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Use of nitrogen compounds for tritium retention and tungsten sputtering control in nuclear fusion reactors

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1 INTRODUCTION 7

1.1 FUSION ENERGY . . . 7

1.1.1 Nuclear fusion reactors . . . 8

1.1.2 Magnetic connement . . . 9

1.1.2.1 Reactor geometry: tokamak and stellarator . . . 9

1.1.2.2 Controlling the plasma shape by solid surfaces . . . 10

1.1.3 Plasma Material Interaction . . . 11

1.1.3.1 Stationary heat and particle loads . . . 12

1.1.3.2 Edge Localized Modes (ELMs) . . . 12

1.1.3.3 O-normal events: disruptions . . . 13

1.1.3.4 Plasma facing materials desired properties . . . 14

1.1.4 Future projects design: ITER and beyond . . . 14

1.2 MATERIAL DAMAGE IN A NUCLEAR FUSION REACTOR . . . 14

1.2.1 Erosion by physical sputtering . . . 15

1.2.2 Erosion by chemical sputtering . . . 16

1.2.2.1 Hydrogen chemical sputtering of carbon materials . . . 17

1.2.2.2 Chemical sputtering of carbon materials by other reactive species . . . 18

1.2.2.3 Total yield . . . 18

1.2.3 Melting and evaporation . . . 19

1.2.4 Neutron irradiation . . . 20

1.2.5 Other damage . . . 20

1.3 PLASMA CONTAMINATION CONTROL . . . 21

1.3.1 Impurities contamination of plasma core . . . 22

1.3.2 Radiative cooling at the plasma edge by impurity seeding . . . 23

1.4 TRITIUM RETENTION CONTROL . . . 23

1.4.1 Bulk retention: implantation and transmutation . . . 24

1.4.2 Codeposition . . . 24

1.4.2.1 Direct codeposition: beryllium . . . 24

1.4.2.2 Indirect codeposition by gaseous molecules: carbon . . . 25

1.4.3 Tritium recovery . . . 26

1.4.3.1 Codeposit inhibition by scavengers injection . . . 26

1.4.3.2 Cold, low pressure reactive plasma erosion . . . 27

1.4.3.3 Baking and thermo-oxidation of codeposits . . . 27

1.4.3.4 Laser removal . . . 28

1.4.3.5 Local plasma generation . . . 28

1.4.3.6 Other tritium removal techniques . . . 29

1.4.3.7 Treatment integration: Good housekeeping . . . 29

1.4.3.8 Real, complex reactor: mixed materials, divertor coating and long term out-gassing . . . 29

1.5 FIRST WALL MATERIALS . . . 31

1.5.1 Carbon . . . 31

1.5.2 Tungsten . . . 32

1.5.2.1 Tungsten nitrides . . . 33

1.5.3 Beryllium . . . 33

1.5.4 Boron . . . 34

1.5.5 Liquid metals: Lithium . . . 34

1.6 OBJECTIVES OF THIS THESIS FOR ITER MATERIALS . . . 35

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2 CARBON CODEPOSITS FORMATION 37

2.1 DIRECT DEPOSITION IN TJ-II . . . 37

2.1.1 Motivation . . . 38

2.1.1.1 Redeposition . . . 38

2.1.1.2 Chemical sputtering yield calculation . . . 39

2.1.2 Experimental . . . 40

2.1.2.1 TJ-II stellarator . . . 40

2.1.2.2 Graphite bar probe experiments . . . 41

2.1.3 Results . . . 43

2.1.3.1 Estimated chemical sputtering and CH emission . . . 43

2.1.3.2 Recovered lms analysis . . . 44

2.1.4 Discussion . . . 45

2.1.5 Summary and future work . . . 46

2.2 CODEPOSITION INHIBITION BY SCAVENGER . . . 47

2.2.1 Motivation . . . 48

2.2.2 Experimental . . . 50

2.2.2.1 Setup . . . 50

2.2.2.2 Experiment phases . . . 51

2.2.2.3 Mass spectra interpretation . . . 51

2.2.3 Results . . . 52

2.2.3.1 Eect of reactor walls . . . 56

2.2.3.2 Eect of sampling arrangement . . . 56

2.2.3.3 Eect of oxygen contamination . . . 59

2.2.4 Discussion . . . 60

2.2.4.1 Reactor wall eects . . . 60

2.2.4.2 Reactor wall and sampling arrangement eects on radicals stability . . . 64

2.2.4.3 Oxygen related eects . . . 65

2.2.5 Summary and future work . . . 66

3 CARBON CODEPOSITS REMOVAL 68 3.1 COLD PLASMA . . . 68

3.1.1 Motivation . . . 69

3.1.2 Experimental . . . 70

3.1.2.1 Castellation gap simulation . . . 70

3.1.2.2 Radical erosion in DC-plasmas by positive biasing . . . 70

3.1.2.3 Radical erosion in RF and MW plasmas . . . 71

3.1.2.4 Laser interferometry2-3 . . . 71

3.1.3 Results and discussion . . . 72

3.1.3.1 Castellation gap simulation . . . 72

3.1.3.2 Radical erosion in DC-plasmas by positive biasing . . . 73

3.1.3.3 Radical erosion in RF and MW plasmas . . . 74

3.1.4 Summary and future work . . . 76

3.2 THERMO-OXIDATION . . . 77

3.2.1 Motivation . . . 77

3.2.2 Experimental . . . 78

3.2.2.1 Samples origin . . . 78

3.2.2.2 Sample treatment . . . 80

3.2.2.3 Sample and thermo-oxidation products analysis . . . 82

3.2.3 Results and discussion . . . 83

3.2.3.1 Erosion of tokamak samples . . . 83

3.2.3.2 Erosion of laboratory samples . . . 84

3.2.3.3 Gas products . . . 88

3.2.4 Summary and future work . . . 92

3.3 LASER ABLATION . . . 93

3.3.1 Motivation . . . 94

3.3.2 Experimental . . . 96

3.3.2.1 Plasma plume and dust ejection in gas . . . 97

3.3.2.2 Dust collection in aerogel . . . 98

3.3.3 Results and discussion . . . 98

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3.3.4 Summary and future work . . . 108

3.4 ATMOSPHERIC PLASMA TORCH . . . 109

3.4.1 Motivation . . . 110

3.4.2 Experimental . . . 112

3.4.3 Results and discussion . . . 112

3.4.3.1 Graphite erosion . . . 112

3.4.3.2 W/a-C:H . . . 113

3.4.4 Summary and future work . . . 114

3.5 INTEGRATED SCENARIO . . . 114

3.5.1 Motivation . . . 115

3.5.2 Summary of techniques for tritium control . . . 115

3.5.3 Good housekeeping . . . 115

3.5.4 Application to ITER . . . 121

4 TUNGSTEN NITRIDES 123 4.1 TUNGSTEN NITRIDES COATING . . . 124

4.1.1 Motivation . . . 124

4.1.2 Experimental . . . 125

4.1.3 Sample characterization . . . 126

4.1.3.1 Characterization of reactive magnetron sputtering (RMS) samples. . . 126

4.1.3.2 Characterization of sequentially deposited and nitrided (SDN) samples. . . . 128

4.1.4 Discussion . . . 130

4.1.5 Summary and future work . . . 132

4.2 TUNGSTEN NITRIDES EROSION BY PLASMA AND FUEL RETENTION . . . 133

4.2.1 Motivation . . . 134

4.2.1.1 Tungsten sputtering . . . 134

4.2.1.2 Hydrogen isotope retention and blistering . . . 136

4.2.2 Experimental . . . 137

4.2.2.1 Low ux: PACVD . . . 138

4.2.2.2 Medium ux: Nano-PSI . . . 138

4.2.2.3 High, reactor-relevant ux: Pilot-PSI . . . 138

4.2.3 Plasma exposure . . . 139

4.2.3.1 Low ux . . . 139

4.2.3.2 Medium ux . . . 139

4.2.3.3 Reactor-relevant uxes . . . 141

4.2.4 Discussion . . . 142

4.2.5 Summary and future work . . . 143

5 SUMMARY 145

6 RESÚMEN 151

7 Glossary, abbreviations and list of Figures and Tables 160

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(AGRADECIMIENTOS)

Esta tesis ha sido extraordinariamente larga, y como consecuencia los agradecimientos se alargarán también. He preferido hacer unos agradecimientos menos formales para poder llegar a todas las personas que me han ayudado todo este tiempo, y también para contar un poco de mi historia, para recordarla cada vez que vuelva a mirar mi tesis. Todos los que empezamos en el mundo de la investigación sabemos que es una carrera de obstáculos, y mi caso ha sido especialmente así, no tanto obstáculos cientícos, sino más bien de índole administrativa. Sin embargo, nunca pensé en abandonar, siempre he tenido claro que quiero ser un cientíco; todos los obstáculos y golpes sólo me animaban a levantarme más fuerte. Por eso quiero dar las gracias a los que han hecho posible que pueda leer esta tesis un poco antes de lo que con una interpretación rígida e inexible de las normas otros exigían: Jose María Iriondo y Rafael García, muchas gracias a los dos por todos vuestros esfuerzos. También quiero aprovechar para agradecer a Javier Sanz por su inestimable ayuda para poder realizar mi estancia postdoctoral en San Diego, ½gracias! Por otro lado, haber viajado tanto me ha permitido hacer muchos buenos amigos en muchas partes del mundo, a lo que también ayuda que el mundo de la fusión nuclear sea pequeño, con mucho contacto entre sus cientícos, y con muchas ayudas a los doctorandos como yo. ½Qué grandes recuerdos del curso de verano Carolus Magnus! ½Cuantos buenos amigos hice allí! (además de unos cuantos contactos que originaron varias colaboraciones). Por todo ello mezclaré varios idiomas en estos agradecimientos, para que los entienda la persona a la que me dirijo.

Lo primero es dar alas gracias a mi director de tesis, Paco Tabares. Él ha sido mi guía todos estos años, dándome una gran libertad para investigar pero siempre cuidándome para que no me desviase del tema central, algo que no siempre conseguía (muy a su pesar). Paco es del tipo de cientícos que siempre está haciendo de todo, sin parar de trabajar, sin parar de viajar, sin parar de pensar en ciencia y en nuevos experimentos (cientícos y musicales con botellines de Mahou), siempre investigando en múltiples áreas, con múltiples colaboraciones con otros grupos y dirigiendo a mucha gente. Sin embargo, siempre conseguía sacar un rato para mí, para ayudarme en el laboratorio cuando era necesario, o para analizar conmigo los resultados, que casi siempre eran los contrarios a los que esperábamos (o más bien a los que deseábamos). Él también ha sido el responsable de mi gran cantidad de viajes y estancias, de mis colaboraciones con otros grupos. Básicamente, he seguido sus pasos, el camino que me ha marcado. También quiero agradecer al resto del grupo de plasma pared su ayuda: a Miguel por la ayuda para el montaje de equipos de vacío y demás temas mecánicos; a David por su ayuda con la espectroscopía y por la tarjeta del comedor; a Alfonso porque sabrá seguir mi trabajo y mejorarlo, como ya está haciendo; a Eider por corregir tantas partes de mi tesis; y a Ana por su amistad y por esos viajes en coche al trabajo desde Alcorcón. Aquí tampoco me quiero olvidar de alguien que me ayudó muchísimo los primeros años: Jose Ferreira. Parte de mi tesis es continuación de la suya, con él también aprendí muchísimo, sobre todo a no cerrarme en mi campo, a aprender de más campos de la física de plasma. Sinceramente, creo que es una de las personas más brillantes que he conocido, parecía siempre saber de todo, hacía el trabajo de varias personas, ayudaba en el TJ-II, en el laboratorio, realizaba diseños de ingeniería para los proyectos de máquinas lineales de plasma, etc. Cuando se marchó al CERN fue una gran pérdida para el grupo y para el departamento, pero sé que hizo lo mejor; ahora es más feliz allí y con menos preocupaciones.

The rst stay abroad is something you will remember forever, and in my case it is precisely in that way. It was on Warsaw, Poland. You are always a little bit scared, it is your rst time abroad for experiments, in a country where not many people talk English (I learned afterwards that some people talk Spanish instead!). I tried to learn a little bit of Polish, but that language is like hell! My god! I spent two months trying just to say hello: cze±¢, it is so complicated... In those two months I visited many places (Auschwitz will always remain in my mind) and I really enjoyed the marvelous polish food. About people there, they are really nice, Agatha, Monika and specially my doppelganger Pawel. Ah! Those beers with him in the morning, the talks about recipes, food, games, the vacuum meters... The experiments were really interesting, we did so many things, we red the laser to many dierent samples, in dierent gases, etc. They were the origin to two

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articles, but the rst one was the best! A collaboration originated during a poster session with my friend Igor Bykov at Carolus Magnus (and now we work together at CER in San Diego! What a coincidence!). I had the idea, Pawel did the experiments, and Igor made the complex analysis. I will always remember our discussion about who should be the rst author. Pawel wrote the paper, but nobody wanted to be rst author! Each one wanted other to take the rst place. In the end Pawel nished in a way I could not rebate: alphabetical order, so I was the rst author and Pawel got what he wanted from the start. Dzi¦kuj¦ Pawel!! Do widzenia! The second stay I made was on Bucharest, Romania. I met such great people there! I made so many friends! Cristi and Alina, Andrada, Gheorghe, Claudia, Tomy, etc. I traveled to so many places, such nice ones! I think Romania is the only place in Europe with so many virgin forests, where food is still really natural (and really good, I continue to use some Romanian recipes), where you can meet people who are really kind and take you as a guest and give you the best bite. It is sad that we have lost all that in the rest of Europe. This is the stay which originated more collaborations, specially with Tomy, we spent so many hours at the laboratory... All the tungsten nitride chapter of this thesis is due to one day when we though: hey! some people is talking about the eect of tungsten nitrides, why don't we try to study it too? The funniest thing happened when I came back later for a congress. I received the Best Poster Award, but I was not present to collect it because I was working with Tomy to analyze some samples! It was so funny that I could not take an award precisely because I was working! Like a scientic of old... Obviously, I did not do that on purpose, we tried to nish before, but we did not realize it was so late! Anyway, for all that: Multumesc!

I will try to be faster acknowledging the rest of the stays, as they were shorter. It was again in Carolus Magnus summer school where I met Sören, from Jülich, another good friend, and another collaboration originated during a discussion at a bar! Ljubljana, Slovenia, a really nice, small country, where all people seem to speak many languages. There I met my very good friend Sa²o. Now, our interchange of mails ends always with Saludos cordiales and Nasvidenje, but I will forever acknowledge that he brought me the marvelous Golden Ghee (well, mostly Patricia acknowledges him!). Our friendship and collaboration continues at JET, Oxford, England, where we work together (he now has become my boss!). I have learned so many things at JET, how complex the real operation of such a big machine is!. During my stays at JET I have to acknowledge the help and friendship from Emilia, Elena, Ana, and specially Daniel, the only one who stays permanently at JET. At JET I also met Sebastijan, my current boss. I really appreciate all the things he has done for me, all the problems we have gone through until I won my current Eurofusion call to work at San Diego. Thanks to all of you! And to nish the abroad acknowledgments, I want to thank the people I have just met in San Diego, where, as I now explain to my friends I am there to re a Plasma Cannon to test how dierent materials are destroyed, as in sci- movies!: Marta, Paco's daughter, who introduced me to a lot of friends; and the people from PISCES-B, Russ, Leo, Jonathan, Saikat, Matthew, Rolando, Michael, etc. You have made that I do not miss (much) my home! Thanks a lot!

Volviendo al español, a la primera persona que quiero agradecer su apoyo y ayuda es a Patricia. La conocí gracias a un amigo común del CIEMAT (Antez) y enseguida encajamos. Su visión tan opuesta a mi pragmatismo me ha ayudado a ver la vida de otra forma, en la que lo meramente bonito también es útil. Su perfeccionismo a la hora de corregir las partes más generales de mi tesis me ha ayudado muchísimo a que cualquiera pueda entenderlas con mucha más facilidad, aunque le ha llevado tiempo hacerlo. Ahora tenemos que afrontar un futuro juntos en San Diego, y aunque no sepa qué me deparará el futuro después de esto, sí que tengo claro con quién quiero estar.

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de la muerte sobre Formula 1 entre Luis, Tim y yo, a pesar del pobre Antez. Es sorprendente la gran amistad que desarrollamos en cada sitio, primero en el edicio 20 con Juan, Marcos, Yupi, Pedro, Sun, Fontdecava, Alfonso, etc., y luego cuando me mudé al edicio 6 al camarote de los hermanos Marx o sinagoga (sigo sin saber muy bien porqué la llaman así) y caí en las garras del correo spam de la gente de allí: Gerardo, Regidor, Kike, Ivan, Elena, Pablo, Iole, Jesus, Raul,etc. Seguro que me dejo a alguien sin nombrar, ½pero después de mencionar a tantos espero que sean ya pocos!.

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INTRODUCTION

1.1 FUSION ENERGY

The development of mankind is invariably bound to a continually rising energy demand. This could not always be coped with a higher eciency as it usually happens in developed countries where the energy consumption per capita diminish as a result of the conversion of a industrial to a service economy. This fact is particularly true for quick growing, highly populated developing countries like China, India and Brazil. The predictions of the World Energy Council for 2050 are about 2-3 times the current energy consumption. They are based in a world population of about ten thousand million people at three scenarios from ecologically and controlled growth to fast technological and economic-driven growth. However, the world energy supply is highly unstable and uncertain, as it depends not only on technological advances, but also on environmental, social and geopolitical issues. The quick developing countries have based their growing energy supply mostly in fossil fuels and nuclear ssion reactors, and hardly in alternative energies like solar and wind. The Global Warming, which is due to the Greenhouse Eect caused mainly by burning fossil fuels, along with the shortage of the latter, turn a theoretically cheap and easily accessed energy source into a very dangerous one, since the limited fossil fuel reserves are only available in a few countries today, and will be extinguished over the course of the next few decades. The cost of adaptation to the Global Warming is subjected to an intense debate, but with a sea water level rise of 38 cm until 2050 it could amount to 0.1 billions dollars per year, from which about 30% are infrastructures like building and improving of dams, roads, etc [1]. The cost of the predicted stronger and more frequent extreme weather phenomena like hurricanes (the estimated cost for hurricane Sandy in 2012 was of 65,000 million dollars in the USA, 18,000 million dollars only in New York city), or the cost in human lives and production due to an extended area of endemic illness like malaria (for example, the expansion of Anopheles mosquito area distribution, as its larvae die at temperatures under 20°C, could aect millions of people in sub-Saharan Africa, not immunized to it) is highly uncertain to say the least, but even the current cost is undoubtedly high. Moreover, those limited reserves, mainly petroleum, have been a source of political instabilities, as they have been used as a political weapon by some countries, have led to wars for access to them, and will probably cause more in the future. Nuclear ssion energy, on the other hand, has an inherent risk, which is unfortunately impossible to avoid as it has been shown in the Fukushima accident in Japan. It is true that Fukushima design was old and awed, the operator company actuation was far from ideal, and such a large tsunami could not be predicted, but even generation III and IV reactors would have had problems to withstand those conditions. Although the possibility that those designs suer a hydrogen explosion or liberate so many radiation products is very low, the necessary shutdown time to be able to operate safely again would be in the range of months. Alternative renewable energies like solar thermal, photovoltaics, wind, etc, have a lower ecological impact, and help to reduce the external energy sources dependence (mainly on fossil fuels), but are usually more expensive than traditional ones (however, in a few years they may be competitive, as it is now wind power), have a large production variability during the day and also along the year, and each country has good eciency just at some of them. Even though these problems can be mostly overcome in a near 100% renewable electricity generation grid by means of a better weather forecast, electrical energy storage in reversible hydraulic dams or in melted salts, a continuous electricity generation baseline by other means is needed to guarantee that the demand is met.

Nuclear fusion energy can be an alternative for the electricity generation baseline due to its characteristics, which will be addressed in the next chapters of this thesis. First, a brief introduction about the reactors being studied for nuclear fusion will be given in section 1.1.1, followed by a description of magnetic connement in section 1.1.2. The main plasma material interactions will be explained in section 1.1.3, nishing with the future projects being studied in section 1.1.4.

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1.1.1 Nuclear fusion reactors

It is in the electricity generation baseline where nuclear fusion energy can be situated in a good position in the mid-long term. The main advantages of nuclear fusion reactors are: almost inexhaustible-fuel, inherent safety and low level radioactive waste. The easiest nuclear reaction in the Earth is between the hydrogen isotopes, deuterium and tritium, called D-T reaction:

2 1D+

3 1T →

4

2He(3.5M eV) + 1

0n(14.1M eV)

Deuterium can be obtained from ordinary water by inexpensive, conventional techniques, about 33 mg per kilogram. The energy contained in these 33 mg is equivalent to 260 liters of gasoline. The oceans are estimated to contain about 4.6·1013tons. Tritium is a radioactive isotope of only 12.3 years of half-life, so it is almost impossible to be found in nature. However, neutrons produced in the nuclear reaction can be used to breed it by bombarding a blanket (see glossary) around the chamber containing lithium.

6 3Li+

1 0n→

4

2He(2.05M eV) + 3

1T(2.73M eV)

7 3Li+

1 0n→

4

2He(2.05M eV) + 3

1T(2.73M eV) + 1

0n−2.47M eV

Only the reaction with6Li is useful, as it reacts with neutrons in the lower energy range (E < 1 MeV).

Its natural abundance is 7.5% of 11 million tons of known reserves together with 200,000 millions tons in sea water. Since only one neutron is produced in each fusion reaction, and each produced tritium requires one neutron it is necessary to provide a small quantity of additional neutrons to balance loses. As neutron multiplier beryllium or lead could be used. There are other nuclear reactions available for controlled fusion like D + D, D +3He, or H +11B which avoid the necessity of tritium production, and have a lower neutron generation, but they have much lower power density, reaction rates, and higher temperature requirements. The prospects for these fuels are too speculative for now, but in the future, with more technological and plasma physics advances they could lead to a cleaner and cheaper energy.

There are many reactor designs based mostly on the connement scheme of the ions in the plasma to minimize their contact and subsequent neutralization at the reactor chamber walls. The usual main objective to make nuclear fusion reaction possible is to maximize the Lawson criterion or triple product: connement time by electron density and temperature. The most successful connement types so far have been inertial connement, through laser radiation or particle beams, maximizing electron density; and magnetic conne-ment, through magnetic elds, as the plasma is composed of charged particles, maximizing connement time and electron temperature. In the USA more eorts have been made towards laser inertial fusion, i.e. at the National Ignition Facility (NIF), where frozen pellets of deuterium and tritium are imploded by focusing 192 lasers to get 500 TW power. Europe, by contrast, is focusing more towards magnetic connement, mainly in the international project ITER (International Thermonuclear Experimental Reactor) signed in 2005 involving the European Union, China, India, Japan, Korea, Russia and the USA. The ITER project implies building the largest fusion experimental device in the world in Cadarache, in the South of France. Nowadays, most eorts in the magnetic connement nuclear fusion community are directed towards the ITER development and previous studies like material resistance, plasma physics, connement, etc. A continuous reactor called DEMO (DEMOnstration power plant) is starting to be conceptually designed and its construction is planned to begin in 2030 based on the ITER results. It is necessary to emphasize something in common between all nuclear fusion reactor schemes: they need hard-to-reach conditions to be able to achieve fusion.

Nuclear fusion has already been demonstrated in 1991 at JET and TFTR tokamaks where several MW of fusion power were obtained by D-T reaction. At JET a peak value of 16 MW was reached with a 25 MW heating power, corresponding to a ratio of 0.6 calledQDT. In 2017-2018 new D-T experiments are planned

in JET to achieve the break-even, QDT = 1, where more power is obtained by nuclear reactions than the

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for renewable and fossil fuel respectively [2]), and even projects as expensive at rst thought like ITER, with an estimated cost of 18.000 million euro, could be paid with the estimated cost of one and a half month of Iraq war [3].

1.1.2 Magnetic connement

As previously mentioned, one of the main areas of research in Europe within nuclear fusion is the magnetic connement reactor. As the plasma is composed of free ions and electrons, they will follow magnetic eld lines. Ideally, in this way, charged particles would never abandon that lines. However, due to many reasons, this is far from being true, resulting in particle banana orbits due to the geometrically inhomogeneous magnetic and electric elds. Additionally, the measured particle diusion is very large, 4 and 2 orders of magnitude larger than non classical or neoclassical calculations respectively. This occurs because the turbulent regime is predominant, contrary to what was thought in the 50's (being the main reason of the initial high hopes for a nuclear fusion reactor development in 30-40 years). So due to plasma turbulence and instabilities some particles will always escape from the connement and impact on the walls. The dierent magnetic coil schemes for the conguration of the magnetic eld lines and their interaction with the walls dene the main types of reactor.

1.1.2.1 Reactor geometry: tokamak and stellarator

A magnetic eld created by a pure solenoid structure is not enough to maintain charged particles trapped, as they are lost at the solenoid ends. The most intuitive geometry consist in closing those magnetic eld lines on themselves by means of a torus. But the charged particles in a magnetic eld applied only in the toroidal direction, called toroidal magnetic eld, suer a large drift towards the mayor radius due to the centrifugal force and the negative gradient of the magnetic eld strength. If those magnetic eld lines are twisted helicoidally, then that drift is avoided. How these lines are twisted helicoidally by means of generating a poloidal eld denes the two main types of reactor, tokamak and stellarator.

Figure 1.1: Schematic view of a tokamak and its magnetic

coils Figure 1.2: Plasma and modular magnetic coils from theWendelstein-7X in Greifswald (Germany).

ˆ Tokamak: is a toroidal device with a strong toroidal magnetic eld generated by a toroidal eld coil system as can be seen in Figure 1.1. The poloidal magnetic eld is created by a toroidal current owing through the plasma. This current is created by means of a transformer, where the plasma itself forms the secondary winding and the primary is wound around an iron core. There are also two loop forces which expand the plasma ring, that are compensated by an external vertical magnetic eld that interacts with the toroidal current to give an inward force. This vertical eld is spatially non-uniform to create a D-shaped plasma, to have elongation and triangularity, and in this way more particles will be on the high eld side. Finally, another external horizontal magnetic eld is used to maintain the plasma well centered. Both, horizontal and vertical magnetic elds, are applied by means of a feedback control to ensure proper plasma positioning.

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the electricity generation depending on its duty cycle. Alternative ways are being studied to maintain the current, known as current-drive methods, with encouraging results. On the other hand, the main advantage of a tokamak, the toroidal current, can be also its main showstopper. A sudden termination of the plasma pulse, called disruption, can lead to the complete release of the energy contained in the plasma to the vessel walls. This is a serious issue, as a disruption can lead to partial wall melting or even breaking the vacuum vessel. Disruption consequences and its mitigation techniques will be treated in section 1.1.3.

ˆ Stellarator: is a group of toroidal devices where the helicoidal twist of the magnetic eld lines are generated by external eld coils. The classical stellarator consists on a set of planar toroidal eld coils and a set of dipole coils that are wound around the torus circumference a number of times. In a heliac the center of the toroidal elds follow a helical line, where a small helical coil can be added to improve the magnetic eld lines twist. The most advanced design is shown in Figure 1.2. The set of toroidal and helical coils is replaced by a set of modular coils that generate approximately the same magnetic eld. Since coil geometry calculations are not a restraint nowadays, the magnetic eld should be easier to optimize.

When compared to tokamaks, the main drawback of stellerators is their complexity: to be designed and built (like the modular stellarators), to design in-vessel components able to withstand large heat loads with few impurity release into the plasma core, to interpret data from the diagnostics and to develop theoretical or semi-empirical codes. Nevertheless, stellarators do not have a toroidal current so they can easily operate continuously and no disruption can occur. These issues are so serious for tokamaks that stellarators can be the chosen design for advanced nuclear fusion reactors, as the current dierences in performance achieved by both concepts are largely due to experimental reactor sizes rather than inherent shortcomings.

1.1.2.2 Controlling the plasma shape by solid surfaces

Controlling the plasma shape in any magnetic connement device is paramount, as the plasma tends by itself to ll the whole inner volume. In that case the realizable plasma parameters would be poor as the wall will be too close to the plasma, and hence the impurities from the wall would enter the plasma unopposed, cooling it down. To solve this problem, a specic solid surface, target tile, must be established where most of the charged particles which escape the plasma impact, and are thus neutralized. Then a large particle ux is established between the plasma (the source) and the solid surface (the sink) because of the charged particles density gradient. In fact this gradient is so large that the particles impacting the solid acquire velocities close to the sound velocity. Then a shell develops around the plasma which dening the power and particle transport, called Scrape O Layer (SOL). At the SOL is where the charged particles escaping from the plasma meet the eld lines that direct them to the target tiles. The thickness of this layer is usually around 3-10 mm, and the number of charged particles is almost nonexistent beyond it, thus greatly reducing the ion bombardment of the walls (except at the target tiles of course), but not the ion-generated energetic neutrals. In this way two zones are created, the plasma core with closed magnetic eld lines where almost all atoms are ionized, and the plasma edge, where the magnetic eld lines pass through a material surface and the number of neutral atoms and molecules is very large. Target tiles will obviously have to withstand large heat and particle loads, and at the same time they have to minimize the generated impurity inux to the plasma. Their design is therefore one of the main parameters in any reactor. Two options are used: limiters and divertor.

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ˆ Limiter: is a piece protruding from the main wall that intercepts the plasma, as can be seen in Figure 1.3. The last magnetic line that does not pass through the limiter is called Last Closed Flux Surface (LCFS). The lines beyond LCFS dene the SOL, in purple color in Figure 1.3, as they guide the escaping ions to the limiter surface. The exposed surface has to be large enough to avoid too large power uxes and at the same time maintain a symmetry with the plasma to support its stability. The most convenient is a limiter along the toroidal direction. The main drawback of the limiter is its direct contact with the plasma, since it receives an unimpeded heat and particle ux from the plasma, which can melt or erode too fast the limiter material if a very powerful plasma wants to be achieved. Furthermore, due to these large heat and particle uxes, a big quantity of limiter material atoms would be expelled and enter the plasma directly, contaminating it. So the limiter has to be suitably cooled and made from a refractory material that minimizes the contamination of the plasma. The main material options will be outlined in section 1.5. In order to avoid direct contact with the plasma and thus, reduce the issues related to it, the divertor concept was developed.

ˆ Divertor: in this conguration an extra set of magnetic coils is placed concentric with the plasma current. Then the magnetic conguration can be divided in two zones by a line named separatrix, and a so-called X-point where the poloidal eld is zero, as it can be seen in Figure 1.3. The charged particles escaping the plasma core are directed to a separate chamber under the X-point where they are neutralized on the target tiles. This conguration has many advantages, as the pressure in the divertor chamber can be high enough to reduce the energy of the incoming particles and thus the damage to the target tiles, and also support the helium ash pumping out of the reactor (which is not an easy task in limiters). It can even reach such a high neutral pressure that it enters the denominated plasma de-tachment regime, where the plasma temperature at the divertor is low and there is a signicant plasma pressure decrease along eld lines close to the target tile. It is usually accompanied by a signicant decrease in the incident power and plasma ux density, hence very desirable operation for material damage control. The contamination from the wall materials to the plasma is also greatly reduced, as impurities are screened out by the plasma pressure, and the magnetic eld lines. Furthermore, the atomic radiation from wall impurities or on-purpose injected ones, like noble gases, leads to a lower mean ion temperature and to a lower particle bombardment energy at the divertor materials, accord-ingly distributing the heat from the plasma all along the divertor surface, not just at the target tiles. This last process is called divertor radiative cooling and is paramount in a future reactor, refer to sec-tion 1.3.2 for more details. Nowadays the main research focus is on the divertor design due to its lower material resistance requirements and better impurity plasma screening. However, their implementation in a stellarator is very arduous due to the intrinsic complexity of its magnetic conguration.

1.1.3 Plasma Material Interaction

In a controlled fusion reactor, the temperature gradients between the plasma and the surrounding walls are the greatest known to humanity. The control of the wall load in nuclear fusion devices in terms of material erosion and migration, fuel trapping and core plasma contamination are therefore key for the successful development of a nuclear fusion power plant. A large variety of processes are involved in the plasma material interaction in a range from electron-volts (eV, see glossary) scale atomic interactions to hundreds of megajoules disruptions, a dierence of about 27 orders of magnitude. Furthermore, the edge plasma and the walls are closely coupled, and at the same time, the edge plasma limits the performance of the core plasma due to its strong inuence in particle transport processes and thus energy connement. For example, if the edge places too much power onto the walls, they will erode and generate impurities that can enter the plasma core if the edge does not screen them out eectively. Those impurities both dilute and cool down the plasma core and might lead to a plasma instability, which could place more power onto the walls, starting a chain of coupled reactions leading to a poor plasma performance or even a disruption in tokamaks. So this subject requires collaboration between several technical and physical disciplines: material physics, chemistry, atomic and molecular physics and plasma physics from cold, low ionized plasmas to keV plasmas.

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1.1.3.1 Stationary heat and particle loads

Regular operation of a plasma device implies the necessity to dissipate the power entering the SOL from the plasma core, 10-15 MW in the largest machines like JET . As the SOL is very thin, the power is translated in a parallel heat ux up to 500 MW/m2 with a width of 5 mm that is directed to the strike point (see glossary), where the separatrix intersect the divertor target tiles (see glossary). This huge heat ux on the divertor plates must be reduced bellow the technological feasible perpendicular heat uxes. For actively cooled surfaces it is in the order of 10 MW/m2. Several strategies are used in present machines to maximize the area where power is loaded to: poloidally inclining the divertor tiles, optimizing the divertor coils to increase the magnetic ux expansion and broadening the SOL heat ux width through increased perpendicular transport. A factor of 10 is expected in ITER, 4 from magnetic ux expansion and 2.5 from target inclination (however, recent calculations point to a very reduced loaded area, around 4 mm, at some conditions like non-detached or attached mode at the divertor [4]). In addition, extrinsic gas impurities like neon, argon and nitrogen can be introduced to increase the capability of the divertor to radiate power (radiative cooling, see section 1.3.2). The necessary radiated power in ITER will be much larger than in present machines, around 50% of the total power.

At the main wall, due to the SOL heat and particle ux, the gas density is between 30 and 300 times lower than at the divertor. Thus the power loads at the main walls compromise mostly plasma and divertor gas radiation, leading to an uniform power loading prole. For example, for ITER half of the total power will be radiated onto the walls leading to a power density of only 0.11 MW/m2, easily extracted through water cooled panels. However, great care must be taken when selecting the main wall material in order to reduce plasma core contamination due to their proximity and dicult impurities screening, as will be explained in section 1.3.1.

1.1.3.2 Edge Localized Modes (ELMs)

ELMs are magnetohydrodynamic related periodic events that occur during a regime of enhanced global energy connement denominated high connement mode. The so-called H-mode is a regime of operation spontaneously attained when the auxiliary heating power is high enough, mainly in divertor devices. Although H-mode is also possible in tokamaks or stellarators without divertors, it is more dicult to achieve at those devices. A sudden improvement in particle connement time (a factor of 2) is detected, leading to increased density and temperature in the core, separating this mode from the normal low mode or L-mode. This makes the H-mode a desirable operation regime, specially for ITER and future nuclear fusion reactors. However, the processes involved in it are not completely understood, and thus object of deep study. It is known that such particle connement improvement is originated by the development of a so-called transport barrier at the edge of the plasma which greatly reduces the transport to the SOL and cause an abrupt step or pedestal in the temperature and density proles, in a process similar to an accumulation of energy at the edge. ELMs involve very rapid expulsion of energy and particles from the edge of the plasma into the SOL and can transiently reduce the temperature and density in this region, decreasing the pedestal, and thereby aect the core connement. They play a benecial part as they help the expulsion of core plasma impurities that usually accumulate at the edge in H-mode. Nevertheless, depending of the ELM type, they can carry up to 15% of the energy accumulated in the pedestal, large enough to melt or quickly erode the divertor tiles. The physics of how the ELM energy reaches the divertor and wall is still too uncertain for modeling the heat ux reaching those surfaces, but experimentally it has been observed that up to half of this energy is transported outside of the divertor. They are mainly two types of ELMs that are studied for ITER H-mode regimes, but also some other regimes without ELMs, or with controlled ELM frequency are good candidates.

ˆ Type I or giant ELMs: they are the most usual in H-mode. They have a low frequency, 0.5-2 Hz, but large energy, up to 0.6 MJ in present machines and 8-20 MJ for ITER. However, the timescale of the heat load at the wall tiles is only about 0.1-1 ms, so a semi-innite solid heat transfer model has to be used, as thermal equilibrium is in the range of some seconds. Therefore the power deposited on the divertor tiles could be in the order of 1 GW/m2, or 120 MJ/m2s0.5, more than enough to ablate or melt the surface, about 40-50 MJ/m2s0.5for most refractory materials. In current machines the temperature excursion is around 2500 °C during the ELM. On the other hand, the connement achieved in type I ELMs H-mode is the highest.

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has been achieved in steady state. In this regime the heat loads at the divertor tiles are very low, <1 MJ/m2s0.5, with temperature increments of 10 °C in present machines. However, due to the high impurity levels at the edge the connement loss lies between 10 and 30 % with respect to the type I ELMs H-mode, but the divertor material lifetime improvement could worth this loss.

ˆ ELM free: They are a group of H-mode regimes with dierent parameters. They are usually dicult to achieve and are not observed at all experimental nuclear fusion device. Furthermore, they reduce too much the connement compared to type I ELM H-mode, and are thus not included in the main ITER design. They include: quiescent H-mode, enhanced D-Alpha, electron cyclotron heating at the edge, etc. Another special regime consist in using in-vessel coils to apply resonant magnetic perturbation on the plasma edge. In this way the plasma edge is ergodized (chaotic eld lines are generated) leading to a mitigation or even suppression of ELMs. These in-vessel coils have already been designed for ITER, although the physics of boundary ergodization are not completely understood as they vary considerably from one device to another. For example in some devices, like DIII-D in San Diego U.S.A., density and temperature in the plasma core are increased in the rst moments, but need an active feedback control. ˆ ELM pacing: they are a group of techniques that are used in type-I ELM regime which are able to trigger an ELM in a controlled way, thus increasing the frequency (even a factor of 15-30) and reducing the ELM energy. They include the application of short plasma vertical displacements (kicks), and frozen deuterium pellets injection as most of them trigger an ELM around 0.2 ms after.

1.1.3.3 O-normal events: disruptions

All existing tokamaks are subjected to occasional rapid plasma termination events, called disruptions. In large size machines disruptions have already caused signicant damage, such as melting or signicant erosion of plasma facing components, short circuits in external supplies and deformation of in-vessel structures. In future nuclear fusion devices, such as ITER, the problems will be more serious, as heat loads and forces will be up to two orders of magnitude larger. Disruptions can generally be divided into two basic categories: major disruptions and loss of equilibrium control by vertical displacement events (VDEs) leading to a disruption.

ˆ Major disruption: the plasma becomes unstable as a result of reaching an operational limit, in density or plasma pressure, which leads to the growth of a large magnetohydrodynamic mode. This may be initiated for many reasons, e.g. a small piece of material falling into the plasma, where the resulting rapid cooling of the plasma periphery can result in an unstable plasma. The large MHD activity breaks the nested magnetic eld surfaces. Thermal energy is rapidly lost, and the current prole attens, causing a drop in the plasma inductance (the source of the poloidal eld generation) and a corresponding upward spike in the current. Finally, the high resistivity of this cold plasma results in a rapid decay of the plasma current, nishing in a VDE where part of the magnetic energy is lost to the main wall. The thermal loss transfers most (80-100%) of the total plasma energy into the divertor walls. The heat loads are indeed very large and occur in a short timescale of 1-10 ms, meaning 10-150 GW/m2, or up to 2 GJ/m2s0.5 in ITER. This huge heat load is more than enough to severely melt and ablate a large part of the divertor tiles. But also the tokamak structure suers during the VDE following the disruption due to the formation of halo currents from the plasma current (up to 100%). They ow along open eld lines surrounding the plasma intersecting the vessel wall and return poloidally through conducting components of the vessel structures. The ow of this return current will be perpendicular to the main magnetic eld, thus exerting a large mechanical force on these structures. Disruptions in ITER are predicted to be around 10 % of the total pulses.

ˆ Loss of equilibrium VDE, the results are similar to disruptions but the sequence is dierent. The rst event is a loss of the vertical position, and the plasma moves vertically with the cross-section and the poloidal magnetic eld decreasing as the plasma scrapes o against the main wall. The plasma then disrupts: thermal energy is rst loss (typically, there is no current spike), followed by a plasma current decay. This VDEs can place the heat loads in a dierent part of the divertor than stationary loads, or other disruptions and ELMs, or even in the upper part of the reactor. Consequently these parts are less protected against so heavy heat loads and the damage could be catastrophic, specially in the upper part as the materials are usually not (so) refractory. This kind of VDE is predicted to aect 1% of ITER pulses, but even so they could cause such a severe damage that, in order to prevent it, the in-vessel coils will have to be used for a fast active feedback realignment of the plasma.

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radiator atom as Argon has to be added). Nowadays the prediction codes only act over the MGI valve, but great progress has been made towards active feedback through identifying the disruption origin and determining how to control it in order to avoid losing the pulse.

1.1.3.4 Plasma facing materials desired properties

The materials to be used in a nuclear fusion reactor must be compatible with ultra-high vacuum, cryogenics because of cryopumps, magneto-hydro dynamics, neutron irradiation and handling of large particle and heat loads. As a consequence, the selected materials are subject to strict properties requirements: high thermal and electrical conductivity, good thermomechanical properties and resilience against thermal shocks, low plasma contamination due to line radiation in the core, low neutron activation and resistance to radiation damage, low retention and low chemical anity of hydrogen isotopes. Also high anity to air molecules, oxygen and nitrogen, leading to the formation of stable and non-volatile compounds, is also important for impurity gettering to reduce plasma contamination and material sputtering. Unfortunately, no material could satisfy all these requirements, only a few could be considered. Their advantages and drawbacks will be treated in section 1.5.

1.1.4 Future projects design: ITER and beyond

Much of the signicant progress in magnetic fusion science has been made in the tokamak concept, which has represented the main approach to magnetic connement fusion. In current tokamaks, improvements of plasma performance and control have occurred owing to remarkable advance in several areas of physics and engineering. For example, superconducting coils have allowed long pulses supporting the achievement of a steady state operation regime, for example, in TRIAM-1M pulses of 2 hours were achieved using non-inductive current drive, although at low density and low power discharges. These advances lead to the development of regimes of operation, with both good connement and magnetohydrodynamic stability, which have enabled the production of fusion power from deuterium-tritium plasmas in the tokamaks TFTR (11 MW) and JET (16 MW). So the next step is to demonstrate a safe and economic reactor operation in long discharges with a burning plasma, where more energy is produced by nuclear fusion than the energy necessary to maintain it. This is the main objective for ITER, but it also will provide a test facility for the development of nuclear component technology, which will be extrapolated to a fusion reactor prototype, DEMO. An example are the breeding blankets (see glossary), which will surround the main wall to produce enough tritium from the reac-tion of lithium with the fusion neutrons to replenish the consumpreac-tion and losses. But they are also essential because they stop neutrons extracting their heat to generate electricity and protecting other vulnerable parts like electronic and superconducting coils from them.

However, there are some essential dierences between today's tokamak research facilities and ITER. The increase in pulse duration and cumulative run time, together with the increase in plasma energy content, will represent a true challenge for the materials lifetime, not only for short heavy heat and particle loads, but also in long term erosion and fatigue, as it will be explained in section 1.2. On the other hand, erosion of the reactor walls is not an issue for current tokamak devices in terms of component lifetime, but poses a problem as a source of impurities in the plasma, and it will be also a serious problem for ITER and new stellarators like Wendelstein-7X, as explained at section 1.3. Similarly, fuel economy has never been an issue in deuterium experiments at present devices, but the incomplete recovery of tritium in TFTR and JET experiments in the 90's has placed the tritium retention in the vessel as one of the main issues, as covered in section 1.4. The main materials used at the rst wall for present devices, or planned for near-future ones as ITER and for nuclear fusion reactor designs will be reviewed in section 1.5. Finally, the objectives of this thesis applied for ITER plasma facing materials will be depicted in section 1.6.

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neutron irradiation will also be an issue. Erosion of wall materials limits the lifetime of the wall components, and, at the same time, it is the source of other problems in the reactor. For example, these eroded materials could enter the plasma core diluting and cooling it down, section 1.3, or be deposited elsewhere creating a problem of dust generation and/or fuel retention, section 1.4.

(a)Macrobrush (b)Monoblock

Figure 1.4: First wall tungsten armor designs for ITER divertor

Current estimations of wall lifetime for ITER are based on extrapolations from present experiments or modeling calculations that imply (relatively) large uncertainties. They can give an idea of the number of pulses before reaching a limit in erosion, mechanical strength, etc., but some limits could be reached after only several tens of discharges. Consequently, it is paramount to reduce that damage by means of a careful design of the device components and operation scheme, but mainly by means of a thoughtful material choice, refer to section 1.5 for more details. However, all those designs are linked, and what could be good against thermal loads, could be bad in other terms. For example, the tiles are divided in castellations against thermal shocks like in Figure 1.4, but that causes other problems like formation of codeposits of eroded material with nuclear fuel inside the gaps. In this way a large amount of tritium could be trapped in a dicult to reach area. This issue could be diminished through a careful operation of the device, by selecting a material which does not originate codeposits with fuel, or developing removal methods inside these gaps, refer to section 1.4 for more details.

The main types of damage for the plasma facing components will now be addressed: rst the erosion by physical and chemical sputtering in sections 1.2.1 and 1.2.2 respectively; followed in section 1.2.3 by thermal damage like melting; neutron radiation will be treated in section 1.2.4; other types of damage like blistering will be enumerated in the last section 1.2.5.

1.2.1 Erosion by physical sputtering

When a projectile, energetic ion or neutral, impacts on a target material, it transfers its momentum (energy and mass) to the surface atoms. Depending on the projectile momentum there are mainly 4 processes from lower to larger energy: backscattering of the projectile; ion-induced desorption of adsorvates and emission of electrons and photons from the target; ejection of atoms from the target via nuclear collision with the projectile sputtering; and projectile implantation into the target. During physical sputtering, if the projectile momentum is large enough to overcome the surface binding energy, a surface atom may be ejected from the material. The rst collisions will direct the target atoms into the surface, but subsequent collisions with other projectiles or between surface atoms (in a cascade regime) can direct some of them out of the surface. This process is depicted in Figure 1.5 where it is shown that, depending on its momentum, the projectile might be implanted into the solid or simply backscattered. The sputtered compounds are mostly neutral atoms, but ions and small clusters of the target material may also be ejected. However, as the projectile momentum has to overcome the surface binding energy, there will exist a threshold energy bellow which the sputtering yield is zero.

In order to characterize the erosion by physical sputtering, a yield (YPhys) is dened as the ratio of the averaged number of sputtered atoms for each incoming projectile. This sputtering yield will depend on the momentum transfer between projectile and surface. It will thus depend on the impact energy and angle, and the atomic mass ratio between the projectile and the target atoms. It is important to note that the sputtering yield does not depend signicantly on the surface temperature.

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Figure 1.5: Physical sputtering process of surface atoms by a projectile

to the surface atoms to be sputtered. Nonetheless, the mechanical properties of both the bulk material and the surface will be degraded. In a plasma the impact energy of ions is determined by the ion and electron temperature (Ti and Te, respectively) by this equation: E∼3·Q·Te+ 2·Ti, where Q is the

charge state of the ion. The rst part of the equation originates from the acceleration of the ions in the sheath, and the second part correspond to the Maxwell distribution of the thermal velocity of ions. ˆ Impact angle and surface roughness: the larger the grazing incidence of the projectiles, the more energy

will be transferred to the surface atoms. After reaching a maximum, usually between 70 and 80°, the sputtering yield suers a strong decrease as the projectiles are more eciently reected. The roughness of the surface can change the local angle of incidence, and the sputtered atoms can be redeposited at the side walls of the valleys of a rough surface. Therefore, the roughness could lower considerably the dependence on the angle for larger nominal incidences and increase the erosion at near normal incidences.

ˆ Atomic mass ratio, self-sputtering and preferential sputtering: the momentum transfer is maximum for identical masses of projectile and surface atoms and so is the sputtering yield. This process, called self-sputtering, could be very important in magnetic connement fusion plasmas when a material is sputtered by its own returning ions due to the applied magnetic eld. In some cases the value of the sputtering yield could be larger than one if the ion energy is high enough, meaning that it could cause a catastrophic chain reaction. For very dierent combinations of projectile and target atoms mass, the sputtering yield could be greatly reduced. For this reason when the target is not a monoatomic material (like stainless steel) the atoms most mass-matched with the projectile will suer a preferential sputtering, and consequently the surface will gradually become relatively rich on poorly mass-matched atoms.

Therefore as the nuclear fusion plasma is made of very light atoms (hydrogen isotopes), for reducing the erosion it could be very interesting to use high Z materials (for example the threshold energy for tungsten sputtering by deuterium is around 200 eV, while for beryllium is around 10 eV). However, these high Z atoms can pose serious operation problems if they enter the plasma core because their large line radiation would cool down the plasma decreasing its eciency and even leading to disruptions, refer to section 1.3 for more details.

1.2.2 Erosion by chemical sputtering

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ˆ Impact energy: the dependence is much lower than for physical sputtering, but qualitatively similar. The exception is that the threshold energy is usually very low, in the order of 2 eV.

ˆ Surface temperature: as any chemical reaction, the chemical sputtering has a great temperature de-pendence. It increases with temperature, but many times it has a maximum where the erosion-related reaction is hampered, for example by the formation of a passivating layer, or because another reaction becomes more important.

ˆ Ion ux: data from various experiments in ion beam devices, linear plasma machines and nuclear fusion experiments show a yield reduction in the yield for very large uxes, for carbon-hydrogen is around 1021 H+/m2s. This eect seems to be related to the increase of the surface temperature due to strong ion bombardment, and hence, as commented in the previous point, the development of another non-erosion related reaction. A displacement in the thermodynamic equilibrium of the reaction towards the reactants like the generation and release of molecular hydrogen could also be the cause of this eect. The initial surface temperature has been found to have an eect in the ux dependence.

1.2.2.1 Hydrogen chemical sputtering of carbon materials

Carbon materials, like graphite and Carbon Fibre Composites CFC, present a large chemical sputtering by hydrogen isotopes. Plasma-generated hydrogen radicals react towards molecular, or radical, hydrocarbons by chemical erosion, or from dangling bonds created previously by impinging hydrogen ions or any other ions (wall materials, helium, and other impurities). The atomic description of the chemical erosion is presented in Figure 1.6. Firstsp2 bonded carbon is hydrogenated to sp3 via an intermediate spx (bottom and

left-hand side of Figure 1.6). Further hydrogen radical bombardment leads to the desorption of H2 through the intermediatespx (top of Figure 1.6). Finally, if the surface temperature is larger than 400 K, the chemical

erosion can continue via desorption of hydrocarbon complexes, closing the cycle at the initial step withsp2

bonding. At temperatures larger than 600 K, the intermediate statespxstarts to dehydrogenate towardssp2,

therefore decreasing the chemical erosion. Accordingly, the chemical bonding at the surface of the carbon material determines its chemical erosion rate. For example, diamond,sp3 with very strong bonds closed at

the surface, presents an erosion rate one order of magnitude lower than graphite,sp2; and more than three

orders of magnitude lower than high-hydrogen amorphous hydrocarbon called soft a-C:H , with a large content of reactivespx bonds. On the other hand, ion bombardment creates dangling bonds which reduce

the disparities among dierent carbon materials, i.e. diamond erosion is largely enhanced whereas soft a-C:H is not so much.

Figure 1.6: Atomistic process of chemical erosion of carbon by hydrogen

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lies between 200 and 400 eV, except for C3H4where it lies at 800 eV. However, a uence dependence of a factor 3-4 lower has been found for C2Hy formation, although its quantication remains unresolved. In a fusion reactor, these hydrocarbons released from the surface, stable or radicals, can be ionized and/or broken within the plasma into more reactive species, and thus develop amorphous hydrocarbon lms elsewhere. This last process generates codeposits of carbon with hydrogen isotopes, i.e. fuel, in dicult-to-reach places, which will be the main drawback of carbon-related materials. This severe problem will be approached in section 1.4.2.

1.2.2.2 Chemical sputtering of carbon materials by other reactive species

Carbon materials can also suer chemical sputtering by other species. The most important ones for nuclear fusion are oxygen and nitrogen. They could be unintended present from air leaks, leading to an undesired erosion of wall tiles, or injected on purpose to eliminate the codeposits of carbon and hydrogen isotopes (a-C:H) due to their high reactivity, refer to section 1.4.3 for full details.

ˆ Oxygen: his strong reaction with carbon materials producing CO and CO2 is well-known. The ion bombarding energy dependance is weak for values between 50 eV and 10 keV, being the yield of 0.7-1 C atoms eroded by impinging O atom, two orders of magnitude larger than hydrogen. The dierent types a-C:H are far more reactive and the yield is usually 10 to 50 times larger than for pure carbon. This huge yield is explained by the simultaneous production of CO, CO2, H2O, hydrocarbons, etc. ˆ Nitrogen: its reaction with carbon materials to produce C2N2is also well-known. The ion bombardment

energy dependance is again weak: between 50 eV and 1 keV the yield is approximately 1 C atom for each incident N atom (comparable to oxygen). Now the yield from a-C:H lms and pure carbon materials is almost the same except for the more reactive soft a-C:H lms (high H content). Furthermore, a clear synergistic eect is seen when atomic hydrogen is injected, as the yield increases 3 to 7 times, due to the large number of broken bonds caused by the large momentum transfer between nitrogen and carbon atoms. The enhanced reaction to hydrogenated carbon-nitrogen products plays an important role as well, being HCN the main product, but also CH3CN, NH3, hydrocarbons, etc.

1.2.2.3 Total yield

The total erosion, or yield (Y), is determined as the ux of eroded particles divided by the ux of incoming projectiles. It is the sum of the physical sputtering and chemical sputtering (if existing). A clear example of both processes on graphite can be found in Figure 1.7. It can be seen how the experimental values (squares) at lower energies are quite large and do not match the physical sputtering model (red line). When a chemical sputtering model (blue lines) is added, then the experimental data are well represented (green line). It is possible to extract some conclusions for carbon erosion by hydrogen as well: total yield is fairly independent from bombardment energy except for very large values; and chemical sputtering is prominent at low bombardment energies, while at energies above 1 keV only physical sputtering has an eect.

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1.2.3 Melting and evaporation

Refractory materials like tungsten and graphite are usually chosen for those places where the largest heat loads are expected. For example, in ITER considering a heat ux of 5 MW/m2to the divertor tiles during steady state, a temperature increase of 400 K can be estimated, easily handled by the proposed heat sink designs: see examples in Figure 1.4. Although melting and evaporation of plasma facing materials could occur in steady state if a heat sink fails, or due to a coolant failure, they would take place during unmitigated type-I ELMs and o-normal events like disruptions due to their short-timescale heat deposition. The temperature increase for unmitigated Type I ELMS could reach 9,600K, and for disruptions 16,000 K, refer to sections 1.1.3.2 and 1.1.3.3 for more details. In ITER the number of these events could be very high during H-mode operation, and also 10% of the pulses during normal operation will nish in a disruption. Hence a characterization of the behavior of divertor materials under heavy transitory heat loads is mandatory.

In metals, the melt layer developed could be in the tens of microns, and will be exposed to various forces such as electromagnetism, gravitation, mechanical vibration, plasma momentum, surface tension and ablation recoil. All these forces leads to the almost completely loss of the melted layer in mainly two ways:

ˆ Horizontal melt layer motion caused by current decay in the liquid metal layer in combination with the strong magnetic eld producing destabilizing Lorentz forces. This process can lead to a bridging of adjacent castellations, and thereby increasing electromagnetic (eddy current) forces, making more susceptible to thermal shocks, etc.

ˆ Droplet ejection (aerosol) due to build-up of vapor bubbles inside the liquid layer, and to the increase of hydrodynamic instabilities caused by the plasma impact momentum at the liquid surface. These droplets could also bridge castellations, but the main problems are the possibility that they could enter the plasma leading to a disruption, and dust generation, which could be radioactive and/or contain trapped fuel.

The amount and rate of melt layer loss is dicult to predict as it depends on many parameters, such as heat ux, impurity and gas content, material properties and disrupting plasma parameters. Getting to know the consequences of the melting, both on the material itself and on the capacity to maintain plasma operation on damaged surfaces is paramount. Devices like plasma guns could simulate the conditions during a disruption thermal quench in ITER. However, those conditions are not completely fullled: expected plasma pressure in ITER is at least one order of magnitude lower; and neither glancing angles of incidence nor strong magnetic elds are possible in plasma guns. In Figure 1.8 the melting of tungsten macrobrush tiles at two dierent pulse conditions in the Quasi-Stationary Plasma Accelerator (QSPA) SRC RF TRINITI are shown. At a power in the order of ITER-like type-I ELMs, Figure 1.8b, the castellations are fully bridged and the tile is rendered useless after only 10 pulses [5]. If lower powers are used, but in the order of ITER-like type-I ELMs, the melting process can be monitored, see Figure 1.8c and 1.8d for a more detailed picture. A crack network appear because of fatigue. Then melting starts in any leading edge due to their lower heat conduction geometry, like those projecting cracks and at the plasma-facing edge of the macrobrush. In order to reduce this eect, a sh-scale conguration has been selected for ITER divertor tiles, where the leading plasma-facing edge of each castellation is shadowed by the previous one.

On the other hand, evaporation losses of metallic plasma facing components are generally smaller, only a few micrometres, about one order of magnitude lower than melting loses. This is combined with a process called vapor shielding. When the heat ux is too large and the metal start to be vaporized, it develops a cloud of vapor just in front of the material, which is heated and ionized by the incoming plasma, and becomes conned by the magnetic eld. The incoming plasma particles are then stopped in the vapor plasma, and its energy is radiated in all directions by the vapor atoms (usually high Z, so large line radiation). In this way the incoming energy towards the metal tile is greatly reduced when a really huge heat ux impacts on the surface. The parameters and the dynamics of vapor shielding depend on the energy ux and the type of target material. A low Z target plasma (e.g. C, Be) expands to larger distances from the surface, whereas vapor shields formed from higher Z materials (e.g. W, Mo) stay closer to the surface.

(22)

(a)Example of initial armor tile (b) Armor tile after 100 pulses at 71

MJ/m2·s0.5

(c)Armor tile after 100 pulses at 44.7 MJ/m2·s0.5 (d)Melt motion and surface cracks

details from c)

Figure 1.8: Tungsten macrobrush armor tile tested in SRC RF TRINITI QSPA to simulate ITER-like ELMs. More details in A. Zhitlukhin et al. [5].

1.2.4 Neutron irradiation

Due to the expected large 14 MeV neutron uence from deuterium-tritium reaction in ITER (∼0.3 MW/m2year), but mainly in DEMO (∼10 MW/m2year), reactor materials are required to have a low neutron activation and to retain their properties like electrical and thermal conductivity, mechanical strength, etc, as much as possible. The energy of the neutrons will be absorbed in the blanket. Nonetheless, the rst wall will suer the severe eects of the neutrons passing through it. In solids the main eects are structural damage (atoms displacements) and nuclear transmutation. They are closely inter-related, synergistic processes. Nuclear transmutation depends greatly on the material, not only on the main atoms, but also on the impurities. As a consequence, great care has to be taken during the design process and material manufacturing in order to avoid any undesired impurity (for example nickel has to be completely avoided in stainless steels as it is transmuted to radioactive60Co). On the other hand, structural damage is usually measured in displace-ments per atoms (dpa), i.e. the number of times an atom is moved from its place in the crystalline structure. The dpa level suered by a material determines its volumetric damage (voids, dislocations, vacancies, etc), which at the same time cause the degradation of material properties (electrical and thermal conductivity, mechanical strength, embrittlement by helium bubbles, etc), and also plasma related eects like swelling, blistering, enhanced tritium trapping in defects, enhanced plasma erosion, etc. All these processes greatly depend on three factors: the material compounds; its structure; and its manufacture. The critical parameter to be controlled by means of these three factors is the ductile-to-brittle transition temperature of the mate-rial. If a rst wall material is in the brittle regime at the steady state temperature, during thermal shocks (ELMs, disruption, etc) it will suer enhanced erosion, severe surface cracking leading to dust production and eventual destruction of the material, increased possibility of catastrophic material failure (for example, loosening of a lamellae from the castellated plasma facing material), etc.

1.2.5 Other damage

ˆ Blistering: this process is observed at high uences of light atoms like helium and hydrogen. These atoms diuse in the surface material and are trapped into voids and vacancies. These voids become lled with more atoms, growing into high-pressure bubbles at the surface of the material. These blisters eventually burst, leading to enhanced erosion by surface aking of the material.

Referencias

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