Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial description of the reactor core. The TH code uses few, typically 5 to 20 TH channels, which represent the core. The NK code uses explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and fine grid NK domain is necessary. However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this thesis is to study the sensitivity of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK-TH mapping in simulation of safety transients – Control Rod Drop, Turbine Trip, Feedwater Transient combined with stability performance (minimum pump speed of recirculation pumps). The research methodology consists of spatial coupling convergence study, as increasing number of TH channels and different mapping approach the reference case. The reference case consists of one TH channel per one fuel assembly. The comparison of results has been done under steady-state and transient conditions. Obtained results and conclusions are presented in this licentiate thesis.

Examensarbetet behandlar problematik som uppstår vid härdsmältningsförlopp i en kärnkraftsreaktor av typen kokvattensreaktor. Resultatet ska användas som riktlinjer till strategier som utvecklas av Severe Accident Management Guidelines (SAMG) där arbetets uppdrag är ett delmoment i framtagning av strategier för att bemästra de olika fenomen som uppstår vid härdsmälta.   Syftet med arbetet är att undersöka begränsningar för att bevara reaktortanken intakt vid haveri, genom att undersöka den minsta mängd vatten som behövs för att undvika tankgenomsmältning. Undersöka fallen som leder till härdsmälta och gruppera dem efter händelsesekvenser. Undersöka metall/vatten-reaktionen som uppstår då härden blir över 800°C och undersök om fallen kan grupperas i händelsesekvenser.  Metoder som används i arbetet är PSA-dokumentation, händelseutvecklingsträd, teoretiska beräkningar och MAAPv5.03. Resultatet beskriver att grupperingar av fallen som slutar i härdsmälta och grupperingar av metall/vatten-reaktionen hos de olika fallen kan genomföras. Resultatet beskriver även ett minsta flöde som kan föras in i reaktortanken för att hindra tankgenomsmältning och flöden upp till 100 kg/s så det finns resultat för olika flöden beroende på vilka kylmedel som är tillgängliga.  Slutsatsen av arbetet är att fall kan grupperas efter händelsesekvenser och påverkan hos metall/vattenreaktion, grupperingarna sparar tid vid ett haveriförlopp. I varje grupp kunde det svåraste fallet beräknas för minsta flöde för att klara tankgenomsmältning och flöden upp till 100 kg/s.  Framtida arbeten bör undersöka trycket och vätgasen som skapas vid vatteninmatning samt dess påverkan på reaktorinneslutningen. / The master thesis deals with problems that arise during nuclear meltdown in a nuclear powerplant of the type boiling water reactor. The work will be used as guidelines for strategies developed by Severe Accident Management Guidelines (SAMG), this master thesis is a sub-element in the development of strategies for mastering the various phenomena that arise during a meltdown.  The purpose of the work is to investigate limitations for maintaining the reactor tank intact during the meltdown by, examining the minimum amount of water needed to avoid the meltdown getting through the reactor tank. Examining the cases that lead to meltdown and group them according to the event sequences. Examine the metal/water-reaction that occurs when the core becomes over 800°C and examine if the cases can be grouped into event sequences.  Methods used in the master thesis is PSA-analysis, event development threes, theoretical calculations and MAAPv5.03.  The result describes the groupings of the cases ending in meltdown and the groupings of the metal/water-reaction of the various cases. The result also describes a minimum flow that is required to prevent meltdown of getting through the reactor tank and flow up to a 100 kg/s.  The conclusion of the master thesis is that cases can be grouped according to event sequences and the influence of the metal/water-reaction, the groupings save time in the event of a breakdown. In each group the most difficult case was calculated so that the lowest flow to prevent the meltdown from getting through the reactor tank was presented among with different flows up to 100 kg/s. Future work should investigate the pressure and hydrogen gas created by the water input and its influence on the reactor inclusion.

No description available.

<p dir="ltr">The research focuses on improving core simulation procedures in Boiling Water Reactors (BWRs) by leveraging machine learning techniques. Aimed at better fuel planning and enhanced safety, a machine learning model has been developed to predict errors in existing low-fidelity, diffusion-based core simulators. The machine learning models have demonstrated the capability to accurately and efficiently predict errors in core eigenvalue and power distribution in BWR Operations. This results in a significant improvement over conventional simulation methods in nuclear reactors without increasing computational complexity.</p>

Für den zielgerichteten Einsatz von präventiven und mitigativen Notfallmaßnahmen sowie zur Beurteilung ihrer Wirksamkeit während auslegungsüberschreitender Unfälle in Druckwasserreaktoren aber auch für Hinweise zum Störfallverlauf und für die Abschätzung der Auswirkungen auf die Anlagenumgebung müssen geeignete Störfallinstrumentierungen vorhanden sein. Insbesondere der Zustand des Reaktordruckbehälterinventars (RDB-Inventar) während der In-Vessel-Phase eines auslegungsüberschreitenden Unfalls lässt sich mit aktuellen Störfallinstrumentierungen nur stark eingeschränkt überwachen, wodurch die o. g. Forderungen nicht erfüllt werden können. Die vorliegende Arbeit beinhaltet detaillierte Untersuchungen für die Entwicklung einer Störfallinstrumentierung, welche eine durchgängige Zustandsüberwachung des RDB-Inventars während der In-Vessel-Phase eines auslegungsüberschreitenden Unfalls ermöglicht. Die Störfallinstrumentierung basiert auf der Messung und Klassifikation von unterschiedlichen Gammaflussverteilungen, welche während der In-Vessel-Phase außerhalb des Reaktordruckbehälters auftreten können. Ausgehend von der Analyse zum Stand von Wissenschaft und Technik wird der modell-basierte Ansatz für Structural Health Monitoring-Systeme genutzt, um eine grundlegende Vorgehensweise für die Entwicklung der Störfallinstrumentierung zu erarbeiten. Anschließend erfolgt eine detaillierte Analyse zu den Vorgängen während der In-Vessel-Phase und eine daraus abgeleitete Definition von Kernzuständen für einen generischen Kernschmelzunfall. Für die definierten Kernzustände werden mittels Simulationen (Monte-Carlo-Simulationen zum Gammastrahlungstransport in einem zu dieser Arbeit parallel laufenden Vorhaben) Gammaflüsse außerhalb des Reaktordruckbehälters berechnet. Die Simulationsergebnisse dienen dem Aufbau von Datenbasen für die Entwicklung und Analyse eines Modells zur Klassifikation von Gammaflussverteilungen. Für die Entwicklung des Klassifikationsmodells kommen drei diversitäre und unabhängig arbeitende Klassifikationsverfahren (Entscheidungsbaum, k-nächste-Nachbarn-Klassifikation, Multilayer Perzeptron) zur Anwendung, um die Identifikationsgenauigkeit und Robustheit der Störfallinstrumentierung zu erhöhen. Die abschließenden Betrachtungen umfassen die Validierung der Störfallinstrumentierung mittels eines Versuchstandes zur Erzeugung unterschiedlicher Gammaflussverteilungen. Im Ergebnis der Untersuchungen konnte die prinzipielle Funktionsweise der entwickelten Störfallinstrumentierung nachgewiesen werden. Unter der Voraussetzung, die Gültigkeit der definierten Kernzustände zu untermauern sowie weitere, nicht in dieser Arbeit betrachtete Kernschmelzszenarien mit in die Entwicklung der Störfallinstrumentierung einzubeziehen, steht somit insbesondere für zukünftige Kernkraftwerke mit Druckwasserreaktoren eine Möglichkeit für die messtechnische Überwachung des RDB-Inventars während auslegungsüberschreitender Unfälle bereit. Die Arbeit leistet einen wesentlichen Beitrag auf dem Gebiet der Reaktorsicherheitsforschung sowie für den sicheren Betrieb von kerntechnischen Anlagen.

Cette étude fondamentale est dédiée à la compréhension de l’influence de la composition chimique, notamment la teneur en chrome, des alliages base de nickel sur leur mécanismes de corrosion et d’oxydation. La corrosion sous contrainte intergranular (CSCIG) est un mode de dégradation qui affecte de nombreux alliages au sein des réacteurs à eau pressurisé. En particulier, les alliages base nickel tubes des générateur de vapeur (GV). La sensibilité à la CSC est désormais dépend de la teneur en chrome, ce qui a conduit au remplacement de l’alliage 600 (Ni-16Cr-9Fe) par l’alliage 690 (Ni-30Cr-9Fe). Cependant le bon comportement de l’alliage 690 en termes de résistance à la corrosion restes mal défini. L’objective de cette thèse est double : i) déterminer l’effet de la teneur en chrome, ii) contribuer à la compréhension de l’effet de fer étant un élément d’addition sur la résistance à la corrosion et l’oxydation généralisée des alliages base nickel en milieu primaire assimilé et en vapeur surchauffée à 700°C. Par ailleurs, des analyses électrochimiques pertinentes dans la température ambiante ont été mené afin d’établir une corrélation entre les propriétés physiques de film passive susceptible de protéger le matériau et de la teneur en chrome. Des alliages modèles binaires Ni-Cr, à teneur de chrome varie entre 14 et 30 % en poids, des alliages ternaires Ni-Cr-8Fe et l’alliage 600 ont été étudies. L’aspect expérimental de cette étude repose sur des techniques conventionnelles: SEM, STEM, EDX, Potentiodynamique, EIS, Chronoamperometrie, Mott-Schottky. La cinétique d’oxydation en vapeur surchauffée a été déterminée en mesurant l’apport de masse. L’impact de l’état de surface sur le processus de la corrosion et l’oxydation a été mis en évidence. Les polissages miroir et électrochimique ont été réalisés afin de découpler l’effet de l’écrouissage développé en subsurface, induit par la préparation de surface, et la composition chimique de l’alliage. La teneur en chrome limite à partir de laquelle l’alliage a un comportement satisfaisant en corrosion a été déterminé à 20% dans le milieu primaire. Cependant les analyses électrochimiques ont décelé l’existence d’une teneur en chrome optimal à 26%. La cinétique d’oxydation des alliages modèles ainsi que la morphologie des oxydes formés sur ces matériaux dans le milieu vapeur surchauffée ont indiqué l’existence d’une teneur en chrome optimal à 24%. Une dégradation des propriétés des films d’oxydes a été observée en augmentant la teneur en chrome au-dessus de l’optimum. En résumé, ce travail se préoccupe de l’optimisation de la teneur en chrome, méthode plus adéquate, pour l’amélioration de la résistance à la corrosion et l’oxydation des alliages base nickel. / This fundamental study is focused on the understanding of the influence of the chemical composition of Ni-based alloys on their corrosion and oxidation mechanisms. This work is not dedicated for a particular application. It is well known for instance that Ni-based alloys are susceptible to intergranular stress corrosion cracking (IGSCC) in primary water. Thus, Alloy600 (Ni-16Cr-9Fe), used in steam generator (SG) tubing, was replaced by higher chromium content material Alloy690 (Ni-30Cr-9Fe). This later shows a better resistance to IGSCC which may be linked to the growth of more protective oxide layer as chromium content is increased to 30 wt.%. The main goal of this study is to investigate: i) the influence of chromium content, ii) impact of iron addition on the corrosion and oxidation resistance of Ni-based alloys in primary water and superheated steam at 700°C. Furthermore, analytical approach in acidic solution is conducted at room temperature. This allowed to establish a relationship between alloying elements and physical properties of the oxide layers. For this purpose, Ni-xCr (14 ≤ x≤ 30 wt.%), Ni-xCr-8Fe (x=14,22 and 30 wt.%) model alloys and industrial material Alloy600 have been studied. To characterize the oxide scales, conventional technics were used: SEM, STEM, EDX, Potentiodynamic, EIS, Chronoamperometry, Mott-Schottky. Furthermore, steam oxidation kinetics was evaluated by means of weight gain measurements. To uncouple the effect of surface cold-work and the chemical composition of the base metal, mirror and electro polishing were carried out. In primary water, critical chromium content (20 wt.%), which corresponds to the minimum amount of chromium required to the transition from non-protective to protective and compact Cr-oxide layer, is determined. However, the analytical approach, using electrochemical technics, at room temperature elucidated the existence of optimum chromium content (26 wt.%) in terms of corrosion resistance. In superheat steam, oxidation kinetics and oxide scale characteristics showed the existence of optimum chromium content (24 wt.%) in terms of oxidation resistance. The corrosion and oxidation resistance is degraded as chromium content was increased more than optimal amount. Iron addition (8 wt.%) had a detrimental effect on the protectivess of the resulting oxide scales. Finally, this study showed that optimizing of chromium content is more appropriate method for enhancing corrosion and oxidation resistance, that increasing chromium content to high level is not necessary beneficial to those parameters. This work provides a useful knowledge to design new alternative materials. For this purpose, more investigations should be conducted to test other parameters such as: weldability, fabricability, thermal conductivity,etc.

<p>Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. The research is also focusing on evaluation of the proposed accident management schemes for mitigating the consequences of such accidents.</p><p>During a hypothetical severe accident, whatever the scenario, there is likelihood that the core material will be relocated and accumulated in the lower plenum in the form of a debris bed or a melt pool. Physical phenomena involved in a severe accident progression are complex. The interactions of core debris or melt with the reactor structures depend very much on the debris bed or melt pool thermal hydraulics. That is why predictions of heat transfer during melt pool formation in the reactor lower head are important for the safety assessment.</p><p>The main purpose of the present study is to advance a method for describing turbulent natural convection heat transfer of a melt pool, and to develop a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of Core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident.</p><p>Given the insights gained from Computational Fluid Dynamics (CFD) simulations, a physics-based model and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a Boiling Water Reactor (BWR) during the late phase of an in-vessel core melt progression. A model is developed for the core debris bed heat up and formation of a melt pool in the lower head of the reactor vessel, and implemented in a commercial CFD code. To describe the natural convection heat transfer inside the volumetrically decay-heated melt pool, we advanced the Effective Convectivity Conductivity Model (ECCM), which was previously developed and implemented in the MVITA code. In the present study, natural convection heat transfer is accounted for by only the Effective Convectivity Model (ECM). The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables simulations of heat transfer of a high Rayleigh melt pool in 3D large dimension geometry.</p><p>In order to describe the phase-change heat transfer associated with core debris, a temperature-based enthalpy formulation is employed in the ECM (the phase-change ECM or so called the PECM). The PECM is capable to represent possible convection heat transfer in a mushy zone. The simple approach of the PECM method allows implementing different models of the fluid velocity in a mushy zone for a non-eutectic mixture. The developed models are validated by a dual approach, i.e., against the existing experimental data and the CFD simulation results.</p><p>The ECM and PECM methods are applied to predict thermal loads to the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heat up and melting in the BWR lower plenum. Applying the ECM and PECM to simulations of reactor-scale melt pool heat transfer, the results of the ECM and PECM calculations show an apparent effectiveness of the developed methods that enables simulations of long term accident transients. It is also found that during severe accident progression, the cooling by water flowing inside the CRGTs plays a very important role in reducing the thermal load on the reactor vessel wall. The results of the CFD, ECM and PECM simulations suggest a potential of the CRGT cooling as an effective mitigative measure during a severe accident progression.</p>

Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents.  In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment.  The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis.   The CFD method, on the one hand, is indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling plays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs. / <p>QC 20100812</p>

L’un des accidents de dimensionnement d’un réacteur à eau pressurisée est l’Accident de Perte de Réfrigérant Primaire (APRP). L’évènement initiateur d’un tel accident est une brèche sur le circuit primaire du réacteur entrainant une perte d’inventaire en eau, et de ce fait conduit à un assèchement des assemblages combustibles. En conséquence, une augmentation considérable de la température surviendrait à l’intérieur du cœur du réacteur. Ainsi, les gaines de combustible peuvent éventuellement se déformer et des zones dites ballonnées apparaitre. Ces zones vont avoir un fort impact sur l’efficacité du refroidissement du cœur du réacteur. Pour contribuer à l’étude thermohydraulique d’un APRP, la présente thèse a pour but la caractérisation expérimentale des interactions entre un écoulement diphasique de vapeur/gouttes et une zone partiellement bouchée. Afin de reproduire un tel scénario, le banc expérimental thermohydraulique COLIBRI a été conçu. Plusieurs configurations géométriques de la zone ballonnée, caractéristiques d’un APRP, sont analysées (longueur et taux de bouchage associés au ballonnement). Afin de caractériser les échanges thermiques paroi/fluide ainsi que la dynamique des gouttes, des diagnostics optiques et thermiques sont utilisés : l’Anémométrie Phase Doppler (PDA) pour mesurer le diamètre et la vitesse des gouttes, la Fluorescence Induite par Laser (LIF) pour mesurer la température des gouttes et la Thermographie Infrarouge (IR) afin d’estimer le flux de chaleur extrait du tube par l’écoulement. En parallèle, une modélisation du problème a été développée afin d’obtenir une approche théorique de la capacité de refroidissement de l’écoulement diphasique. Le système d’équations décrivant la conservation de la masse, de la quantité de mouvement et de l’énergie permettra d’estimer l’impact respectif des différents mécanismes de transferts thermiques mis en jeu ainsi que l’évolution spatio-temporelle des paramètres thermohydrauliques / During a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), caused by a break or a leakage on the primary circuit, partial or even complete drying of the fuel assemblies may occur. In these conditions, the fuel temperature increases, leading to a significant deformation and rupture of the fuel rod cladding. The cooling flow might be impaired, according to the size and distribution of the deformed zones within the fuel assemblies during the emergency cooling phase (Reflooding phase). To contribute to the thermalhydraulic study of the reflooding phase, this study aims to characterize experimentally the coolability of a representative deformed sub-channel by a steam-droplets flow under LOCA conditions. In order to reproduce such a scenario, the experimental thermal-hydraulic set-up COLIBRI was designed. Several geometrical blockage configurations are analyzed (Blockage ratios and axial lengths). Three measurement techniques are set up to follow the cooling transient phase of each experience: Phase Doppler Anemometry (PDA) in order to obtain both velocity and diameter of droplets, Laser Induced Fluorescence (LIF) to measure the mean droplet temperature and Infrared thermography to estimate the heat flux removed by the two-phase flow. Additionally, a one-dimensional mechanistic model, taking into account of the heat transfers mechanisms in the post-dry out region, is developed in order to analyze the experimental data and identify each one of the wall-to-fluid heat transfers (radiation with vapor and droplets, convection with vapor, evaporation, and droplet impact)

Fundamental understanding of the behaviour of nuclear fuel has been of great importance. Enhancing this knowledge not only by means of experimental observations, but also via multi-scale modelling is of current interest. The overall goal of this thesis is to understand the impact of atomic interactions on the nuclear fuel material properties. Two major topics are tackled in this thesis. The first topic deals with non-stoichiometry in uranium dioxide (UO2) to be addressed by empirical potential (EP) studies. The second fundamental question to be answered is the effect of the atomic fraction of americium (Am), neptunium (Np) containing uranium (U) and plutonium (Pu) mixed oxide (MOX) on the material properties.UO2 has been the reference fuel for the current fleet of nuclear reactors (Gen-II and Gen-III); it is also considered today by the Gen-IV International Forum for the first cores of the future generation of nuclear reactors on the roadmap towards minor actinide (MA) based fuel technology. The physical properties of UO2 highly depend on material stoichiometry. In particular, oxidation towards hyper stoichiometric UO2 – UO2+x – might be encountered at various stages of the nuclear fuel cycle if oxidative conditions are met; the impact of physical property changes upon stoichiometry should therefore be properly assessed to ensure safe and reliable operations. These physical properties are intimately linked to the arrangement of atomic defects in the crystalline structure. The first paper evaluates the evolution of defect concentration with environment parameters – oxygen partial pressure and temperature by means of a point defect model, with reaction energies being derived from EP based atomic scale simulations. Ultimately, results from the point defect model are discussed, and compared to experimental measurements of stoichiometry dependence on oxygen partial pressure and temperature. Such investigations will allow for future discussions about the solubility of different fission products and dopants in the UO2 matrix at EP level.While the first paper answers the central question regarding the dominating defects in non-stoichiometry in UO2, the focus of the second paper was on the EP prediction of the material properties, notably the lattice parameter of Am, Np containing U and Pu MOX as a function of atomic fractions.The configurational space of a complex U1-y-y’-y’’PuyAmy’Npy’’O2 system, was assessed via Metropolis-Monte Carlo techniques. From the predicted configuration, the relaxed lattice parameter of Am, Np bearing MOX fuel was investigated and compared with available literature data. As a result, a linear behaviour of the lattice parameter as a function of Am, Np content was observed, as expected for an ideal solid solution. These results will allow to support and increase current knowledge on Gen-IV fuel properties, such as melting temperature, for which preliminary results are presented in this thesis, and possibly thermal conductivity in the future. / Doctorat en Sciences de l'ingénieur et technologie / info:eu-repo/semantics/nonPublished