The major goal of the work is to develop effective additives for U-10Zr (wt.%) metallic fuel to mitigate the fuel-cladding chemical interactions (FCCIs) due to fission product lanthanides and to optimize the fuel phase mainly by lowering the gamma-onset temperature. The additives Sb, Mo, Nb, and Ti have been investigated. Metallic fuels with one or two of the additives and with or without lanthanide fission products were fabricated. In this study, Ce was selected as the representative lanthanide fission product. A series of tests and characterizations were carried out on the additive-bearing fuels, including annealing, diffusion coupling, scanning electron microscopy (SEM), X-ray powder diffraction (XRD), and differential scanning calorimetry (DSC). Sb was investigated to mitigate FCCIs because available studies show its potential as a lanthanide immobilizer. This work extends the knowledge of Sb in U-10Zr, including its effect in the Zr-free region. Sb forms precipitates with fuel constituents, either U or Zr. However, it combines with the lanthanide fission product Ce when Ce is present. Those Sb-precipitates are found to be stable upon annealing, and are compatible with the cladding. The additive does not change the phase transition of U-10Zr. Mo, Nb, and Ti have been investigated for phase optimization based on the known characteristics shown in the binary phase diagrams. The quaternary alloys, i.e., two Mo-bearing alloys and two Nb-bearing alloys, were investigated. Compared to U-10Zr, a few weight percentages of Zr are replaced by those additives in the quarternary alloys. The solid-state phase transitions were determined (alpha and U2Ti transfer into gamma). The transition temperature varies depending on the compositions. The Mo-bearing alloys have lower -onset temperatures than the Nb-bearing alloys. All of them have lower gamma-onset temperatures than that of U-10Zr. Since low gamma-onset temperature is favorable, the results indicate that the fuel phase can be optimized by the replacement of a few weight percentages of Zr into those additives. All the experiments were out-of-pile tests. Therefore, in-pile experiments will be necessary to fully evaluate the performance of the additives in the future. / Doctor of Philosophy / Fuel is the "heart" of a nuclear reactor, and fuel development is a key to improving the performance and reliability of a nuclear reactor. This study investigated the effects of metallic fuel additives in a sodium-cooled fast reactor (SFR). SFRs are an advanced reactor design. Metallic fuel, e.g., U-10Zr (wt.%), is one of the common candidates for SFR fuel. The aim of this study is to develop effective additives for U-10Zr metallic fuel to improve fuel performance. The study has two main objectives. The first one is to mitigate the fuel-cladding chemical interactions (FCCIs), while the second one is to optimize the fuel phase. Four additives, i.e., Sb, Mo, Nb, and Ti have been investigated. The study is a pioneer for the application, thus, the experiments were performed without considering the irradiation effect. Metallic fuels with one or two additives were fabricated, with a series of tests being performed at a laboratory scale. The additive, Sb, was used to mitigate the FCCIs, since FCCIs are a limitation of fuel utilization (i.e., burnup). Lanthanides are produced during fuel operation and attack cladding, being one of the reasons for FCCIs. It is known that the additive Sb has the potential to bind lanthanides into stable precipitates. This work brings the investigation a step further, providing more evidence to demonstrate the stability of the precipitates and the compatibility with cladding. The results are favorable as they demonstrate that the lanthanides will not attack the cladding if they can be caught by the additive Sb in the fuel. The additives Mo, Nb, and Ti were investigated to optimize the phase. One of the favorable phase properties is the gamma-onset temperature - the lower the better. For example, the gamma-onset temperature is 776°C in pure U, while it is 680°C in U-10Zr (meaning that 10 wt.% Zr lowers the gamma-onset temperature by 96°C). In this work, the exploration moves forward by replacing a few percentages of Zr with Mo+Ti, or Nb+Ti. After the change, the gamma-onset temperatures are further decreased, with the temperatures decreasing more in the Mo-bearing fuels than in the Nb-bearing fuels. The significance of this work is twofold. Firstly, it extends the knowledge of Sb as an additive for mitigating FCCIs; secondly, it shows that Mo, Nb, and Ti can optimize the fuel to achieve a favorable phase property. The results provide strong reasons for additional irradiation tests in the future.

Le choix de la filière SFR (Sodium Fast Reactor : Réacteurs à neutrons rapides refroidis par du sodium liquide), par la France conduit à la réalisation d’un prototype de quatrième génération nommé ASTRID. Le développement de ce type de réacteurs présente plusieurs défis, en particulier du point de vue de l’amélioration de la démonstration de la sûreté et de la surveillance du fonctionnement. Cette dernière passe, entre autres, par la caractérisation de l’engazement du sodium liquide (présence de microbulles de gaz). La caractérisation de l’engazement est l’objet de cette étude, elle implique la détermination du taux de vide (fraction volumique de gaz) et de l’histogramme des rayons des microbulles. Le travail bibliographique réalisé a montré que les techniques acoustiques linéaires de caractérisation des nuages de bulles ne permettaient pas de répondre pleinement à cette problématique, en revanche des pistes prometteuses ont été identifiées en étudiant les techniques acoustiques non linéaires. Cette dernière voie a par conséquent été explorée. Un banc expérimental en eau permettant la génération et le contrôle optique de nuages de microbulles nous a permis de valider finement la reconstruction d’histogrammes des rayons grâce à une technique de mixage nonlinéaire d’une haute fréquence avec une basse fréquence. La potentialité du mixage de deux hautes fréquences, plus intéressante d’un point de vue industriel, a par ailleurs été démontrée. Enfin, les bases de la transposition originale d’une technique de spectroscopie de résonance non linéaire appliquée à un nuage de bulles ont été posées, grâce à la mise en place de résonateurs acoustiques. Les résultats obtenus offrent de nombreuses perspectives, tant du point de vue des applications industrielles que du point de vue plus fondamental de la compréhension du comportement acoustique non linéaire d’une bulle excitée par plusieurs fréquences et d’un nuage de bulles excité à basse fréquence. / The SFR system chosen (Sodium Fast Reactor: fast neutron reactors cooled by liquid sodium) by France led to afourth-generation prototype named ASTRID. The development of this kind of reactors presents several challenges, particularly in terms of improving the safety and monitoring operation. This involves, among other things, characterization of the bubbles presence in liquid sodium. The characterization of the bubbles presence is the subject of this thesis. It involves the determination of void fraction (gas volume fraction) and histogram of the radiiof bubbles. The bibliographic work done has shown that linear acoustic techniques for the characterization of bubble clouds are inadequate to achieve this. However promising leads have been identified by studying nonlinear acoustic techniques. This last idea has therefore been explored. An experimental water bench for the generation and optical control of microbubbles cloudallowed us to validate finely the reconstruction of histograms of radii through a technique of nonlinear mixing of a high frequency with a low frequency. The potential of the mixing of two high frequencies, more interesting for the industrial point of view has also been demonstrated. Finally, the bases of the transposition of an original technique of nonlinear resonance spectroscopy applied to a bubbles cloud were explored through the introduction of acoustic resonators. The results offer many interesting opportunities, both in terms of industrial applications and formore fundamental understanding of non-linear behavior of a bubble excited by multiple frequencies and of bubbles clouds excited at low frequency.

Spent Fuel Repository for low and intermediate radioactive waste, SFR, is owned by SKB (Swedish Nuclear Fuel and Waste Management Co). The repository needs more monitoring of groundwater seepage and improvement of the monitoring methods. Seepage is currently managed by a system that collects the seepage in 2 levels before pumping it to the surface. The current method only allows for the continuous measurement of groundwater seepage in one location in the system. Seepage is measured once per year in different locations in the repository to monitor the amount of seepage in the different parts of the repository. The seepage is primarily caused by deformation zones in the vicinity of the facility. The suggested improvements to the measuring system are to increase the number of places that measure groundwater seepage and up dated the current system in order to increase the measuring accuracy.

No description available.

Dans le cadre du projet ASTRID ((Advanced Sodium Technological Reactor for Industrial Demonstration), les interactions fluide-structure mettant en jeu la dynamique du coeur (gerbage), tels qu'elles peuvent survenir lors d'un séisme, sont d'un grand intérêt. Le gerbage du coeur est également reconnu comme l'événement le plus plausible pour expliquer les quatre AURN (Arrêt d'Urgence pour Radioactivité Négative) survenus dans le réacteur Phénix, durant les années 1989 et 1990. L'objectif poursuivi est d'améliorer, pour leurs aspects dynamiques, la compréhension des interactions fluide-structure susceptibles de se produire dans un SFR (Sodium Fast Reactor). Le centre d'intérêt principal étant phénomène de dissipation visqueuse, cette thèse entreprend trois approches expérimentales, numérique et analytique, en s'appuyant sur des expériences de vibrations libres menées sur deux installations, PISE1A, mono-assemblage et PISE2C, multi-assemblages. Deux séries d'expériences de vibrations libres ont été menées sur PISE1A, en faisant varier la hauteur d'eau et en utilisant un mélange d'eau et de glycérol, dans des proportions variables. Le but est d'examiner l'influence des variations de masse ajoutée et de viscosité sur la dynamique des oscillations de l'assemblage. Les simulations numériques correspondantes, développées dans le code CAST3M, se sont appuyées sur la résolution des équations de Navier-Stokes 3D. Les écarts entre les résultats numériques et expérimentaux sont présentés et analysés. En particulier, les effets d'extrémité se sont révélés être d'une importance marginale. Des expériences de vibrations libres ont également été effectuées sur PISE2C en sollicitant l'installation de trois façons différentes : mise en mouvement globale, mise en mouvement par la couronne externe puis par la couronne interne. Un modèle réticulé, fondé sur des hypothèses de symétrie et de linéarité a été développé parallèlement. Les résultats expérimentaux ont permis de confirmer les symétries mais ont remis en cause les hypothèses de linéarité. Ce résultat encourage à persévérer dans la voie des modèles déterministes. / In the scheme of French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project, fluid-structure interaction phenomena involved in the dynamic behaviour of core flowering, which could happen during seismic events, are of high interest. Also core flowering behaviour is considered as the main initiating event for the four SCRAMs that happened in Phénix reactor during 1989 and 1990. In objective to improve the knowledge of fluid-structure interaction phenomena of dynamic issues in a SFR core, especially focused on damping, this Ph.D. thesis have been conducted in experimental, numerical and analytical approaches based on free-vibration experiments on mono-assembly test facility PISE-1A and multi-assembly test facility PISE-2C. Two series of free-vibration experiments have been performed on PISE-1A with different water heights and different mass fractions of water-glycerol mixtures to examine the dynamic behaviours with respect to different added mass, different densities and viscosities. Corresponding numerical interpretations have been conducted with 3D Navier-Stokes model in CAST3M code. Sources of uncertainties are discussed to explain the discrepancies between the numerical computation and experimental results. Edge effects are not found to have an important impact on the dynamic behaviours of the system. On PISE-2C, free-vibration experiments with different modes of excitations have been conducted, including total flowering, partial flowering with internal crown excited and partial flowering with external crown excited. A reticulate model with homogenised linear hypothesis has been developed to interpret PISE-2C experiments. Good symmetries are found in PISE-2C suggesting that the deterministic tool is valid for the analysis.

In the past years, more concerns are focused on the nuclear waste management due to the very long half-lives of various actinides produced in Light Water Reactors (LWRs). Sodium Fast Reactors (SFRs) are thus becoming more attractive since they are known to be very efficient to transmute long-lived radionuclides present in spent fuel. However, the current simulation tools (thermal-hydraulics code with point kinetics) and safety assessment methods are not as mature as for LWR applications and need to be enhanced. This thesis aims at filling the gap in safety analysis of SFR cores to reach a standard similar to LWR applications by applying multi-physics modelling. In contrast to LWRs, the reactivity in SFRs is affected by three main feedback: the Doppler broadening reactivity effect, the sodium density change reactivity effect and the thermal expansion of several mechanical components of the reactor. In this thesis, the thermal-hydraulic system code ATHLET is coupled with the three-dimensional neutron-physics code PARCS for transient analysis. Developed at GRS, ATHLET was recently upgraded for sodium coolant properties. The nodal diffusion codes PARCS, developed at the University of Michigan, can solve the multi-group diffusion equation in hexagonal geometry. While both codes already have the main features to simulate SFRs, the development of models dedicated to the thermal expansion effect of reactivity is necessary. The latter has three main origins i.e. the core axial thermal expansion effect (caused by the fuel and the cladding axial thermal expansion), the core radial thermal expansion effect (caused by the diagrid thermal expansion), the control rod displacement due to the thermal expansion of the Control Rod Drive Lines (CRDLs), the strongback and the reactor vessel. Thus, the three main new developments achieved in the scope of this work are: - Development of a method to generate homogenized multi-energy-group neutron macroscopic cross sections (needed by PARCS) for SFR applications which consider not only the Doppler temperature and sodium density but also the core axial and radial thermal expansion. - Development of a three-dimensional core radial thermal expansion model and its implementation in PARCS. A core axial thermal expansion model has already been developed for PARCS prior to this work. - Development of a module in ATHLET for modelling the control rod displacement as a result of the influence of the reactor structures thermal expansion. The parametrized homogenized multi-energy-group neutron macroscopic cross section libraries for PARCS applications are generated with the Monte Carlo reactor-physics code Serpent. For all materials contained in fuel assemblies, a three-dimensional model is used while the SPH method is applied to materials contained in non-fuel assemblies (e.g. control rods, etc.). The cross section libraries are collapsed into a 12-energy-group structure. Furthermore, a dedicated module was successfully developed and implemented within the core simulator KMACS (developed at GRS). The core radial thermal expansion effect is implemented in PARCS using a coordinate transformation of the diffusion equation from the expanded state to the nominal geometry. The core radial thermal expansion depends on the diagrid temperature. It is calculated by ATHLET and transferred to PARCS by the extended interface between both codes. The modelling of the control rod displacement as a result of the reactor structures thermal expansion is performed by a module linked to ATHLET. The strongback, the reactor vessel and the CRDLs are modelled as heated structures in ATHLET, which calculates their respective temperature. The module can compute the thermal expansion of each structure as well as the total control rod banks displacement. The new techniques are verfied on a selected case study, the ASTRID core design. First, full core criticality simulations are performed with the Monte Carlo reactor-physics code Serpent (considered as reference calculations) and with PARCS. Good agreement between the two codes is achieved in terms of multiplication factors and power distribution. This allows to conclude that the developed method for neutron cross section libraries can be used for SFR applications. The newly implemented core radial expansion model in PARCS is successfully verified on the ASTRID core with the standalone version of PARCS. Then, various transient simulations are performed in order to separately analyse the different contributions to the reactivity by: the Doppler broadening effects, the sodium density change effect, the core radial and axial thermal expansion effect and the control rod displacement effect. It is demonstrated that the core power responses are plausible which allows the conclusion that all the different thermal expansion models are properly implemented. Furthermore, the presented simulations show very different core power responses. It appears that the effect of the sodium density change on reactivity is a parameter that is strongly heterogeneous (depending on the core location). This shows the importance of using a three-dimensional neutron kinetics model rather than a point-kinetic model for transient simulations with thermal-hydraulic codes. Moreover, the time-scale of the various effects are ranging from few seconds to several hundred seconds. While the Doppler broadening, the sodium density change, as well as the core axial and radial thermal expansion effects on reactivity are fast, the thermal expansion of the strongback and the vessel only appears after several hundred seconds. This emphasizes the importance of considering all thermal expansion effects in addition to the usual thermal-hydraulic feedback parameters (e.g. fuel temperature, coolant density etc.) to be able to compute the core behavior realistically.:Contents Abstract II List of Figures VII List of Tables X List of Acronyms XI Acknowledgments XIII 1 Introduction 1.1 Sodium cooled fast reactors 1.1.1 Fast reactor development 1.1.2 Comparison of sodium fast reactor and pressurized water reactor designs 1.1.2.1 Neutron spectrum 1.1.2.2 Breeding 1.1.2.3 Partitioning and Transmutation 1.1.2.4 Control of the reactivity in the core 1.1.2.5 Coolant properties 1.1.2.6 Reactivity feedback 1.1.2.7 Comparison summary 1.2 Objectives and structure of the thesis 1.2.1 Objectives 1.2.2 Structure of the thesis 2 State of the art of Sodium Fast Reactor safety assessment 2.1 Relevant safety events to consider for Sodium Fast Reactors 2.2 Major reactivity feedback mechanisms 2.3 State of the art of safety analysis methods for Sodium Fast Reactor 3 Methods and codes for safety assessment of sodium cooled fast reactors 3.1 Neutronics core calculations 3.1.1 Core calculations with the diffusion code PARCS 3.1.2 Generation of nodal few-group cross sections with the Monte Carlo code Serpent 3.1.3 Core simulator KMACS 3.2 Thermal-hydraulics simulations with the system code ATHLET 3.3 Coupled three-dimensional thermal-hydraulics / neutronics calculations 4 Development of three-dimensional thermal expansion models 4.1 General calculation approach proposed for safety assessment 4.2 Thermal expansion in solids 4.3 Model for generating nodal few-energy-group cross sections for deterministic core analysis 4.3.1 Energy group structure 4.3.2 Full-scale three-dimensional fuel assembly models in Serpent 4.3.3 Two-dimensional non-fuel assembly models in Serpent 4.3.4 Super homogenization method for non-multiplying media 4.3.5 Automated creation of Serpent models for parametrized cross section generation with KMACS 4.4 Core radial thermal expansion effect 4.4.1 Description of the core radial thermal expansion phenomenon 4.4.2 Coordinate transformation of the diffusion equation 4.4.3 Implementation of the coordinates transformation in PARCS 4.4.4 Adapted cross section parametrization scheme for the core radial expansion model 4.4.5 Diagrid model in ATHLET and temperature transfer 4.5 Core axial thermal expansion effect 4.5.1 Description of the core axial thermal expansion phenomenon 4.5.2 Implementation of a core axial thermal expansion model in PARCS 4.5.3 Appropriate cross section parametrization scheme 4.6 Control rod displacement due to reactor structures thermal expansion effects 4.6.1 Modelling scheme 4.6.2 Strongback model in ATHLET 4.6.3 Vessel model in ATHLET 4.6.4 Control rods drive lines ATHLET model 5 Verification on a case study 5.1 Description of the ASTRID reactor 5.2 Full core models 5.2.1 Full core Serpent reference models of the ASTRID core 5.2.2 Three-dimensional neutron kinetics model of ASTRID core in PARCS 5.2.3 Generation of appropriate few-group cross sections 5.2.4 Thermal-hydraulic model in ATHLET and ATHLET-PARCS feedback mapping 5.3 Verfications of the radial core expansion model 5.4 Assessment of the Doppler and sodium density effects 5.4.1 Assessment of the Doppler effect 5.4.2 Assessment of the sodium density effect 6 Coupled three-dimensional thermal-hydraulics/neutron-physics transient simulations with ATHLET-PARCS 6.1 Description of the models and transient simulations 6.2 Simulation 1: Doppler effect 6.2.1 Description 6.2.2 Results 6.3 Simulation 2: Sodium density effect 6.3.1 Description 6.3.2 Results 6.4 Simulation 3: Doppler and sodium density effects 6.4.1 Description 6.4.2 Results 6.5 Simulation 4: Core radial thermal expansion effect 6.5.1 Description 6.5.2 Results 6.6 Simulation 5: Doppler, Sodium density and core radial thermal expansion effects 6.6.1 Description 6.6.2 Results 6.7 Simulation 6: Core axial thermal expansion effect 6.7.1 Description 6.7.2 Results 6.8 Simulation 7: Doppler, Sodium density and core axial thermal expansion effects 6.8.1 Description 6.8.2 Results 6.9 Simulation 8: Doppler effect, Sodium density effect, core radial thermal expansion effect and core axial thermal expansion effect 6.9.1 Description 6.9.2 Results 6.10 Simulation 9: Doppler effect, Sodium density effect, core radial thermal expansion effect, core axial thermal expansion effect and control rod displacement due to reactor structures thermal expansion effect 6.10.1 Description 6.10.2 Results 6.11 Preliminary conclusions of the test calculations 7 Conclusion and outlook for future developments 7.1 Summary and conclusions 7.2 Suggestions for future work Appendices A The Boltzmann equation B Macro-group structure Bibliography

Afin de limiter les conséquences d'un accident grave avec la fusion du cœur dans un réacteur à neutrons rapides de génération IV refroidi au sodium, la recriticité doit être évitée au sein du mélange de combustible oxyde et de structures fondus, appelé corium. Pour cela, des matériaux absorbants, tels que le carbure de bore B4C, seront utilisés dans ou près du cœur, et des matériaux diluants dans le récupérateur de corium. L'objectif de ce travail est de présélectionner des matériaux parmi ces deux types de familles et de comprendre leur comportement au contact avec le corium. Concernant le B4C, des calculs thermodynamiques et des expériences ont permis de conclure à la formation de deux phases immiscibles dans le système UO2 - B4C à haute température, une oxyde et une borure, ainsi qu'à la volatilisation d'une partie de l'élément absorbant bore. Cette séparation de phases pourra réduire l'efficacité de l'absorption neutronique au sein de la phase oxyde. Une solution à ce comportement serait d'augmenter la quantité de B4C ou d'utiliser un absorbant oxyde miscible avec le combustible. Eu2O3 ou HfO2 pourraient convenir car il a été montré qu'ils forment une solution solide avec UO2. Concernant le matériau diluant, les oxydes mixtes Al2O3 - HfO2 et Al2O3 - Eu2O3 ont été étudiés. L'interaction de ces systèmes avec UO2 étant inconnue à ce jour, les premiers points ont été recherchés sur les diagrammes ternaires correspondants. Contrairement au système Al2O3 - Eu2O3 - UO2, le mélange Al2O3 - HfO2 - UO2 présente un seul eutectique et donc un seul chemin de solidification ce qui permet de prévoir plus facilement la manière dont le corium solidifierait dans le récupérateur.

Sweep Frequency Response (SFR) measurement on a transformer is a low voltage, offline exercise. So, it virtually permits determination of any network or system function, by imposing any desired terminal condition for the nontested windings and terminals. The terminal conditions employed have significant influence on the achievable fault detection ability, and maximizing this ability should obviously be one of the main aims of frequency response measurements. Simply stated, this requirement translates to the ability to identify/measure as many natural frequencies as possible. However, there is a practical limitation that not all system functions can exhibit all natural frequencies. Hence, it is necessary to determine the most appropriate combination of terminal connection and system function for achieving this objective. The growing popularity of SFR measurements has led to a new IEEE Guide. This document (IEEE Std PC57.149TM/D1) on SFR measurement lists out most of the possible terminal connections and system functions, for both 1φ and 3φ transformers. Surprisingly, it does not identify and recommend any one of them as preferred for maximizing this objective. Initially, considering the high frequency equivalent circuit representation of a 1φ, twowinding transformer, system function for different terminal conditions were computed. Depending on the number of natural frequencies distinguishable in the amplitude frequency response of a system function, each measuring condition was ranked. Thus, it led to identification of the best configuration. Later, these findings were verified on an actual 1φ, two-winding transformer. However, 3φ transformers are quite different in construction compared to 1φ transformers. So, whether the same configuration would also be applicable for SFR measurements on 3φ transformers had to be ascertained. So, the study was next extended to 3φ transformers. Performance of best configuration identified during this investigation are compared with currently employed low-voltage impulse test (used during short-circuit testing of transformers) and currently practiced SFR measurement test conditions, and found to be better. In conclusion, it is believed that after adequate field verifications, the identified configuration can be declared as the preferred way of making SFR measurement on transformers.

Power systems in many countries are stressed towards their stability limit. If these stable systems experience any unexpected contingencies, or disturbances, there is a significant risk of instability, which may lead to wide-spread blackout. Existing methods to minimize the risk of stability and excessive frequency decline; need to be redeveloped to address these new challenges. This research project will develop a new emergency control methodology, which can more effectively prevent power system frequency and voltage instability under emergency conditions. Frequency and voltage instability are two major concerns in power system operation. The primary aim of this project is to develop new optimal load shedding techniques, which are able to better address various voltage and frequency instability issues for power systems emergency control purpose. In this thesis, new approach of load shedding for frequency and voltage stability are presented. For the load shedding to prevent frequency collapse, System Frequency Respond – Under Frequency Load Shedding (SFR-UFLS) from the previous approach has been redeveloped to compute an optimal load shedding scheme. The limitation of previous optimal load shed method is that they only shed load following one particular contingency event. As an improvement of this method, we developed a technique that protects against a range of contingencies. For the load shedding to prevent voltage collapse, The proposed method is then tested on the 39-bus New England test system. Generators are of different importance to the system in terms of voltage stability. It is essential to investigate generators’ impact on system voltage stability. The theory of the normal forms of diffeomorphism is used to analyze the power flow equations, and then nonlinear active participation factor is obtained and is used to determine the influence of generators on voltage stability. By using this method, the nonlinearity of power systems can be taken into consideration while the computational efficiency is maintained. Therefore, the impact of generators can be measured with more accuracy even for the cases in which the system is characterized with strong nonlinearity. In order to show the validity of the proposed method, the IEEE 14-bus test system and the New England 39-bus power system are used as case studies. The steady-state voltage stability index verifies the proposed method. The results show that nonlinear active participation factor can describe the characteristics even when power systems are operating at a highly stressed condition.

Power systems in many countries are stressed towards their stability limit. If these stable systems experience any unexpected contingencies, or disturbances, there is a significant risk of instability, which may lead to wide-spread blackout. Existing methods to minimize the risk of stability and excessive frequency decline; need to be redeveloped to address these new challenges. This research project will develop a new emergency control methodology, which can more effectively prevent power system frequency and voltage instability under emergency conditions. Frequency and voltage instability are two major concerns in power system operation. The primary aim of this project is to develop new optimal load shedding techniques, which are able to better address various voltage and frequency instability issues for power systems emergency control purpose. In this thesis, new approach of load shedding for frequency and voltage stability are presented. For the load shedding to prevent frequency collapse, System Frequency Respond – Under Frequency Load Shedding (SFR-UFLS) from the previous approach has been redeveloped to compute an optimal load shedding scheme. The limitation of previous optimal load shed method is that they only shed load following one particular contingency event. As an improvement of this method, we developed a technique that protects against a range of contingencies. For the load shedding to prevent voltage collapse, The proposed method is then tested on the 39-bus New England test system. Generators are of different importance to the system in terms of voltage stability. It is essential to investigate generators’ impact on system voltage stability. The theory of the normal forms of diffeomorphism is used to analyze the power flow equations, and then nonlinear active participation factor is obtained and is used to determine the influence of generators on voltage stability. By using this method, the nonlinearity of power systems can be taken into consideration while the computational efficiency is maintained. Therefore, the impact of generators can be measured with more accuracy even for the cases in which the system is characterized with strong nonlinearity. In order to show the validity of the proposed method, the IEEE 14-bus test system and the New England 39-bus power system are used as case studies. The steady-state voltage stability index verifies the proposed method. The results show that nonlinear active participation factor can describe the characteristics even when power systems are operating at a highly stressed condition.