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Improved feedback control of MHD instabilities and errors fields in reversed-field pinch and tokamak
Full text linkThis Thesis presents a series of results on the development of advanced magnetic feedback schemes for the active control of magnetohydrodynamic (MHD) instabilities and error elds obtained in two magnetically conned toroidal experiments:
the RFX-mod reversed-eld pinch (RFP) in Padova, Italy, and the DIII-D tokamak at General Atomics, San Diego, CA, USA.
In the last years, these two devices have explored dierent types of highperformance regimes, also thanks to their sophisticated active control systems.
In RFX-mod, high-plasma current experiments, up to 2MA, were performed for the rst time in a RFP. These experiments allowed for the discovery of a new self-organized helical equilibrium with good connement properties [40]. Instead, in DIII-D, steady-state, high-performance tokamak operations are being explored.
The scientic programs of these experiments, in particular on error eld and MHD mode control, can give precious contributions to the International Thermonuclear Experimental Reactor (ITER) and to magnetic fusion research in general.
The RFP and the tokamak are toroidal devices for the magnetic connement of thermonuclear plasmas. An introduction to thermonuclear fusion, the main requests to exploit fusion as a future energy source, the magnetic connement of the plasma, and the MHD model which describes the plasma behavior in many cases of interest will be given in Chapter 1. The role of magnetic feedback control for the development of advanced operational regimes in RFX-mod and in DIII-D will be also discussed in this Chapter.
Chapter 2 describes the two experiments above mentioned and their magnetic feedback control systems. They are in fact equipped with very exible systems devoted to the control of MHD instabilities and error elds. In particular the feedback control strategies that are crucial for the work discussed in this Thesis will be presented.
The rst important result of this Thesis is reported in Chapter 3 and regards the optimization of multi-mode control of tearing instabilities in RFX-mod.
Tearing modes, which sustain the reversed-eld conguration typical of RFP experiments through a dynamo mechanism, can not be completely suppressed by magnetic feedback control. Nonetheless, it is important to reduce their edge radial magnetic eld amplitude to the lowest possible value, since it produces a deformation of the last closed ux surface, enhancing the plasma-wall interaction. In this work, the control of tearing modes has been optimized by using a non-linear model of the tearing mode dynamics in presence of the multiple-shell layout of RFX-mod and of the magnetic feedback system. This model of tearing modes has been implemented in a code named RFXlocking and previously described in [84]. Given the good match between the model predictions and the experimental mode behavior, the RFXlocking code has been used as a tool to identify a new set of mode control parameters (i.e. feedback gains), which allow to reduce the radial magnetic eld of multiple tearing modes at the plasma edge, maintaing at the same time
the modes into rotation and avoiding coil current saturations. The optimization approach consisted in simulating the mode dynamics varying the feedback gains and identifying the gain set, which fullls the requirements above described. Once the "model-based" gain set was identied, an extensive experimental campaign was performed on RFX-mod, obtaining satisfactory results in terms of edge radial magnetic eld reduction and also conrming the code predictions.
The magnetic feedback optimization performed during this Thesis work concerned not only tearing modes, but also the main magnetic eld errors present in RFX-mod. The presence of poloidal gaps in the RFX-mod wall modies the pattern of eddy-currents induced in it by the vertical magnetic eld during the plasma current ramp-up, thus forming toroidally-localized error elds to which tearing modes are phase-locked. Two advanced feedback control strategies have been applied to correct these error elds: a multi-mode control scheme and a
dynamic decoupling scheme.
Regarding the rst feedback control strategy, a Simulink model of the RFXmod magnetic feedback system has been used to identify the feedback gains, which allow a signicant reduction of the error eld amplitude, avoiding coil current saturations.
A dynamic decoupler has also been used to compute oine the feedback currents needed to cancel the error elds. As will be described in Chapter 4, these two techniques have been tested during a dedicated experimental campaign. The best result in terms of error eld reduction has been obtained when both multimode control and the decoupler are used. With error eld correction during the plasma current ramp-up, the phase-locking among tearing modes is no more localized near the poloidal gaps of the wall, thus reducing the plasma-wall interaction at these positions.
As mentioned above, the high-current RFX-mod experiments have disclosed a promising physics regime, where the RFP spontaneously evolves towards an Ohmic helical equilibrium. This new magnetic equilibrium is characterized by a single helical magnetic axis and helical magnetic surfaces in the plasma core. This leads to a signicant decrease in the stochastic transport and to the formation of core electron temperature barriers. During the last experimental campaign, it has been demonstrated that a (1;-7) helical equilibrium can be sustained and controlled by applying helical boundary conditions at the plasma edge through magnetic feedback. In this Thesis work, Chapter 5 and Chapter 6 deal with the optimization of the helical boundary conditions used to control the helical equilibrium. The optimization procedure uses control strategies analogous to those described above and adopted to improve the control of tearing modes and error elds. The RFXlocking code has been modied by adding the possibility to apply a helical boundary
conditions. In this way, the mode dynamics has been simulated with this new helical boundary, by varying the feedback gains and the amplitude and phase of the helical magnetic eld perturbations imposed at the plasma edge. A model-based
optimization approach similar to the one described in Chapter 3 has been adopted here to identify the feedback gains that allow to produce the requested radial eld pattern at the edge with the lowest possible coil current. A partial gain scan has been performed in the experiment and the results conrm the model predictions.
The main outcomes of the model-based optimization and an analysis of the effects on the plasma performance of the applied helical boundary conditions are described in Chapter 5.
Vacuum eld analyses described in Chapter 6 reveal that, when rotating magnetic eld perturbations are applied through magnetic feedback, as in the case of the helical equilibria above described, error elds are induced by the frequency response of the wall to external magnetic elds varying in time. These error elds, that are mainly introduced by the presence of the toroidal and poloidal gaps in the wall structure, may somehow aect the good connement properties of the helical equilibrium. For this reason, a dynamic decoupler similar to the one used to correct the error elds in the current ramp-up phase of the plasma discharges has been applied. Encouraging results in terms of error eld reduction are obtained.
The frequency-response of the wall to any external time-varying magnetic eld has been investigated also in the DIII-D tokamak, in the framework of a collaboration between the RFX-mod and DIII-D teams. In the DIII-D control algorithm, the magnetic feedback measurements are usually real-time compensated for spurious magnetic elds, due for instance to the feedback and axi-symmentric coils.
These contributions are calculated from the zero-frequency coupling coecients between each actuator and sensor. In this way the eects of eddy-currents induced in the wall are neglected. The relevance of these frequency-dependent, or
AC eects, on RWM and error eld control has been evaluated by analyzing past experiments. The analyses suggested that, if the wall frequency response is not taken into account in the feedback compensation scheme, error elds can be introduced
when doing magnetic feedback. These can be important especially at high β, where uncorrected error elds can be strongly amplied by the plasma.
An AC compensation algorithm has been implemented and tested in real-time in dry-shots and Ohmic plasmas. More tests of this algorithm at high β have been proposed for the next experimental campaign to assess its relevance on plasma
performance in scenarios where the plasma is less resilient to error elds. The main outcomes of this Thesis work is reported in Chapter 7.
Chapter 8 summarizes the main conclusions of this work and describes a series of experiments that could be made both in RFX-mod and DIII-D in the near future, to further develop the studies started with this Thesis.Questa Tesi presenta una serie di risultati sullo sviluppo di avanzati schemi di feedback per il controllo di instabilità magnetoidrodinamiche (MHD) e di campi errori ottenuti in due esperimenti toroidali a connamento magnetico: il reversed-eld pinch (RFP) RFX-mod, a Padova, Italia, e il tokamak DIII-D, presso la General Atomics, San Diego, CA, USA.
Negli ultimi anni, questi due esperimenti hanno esplorato differenti tipi di regimi ad alte prestazioni, anche grazie ai loro sofisticati sistemi di controllo attivo.
Ad RFX-mod, esperimenti ad alta corrente di plasma, sino a 2MA, sono stati realizzati per la prima volta in un RFP. Questi esperimenti hanno permesso la scoperta di un nuovo equilibrio elicoidale, auto-organizzato e con buone proprietà di connamento [40]. Invece, a DIII-D scenari stazionari ed operazioni ad alte prestazioni vengono investigati. I programmi scientifici di questi esperimenti, in particolar modo il controllo di campi errori e di instabilità MHD, possono dare preziosi contributi all'International Thermonuclear Experimental Reactor (ITER) e, più in generale, alla ricerca nel campo della fusione.
Il RFP e il tokamak sono esperimenti toroidali per il connamento magnetico di plasmi termonucleari. Un'introduzione alla fusione termonucleare, i principali requisiti per sfruttare la fusione come una sorgente di energia per il futuro, il connamento magnetico del plasma, e il modello MHD che descrive il comportamento
del plasma in molti casi di interesse, verranno descritti nel Capitolo 1. Il ruolo del controllo dei campi magnetici in feedback per lo sviluppo di regimi operazionali avanzati in RFX-mod e a DIII-D verrà anche discusso in questo Capitolo.
Il Capitolo 2 descrive gli esperimenti sopra citati e i relativi sistemi di controllo dei campi magnetici in feedback. Questi infatti sono muniti di sistemi molto essibili per il controllo di instabilita MHD e di campi errori. In particolar modo, verranno presentate le strategie di controllo in feedback che sono cruciali per il lavoro discusso in questa Tesi.
Il primo risultato importante di questa Tesi è riportato nel Capitolo 3 e riguarda l'ottimizzazione del controllore a multi-modo delle instabilita tearing di RFX-mod.
I modi tearing, che sostengono la congurazione a campo rovesciato tipica degli esperimenti RFP attraverso un meccanismo di dinamo, non possono essere completamente
soppressi dal controllo dei campi magnetici in feeedback. Ciò nonostante, è importante ridurre la loro ampiezza di campo magnetico radiale al bordo del plasma al più piccolo valore possibile, dal momento che questa produce una deformazione
dell'ultima superficie chiusa di flusso, aumentando l'interazione plasmaparete.
In questo lavoro, il controllo dei modi tearing è stato ottimizzato utilizzando un modello non lineare che simula la dinamica dei modi tearing in presenza del layout a multipla shell di RFX-mod e del sistema per il controllo di campi magnetici in feedback. Questo modello dei modi tearing è stato sviluppato in un codice chiamato RFXlocking che è stato descritto in [84]. Dato il buon accordo tra le predizioni del modello e il comportamento dei modi nell'esperimento, il codice RFXlocking è stato usato per identicare un nuovo set di parametri di controllo (i.e. i guadagni del feedback), che permetta di ridurre l'ampiezza radiale dei modi tearing al bordo del plasma, mantenendo allo stesso tempo i modi in rotazione ed evitando saturazioni di corrente nelle bobine di controllo. L'approccio per l'ottimizzazione prevede di simulare la dinamica dei modi tearing al variare dei guadagni del feedback, e di identicare un set di guadagni che soddisfa le richieste sopra descritte. Una volta trovato il set di guadagni ispirato dal modello, una lunga campagna sperimentale è stata fatta ottenendo soddisfacenti risultati in termine di riduzione di campo magnetico radiale al bordo del plasma e confermando
così le previsioni del codice.
L'ottimizzazione del controllo dei campi magnetici in feedback svolta in questo lavoro di Tesi non ha solo riguardato i modi tearing ma anche i principali campi errori in RFX-mod. La presenza di tagli in direzione poloidale nella struttura conduttiva di RFX-mod modica il pattern delle correnti immagine indotte dal
campo magnetico verticale durante la fase di salita di corrente di plasma, inducendo campi errori localizzati dove i modi tearing hanno un'inteferenza costruttiva. Due tecniche di controllo in feedback sono state utilizzate per sopprimere questi campi
errori: uno schema di controllo a multi modo e il disaccoppiatore dinamico.
Per quanto riguarda la prima strategia di controllo, un modello di Simulink del sistema magnetico di feedback di RFX-mod è stato usato per identicare i guadagni del feedback, che permettono di ridurre signicativamente l'ampiezza del campo errore, evitando saturazioni di corrente nelle bobine di controllo. Un disaccopiatore dinamico è stato usato per calcolare le correnti nelle bobine di controllo necessarie a cancellare i campi errore. Come verrà descritto nel Capitolo 4, durante una campagna sperimentale dedicata, queste due tecniche sono state testate.
Il miglior risultato in termini di riduzione dei campi errore è stato ottenuto quando il controllore a multi-modo e il disaccopiatore sono stati usati contemporaneamente.
Quando la correzione del campi errori è applicata durante la fase
di salita di corrente di plasma, l'interferenza costruttiva tra i modi tearing non è piu localizzata vicino ai tagli in direzione poloidale della struttura conduttrice, in questo modo viene ridotta l'interazione plasma-parete in queste zone.
Come accennato precedentemente, gli esperimenti ad alta corrente in RFX-mod hanno rivelato un nuovo regime promettente, in cui l'RFP evolve spontaneamente in uno stato elicoidale Ohmico. Questo nuovo equilibrio magnetico è caratterizzato da un singolo asse magnetico elicoidale e da superci magnetiche elicoidali all'interno del plasma. Questo produce una diminuzione del trasporto stocastico e la formazione di proli di temperatura elettronica molto ripidi. Nell'ultima campagnia sperimentale è stato dimostrato che un equilibrio elicoidale con elicità (1,-7) può essere indotto e controllato applicando condizioni elicoidali al bordo del plasma per mezzo del sistema magnetico di feedback.
In questo lavoro di Tesi, il Capitolo 5 e il Capitolo 6 descrivono l'ottimizzazione delle perturbazioni magnetiche elicoidali usate per controllare l'equilibrio elicoidale.
La procedura di ottimizzazione usa le stesse strategie di controllo che sono state adottate per migliorare il controllo dei modi tearing e dei campi errori. Il codice RFXlocking è stato modificato permettendo di applicare condizioni elicoidali al bordo del plasma. In questo modo, la dinamica dei modi può essere simulata con questo nuovo boundary elicoidale, al variare dei guadagni di feedback, dell'ampiezza e della fase delle perturbazioni elicoidali imposte al bordo del plasma. Un approccio
di ottimizzazione ispirato dal modello, simile a quello descritto nel Capitolo 3, è stato adottato in questo caso per identicare i guadagni di feedback che permettono di produrre il richiesto pattern di campo radiale al bordo del plasma con la minor richiesta di corrente nelle bobine di controllo. Uno scan parziale dei guadagni è stato svolto nell'esperimento e i risultati confermano le predizioni del modello. I risultati salienti dell'ottimizzazione ispirata dal modello e le prestazioni del plasma negli stati elicoidali sostenuti imponendo condizioni elicoidali al bordo plasma sono descritti nel Capitolo 5.
Analisi di esperimenti a vuoto, descritte nel Capitolo 6, rivelano che, quando una perturbazione rotante di campo magnetico viene applicata dal sistema di feedback, come nel caso degli stati elicoidali descritti sopra, campi errori vengono indotti dalla risposta in frequenza della struttura conduttrice ai campi magnetici esterni che variano nel tempo. Questi campi errori, che sono indotti principalmente dalla presenza di tagli in direzione toroidale e poloidale nella struttura conduttrice, possono in qualche modo influenzare le proprietà di buon connamento degli stati elicoidali. Per questo motivo un disaccoppiatore dinamico, simile a quello usato per correggere i campi errori durante la fase di salita delle corrente, è stato utilizzato.
Esperimenti a vuoto mostrano risultati incoraggianti in termini di riduzione dell'ampiezza del campo errore.
La risposta in frequenza della struttura conduttiva ad un campo magnetico variabile nel tempo è stata esaminata anche nell'esperimento DIII-D, durante una collaborazione tra i gruppi di ricerca di RFX-mod e di DIII-D. Nell'algoritmo di controllo di DIII-D, le misure di campo magnetico sono compensate in tempo reale dai campi magnetici spuri, che possono essere indotti dalle bobine di controllo o dalle bobine assialsimmetriche. Questi contributi esterni sono calcolati dai coefficienti di accoppiamento tra ciascun attuatore e sensore, a frequenza nulla. In questo approccio, gli effetti delle correnti immagine indotte nella struttura conduttiva vengono trascurati. L'importanza di questi effetti che dipendono dalla frequenza, o effetti AC, per il controllo di RWM e di campi errori è stata valutata analizzando esperimenti precedenti. Le analisi suggeriscono che campi errori possono essere indotti quando viene applicato il controllo in feedback se la risposta in frequenza delle strutture conduttive non è inclusa nell'algoritmo di feedback di compensazione. Questi possono risultare importanti specialmente ad alto β regime in cui i campi errori residui possono essere amplicati dal plasma. Un algoritmo di compensazione AC e stato implementato e testato in tempo reale in spari a vuoto e in spari Ohmici. Un maggior numero di test di questo algoritmo ad alto β è stato proposto per la prossima campagna sperimentale per testare la sua rilevanza nella performance del plasma quando questo è più soggetto ai campi errori. I risultati più salienti di questo lavoro di Tesi sono discussi nel Capitolo 7.
Il Capitolo 8 riassume le conclusioni principali di questo lavoro e presenta degli esperimenti che posso essere eseguiti ad RFX-mod e a DIII-D nel futuro prossimo che possono indagare ulteriormente gli studi iniziati in questo lavoro di Tesi
Going Beyond Counting First Authors in Author Co-citation Analysis
Full text linkThe present study examines one of the fundamental aspects of author co-citation analysis (ACA) - the way co-citation
counts are defined. Co-citation counting provides the data on which all subsequent statistical analyses and mappings
are based, and we compare ACA results based on two different types of co-citation counting - the traditional type that
only counts the first one among a cited work's authors on the one hand and a non-traditional type that takes into
account the first 5 authors of a cited work on the other hand. Results indicate that the picture produced through this non-traditional author co-citation counting contains more coherent author groups and is therefore considerably clearer. However, this picture represents fewer specialties in the research field being studied than that produced through the traditional first-author co-citation counting when the same number of top-ranked authors is selected and analyzed. Reasons for these effects are discussed
Variations on the Author
Full text link“Variations on the Author” discusses two of Eduardo Coutinho’s recent films (Um Dia na Vida, from 2010, and Últimas Conversas, posthumously released in 2015) and their contribution to the general question of documentary authorship. The director’s filmography is characterized by a consistent yet self-effacing form of authorial self-inscription: Coutinho often features as an interviewer that rather than express opinions propels discourses; an interviewer that is good at listening. This mode of self-inscription characterizes him as an author who is not expressive but who is nonetheless markedly present on the screen. In Um Dia na Vida, however, Coutinho is completely absent form the image, while Últimas Conversas, on the contrary, includes a confessional prologue that moves the director from the margins to the center of his films. This article examines the ways in which these works stand out in the filmography of a director who offers new insights into the notion of cinematic authorship
An active feedback recovery technique from disruption events induced by m= 2, n=1 tearing modes in ohmically heated tokamak plasmas
No full textAppropriate Similarity Measures for Author Cocitation Analysis
Full text linkWe provide a number of new insights into the methodological discussion about author cocitation analysis. We first argue that the use of the Pearson correlation for measuring the similarity between authors’ cocitation profiles is not very satisfactory. We then discuss what kind of similarity measures may be used as an alternative to the Pearson correlation. We consider three similarity measures in particular. One is the well-known cosine. The other two similarity measures have not been used before in the bibliometric literature. Finally, we show by means of an example that our findings have a high practical relevance.information science;Pearson correlation;cosine;similarity measure;author cocitation analysis
Real-time modelling of DTT plasma scenarios using the RAPTOR transport simulator code
Full text linkopenControlled nuclear fusion is a promising solution to our current energy crisis that aims to provide clean energy without greenhouse gas emissions or long lasting highly radioactive waste production. Several projects are currently in development to obtain the knowledge necessary to build a commercially viable fusion power plant in the next decades, first and foremost the ITER reactor in construction in Cadarache (France) that will in turn pave the way for the following DEMO reactor, the first hopefully capable of net production of electrical energy. Despite the ambitious aims of ITER, it will not be able to fully explore several key physical and engineering aspects required for the subsequent generation of reactors.
Among these of great importance is the design of their divertors: these devices are able to open the flux surfaces of the containment magnetic field in the reactor and by doing so redirect the flux of particles coming from the plasma toward suitable targets. Among the several advantages of this configuration is that the interaction between the plasma and the chamber wall is kept in a region separated from the core plasma, greatly reducing the influx of impurities, and that the power exhaust is directed toward a specific section of the wall that can thus be the only one designed to sustain extremely high energy fluxes. Despite their importance, ITER will not implement a divertor design suitable for the later DEMO and this brought forth the necessity for a dedicated experiment to study the viability of various divertor configurations in more demanding conditions. This experiment will be DTT (Divertor Tokamak Test), currently under construction at the ENEA Research Center in Frascati (Italy).
To control the plasma dynamics, DTT will make use of an integrated approach that requires a faster-than-real-time physical modeling of the plasma, of which one key component is the simulation of transport phenomena and the corresponding radially dependent physical profiles of the plasma. This will be performed by a 1-dimensional simulator code optimized for control problems and rapid iteration called RAPTOR (RApid Plasma Transport simulatOR). The aim of this thesis is to validate RAPTOR in this application by simulating different plasma scenarios of DTT, following a full time evolution and with both low magnetic field and low externally injected power, as is expected to operate during its first years after commissioning, and with a full suite of external heating systems and higher fields that will be needed for the full scale test of DEMO-like divertor conditions. To improve the results and check their validity, our simulations have been confronted with those obtained with other codes, more specifically METIS, another fast transport simulation code, and ASTRA, a much more complex and computationally demanding simulator. RAPTOR also provides modeling of sawtooth instabilities, an important phenomena that results in periodic crashes of the core temperature of the plasma, caused by topological effects that trigger sudden changes of resistivity in the plasma. Characterizing these sawteeth and how their onset and period are affected by changes in the external heating is useful for controlling them and avoiding negative effects linked to the onset of other types of more problematic instabilities. A double sawtooth sweeping experiment, involving the variation of the radial deposition depth of the external sweeping system while monitoring the sawteeth's period, has been successfully simulated using RAPTOR, showing the robustness of its modelling of this kind of instability.
RAPTOR has thus being tested in different ways in its ability to simulate scenarios that more closely model the actual operation conditions of DTT, paving the way for its future use both for integrated control of the experiment and for obtaining quick preliminary modeling results without using more time consuming codes.Controlled nuclear fusion is a promising solution to our current energy crisis that aims to provide clean energy without greenhouse gas emissions or long lasting highly radioactive waste production. Several projects are currently in development to obtain the knowledge necessary to build a commercially viable fusion power plant in the next decades, first and foremost the ITER reactor in construction in Cadarache (France) that will in turn pave the way for the following DEMO reactor, the first hopefully capable of net production of electrical energy. Despite the ambitious aims of ITER, it will not be able to fully explore several key physical and engineering aspects required for the subsequent generation of reactors.
Among these of great importance is the design of their divertors: these devices are able to open the flux surfaces of the containment magnetic field in the reactor and by doing so redirect the flux of particles coming from the plasma toward suitable targets. Among the several advantages of this configuration is that the interaction between the plasma and the chamber wall is kept in a region separated from the core plasma, greatly reducing the influx of impurities, and that the power exhaust is directed toward a specific section of the wall that can thus be the only one designed to sustain extremely high energy fluxes. Despite their importance, ITER will not implement a divertor design suitable for the later DEMO and this brought forth the necessity for a dedicated experiment to study the viability of various divertor configurations in more demanding conditions. This experiment will be DTT (Divertor Tokamak Test), currently under construction at the ENEA Research Center in Frascati (Italy).
To control the plasma dynamics, DTT will make use of an integrated approach that requires a faster-than-real-time physical modeling of the plasma, of which one key component is the simulation of transport phenomena and the corresponding radially dependent physical profiles of the plasma. This will be performed by a 1-dimensional simulator code optimized for control problems and rapid iteration called RAPTOR (RApid Plasma Transport simulatOR). The aim of this thesis is to validate RAPTOR in this application by simulating different plasma scenarios of DTT, following a full time evolution and with both low magnetic field and low externally injected power, as is expected to operate during its first years after commissioning, and with a full suite of external heating systems and higher fields that will be needed for the full scale test of DEMO-like divertor conditions. To improve the results and check their validity, our simulations have been confronted with those obtained with other codes, more specifically METIS, another fast transport simulation code, and ASTRA, a much more complex and computationally demanding simulator. RAPTOR also provides modeling of sawtooth instabilities, an important phenomena that results in periodic crashes of the core temperature of the plasma, caused by topological effects that trigger sudden changes of resistivity in the plasma. Characterizing these sawteeth and how their onset and period are affected by changes in the external heating is useful for controlling them and avoiding negative effects linked to the onset of other types of more problematic instabilities. A double sawtooth sweeping experiment, involving the variation of the radial deposition depth of the external sweeping system while monitoring the sawteeth's period, has been successfully simulated using RAPTOR, showing the robustness of its modelling of this kind of instability.
RAPTOR has thus being tested in different ways in its ability to simulate scenarios that more closely model the actual operation conditions of DTT, paving the way for its future use both for integrated control of the experiment and for obtaining quick preliminary modeling results without using more time consuming codes
Machine Learning for the interpretation of the main ion charge exchange recombination spectra at the ASDEX Upgrade tokamak
No full textnon
Magnetohydrodynamic stability analysis of the pedestals of ASDEX Upgrade plasmas
Full text linkopenThe high confinement mode (H-mode) dramatically improves the confinement properties of present tokamak plasmas and is therefore the scenario envisioned for future fusion reactors.
The main characteristic of this scenario is the formation of a pedestal, a zone of steep temperature and density gradients, at the edge of the plasma, by means of a transport barrier. The height of the pedestal is limited by the onset of edge localised modes (ELMs), quasi-periodic explosive instabilities at the plasma edge which expel particles and energy on millisecond time-scales.
While ELMs in present day machines pose no danger, when scaled to a fusion reactor device they are predicted to cause significant damage to the machine components. As such, the understanding and exploitation of alternative regimes with high confinement, but without ELMs, is of significant interest.
The onset of an ELM can be described by magnetohydrodynamic (MHD) stability codes.
The aim and project of the current thesis carried out at Max-Planck-Institut für Plasmaphysik (IPP) in Garching (Germany), involves the automation of a workflow which runs codes to test the pedestal MHD stability, such as MISHKA, starting from a standardised set of experimental information. In addition, the HELENA code is employed as a high resolution equilibrium solver through the calculation of the Grad-Shafranov equation for a toroidal axisymmetric plasma.
Once the workflow is implemented, it is applied to a database of experimental data from the ASDEX Upgrade tokamak to study the properties of the pedestal through stability diagrams. It is particularly important to provide an estimate of the distance to the MHD stability boundary in the various ELM-free regimes to understand how robust these regimes are and the margin a given regime has before a large ELM is triggered.
Various deuterium and helium plasma discharges are studied in this regard.The high confinement mode (H-mode) dramatically improves the confinement properties of present tokamak plasmas and is therefore the scenario envisioned for future fusion reactors.
The main characteristic of this scenario is the formation of a pedestal, a zone of steep temperature and density gradients, at the edge of the plasma, by means of a transport barrier. The height of the pedestal is limited by the onset of edge localised modes (ELMs), quasi-periodic explosive instabilities at the plasma edge which expel particles and energy on millisecond time-scales.
While ELMs in present day machines pose no danger, when scaled to a fusion reactor device they are predicted to cause significant damage to the machine components. As such, the understanding and exploitation of alternative regimes with high confinement, but without ELMs, is of significant interest.
The onset of an ELM can be described by magnetohydrodynamic (MHD) stability codes.
The aim and project of the current thesis carried out at Max-Planck-Institut für Plasmaphysik (IPP) in Garching (Germany), involves the automation of a workflow which runs codes to test the pedestal MHD stability, such as MISHKA, starting from a standardised set of experimental information. In addition, the HELENA code is employed as a high resolution equilibrium solver through the calculation of the Grad-Shafranov equation for a toroidal axisymmetric plasma.
Once the workflow is implemented, it is applied to a database of experimental data from the ASDEX Upgrade tokamak to study the properties of the pedestal through stability diagrams. It is particularly important to provide an estimate of the distance to the MHD stability boundary in the various ELM-free regimes to understand how robust these regimes are and the margin a given regime has before a large ELM is triggered.
Various deuterium and helium plasma discharges are studied in this regard
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