DOCSLIB.ORG

Theoretical Review of M N=1 1 Sawtooth

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Theoretical Review of M N=1 1 Sawtooth 저작자표시-비영리-변경금지 2.0 대한민국 이용자는 아래의 조건을 따르는 경우에 한하여 자유롭게 l 이 저작물을 복제, 배포, 전송, 전시, 공연 및 방송할 수 있습니다. 다음과 같은 조건을 따라야 합니다: 저작자표시. 귀하는 원저작자를 표시하여야 합니다. 비영리. 귀하는 이 저작물을 영리 목적으로 이용할 수 없습니다. 변경금지. 귀하는 이 저작물을 개작, 변형 또는 가공할 수 없습니다. l 귀하는, 이 저작물의 재이용이나 배포의 경우, 이 저작물에 적용된 이용허락조건 을 명확하게 나타내어야 합니다. l 저작권자로부터 별도의 허가를 받으면 이러한 조건들은 적용되지 않습니다. 저작권법에 따른 이용자의 권리는 위의 내용에 의하여 영향을 받지 않습니다. 이것은 이용허락규약(Legal Code)을 이해하기 쉽게 요약한 것입니다. Disclaimer Master's Thesis Theoretical Review of m/n=1/1 Sawtooth Instability and Comparison with Experimental Observations in KSTAR Soomin Lee Department of Physics Graduate School of UNIST 2017 Theoretical Review of m/n=1/1 Sawtooth Instability and Comparison with Experimental Observations in KSTAR Soomin Lee Department of Physics Graduate School of UNIST Theoretical Review of m/n=1/1 Sawtooth Instability and Comparison with Experimental Observations in KSTAR A thesis/dissertation submitted to the Graduate School of UNIST in partial fulfillment of the requirements for the degree of Master of Science Soomin Lee 06. 09. 2017 Approved by _________________________ Advisor Hyeon-Keo Park Theoretical Review of m/n=1/1 Sawtooth Oscillation and Comparison with Experimental Result in KSTAR Soomin Lee This certifies that the thesis/dissertation of Soomin Lee is approved. 06. 09. 2017 signature ___________________________ Advisor: Hyeon-Keo Park signature ___________________________ Kyujin Kwak: Thesis Committee Member #1 signature ___________________________ Min Sup Hur: Thesis Committee Member #2 signature Abstract One of the main problems in nuclear fusion is plasma instability. This instability causes the loss of energy in tokamak, and results in shorter confinement time. By studying plasma instability, it can be predicted in more detail, and therefore, better controlled. In this thesis, sawtooth instability at the core of the tokamak is discussed. At the core of tokamak, the plasma density and temperature shows periodic temporal behavior. When the plasma is heated, the temperature increases linearly. As the core temperature increases and reaches a critical value, the core density and temperature rapidly drop and start to increase again. The periodic relaxation of the core plasma is observed as the sawtooth-like signals of the density and temperature. In this thesis, two famous models by Kadomtsev and Wesson to explain the “sawtooth instability” are introduced. Kadomtsev’s model explains that the fast crash is due to the magnetic reconnection. This model is called ‘the fast reconnection model’. Due to the internal kink instability, the plasma is concentrated near 푞 = 1 surface, and pressure increases in this region. Then the magnetic reconnection takes place near 푞 = 1 surface, and the core heat and temperature collapses through this region. Wesson’s model explains the sawtooth instability assuming a flat 푞-profile inside the 푞 = 1 surface. Due to the flat 푞-profile, magnetic shear near the core is very small, and that results in the formation of ‘hot crescent, cold bubble’ inside the 푞 = 1 surface. This model is called ‘the quasi-interchange’ model. In order to explain both models theoretically, the change of potential energy, 훿푊 for a finite plasma displacement 훏, is derived to determine stability. When 훿푊 > 0, the plasma equilibrium is determined to be stable whereas when 훿푊 < 0, the plasma equilibrium is determined to be unstable. In 푚/푛 = 1/1 internal kink mode, 훿푊 is composed of a second order term, 훿푊2, and a fourth order term, 훿푊4 . Kadomtsev’s model assumes d휉/훿푟 → 0 except 푞 = 1 surface, and Wesson’s model assumes 푞 ~ 1 inside the 푞 = 1 surface. These models have different approaches for explaining the fast disruption in the core. To validate the theoretical models, experimental observation is conducted by measuring several parameters such as temperature or density. In tokamak, it is impossible to measure the temperature, because there is no material to measure the plasma temperature at that high temperature. To measure the temperature, an indirect method by I measuring radiation intensity emitted by gyrating electrons is used in which the radiation is proportional to electron temperature. The ECEI system is a diagnostic system which can directly visualize the MHD instabilities by measuring electron temperature fluctuation. The KSTAR ECEI system measures 2- D/quasi 3-D electron temperature fluctuation which is proportional to the intensity of radiation emitted by electrons. According to Rayleigh-Jean’s law, the optical depth of the plasma is thick enough, the plasma temperature is proportional to the intensity of radiation. Sawtooth oscillations were observed in this study using ECEI system by measuring electron temperature fluctuations in 3D. The sawtooth period was observed as 휏sawtooth ~ 10 ms and collapse time was observed as 휏푐 ~ 100 휇s. The observed spatial structure at the core of the sawtoothing plasma resembles the 푚/푛 = 1/1 internal kink mode, which concludes that the Kadomtsev’s model gives a more relevant explanation of the sawtooth instability in the experiment than Wesson’s model. Keywords: Sawtooth Oscillation, MHD instability, KSTAR ECEI system II III Contents I. Introduction ---------------------------------------------------------------------------------------------- 1 1.1 Motivation ------------------------------------------------------------------------------------------- 1 1.2 Fusion energy and tokamak ------------------------------------------------------------------------ 2 1.3 Background of this study -------------------------------------------------------------------------- 6 1.4 Thesis Outline --------------------------------------------------------------------------------------- 9 II. Theoretical & Mathematical Development --------------------------------------------------------- 10 2.1 MHD Physics Review in Tokamak-------------------------------------------------------------- 10 2.1.1 MHD equations ------------------------------------------------------------------------- 10 2.1.2 Mathematical Description in Tokamak ----------------------------------------------- 13 2.1.3 The 푧-pinch configuration ------------------------------------------------------------ 15 2.1.4 The Screw Pinch Configuration------------------------------------------------------- 17 2.1.5 The Safety factor, 푞 ------------------------------------------------------------------- 19 2.1.6 The Beta, 훽 ----------------------------------------------------------------------------- 22 2.1.7 Properties of large aspect ratio in tokamak ------------------------------------------ 24 2.2 Reviews of MHD instability in tokamak ------------------------------------------------------- 27 2.2.1 Basic idea of MHD stability ----------------------------------------------------------- 27 2.2.2 Energy principle ------------------------------------------------------------------------ 28 2.3 Theories of sawtooth instability ----------------------------------------------------------------- 39 2.3.1 Internal kink, 푚/푛 = 1/1 mode ---------------------------------------------------- 39 2.3.2 Kadomtsev’s model -------------------------------------------------------------------- 41 2.3.3 Wesson’s model ------------------------------------------------------------------------- 46 2.3.4 Comparison of the two models -------------------------------------------------------- 49 IV III. Experimental Methods -------------------------------------------------------------------------------- 50 3.1 KSTAR ECEI System ---------------------------------------------------------------------------- 50 3.1.1 Physical principle of ECE ------------------------------------------------------------- 50 3.1.2 KSTAR ECEI System ------------------------------------------------------------------ 52 3.2 Experimental Set-up ------------------------------------------------------------------------------ 53 IV. Results & Discussions --------------------------------------------------------------------------------- 55 4.1 Experimental results ------------------------------------------------------------------------------- 55 4.1.1 Observation of periodic behavior ----------------------------------------------------- 55 4.1.2 Observation of precursor phase ------------------------------------------------------- 59 4.1.3 Observation of fast crash --------------------------------------------------------------- 62 4.2 Discussions ----------------------------------------------------------------------------------------- 63 V. Conclusion ---------------------------------------------------------------------------------------------- 65 Appendix A. Plasma Parameters --------------------------------------------------------------------------- 66 Appendix B. Vector Relations ------------------------------------------------------------------------------ 67 Appendix B1. Vector formulas ----------------------------------------------------------------------- 67 Appendix B2. Applications to the cylindrical coordinates (푟, 휃, 푧) ---------------------------- 68 Appendix C. Detail derivations ---------------------------------------------------------------------------- 69 Appendix C1. Integration of trigonometric function in section 2.1.4 ---------------------------- 69 Appendix C2. Kadomtsev’s collapse time calculation process ----------------------------------- 70 V References ---------------------------------------------------------------------------------------------------- 73 Acknowledgement ------------------------------------------------------------------------------------------ 77 VI List of figures
Load more

Recommended publications
  • Novel Concept for Disruption Mitigation in the ADITYA
    28th IAEA Fusion Energy Conference (FEC 2020) Contribution ID: 1290 Type: Regular Poster Novel Concept for Disruption Mitigation in the ADITYA - U tokamak by Fast Time Response Electromagnetic Driven Pellet Impurity Injector Wednesday, 12 May 2021 17:00 (20 minutes) Implementation of suitable disruption mitigation technique remains the topmost priority for larger tokamaks including the ITER. The spontaneous disruption in ITER may probably be an unavoidable part, while operated with high performance D-T fuel [1]. Disruptions in ITER could produce very large heat loads on divertor tar- gets and other Plasma Facing Components (PFC), and large electromagnetic forces on the Vacuum Vessel (VV) can lead to structural damages [2]. In order to avoid these detrimental consequences, disruption mitigation is an essential part of tokamak research. The disruption mitigation system (DMS) in ITER is based on massive gas injection (MGI) of impurities, in order to radiate the plasma stored energy and mitigate the potentially damaging effects of disruptions [3]. Although, MGI based technique is suitable for most disruption mitiga- tion, there will be issues in case of having warning time less than 10 ms as may be the case for the onset of some disruptions in ITER [4]. This is due to the slow thermal velocity of the heavier impurity gas molecules, which limits the time needed to travel the distances before it reaches the plasma edge. At present, none of the currently planned disruption mitigations systems for ITER can respond on this time scale. The radiative dissipation of the plasma stored energy during major disruption in ITER by fast injection of massive pellets of low Z impurities, such as Li and Be pellet injection has been numerically modelled by Lukash et al.
    [Show full text]
  • Overview of Recent Experimental Results from ADITYA Tokamak R
    Tanna DOI:10.1088/1741-4326/aa6452 OV/4-3Rb Overview of Recent Experimental Results from ADITYA Tokamak R. Tanna1, J. Ghosh1, P. K. Chattopadhyay1, H. Raj1, S. Patel1, C. Gupta1, P. Dhyani2, K. A. Jadeja1, K. M. Patel1, S. B. Bhatt1, V. K. Panchal1, N. C. Patel1, C. Chavda1, E. V. Praveenlal1, K. S. Shah1, M. N. Makwana1, S. K. Jha1, M. V. Gopalakrishna1, K. Tahiliani1, D. Sangwan1, D. Raju1, U. C. Nagora1, S. K. Pathak1, P. K. Atrey1, S. Purohit1, Y. S. Joisa1, J. V. Raval1, C. V. S. Rao1, M. B. Chowdhuri1, S. Banerjee1, J. Thomas1, N. K. Ramaiya1, R. Manchanda1, A. Kumar1, P. K. Sharma1, S. V. Kulkarni1, K. Sathyanarayana1, B. K. Shukla1, A. Das1, R. Jha1, Y. C. Saxena1, A. Sen1, P. K. Kaw1, and D. Bora1 Rapporteured by: S. Pradhan 1Institute for Plasma Research (IPR), Bhat, Gandhinagar, India 2Ulsan National Institute of Science and Technology (UNIST), Ulsan, Republic of Korea Corresponding Author: R. Tanna, [email protected] Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of maximum plasma current of „160 kA and discharge duration beyond „250 ms with plasma current flattop duration of „140 ms has been obtained for the first time in ADITYA. The discharge reproducibility has been improved considerably with Lithium wall conditioning and improved plasma discharges are obtained by precisely controlling the plasma position. In these discharges, chord-averaged electron density „3:0–4:0 ˆ 1019{m3 using multiple hydrogen gas puffs, electron temperature of the order of „500– 700 eV have been achieved.
    [Show full text]
  • Modeling of the Hα Emission from ADITYA Tokamak Plasmas
    atoms Article Modeling of the Hα Emission from ADITYA Tokamak Plasmas Ritu Dey 1,*, Malay B. Chowdhuri 1 , Joydeep Ghosh 1,2, Ranjana Manchanda 1, Nandini Yadava 3 , Umeshkumar C. Nagora 1,2, Parveen K. Atrey 1, Jayesh V. Raval 1, Y. Shankara Joisa 1, Rakesh L. Tanna 1 and ADITYA Team 1 1 Institute for Plasma Research, Bhat, Gandhinagar 382 428, India; [email protected] (M.B.C.); [email protected] (J.G.); [email protected] (R.M.); [email protected] (U.C.N.); [email protected] (P.K.A.); [email protected] (J.V.R.); [email protected] (Y.S.J.); [email protected] (R.L.T.) 2 Homi Bhaba National Institute, Anushakti Nagar, Mumbai 400 094, India 3 The National Institute of Engineering, Mysuru 570 008, Karnataka, India; [email protected] * Correspondence: [email protected] Received: 30 June 2019; Accepted: 3 October 2019; Published: 5 October 2019 Abstract: The spatial profile of Hα spectrum is regularly measured using a high-resolution multi-track spectrometer in ADITYA tokamak to study the neutral particle behavior. The Monte Carlo neutral particle transport code DEGAS2 is used to model the experimental Hα spectral emissions. Through the modeling of the spectral line profile of Hα, it is found that the neutral hydrogen, which is produced from molecular hydrogen and molecular hydrogen ion dissociation processes contributes 56% to the total Hα emission, and the atoms which are produced from charge-exchange process have 30% contribution. Furthermore, the experimentally measured spatial profile of chord integrated brightness was modeled for the two plasma discharges having relatively high and low density to understand the neutral particle penetration.
    [Show full text]
  • An Application of Survival Analysis to Disruption Prediction Via Random Forests
    An application of survival analysis to disruption prediction via Random Forests R.A. Tinguelyy, K.J. Montes, C. Rea, R. Sweeney, and R.S. Granetz Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA, USA 02139 Abstract. One of the most pressing challenges facing the fusion community is adequately mitigating or, even better, avoiding disruptions of tokamak plasmas. However, before this can be done, disruptions must first be predicted with sufficient warning time to actuate a response. The established field of survival analysis provides a convenient statistical framework for time-to-event (i.e. time-to-disruption) studies. This paper demonstrates the integration of an existing disruption prediction machine learning algorithm with the Kaplan-Meier estimator of survival probability. Specifically discussed are the implied warning times from binary classification of disruption databases and the interpretation of output signals from Random Forest algorithms trained and tested on these databases. This survival analysis approach is applied to both smooth and noisy test data to highlight important features of the survival and hazard functions. In addition, this method is applied to three Alcator C-Mod plasma discharges and compared to a threshold-based scheme for triggering alarms. In one case, both techniques successfully predict the disruption; although, in another, neither warns of the impending disruption with enough time to mitigate. For the final discharge, the survival analysis approach could avoid the false alarm triggered by the threshold method. Limitations of this analysis and opportunities for future work are also presented. Keywords: tokamak plasma, disruption prediction, survival analysis, machine learning, Random Forest, binary classification 1.
    [Show full text]
  • Evaluation of an Oxygen Transport Coefficient in the Aditya Tokamak
    atoms Article Evaluation of an Oxygen Transport Coefficient in the Aditya Tokamak Using the Radial Profile of O4+ Emissivity and the Importance of Atomic Data Used Therein Malay Bikas Chowdhuri 1,* , Joydeep Ghosh 1, Ritu Dey 1, Sharvil Patel 2, Nandini Yadava 3 , Ranjana Manchanda 1, Amrita Bhattacharya 4 , Izumi Murakami 5 and Aditya Team 1 1 Institute for Plasma Research, Bhat 382 428, Gandhinagar, India 2 Birla Institute of Technology-Mesra, Jaipur Campus, Jaipur 302 017, Rajasthan, India 3 The National Institute of Engineering, Mysuru 570 008, Karnataka, India 4 Nuclear Engineering and Technology Programme, Indian Institute of Technology Kanpur, Kanpur 208 016, Uttar Pradesh, India 5 National Institute for Fusion Science, Toki 509-5292, Gifu, Japan * Correspondence: [email protected] Received: 30 June 2019; Accepted: 30 August 2019; Published: 9 September 2019 Abstract: Oxygen impurity transport in the typical discharges of the Aditya tokamak was investigated 3 3 using emissivity radial profile of emissivity of the spectral line (2p3p D3–2p3d F4) at 650.024 nm from the Be-like oxygen ion. This O4+ spectral line was recorded using a 1.0 m multi-track spectrometer capable of simultaneous measurements from eight lines of sight passing through the plasma. The oxygen transport coefficients were determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional impurity transport code, STRAHL, and photon emissivity coefficient (PEC) belonging to that transition. The PEC values were obtained from both ADAS and NIFS atomic databases. Using both the databases, much higher values of diffusion coefficients compared to the neo-classical values were observed in both high and low magnetic field edge regions of typical Aditya tokamak Ohmic plasma.
    [Show full text]
  • Biomedical Applications
    Programme and Abstract Book 44th IOP Plasma Physics Conference 3–6 April 2017 University of Oxford, Oxford, UK Organised by the IOP Plasma Physics Group Contents Sponsors 2 Organising committee 2 Programme 3 Poster programme 5 Abstracts 7 Monday 3 April 7 Tuesday 4 April 11 Wednesday 5 April 15 Thursday 6 April 25 Posters 31 1 Sponsors Organising committee Professor Philippa Browning (Chair PPG), University of Manchester Dr Christopher Ham (2017 Conference Local organizer, Secretary PPG), Culham Centre for Fusion Energy Dr Ceri Brenner (Treasurer PPG), Rutherford Appleton Laboratory Dr Josephine Coltman, AWE plc Dr Benjamin Dudson, University of York Dr Bogdan Hnat, University of Warwick Professor Paul Maguire, University of Ulster Dr Andrew Thornton, Culham Centre for Fusion Energy Dr Felipe Iza, Loughborough University Dr Bengt Eliasson, University of Strathclyde Dr Felix Parra Diaz, Worcester College 2 44th IOP Plasma Physics Conference Monday 3 April 11:00 Registration 12:00 Lunch 13:00 Welcome 13:10 (Invited) Some remarks on collisionless current sheet equilibria Thomas Neukirch, University of St Andrews, UK 13:50 Temporally resolved optical probing of picosecond laser propagation in underdense and near-critical density plasmas Zoë Davidson, University of Strathclyde, UK 14:10 (Culham Thesis Prize) Diagnosis and applications of laser wakefield accelerators Jason Cole, Imperial College London, UK 14:50 Refreshment break 15:10 (Invited) The role of turbulence in tokamak edge transport Istvan Cziegler, University of York, UK 15:50 Vlasov
    [Show full text]
  • Analyzing Fusion Energy 2017 Date: March 2017 Pages: 185 Price: US$ 850.00 (Single User License) ID: AAF7C2E2BB4EN
    +44 20 8123 2220 [email protected] Analyzing Fusion Energy 2017 https://marketpublishers.com/r/AAF7C2E2BB4EN.html Date: March 2017 Pages: 185 Price: US$ 850.00 (Single User License) ID: AAF7C2E2BB4EN Abstracts Fusion power is the power generated by nuclear fusion reactions. In this kind of reaction, two light atomic nuclei fuse together to form a heavier nucleus and in doing so, release a large amount of energy. In a more general sense, the term can also refer to the production of net usable power from a fusion source, similar to the usage of the term "steam power." Most design studies for fusion power plants involve using the fusion reactions to create heat, which is then used to operate a steam turbine, which drives generators to produce electricity. Except for the use of a thermonuclear heat source, this is similar to most coal, oil, and gas-fired power stations as well as fission-driven nuclear power stations. Aruvian Research brings a research report on the lucrative field of fusion energy – Analyzing Fusion Energy 2017. The research report takes a look at the importance of fusion power in an era where the usage of renewable energy has become a norm. The report begins with an analysis of the basics of fusion energy, the various stages of development in fusion power, the magnetic concept, the Z-pinch concept, inertial confinement concept and many others. The role of tokamaks are analyzed in the report, along with the physics of fusion reactors. The concept of plasma heating is analyzed so as to provide a better understanding of fusion power.
    [Show full text]
  • Overview of Recent Experimental Results in the ADITYA-U Tokamak J
    MF1-I15 AAPPS-DPP2020 4th Asia-Pacific Conference on Plasma Physics, 26-31Oct, 2020, Remote e-conference Overview of recent experimental results in the ADITYA-U Tokamak J. Ghosh*1, R.L. Tanna1, Rohit Kumar1, Tanmay Macwan1, Harshita Raj1, K.A. Jadeja1, K.M. Patel1, Suman Aich1, Kaushlender Singh1, Suman Dolui1, D.S. Varia1, D.H. Sadharakiya1, B.K. Shukla1, M.N. Makawana1, K.S. Shah1, Shivam Gupta1, V. Balakrishnan1, C.N. Gupta1, V.K. Panchal1, Praveenlal E. V.1, B. Arambhadiya1, Minsha Shah1, Pramila Gautam1, V. Raulji1, Praveena Kumari1, R. Rajpal1, Nandini Yadava1, Sharvil Patel, Nilam Ramaiya1, M.B. Chowdhuri1, R. Manchanda1, R. Dey1, G. Shukla, K. Shah, N. Bisai1, P.K. Atrey1, S.K. Pathak1, U. Nagora1, Kiran Patel1, Varsha Siju1, J.V. Raval1, S. Purohit1, Manoj Kumar1, K. Tahiliani1, D. Kumawat1, S.K. Jha1, M.V. Gopalkrishana1, Bhavesh Kadia1, D. Raju1, B. R. Doshi1, P. K. Chattopadhyay1, Y.C. Saxena1, A. Sen1, S. Chaturvedi1, Sambaran Pahari2, Adityanandan Savita2, P. Rahulnath2, P. K. Maurya2, Saroj jha2, K. Raghvendra2, Neeraj Shiv2, B. Nagaraju2, Sukant Mahar2, I.V.V. Suryaprasad2 and R. Pal3 1Institute for Plasma Research, Gandhinagar 382428 India 2 CAD, Bhabha Atomic Research Centre, Visakhapatnam 531011 India 3Saha Institute of Nuclear Physics, Kolkata 700064 India e-mail (speaker): [email protected] 19 -3 The ADITYA Upgrade (ADITYA-U) (R 0 = 75 cm, a= 25 x 10 m ) of plasma density as compared to the H2 puffed cm) tokamak is capable of producing circular and shaped Ohmic discharges (~3 x 1019 m-3), indicating lower electron plasma with single and double null open divertor density threshold for toroidal rotation reversal with neon configurations.
    [Show full text]
  • Fusion Energy Conference 2006 Book of Abstracts
    Fusion Energy Conference 2006 Book of Abstracts Contents Overview (OV) 1 OV/1-1 · Overview progress and future plan of EAST Project . 2 OV/1-2 · Overview of JT-60U Results for Development of Steady-State Advanced Toka- mak Scenario . 2 OV/1-3 · Overview of JET Results . 2 OV/1-4 · Development in the DIII-D Tokamak of Advanced Operating Scenarios and Associated Control Techniques for ITER . 3 OV/2-1 · Extended Steady-State and High-Beta Regimes of Net-Current Free Heliotron Plasmas in the Large Helical Device . 3 OV/2-2 · Overview of ASDEX Upgrade Results . 4 OV/2-3 · Overview of Physics Results from MAST . 4 OV/2-4 · Recent Physics Results from the National Spherical Torus Experiment . 5 OV/3-1 · Integration of High Power, Long Pulse Operation in Tore Supra in Preparation for ITER . 5 OV/3-2 · Overview of Alcator C-Mod Research Program . 6 OV/3-3 · Overview of TCV Results . 6 OV/3-4 · Overview of the FTU results . 7 OV/4-1 · Overview of HL-2A Experiment Results . 7 OV/4-2 · Overview of TJ-II experiments . 8 OV/4-3 · Overview of T-10 Results . 8 OV/4-4 · Experimental Progress on Zonal Flow Physics in Toroidal Plasmas . 8 OV/5-1 · Recent Progress on FIREX Project and Related Fusion Researches at ILE, Osaka 9 OV/5-2 · Overview of Inertial Fusion Research in the United States . 9 OV/5-3 · Theory of Alfv´enwaves and energetic particle physics in burning plasmas . 9 OV/5-4 · Status of R&D Activities on Materials for Fusion Power Reactors .
    [Show full text]
  • Status of SST-1 Project and Fusion Research in India
    Status of SST-1 Project and Fusion Research in India Yogesh C. Saxena Institute for Plasma Research Gandhinagar, INDIA Plan of Talk • Fusion Research In India – ADITYA Tokamak – SINP Tokamak • SST-1 tokamak – Objectives – Parameters – Description of Sub-systems & Fabrication Status Indian Tokamaks ADITYA SINP Tokamak Major Radius R0 0.75 m 0.30 m Minor Radius a 0.25 m 0.075 m Toroidal Field BT 1.50 T 2.00 T Plasma Current Ip 250 kA 75 kA Pulse Duration ? 250 ms 20-30 ms Plasma Cross-section Circular Circular Configuration Poloidal Limiters Poloidal Limiters Coils Type (TF & PF) Copper Water cooled Copper Current Drive & Heating Ohmic Transformer Ohmic transformer (Air Core) (Iron Core) Vacuum vessel Vessel with Electrical break Conducting Shell Design & Fabrication Indigenous M/S Toshiba, Japan Installation 1989 1987 ADITYA TOKAMAK • Regular operations with transformer-Converter power system. 80-100 kA ; 100 ms plasma discharges at 8.0 kG field are being studied. Standard Diagnostics have been developed and deployed. • Magnetic flutuations have been studied using Single value decomposition technique. • Edge fluctuations and turbulence in density and potential have been investigated. • Pdf are found to be non-Gaussian ; turbulence exhibits intermittency • Low frequency (< 50 kHz) contribute to convective particle transport • Turbulence exhibits complex, chaotic spatial and temporal variation. • Coherent structures have been obtained.in potential fluctuations using conditional averaging techniques. The dynamics of the structures suggest mechanism of “bursty” transport of edge plasma. • Coupling coefficients and coherencies have been obtained using Wavlet analysis to study the power transfer due to linear,mixed and quadratic coupling.
    [Show full text]
  • EPS2005, Session "P-1" Abstracts
    EPS2005, Session "P-1" Abstracts Session Author PosterTitle P-1.001 P.Träskelin Molecular dynamics simulation of erosion of tungsten carbide by deuterium bombardment P-1.002 B.N.Bazylev Erosion of dome armour after multiple disruptions and ELMs in ITER P-1.003 I.S.Landman Contamination and radiation losses in post-ELM tokamak plasma P-1.004 T.Lunt Experimental Investigation on the Plasma-Wall Transition P-1.005 TilmannLunt Ion temperature measurements by means of a combined force - Mach - Langmuir probe P-1.006 A.Herrmann Filamentary heat deposition to the first wall in ASDEX Upgrade Fine Structure of Type-I Edge Localized Modes in the Steep Gradient Region in ASDEX P-1.007 B.Kurzan Upgrade P-1.008 D.P.Coster Edge simulations of an ASDEX Upgrade Ohmic shot P-1.009 H.W.Mueller Plasma flow in the scrape-off layer of ASDEX Upgrade P-1.010 R.Dux Tungsten Erosion at Auxiliary Limiters in ASDEX Upgrade P-1.011 V.Rohde Carbon migration at the divertor of ASDEX Upgrade P-1.012 Y.Feng Role of recycling in W7-AS divertor plasmas P-1.013 A.Kirschner Modelling of tritium retention and target lifetime of the ITER divertor Investigation of carbon transport by 13CH4 injection through graphite and tungsten test P-1.014 A.Kreter limiters in TEXTOR Carbon deposition and fuel accumulation in castellated limiters exposed in the SOL of P-1.015 A.Litnovsky TEXTOR P-1.016 C.Busch Impact of the DED on ion transport and poloidal rotation in TEXTOR Modelling of hydrocarbon transport and emission after methane injection into the TEXTOR P-1.017 D.Borodin boundary
    [Show full text]
  • Effect of Periodic Gas-Puffs on Drift-Tearing Modes in ADITYA/ADITYA-U Tokamak Discharges
    Effect of periodic gas-puffs on drift-tearing modes in ADITYA/ADITYA-U Tokamak discharges Harshita Raj1*,2, Tanmay Macwan1,2, Kaushalendra Singh1,2, Suman Dolui1,2, Joydeep Ghosh1,2, Nirmal K Bisai1, 2, K. A. Jadeja2, K. M. Patel2, N. C. Patel2, R. L. Tanna2, D. Raju1,2, S. K. Jha2, P. K. Chattopadhyay1, 2, Abhijit Sen1, 2, Y, C, Saxena1, 2, R. Pal3, and ADITYA-U Team2 1HBNI University, Mumbai, India 2Institute for Plasma Research, Bhat, Gandhinagar 382428, India 3Saha Institute of Nuclear Physics, Kolkata, India Abstract The effect of a periodic train of short gas-puff pulses on the rotation frequency and amplitude of drift-tearing modes has been studied in ADITYA/ADITYA-U tokamak. The short gas puffs, injecting approximately ~1017 – 1018 molecules of fuel gas (hydrogen) at one toroidal location, are found to concomitantly decrease the drift-tearing mode rotation frequency and the mode amplitude during the period of injection and then recover back to its initial values when the gas pulse is over. This leads to a periodic modulation of the rotation frequency and amplitude of the drift-tearing modes that is correlated with the periodicity of the gas pulse injection. The underlying mechanism for this change in the mode characteristic appears to be related to gas puff induced change in the radial profile of the plasma pressure in the edge region that brings about a reduction in the diamagnetic drift frequency. Detailed experimental measurements and BOUT++ code simulations support such a reduction in diamagnetic drift frequency. Our results reveal a close interaction between the edge dynamics and core MHD phenomena in a tokamak that could help us better understand the rotation dynamics and amplitude pulsations of magnetic islands.
    [Show full text]