Doctoral Thesis
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DOCTORAL THESIS Diana Naydenkova Study of plasma in tokamak-type devices with spectroscopic methods Department of Surface and Plasma Science Supervisor of the doctoral thesis: RNDr. Jan Stöckel, CSc. Study programme: Physics Study branch: F2, Physics of Plasmas and Ionized Media Prague 2017 I declare that I carried out this doctoral thesis independently, and only with the cited sources, literature and other professional sources. I understand that my work relates to the rights and obligations under the Act No. 121/2000 Coll., the Copyright Act, as amended, in particular the fact that the Charles University has the right to conclude a license agreement on the use of this work as a school work pursuant to Section 60 paragraph 1 of the Copyright Act. In…...... date............ signature I would like to thank to my supervisor, RNDr. Jan Stöckel, CSc. for proposing this subject and for many helpful suggestions. My thanks also to my advisor, Mgr. Vladimír Weinzettl, PhD for guiding and supporting me over the years. I would also like to thank to all COMPASS tokamak team for patient cooperation in experiments. I would like to thank and dedicate this thesis to my grandparents and to my aunt for giving possibility to get high education. I would like to thank to my faculty for perfect basic received knowledge. Title: Study of plasma in tokamak-type devices with spectroscopic methods Author: Diana Naydenkova Department / Institute: Department of surface and plasma science Supervisor of the doctoral thesis: RNDr. Jan Stöckel, CSc., Tokamak Department, Institute of Plasma Physics of the Czech Academy of Science Abstract: In Tokamak department of Institute of Plasma Physics, radiation of high- temperature plasma is investigated using spectroscopic methods in visible, ultraviolet and infrared regions. The radiation gives information regarding tokamak plasma parameters and their changes, which is necessary for future realization of fusion reactor. In the frame of this doctoral thesis the development of spectroscopic diagnostics for observing of plasma radiation and its temporal evolution in COMPASS tokamak was performed. The absolute calibration of developed systems in order to recalculate measured signal to units of radiation was done. The sources of imprecisions of absolute measurements in tokamak conditions are properly discussed in the first part of the dissertation. Plasma radiation in the range 257-1083 nm was measured and interpreted using NIST database and FLYCHK code. Ion density for the most significant impurities was estimated. IDL code for effective ion charge estimation as a ratio of real and hydrogen plasma Bremsstrahlung radiation near 523 nm was developed. Profiles of electron density and temperature measured by Thomson scattering system were used for hydrogen plasma Bremsstrahlung radiation calculations. The example of applying of spectroscopic data for studying of COMPASS plasma heating using neutral beam injections is given in the last part of this work. Keywords: COMPASS, spectroscopy, calibration, Zeff, NBI Contents Introduction 1 1. Development and absolute calibration of the spectroscopic system for COMPASS tokamak 4 1.1. The initial state 4 1.2. Diagnostic design 6 1.3. Absolute calibration 9 1.3.1. Method 10 1.3.2. Calibration results and their imprecisions 13 1.3.2.1. Alignment 13 1.3.2.2. Quantum noise 15 1.3.2.3. Wavelength dependence of quantum sensitivity 16 1.3.2.4. Wavelength dependence of calibration source irradiance 17 1.3.2.5. Wavelength dependence of transmittance of the system components 18 1.3.2.6. Finite width of interference filter transmittance range used for PMT based system 19 1.3.2.7. Temporal dependence of vacuum window transmittance 24 1.3.2.8. Errors caused by data collection and data processing 28 1.3.2.9. Influence of measurement geometry and plasma wall interaction to the measured signal 30 2. Data processing and results 34 2.1. Spectrally resolved plasma radiation measurements 34 2.1.1. Spectra interpretation methods 35 2.1.1.1. NIST data base 36 2.1.1.2. FLYCHK code 38 2.1.2. Radiation shells 42 2.1.3. Ion density estimation method 44 2.2. Effective ion charge estimation 45 2.2.1. Measurements of Bremsstrahlung radiation of real plasma 46 2.2.2. Calculation of Bremsstrahlung radiation of pure Hydrogen plasma 48 2.2.3. Typical results of Zeff estimation for real Hydrogen plasma 53 2.2.4. Precision of Zeff estimation 55 2.3. Visible spectroscopy at the Neutral Beam Injection on the COMPASS tokamak 65 2.3.1. Neutral Beam Injector on the COMPASS tokamak 65 2.3.2. Main channels of power losses at Neutral Beam Injection 67 2.3.2.1. Power losses in the beam duct, PBD 68 2.3.2.2. Power losses by shine through, PST 68 2.3.2.3. Orbit losses, POL 68 2.3.2.4. Charge exchange losses 68 2.3.3. Behaviour of impurities at a medium density discharge with NBI heating 69 2.3.4. Behaviour of impurities at a low density discharge with NBI heating 71 2.3.5. Impact of the plasma position on impurity content 72 Conclusion 76 Bibliography 79 List of Tables 85 List of Abbreviations 86 Attachment 1. Table of spectral lines in wavelength range 247-1083 nm for D-shape discharge at COMPASS tokamak 88 Attachment2. Research projects 97 Attachment 3. Publications 98 Attachment 4. Summer schools and conferences contributions 101 Introduction Demand for energy is expected to increase three times till the end of 21st century under the combined pressure of population growth, increased urbanization and expanding access to electricity in developing countries. A new source of energy is urgently needed. Fusion is a reaction between two or more atomic nuclei which come close enough to form one or more different atomic nuclei. Fusion is potentially attractive way of energy generation because of several advantages. Typical fusion reaction releases nearly four million times more energy than a typical chemical reaction during burning of coal, oil or gas and four times more than fission reaction (at equal mass). Fusion fuels are widely available and nearly inexhaustible. Nuclear fusion reactors produce no high activity, long-lived nuclear waste. A nuclear accident is not possible in a fusion device. If any disturbance occurs, the plasma cools within seconds and the reaction stops. The quantity of fuel present in the vessel at any one time is enough for a few seconds only and there is no risk of a chain reaction. The power output of fusion reactor will be similar to that of a fission reactor. The average cost per kilowatt of electricity is also expected to be similar. Fusion could make a positive contribution to the reducing of CO2 level. One of the devices where fusion reaction can be reached is tokamak. Experimental research of tokamak systems started in 1956 in Kurchatov Institute, Moscow by a group of Soviet scientists led by Lev Artsimovich. The group constructed the first doughnut-shaped magnetic confinement device called a tokamak, the most successful being T-3 and its larger version T-4. T-4 was tested in 1968 in Novosibirsk, conducting the first ever quasistationary thermonuclear fusion reaction. From this time the tokamaks became the dominant concept in fusion research, and tokamak devices multiplied across the wold. The amount of fusion energy which a tokamak is capable to produce correlates directly to the number of fusion reactions taking place in its core. It means in order to get more energy we need bigger plasma. The world record for fusion power is held by the tokamak Joint European Torus (JET). In 1997, JET produced 16 MW of fusion power from a total input power of 24 MW. It means return of energy defined by so called fusion energy gain factor Q was 0.67. Ten times larger device – ITER (International Thermonuclear Experimental Reactor) tokamak is 1 under construction now. ITER is designed to produce a ten-fold return of energy (Q=10), or 500 MW of fusion power from 50 MW of input power, for long pulses (400-600 s). ITER will not capture the energy it produces as electricity and will be used to study plasmas under conditions similar to those expected in a future power plant and test technologies such as heating, control, diagnostics, cryogenics and remote maintenance in an integrated way. The main goal of existing tokamaks research programs is steered towards basic challenges of physics and technology linked to the construction and operation of the international experimental reactor ITER and future fusion reactors. The COMPASS (COMPact ASSembly) tokamak is the smallest tokamak which has ITER-like plasma shape. The principal parameters of COMPASS tokamak are shown in table 1. Volume of its plasmas correspond to one tenth (in the linear scale) of the ITER plasmas. The flexible neutral beam heating system and ability to achieve H-mode allow COMPASS to contribute to ITER-relevant plasma physics studies. High flexibility in tokamak operation and low-cost of experiments make it especially attractive in fusion program. Nowadays, besides COMPASS there are only two tokamaks in Europe with ITER-like configuration which are able to get the High plasma confinement. They are JET and ASDEX-U (Axially Symmetric Divertor Experiment - Upgrade). Parameters Values Major radius R 0.56 m Minor radius a 0.23 m Plasma current Ip (max) 400 kA Magnetic field BT 0.9 - 2.1 T Vacuum pressure 1x10-6 Pa Elongation 1.8 Plasma shape D, SND, elliptical, circular Pulse length ~ 1 s Beam heating PNBI 40 keV 2 x 0.4 MW Table 1. The main parameters of COMPASS tokamak. The COMPASS tokamak has been designed and operated in 90th in UKAEA (United Kingdom Atomic Energy Authority) in Culham.