used to predict two-phase pressure drops across U-bend with some modifications. Variance of void fraction represents U-bend strength, dissipating exponentially with dissipation length determined by the dissipation rate. A modified Froude number is used to model variance of void fraction, dissipation rate, U-bend dissipation length and bubble velocity, with predictions generally within ±10% accuracy. Other closure models needed in the Interfacial Area Transport Equation (IATE) were also developed. Experimental data revealed strong correlation between variance of void fraction and covariance of Random Collision, modeled using the modified Froude number. Model coefficients for bubble interaction terms were determined by evaluating each region (i.e., vertical upward, U-bend, U-bend dissipation, vertical downward) using experimental data. The one-group interfacial area transport along the whole test section was evaluated using all these closure models and U-bend effects models, with deviations generally within ±15%. The study also identified limitations of existing Multiphase Computational Fluid Dynamics (MCFD) models in simulating bubbly flow across U-bend.</p>

Les sels fondus ont récemment été proposés comme milieux caloporteurs à haute température. Dans l'industrie nucléaire, le concept de réacteur à sels fondus (MSR en anglais) est le seul concept de quatrième génération qui propose l'utilisation d'un sel fondu liquide comme combustible nucléaire. Cette innovation présente des aspects positifs pour la conception et la sûreté nucléaire, mais impose de nouveaux défis. Le réacteur rapide à sels fondus (MSFR en anglais) est un concept qui est actuellement étudié dans le projet européen H2020 SAMOFAR, incluant le développement et la validation expérimentale (dans la plateforme expérimentale SWATH) de modèles plus performants pour les sels fondus : tel est l'objectif de ce travail de thèse. En outre les modèles développés peuvent s'appliquer à d'autres MSRs et à d'autres applications énergétiques utilisant des sels fondus comme milieux caloporteurs.La thèse suivante est divisée en trois parties :Premièrement, le développement de modèles pour décrire de façon réaliste certains des phénomènes thermiques microscopiques et macroscopiques associés à l’utilisation de sels liquides fondus comme milieux caloporteurs. Cette partie comprend l’utilisation et le développement de nouveaux modèles neutroniques pour étudier la production d'énergie nucléaire, ainsi que la modélisation des phénomènes turbulents dans les sels fondus, l’étude de l’interaction du rayonnement thermique et la turbulence dans les sels fondus. Enfin, cette partie traite également du développement d’une approche multi-échelle pour l'étude précise de la solidification/fusion dans les sels.Deuxièmement, la conception et la mise en œuvre d’expériences dédiées à la validation de ces modèles. Deux expériences clés ont été conçues au cours de cette thèse et ont été implémentées dans la plate-forme SWATH. L'objectif de ces expériences est d'étudier le comportement de différents modèles de turbulence et de tester les modèles de solidification développés dans les sels fondus.Troisièmement, les modèles développés ont été couplés dans une plateforme multi-physique pour l'étude précise du transitoire drainant du MSFR. / Molten salts have been recently proposed as high-temperature heat carrier media for energy applications. In the nuclear industry, the Molten Salt Reactors (MSRs) are the only fourth generation concept proposing the usage of a liquid nuclear fuel. This innovative aspect allows proposing improved safety and design features, but it leads to novel challenges. In particular, the Molten Salt Fast Reactor (MSFR) is a MSR concept that is currently being studied in the H2020 European project SAMOFAR. Among the project activities, there are the development of more performant molten salts models and their experimental validation through the SWATH platform. This is the objective of the present thesis. However, the models developed are appropriate for other MSRs and other energy applications using molten salts as heat carrier media.The following thesis is divided into three parts.The first part is dedicated to the development of models for describing realistically some of the microscopic and macroscopic thermal phenomena associated with the usage of liquid molten salts as heat carrier media. This part includes the development and implementation of neutronic models to study nuclear power production in the MSFR, the study of turbulence and turbulence-radiation interaction in molten salt flows and the development of a multiscale approach to model the solidification/melting phenomena in salts.The second part is devoted to the design and implementation of dedicated experiments for validating these models. Two key experiments are addressed: an experiment to study the behavior of different turbulence models after a boundary layer detachment and one to test the multiscale solidification models developed for molten salts.The third part is committed to the coupling of the models developed into a multiphysics platform for the precise study of the draining transient of the MSFR.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

Das Vorhaben bestand in der Validierung des gekoppelten neutronenkinetisch-thermohydraulischen Programmkomplexes ATHLET/DYN3D für Siedewasserreaktoren durch Teilnahme an dem OECD/NRC Benchmark zum Turbinenschnellschluss. Das von der OECD und der amerikanischen NRC definierte Benchmark basiert auf einem Experiment mit Schließens des Turbinenschnellschlussventils, das 1977 im Rahmen einer Serie von 3 Experimenten im Kernkraftwerk Peach Bottom 2 durchgeführt wurde. Im Experiment erzeugte das Schließen des Ventils eine Druckwelle, die sich unter Abschwächung bis in den Reaktorkern ausbreitete. Die durch den Druckanstieg bewirkte Kondensation von Dampf im Reaktorkern führte zu einem positiven Reaktivitätseintrag. Der folgende Anstieg der Reaktorleistung wurde durch die Rückkopplung und das Einfahren der Regelstäbe begrenzt. Im Rahmen des Benchmarks konnten die Rechenprogramme durch Vergleiche mit den Messergebnissen und den Ergebnissen der anderen Teilnehmer an dem Benchmark validiert werden. Das Benchmark wurde in 3 Phasen oder Exercises eingeteilt. Die Phase I diente der Überprüfung des thermohydraulischen Modells für das System bei vorgegebener Leistungsfreisetzung im Kern. In der Phase II wurden 3-dimensionale Berechnungen des Reaktorkerns für vorgegebene thermohydraulische Randbedingungen durchgeführt. Die gekoppelten Rechnungen für das ausgewählte Experiment und für 4 extreme Szenarien erfolgten in der Phase III. Im Rahmen des Projekts nahm FZR an Phase II und Phase III des Benchmarks teil. Die Rechnungen für Phase II erfolgten mit dem Kernmodell DYN3D unter Berücksichtigung der Heterogenitätsfaktoren und mit 764 thermohydraulischen Kanälen (1 Kanal/Brennelement). Der ATHLET-Eingabedatensatz für die Reaktoranlage wurde von der Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) übernommen und für die Rechnungen zu Phase III, die mit der parallelen Kopplung von ATHLET mit DYN3D erfolgten, geringfügig modifiziert. Für räumlich gemittelte Parameter wurde eine gute Übereinstimmung mit den Messergebnissen und den Resultaten anderer Codes erzielt. Der Einfluss der Modellunterschiede wurde mit Hilfe von Variantenrechnungen zu Phase II untersucht. So können Unterschiede in der Leistungs- und Voidverteilung in einzelnen Brennelementen auf die unterschiedliche neutronenkinetische und thermohydraulische Modellierung des Reaktorkerns zurückgeführt werden. Vergleiche zwischen ATHLET/DYN3D (parallele Kopplung) und ATHLET/QUABOX-CUBBOX (interne Kopplung) zeigen für räumlich gemittelte Parameter nur geringe Unterschiede. Abweichungen in den lokalen Parametern können im wesentlichen mit der unterschiedlichen Modellierung des Reaktorkerns erklärt werden (geringere Anzahl von modellierten Kühlkanälen, keine Berücksichtigung der Heterogenitätsfaktoren und ein anderes Siedemodell in der Rechnung mit ATHLET/QUABOX-CUBBOX). Die Rechnungen für die extremen Szenarien von Phase III zeigen die Anwendbarkeit des gekoppelten Programms ATHLET/DYN3D für die Bedingungen bei Störfällen, die weit über das Experiment hinausgehen.