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Sample records for aditya tokamak research

  1. Diamagnetic flux measurement in Aditya tokamak

    SciTech Connect

    Kumar, Sameer; Jha, Ratneshwar; Lal, Praveen; Hansaliya, Chandresh; Gopalkrishna, M. V.; Kulkarni, Sanjay; Mishra, Kishore

    2010-12-15

    Measurements of diamagnetic flux in Aditya tokamak for different discharge conditions are reported for the first time. The measured diamagnetic flux in a typical discharge is less than 0.6 mWb and therefore it has required careful compensation for various kinds of pick-ups. The hardware and software compensations employed in this measurement are described. We introduce compensation of a pick-up due to plasma current of less than 20 kA in short duration discharges, in which plasma pressure gradient is supposed to be negligible. The flux measurement during radio frequency heating is also presented in order to validate compensation.

  2. Novel approaches for mitigating runaway electrons and plasma disruptions in ADITYA tokamak

    NASA Astrophysics Data System (ADS)

    Tanna, R. L.; Ghosh, J.; Chattopadhyay, P. K.; Dhyani, Pravesh; Purohit, Shishir; Joisa, S.; Rao, C. V. S.; Panchal, V. K.; Raju, D.; Jadeja, K. A.; Bhatt, S. B.; Gupta, C. N.; Chavda, Chhaya; Kulkarni, S. V.; Shukla, B. K.; Praveenlal E., V.; Raval, Jayesh; Amardas, A.; Atrey, P. K.; Dhobi, U.; Manchanda, R.; Ramaiya, N.; Patel, N.; Chowdhuri, M. B.; Jha, S. K.; Jha, R.; Sen, A.; Saxena, Y. C.; Bora, D.; the ADITYA Team

    2015-06-01

    This paper summarizes the results of recent dedicated experiments on disruption control and runaway mitigation carried out in ADITYA, which are of the utmost importance for the successful operation of large size tokamaks, such as ITER. It is quite a well-known fact that disruptions in tokamaks must be avoided. Disruptions, induced by hydrogen gas puffing, are successfully avoided by two innovative techniques in ADITYA using a bias electrode placed inside the last closed flux surface and applying an ion cyclotron resonance pulse with a power of ∼50 to 70 kW. These experiments led to better understanding of the disruption avoidance mechanisms and also can be thought of as one of the options for disruption avoidance in ITER. In both cases, the physical mechanism seems to be the control of magnetohydrodynamic modes due to increased poloidal rotation of edge plasma generated by induced radial electric fields. Real time avoidance of disruption with identifying proper precursors in both the mechanisms is successfully attempted. Further, analysing thoroughly the huge database of different types of spontaneous and deliberately-triggered disruptions from ADITYA, a significant contribution has been made to the international disruption database (ITPA). Furthermore, the mitigation of the runaway electron generated mainly during disruptions remains a challenging topic in present tokamak research as these high-energy electrons can cause severe damage to in-vessel components and the vacuum vessel. A simple technique has been implemented in ADITYA to mitigate the runaway electrons before they can gain high energies using a localized vertical magnetic field perturbation applied at one toroidal location to extract runaway electrons.

  3. Multidirectional plasma flow measurement by Gundestrup Probe in scrape-off layer of ADITYA tokamak

    SciTech Connect

    Sangwan, Deepak; Jha, Ratneshwar; Tanna, Rakesh L.

    2015-11-15

    Multidirectional plasma flow measurements by using Gundestrup Probe in the scrape-off layer of ADITYA tokamak are presented. The ADITYA Gundestrup Probe-head consists of eight plates arranged around the ceramic rod and three pins normal to side plates. Plates are used to measure both parallel and perpendicular flows simultaneously and pins are used to measure plasma density and floating potential. A comparison of direct perpendicular flow measurement and by two other plates of Gundestrup Probe is presented. Possible causes of perpendicular flows are identified and compared with the measured flows. It is observed that the mechanism of the parallel flow and the perpendicular flow is different only at high parallel Mach number. A puff of the working gas is used to study its effect on the perpendicular flows and its reversal with the gas puff is observed.

  4. A set-up for a biased electrode experiment in ADITYA Tokamak

    NASA Astrophysics Data System (ADS)

    Dhyani, Pravesh; Ghosh, Joydeep; Sathyanarayana, K.; Praveenlal, V. E.; Gautam, Pramila; Shah, Minsha; Tanna, R. L.; Kumar, Pintu; Chavda, C.; Patel, N. C.; Panchal, V.; Gupta, C. N.; Jadeja, K. A.; Bhatt, S. B.; Kumar, S.; Raju, D.; Atrey, P. K.; Joisa, S.; Chattopadhyay, P. K.; Saxena, Y. C.

    2014-10-01

    An experimental set-up to investigate the effect of a biased electrode introduced in the edge region on ADITYA tokamak discharges is presented. A specially designed double-bellow mechanical assembly is fabricated for controlling the electrode location as well as its exposed length inside the plasma. The cylindrical molybdenum electrode is powered by a capacitor-bank based pulsed power supply (PPS) using a semiconductor controlled rectifier (SCR) as a switch with forced commutation. A Langmuir probe array for radial profile measurements of plasma potential and density is fabricated and installed. Standard results of improvement of global confinement have been obtained using a biased electrode. In addition to that, in this paper we show for the first time that the same biasing system can be used to avoid disruptions through stabilisation of magnetohydrodynamic (MHD) modes. Real time disruption control experiments have also been carried out by triggering the bias-voltage on the electrode automatically when the Mirnov probe signal exceeds a preset threshold value using a uniquely designed electronic comparator circuit. Most of the results related to the improved confinement and disruption mitigation are obtained in case of the electrode tip being kept at ~3 cm inside the last closed flux surface (LCFS) with an exposed length of ~20 mm in typical discharges of ADITYA tokamak.

  5. Improvement of Plasma Performance with Lithium Wall Conditioning in Aditya Tokamak

    NASA Astrophysics Data System (ADS)

    B. Chowdhuri, M.; Manchanda, R.; Ghosh, J.; B. Bhatt, S.; Ajai, Kumar; K. Das, B.; A. Jadeja, K.; A. Raijada, P.; Manoj, Kumar; Banerjee, S.; Nilam, Ramaiya; Aniruddh, Mali; Ketan, M. Patel; Vinay, Kumar; Vasu, P.; Bhattacharyay, R.; L. Tanna, R.; Y. Shankara, Joisa; K. Atrey, P.; V. S. Rao, C.; Chenna Reddy, D.; K. Chattopadhyay, P.; Jha, R.; C. Saxena, Y.; Aditya Team

    2013-02-01

    Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ~100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ~40% to 50%) and a 20% to 30% decrease of Hα signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Zeff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

  6. Measurement of spatial and temporal behavior of H{sub α} emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array

    SciTech Connect

    Chowdhuri, M. B. Ghosh, J.; Manchanda, R.; Banerjee, S.; Ramaiya, N.; Virani, Niral; Mali, Aniruddh; Amardas, A.; Kumar, Pintu; Tanna, R. L.; Gupta, C. N.; Bhatt, S. B.; Chattopadhyay, P. K.; Kumar, Ajay

    2014-11-15

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ∼3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of H{sub α} emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting H{sub α} emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation.

  7. Measurement of spatial and temporal behavior of H(α) emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array.

    PubMed

    Chowdhuri, M B; Ghosh, J; Manchanda, R; Kumar, Ajay; Banerjee, S; Ramaiya, N; Virani, Niral; Mali, Aniruddh; Amardas, A; Kumar, Pintu; Tanna, R L; Gupta, C N; Bhatt, S B; Chattopadhyay, P K

    2014-11-01

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ∼3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of Hα emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting Hα emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation.

  8. Transport driven plasma flows in the scrape-off layer of ADITYA Tokamak in different orientations of magnetic field

    SciTech Connect

    Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.

    2014-06-15

    Parallel plasma flows in the scrape-off layer of ADITYA tokamak are measured in two orientations of total magnetic field. In each orientation, experiments are carried out by reversing the direction of the toroidal magnetic field and the plasma current. The transport-driven component is determined by averaging flow Mach numbers, measured in two directions of the toroidal magnetic field and the plasma current for the same orientation. It is observed that there is a significant transport-driven component in the measured flow and the component depends on the field orientation.

  9. Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak

    SciTech Connect

    Purohit, S. Joisa, Y. S.; Raval, J. V.; Ghosh, J.; Tanna, R.; Shukla, B. K.; Bhatt, S. B.

    2014-11-15

    Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.

  10. Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.

    PubMed

    Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B

    2014-11-01

    Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.

  11. Direct Electron Heating Observed by Fast Waves in ICRF Range on a Low-Density Low Temperature Tokamak ADITYA

    SciTech Connect

    Mishra, K.; Kulkarni, S.; Rathi, D.; Varia, A.; Jadav, H.; Parmar, K.; Kadia, B.; Joshi, R.; Srinivas, Y.; Singh, R.; Kumar, S.; Dani, S.; Gayatri, A.; Yogi, R.; Singh, M.; Joisa, Y.; Rao, C.; Kumar, S.; Jha, R.; Manchanda, R.

    2011-12-23

    Fast wave electron heating experiments are carried out on Aditya tokamak [R = 0.75 m, a = 0.25m,Bt = 0.75T,ne{approx}1-3E13/cc,Te{approx}250eV] with the help of indigenously developed 200 kW, 20-40 MHz RF heating system. Significant direct electron heating is observed by fast waves in hydrogen plasma with prompt rise in electron temperature with application of RF power and it increases linearly with RF power. A corresponding increase in plasma beta and hence increase in stored diamagnetic energy is also observed in presence of RF. We observe an improvement of energy confinement time from 2-4msec during ohmic heating phase to 3-6msec in RF heating phase. This improvement is within the ohmic confinement regime for the present experiments. The impurity radiation and electron density do not escalate significantly with RF power. The direct electron heating by fast wave in Aditya is also predicted by ion cyclotron resonance heating code TORIC.

  12. Core-ion temperature measurement of the ADITYA tokamak using passive charge exchange neutral particle energy analyzer

    SciTech Connect

    Pandya, Santosh P.; Ajay, Kumar; Mishra, Priyanka; Dhingra, Rajani D.; Govindarajan, J.

    2013-02-15

    Core-ion temperature measurements have been carried out by the energy analysis of passive charge exchange (CX) neutrals escaping out of the ADITYA tokamak plasma (minor radius, a= 25 cm and major radius, R= 75 cm) using a 45 Degree-Sign parallel plate electrostatic energy analyzer. The neutral particle analyzer (NPA) uses a gas cell configuration for re-ionizing the CX-neutrals and channel electron multipliers (CEMs) as detectors. Energy calibration of the NPA has been carried out using ion-source and {Delta}E/E of high-energy channel has been found to be {approx}10%. Low signal to noise ratio (SNR) due to VUV reflections on the CEMs was identified during the operation of the NPA with ADITYA plasma discharges. This problem was rectified by upgrading the system by incorporating the additional components and arrangements to suppress VUV radiations and improve its VUV rejection capabilities. The noise rejection capability of the NPA was experimentally confirmed using a standard UV-source and also during the plasma discharges to get an adequate SNR (>30) at the energy channels. Core-ion temperature T{sub i}(0) during flattop of the plasma current has been measured to be up to 150 eV during ohmically heated plasma discharges which is nearly 40% of the average core-electron temperature (typically T{sub e}(0) {approx} 400 eV). The present paper describes the principle of tokamak ion temperature measurement, NPA's design, development, and calibration along with the modifications carried out for minimizing the interference of plasma radiations in the CX-spectrum. Performance of the NPA during plasma discharges and experimental results on the measurement of ion-temperature have also been reported here.

  13. The Aditya-L1 Mission of Indian Space Research Organization

    NASA Astrophysics Data System (ADS)

    Tripathi, Durgesh

    2016-07-01

    The Aditya-L1 is the first mission of the Indian Space Research Organization (ISRO) dedicated to solar observations. The spacecraft will be located at the first Lagrangian point and will provide continuous observations of the Sun using remote sensing as well as in-situ measurements. The spacecraft will carry 7 payloads including a coronagraph that will image the corona in visible and IR wavelength and will provide measurements of coronal magnetic field and will study the dynamics of coronal mass ejections; a NUV imaging telescope to study the coupling between solar photosphere and chromosphere and to measure spatially resolved solar spectral irradiance and its variation. There will be two payloads to study the soft X-ray and hard X-ray emission from the Sun, two payloads for in-situ measurements of the charged particles and a magnetometer to study the magnetic field variations during energetic events. Some of the salient features of the experiments on board Aditya-L1 mission will be discussed.

  14. Influence of Wall Conditioning on ADITYA Plasma Discharges

    NASA Astrophysics Data System (ADS)

    Tanna, R. L.; Jadeja, K. A.; Bhatt, S. B.; Bawankar, P. S.; Gupta, C. N.; Joisa, Y. S.; Atrey, P. K.; Manchanda, R.; Ramaiya, Nilam; Ghosh, J.; Raju, D.; Chattopadhyay, P. K.; Jha, R.; the Aditya Team

    2012-11-01

    ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent operational campaign, various experiments have been carried out to enhance the discharge performance as well as improve the plasma parameters. A comparative plasma discharges study with SiC and Graphite limiter was carried out to increase the plasma heating and reduce runaways. Excellent plasma heating has been observed in many discharges using Graphite limiter. Good repeatability of low hard X-rays, high temperature discharges was obtained. The control of plasma impurities and hydrogen recycling is very much essential for high performance discharges. The wall conditioning in ADITYA tokamak is carried out by hydrogen glow discharge cleaning (GDC), Pulse discharge cleaning and electron cyclotron resonance (ECR) discharge cleaning techniques with and without lithium wall coating. GDC assisted Lithiumization was found to be the most effective technique for substantial reduction in Ha and low Z (CIII & O-I) impurities. The partial pressure of mass number 18 (H2O) and 28 (N2/C2H4/CO) were regularly monitored before plasma discharge operation. Furthermore, experiment on optimization of pulse gas feed was helped in reducing wall loading and recycling. However, hard X-rays suppression with the application of multiple gas puff has been successfully achieved during negative converter operation, which led to the extension of plasma pulse length up to ~ 250 ms. All the supporting facts and operation aspects are reported.

  15. Study of Laser Produced Plasma of Limiter of the Aditya Tokomak for Detection of Molecular Bands

    NASA Astrophysics Data System (ADS)

    Rai, Awadhesh Kumar

    2016-06-01

    The tokamak wall protection is one of the prime concerns, and for this purpose, limiters are used. Graphite is commonly used as a limiter material and first wall material for complete coverage of the internal vacuum vessel surfaces of the tokamak. From the past few years, we are working to identify and quantify the impurities deposited on the different part of Aditya Tokamak in collaboration with the Scientists at Institute of Plasma Research, Ahmedabad, India using Laser Induced Breakdown Spectroscopy (LIBS) [1-3]. Laser induced breakdown spectroscopy (LIBS) spectra of limiter of Aditya Tokamak have been recorded in the spectral range of 200-900 nm in open atmosphere. Along with atomic and ionic spectral lines of the constituent elements of the limiter (1-3), LIBS spectra also give the molecular bands. When a high power laser beam is focused on the sample, laser induced plasma is produced on its surface. In early stage of the plasma Back ground continuum is dominated due to free-free or free-bound emission. Just after few nanoseconds the light from the plasma is dominated by ionic emission. Atomic emission spectra is dominated from the laser induced plasma during the first few microsecond after an ablation pulse where as molecular spectra is generated later when the plasma further cools down. For this purpose the LIBS spectra has been recorded with varying gate delay and gate width. The spectra of the limiter show the presence of molecular bands of CN and C2. To get better signal to background ratios of the molecular bands, different experimental parameters like gate delay, gate width, collection angle and collection point (spatial analysis off the plasama) of the plasma have been optimized. Thus the present paper deals with the variation of spectral intensity of the molecular bands with different experimental parameters. Keywords: Limiter, Molecular bands, C2, CN. References: 1. Proof-of-concept experiment for On-line LIBS Analysis of Impurity Layer Deposited on

  16. (High beta tokamak research and plasma theory)

    SciTech Connect

    Not Available

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report.

  17. Summer Research Experiences with a Laboratory Tokamak

    NASA Astrophysics Data System (ADS)

    Farley, N.; Mauel, M.; Navratil, G.; Cates, C.; Maurer, D.; Mukherjee, S.; Shilov, M.; Taylor, E.

    1998-11-01

    Columbia University's Summer Research Program for Secondary School Science Teachers seeks to improve middle and high school student understanding of science. The Program enhances science teachers' understanding of the practice of science by having them participate for two consecutive summers as members of laboratory research teams led by Columbia University faculty. In this poster, we report the research and educational activities of two summer internships with the HBT-EP research tokamak. Research activities have included (1) computer data acquisition and the representation of complex plasma wave phenomena as audible sounds, and (2) the design and construction of pulsed microwave systems to experience the design and testing of special-purpose equipment in order to achieve a specific technical goal. We also present an overview of the positive impact this type of plasma research involvement has had on high school science teaching.

  18. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

    SciTech Connect

    Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar; Kumar, Ajai

    2015-12-15

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.

  19. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

    NASA Astrophysics Data System (ADS)

    Maurya, Gulab Singh; Kumar, Rohit; Kumar, Ajai; Rai, Awadhesh Kumar

    2015-12-01

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as "back collection method" to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.

  20. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak.

    PubMed

    Maurya, Gulab Singh; Kumar, Rohit; Kumar, Ajai; Rai, Awadhesh Kumar

    2015-12-01

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as "back collection method" to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.

  1. DIII-D Advanced Tokamak Research Overview

    SciTech Connect

    V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler

    1999-12-01

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.

  2. Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak

    NASA Astrophysics Data System (ADS)

    Seo, Seong-Heon; Park, Jinhyung; Wi, H. M.; Lee, W. R.; Kim, H. S.; Lee, T. G.; Kim, Y. S.; Kang, Jin-Seob; Bog, M. G.; Yokota, Y.; Mase, A.

    2013-08-01

    Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 1019 m-3 when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.

  3. Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak

    SciTech Connect

    Seo, Seong-Heon; Wi, H. M.; Lee, W. R.; Kim, H. S.; Lee, T. G.; Kim, Y. S.; Park, Jinhyung; Kang, Jin-Seob; Bog, M. G.; Yokota, Y.; Mase, A.

    2013-08-15

    Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6–54 GHz), V band (48–72 GHz), and W band (72–108 GHz) to measure the density up to 7 × 10{sup 19} m{sup −3} when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.

  4. Overview of physics research on the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Fasoli, A.; TCV Team

    2009-10-01

    The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, 'bootstrap-aligned' equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch-Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears

  5. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    SciTech Connect

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-15

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  6. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    NASA Astrophysics Data System (ADS)

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  7. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method.

    PubMed

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  8. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method.

    PubMed

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method. PMID:26724028

  9. Generation of multiple analog pulses with different duty cycles within VME control system for ICRH Aditya system

    NASA Astrophysics Data System (ADS)

    Joshi, Ramesh; Singh, Manoj; Jadav, H. M.; Misra, Kishor; Kulkarni, S. V.; ICRH-RF Group

    2010-02-01

    Ion Cyclotron Resonance Heating (ICRH) is a promising heating method for a fusion device due to its localized power deposition profile, a direct ion heating at high density, and established technology for high RF power generation and transmission at low cost. Multiple analog pulse with different duty cycle in master of digital pulse for Data acquisition and Control system for steady state RF ICRH System(RF ICRH DAC) to be used for operating of RF Generator in Aditya to produce pre ionization and second analog pulse will produce heating. The control system software is based upon single digital pulse operation for RF source. It is planned to integrate multiple analog pulses with different duty cycle in master of digital pulse for Data acquisition and Control system for RF ICRH System(RF ICRH DAC) to be used for operating of RF Generator in Aditya tokamak. The task of RF ICRH DAC is to control and acquisition of all ICRH system operation with all control loop and acquisition for post analysis of data with java based tool. For pre ionization startup as well as heating experiments using multiple RF Power of different powers and duration. The experiment based upon the idea of using single RF generator to energize antenna inside the tokamak to radiate power twise, out of which first analog pulse will produce pre ionization and second analog pulse will produce heating. The whole system is based on standard client server technology using tcp/ip protocol. DAC Software is based on linux operating system for highly reliable, secure and stable system operation in failsafe manner. Client system is based on tcl/tk like toolkit for user interface with c/c++ like environment which is reliable programming languages widely used on stand alone system operation with server as vxWorks real time operating system like environment. The paper is focused on the Data acquisition and monitoring system software on Aditya RF ICRH System with analog pulses in slave mode with digital pulse in

  10. Overview of recent and current research on the TCV tokamak

    NASA Astrophysics Data System (ADS)

    S. Codathe TCV Team

    2013-10-01

    Through a diverse research programme, the Tokamak à Configuration Variable (TCV) addresses physics issues and develops tools for ITER and for the longer term goals of nuclear fusion, relying especially on its extreme plasma shaping and electron cyclotron resonance heating (ECRH) launching flexibility and preparing for an ECRH and NBI power upgrade. Localized edge heating was unexpectedly found to decrease the period and relative energy loss of edge localized modes (ELMs). Successful ELM pacing has been demonstrated by following individual ELM detection with an ECRH power cut before turning the power back up to trigger the next ELM, the duration of the cut determining the ELM period. Negative triangularity was also seen to reduce the ELM energy release. H-mode studies have focused on the L-H threshold dependence on the main ion species and on the divertor leg length. Both L- and H-modes have been explored in the snowflake configuration with emphasis on edge measurements, revealing that the heat flux to the strike points on the secondary separatrix increases as the X-points approach each other, well before they coalesce. In L-mode, a systematic scan of the auxiliary power deposition profile, with no effect on confinement, has ruled it out as the cause of confinement degradation. An ECRH power absorption observer based on transmitted stray radiation was validated for eventual polarization control. A new profile control methodology was introduced, relying on real-time modelling to supplement diagnostic information; the RAPTOR current transport code in particular has been employed for joint control of the internal inductance and central temperature. An internal inductance controller using the ohmic transformer has also been demonstrated. Fundamental investigations of neoclassical tearing mode (NTM) seed island formation by sawtooth crashes and of NTM destabilization in the absence of a sawtooth trigger were carried out. Both stabilizing and destabilizing agents

  11. Aditya: India’s First Observatory in Space to Study the Sun

    NASA Astrophysics Data System (ADS)

    Nandi, Dibyendu

    2015-08-01

    Recognizing the need and advantages of continuous solar observations from space, and to further its goal of supporting scientific and technological advances, the Indian Space Research Organization is planning India’s first space mission to observe the Sun. Nicknamed Aditya, this ambitious project aims to place a comprehensive solar observatory at the Lagrange point L1 which will allow uninterrupted views of the Sun. The diverse set of instruments being planned to fly onboard this mission includes a visible emission line coronagraph, a solar ultraviolet imaging telescope, high- and low-energy X-ray spectrometers, a plasma analyzer and a particle detector package for in-situ measurements. In this talk I will provide a brief overview of these instruments and discuss the science objectives of this mission.

  12. Indian Solar mission to study inner solar corona: Aditya 1

    NASA Astrophysics Data System (ADS)

    Singh, Jagdev; Banerjee, Dipankar; Venkatakrishnan, Parameswaran; Kasiviswanathan, Sankarasubramanian; Prasad B, Raghavendra

    2012-07-01

    Aditya-I is India's first dedicated scientific mission to study the sun. This is a low-earth orbit (LEO) mission at an altitude of 800 km. A visible emission line space solar coronagraph (VELC) has been selected as a payload under the small-satellite program of ISRO. It will provide high time cadence sharp images of the solar corona in the Green and Red Emission lines. These images will be used to study the highly dynamic nature of the solar corona including the small-scale coronal loops and large-scale Coronal Mass Ejections (CMEs). The uniqueness of this payload compared to previously flown space instruments are: (a) Observations in the visible wavelength closer to the disk (down to 1.05 solar radii), (b) high time cadence capability (better than 2-images per second), and (c) Simultaneous observations of at least two spectral windows all the time and three spectral windows for short durations. I will update the current status of the project and will point out the complimentary role Aditya can play in conjunction with other solar big missions like SDO.

  13. Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

    SciTech Connect

    Wootton, A.J.

    1994-07-01

    The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described.

  14. Analysis of line integrated electron density using plasma position data on Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Nam, Y. U.; Chung, J.

    2010-10-15

    A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.

  15. Optical system design for the charge exchange spectroscopy of the Korea superconducting tokamak advanced research device

    NASA Astrophysics Data System (ADS)

    Oh, Seungtae; Ko, Won-Ha

    2011-04-01

    The collective optical design is described for the charge exchange spectroscopy (CES) of the Korea superconducting tokamak advanced research (KSTAR) device. The CES diagnostic measures the ion temperature of carbon and other impurities, in conjunction with the neutral heating beam in KSTAR. The visible light from the plasma is concentrated via collection optics and imaged onto quartz fibers. The collection optics in the system is the key component for the CES system. The final design is derived through four steps and its performance is examined in a simulation step. In this paper, the design details of the collective optical system for the KSTAR CES are discussed.

  16. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    SciTech Connect

    Wootton, A.J.

    1995-08-01

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.

  17. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research.

    PubMed

    Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S

    2015-07-01

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties. PMID:26233377

  18. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    NASA Astrophysics Data System (ADS)

    Lampert, M.; Anda, G.; Czopf, A.; Erdei, G.; Guszejnov, D.; Kovácsik, Á.; Pokol, G. I.; Réfy, D.; Nam, Y. U.; Zoletnik, S.

    2015-07-01

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  19. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Lampert, M.; Anda, G.; Réfy, D.; Zoletnik, S.; Czopf, A.; Erdei, G.; Guszejnov, D.; Kovácsik, Á.; Pokol, G. I.; Nam, Y. U.

    2015-07-15

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  20. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research.

    PubMed

    Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S

    2015-07-01

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  1. 20 years of research on the Alcator C-Mod tokamak

    SciTech Connect

    Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; and others

    2014-11-15

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  2. Twenty Years of Research on the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin

    2013-10-01

    Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

  3. Investigation of relativistic runaway electrons in electron cyclotron resonance heating discharges on Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Kang, C. S.; Lee, S. G.

    2014-07-15

    The behavior of relativistic runaway electrons during Electron Cyclotron Resonance Heating (ECRH) discharges is investigated in the Korea Superconducting Tokamak Advanced Research device. The effect of the ECRH on the runaway electron population is discussed. Observations on the generation of superthermal electrons during ECRH will be reported, which will be shown to be consistent with existing theory for the development of a superthermal electron avalanche during ECRH [A. Lazaros, Phys. Plasmas 8, 1263 (2001)].

  4. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    SciTech Connect

    none,

    1996-05-31

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in – vii – DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  5. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Yu Yaowei; Kim, Young-Ok; Kim, Hak-Kun; Kim, Hong-Tack; Kim, Woong-Chae; Kim, Kwang-Pyo; Son, Soo-Hyun; Bang, Eun-Nam; Hong, Suk-Ho; Yoon, Si-Woo; Zhuang Huidong; Chen Zhongyong

    2012-12-15

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is {approx}3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2-3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1-1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D{sub 2} MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  6. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research

    NASA Astrophysics Data System (ADS)

    Yu, Yaowei; Kim, Young-Ok; Kim, Hak-Kun; Kim, Hong-Tack; Kim, Woong-Chae; Kim, Kwang-Pyo; Son, Soo-Hyun; Bang, Eun-Nam; Zhuang, Huidong; Chen, Zhongyong; Hong, Suk-Ho; Yoon, Si-Woo

    2012-12-01

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is ˜3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2-3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1-1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D2 MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  7. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research.

    PubMed

    Yu, Yaowei; Kim, Young-Ok; Kim, Hak-Kun; Kim, Hong-Tack; Kim, Woong-Chae; Kim, Kwang-Pyo; Son, Soo-Hyun; Bang, Eun-Nam; Zhuang, Huidong; Chen, Zhongyong; Hong, Suk-Ho; Yoon, Si-Woo

    2012-12-01

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is ~3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2-3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1-1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D(2) MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  8. Bibliography of fusion product physics in tokamaks

    SciTech Connect

    Hively, L. M.; Sigmar, D. J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category.

  9. The high beta tokamak-extended pulse magnetohydrodynamic mode control research program

    NASA Astrophysics Data System (ADS)

    Maurer, D. A.; Bialek, J.; Byrne, P. J.; De Bono, B.; Levesque, J. P.; Li, B. Q.; Mauel, M. E.; Navratil, G. A.; Pedersen, T. S.; Rath, N.; Shiraki, D.

    2011-07-01

    The high beta tokamak-extended pulse (HBT-EP) magnetohydrodynamic (MHD) mode control research program is studying ITER relevant internal modular feedback control coil configurations and their impact on kink mode rigidity, advanced digital control algorithms and the effects of plasma rotation and three-dimensional magnetic fields on MHD mode stability. A new segmented adjustable conducting wall has been installed on the HBT-EP and is made up of 20 independent, movable, wall shell segments instrumented with three distinct sets of 40 saddle coils, totaling 120 in-vessel modular feedback control coils. Each internal coil set has been designed with varying toroidal angular coil coverage of 5, 10 and 15°, spanning the toroidal angle range of an ITER port plug based internal coil to test resistive wall mode (RWM) interaction and multimode MHD plasma response to such highly localized control fields. In addition, we have implemented 336 new poloidal and radial magnetic sensors to quantify the applied three-dimensional fields of our control coils along with the observed plasma response. This paper describes the design and implementation of the new control shell incorporating these control and sensor coils on the HBT-EP, and the research program plan on the upgraded HBT-EP to understand how best to optimize the use of modular feedback coils to control instability growth near the ideal wall stabilization limit, answer critical questions about the role of plasma rotation in active control of the RWM and the ferritic resistive wall mode, and to improve the performance of MHD control systems used in fusion experiments and future burning plasma systems.

  10. DIII-D research towards resolving key issues for ITER and steady-state tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; the DIII-D Team

    2013-10-01

    The DIII-D research program is addressing key ITER research needs and developing the physics basis for future steady-state tokamaks. Pellet pacing edge-localized mode (ELM) control in the ITER configuration reduces ELM energy loss in proportion to 1/fpellet by inducing ELMs at up to 12× the natural ELM rate. Complete suppression of ELMs with resonant magnetic perturbations has been extended to the q95 expected for ITER baseline scenario discharges, and long-duration ELM-free QH-mode discharges have been produced with ITER-relevant co-current neutral-beam injection (NBI) using external n = 3 coils to generate sufficient counter-Ip torque. ITER baseline discharges at βN ˜ 2 and scaled NBI torque have been maintained in stationary conditions for more than four resistive times using electron cyclotron current drive (ECCD) for tearing mode suppression and disruption avoidance; active tracking with steerable launchers and feedback control catch these modes at small amplitude, reducing the ECCD power required to suppress them. Massive high-Z gas injection into disruption-induced 300-600 kA 20 MeV runaway electron (RE) beams yield dissipation rates ˜10× faster than expected from e-e collisions and demonstrate the possibility of benign dissipation of such REs should they occur in ITER. Other ITER-related experiments show measured intrinsic plasma torque in good agreement with a physics-based model over a wide range of conditions, while first-time main-ion rotation measurements show it to be lower than expected from neoclassical theory. Core turbulence measurements show increased temperature fluctuations correlated with sharply enhanced electron transport when \

  11. Role of neutral gas in scrape-off layer tokamak plasma

    SciTech Connect

    Bisai, N.; Jha, R.; Kaw, P. K.

    2015-02-15

    Neutral gas in scrape-off layer of tokamak plasma plays an important role as it can modify the plasma turbulence. In order to investigate this, we have derived a simple two-dimensional (2D) model that consists of electron continuity, quasi-neutrality, and neutral gas continuity equations using neutral gas ionization and charge exchange processes. Simple 1D profile analysis predicts neutral penetration depth into the plasma. Growth rate obtained from the linear theory has been presented. The 2D model equations have been solved numerically. It is found that the neutral gas reduces plasma fluctuations and shifts spectrum of the turbulence towards lower frequency side. The neutral gas fluctuation levels have been presented. The numerical results have been compared with Aditya tokamak experiments.

  12. Texas Experimental Tokamak

    SciTech Connect

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.

  13. ARIES tokamak reactor study

    SciTech Connect

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein.

  14. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    NASA Astrophysics Data System (ADS)

    Lee, H. Y.; Hahn, S. H.; Ghim, Y.-C.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho

    2015-12-01

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2-3 and the ELM size, which was estimated from the Dα amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34-0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.

  15. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research

    NASA Astrophysics Data System (ADS)

    Lee, W.; Park, H. K.; Lee, D. J.; Nam, Y. U.; Leem, J.; Kim, T. K.

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm-1. The upper limit corresponds to the normalized wavenumber kθρe of ˜0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.

  16. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research.

    PubMed

    Lee, W; Park, H K; Lee, D J; Nam, Y U; Leem, J; Kim, T K

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm(-1). The upper limit corresponds to the normalized wavenumber kθρe of ∼0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.

  17. ECH tokamak

    SciTech Connect

    Firestone, M.A.; Mau, T.K.; Conn, R.W.

    1985-04-01

    A small steady-state tokamak capable of producing power in the 100 to 300 MWe range and relying on electron cyclotron RF heating (ECH) for both heating and current drive is described. Working in the first MHD stability regime for tokamaks, the approach adheres to the recently discovered maximum beta limit. An appropriate figure of merit is the ratio of the fusion power to absorbed RF power. Efficient devices are feasible at both small and large values of fusion power, thereby pointing to a development path for an attractive commercial fusion reactor.

  18. Fluid-particle hybrid simulation on the transports of plasma, recycling neutrals, and carbon impurities in the Korea Superconducting Tokamak Advanced Research divertor region

    NASA Astrophysics Data System (ADS)

    Kim, Deok-Kyu; Hong, Sang Hee

    2005-06-01

    A two-dimensional simulation modeling that has been performed in a self-consistent way for analysis on the fully coupled transports of plasma, recycling neutrals, and intrinsic carbon impurities in the divertor domain of tokamaks is presented. The numerical model coupling the three major species transports in the tokamak edge is based on a fluid-particle hybrid approach where the plasma is described as a single magnetohydrodynamic fluid while the neutrals and impurities are treated as kinetic particles using the Monte Carlo technique. This simulation code is applied to the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak [G. S. Lee, J. Kim, S. M. Hwang et al., Nucl. Fusion 40, 575 (2000)] to calculate the peak heat flux on the divertor plate and to explore the divertor plasma behavior depending on the upstream conditions in its base line operation mode for various values of input heating power and separatrix plasma density. The numerical modeling for the KSTAR tokamak shows that its full-powered operation is subject to the peak heat loads on the divertor plate exceeding an engineering limit, and reveals that the recycling zone is formed in front of the divertor by increasing plasma density and by reducing power flow into the scrape-off layer. Compared with other researchers' work, the present hybrid simulation more rigorously reproduces severe electron pressure losses along field lines by the presence of recycling zone accounting for the transitions between the sheath limited and the detached divertor regimes. The substantial profile changes in carbon impurity population and ionic composition also represent the key features of this divertor regime transition.

  19. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  20. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho; Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H.; Ghim, Y.-C.

    2015-12-15

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the D{sub α} amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.

  1. Planned Visible Emission Line Space Solar Coronagraph on-board Aditya-1

    NASA Astrophysics Data System (ADS)

    Singh, Jagdev

    2012-07-01

    An imaging visible emission line internally occulted coronagraph using 20 cm off axis parabolic mirror has been designed and planned to be launched in 2014. The coronagraph will have the facility to take images of the solar simultaneously, in the green [Fe xiv] and the red [Fe x] emission lines up to 1.5 solar radii with a frequency of about 3 Hz using 0.5 nm pass band filters and the images in continuum at 580 nm up to 3 solar radii. The satellite has been named as Aditya-1 and the scientific objectives of this payload are: (i) to investigate the existence of intensity oscillations for the study of wave driven coronal heating, (ii) to study the dynamics and formation of coronal loops and temperature structure of the coronal features, (iii) to study the origin, cause and acceleration of Coronal Mass Ejections (CME's) and other solar active features, and (iv) Coronal magnetic field topology and the 3-dimensional structures of the CMEs using polarization information. The fabrication of the pay load will be done in the laboratories of LEOS, SAC, ISAC, IIA and USO and launched by ISRO. Here we shall discuss the design and the realization of the mission.

  2. Power and particle exhaust in tokamaks

    SciTech Connect

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  3. Tokamak Systems Code

    SciTech Connect

    Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged.

  4. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  5. Progress and prospects in understanding the physics of tokamak experiments

    SciTech Connect

    Hutchinson, I.

    1992-12-01

    A whistle-stop tour of the diverse physics of tokamak plasma confinement. This talk will illustrate the way in which fusion research on tokamaks has led to important and interesting physics results, and discuss some of the scientific challenges still ahead before fusion`s potential can be established.

  6. Do spherical tokamaks have a thermonuclear future?

    NASA Astrophysics Data System (ADS)

    Mirnov, S. V.

    2012-12-01

    This work has been initiated by the publication of a review by B.V.Kuteev et al., "Intense Fusion Neutron Sources" [Plasma Physics Reports 36, 281 (2010)]. It is stated that the key thesis of the above review that a spherical tokamak can be recommended for research neutron sources and for demonstration hybrid systems as an alternative to expensive "classical" tokamaks of the JET and ITER type is inconsistent. The analysis of the experimental material obtained during the last 10 years in the course of studies on the existing spherical tokamaks shows that the TIN-ST fusion neutron source spherical tokamak proposed by the authors of the review and intended, according to the authors' opinion, to replace "monsters" in view of its table-top dimensions (2 m3) and laboratory-level energetics cannot be transformed into any noticeable stationary megawatt-power neutron source competing with the existing classical tokamaks (in particular, with JET with its quasi-steady DT fusion power at a level of 5 MW). Namely, the maximum plasma current in the proposed tokamak will be not 3 MA, as the authors suppose erroneously, but, according to the present-day practice of spherical tokamaks, within 0.6-0.7 MA, which will lead to a reduction on the neutron flux by two to three orders of magnitude from the expected 5 MW. The possibility of the maintenance of the stationary process itself even in such a "weakened" spherical tokamak is very doubtful. The experience of the largest existing devices of this type (such as NSTX and MAST) has shown that they are incapable of operating even in a quasi-steady operating mode, because the discharge in them is spontaneously interrupted about 1 s after the beginning of the current pulse, although its expected duration is of up to 5 s. The nature of this phenomenon is the subject of further study of the physics of spherical tokamaks. This work deals with a critical analysis of the available experimental data concerning such tokamaks and a discussion of

  7. Tokamak ARC damage

    SciTech Connect

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  8. Quantify Plasma Response to Non-Axisymmetric (3D) Magnetic Fields in Tokamaks, Final Report for FES (Fusion Energy Sciences) FY2014 Joint Research Target

    SciTech Connect

    Strait, E. J.; Park, J. -K.; Marmar, E. S.; Ahn, J. -W.; Berkery, J. W.; Burrell, K. H.; Canik, J. M.; Delgado-Aparicio, L.; Ferraro, N. M.; Garofalo, A. M.; Gates, D. A.; Greenwald, M.; Kim, K.; King, J. D.; Lanctot, M. J.; Lazerson, S. A.; Liu, Y. Q.; Lore, J. D.; Menard, J. E.; Nazikian, R.; Shafer, M. W.; Paz-Soldan, C.; Reiman, A. H.; Rice, J. E.; Sabbagh, S. A.; Sugiyama, L.; Turnbull, A. D.; Volpe, F.; Wang, Z. R.; Wolfe, S. M.

    2014-09-30

    The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10-4 of the main axisymmetric field, such “3D” fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data, in

  9. Modular tokamak magnetic system

    DOEpatents

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  10. Bifurcated helical core equilibrium states in tokamaks

    NASA Astrophysics Data System (ADS)

    Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

    2013-07-01

    Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

  11. A rotating directional probe for the measurements of fast ion losses and plasma rotation at Tokamak Experiment for Technology Oriented Research.

    PubMed

    Rack, M; Liang, Y; Jaegers, H; Assmann, J; Satheeswaran, G; Xu, Y; Pearson, J; Yang, Y; Denner, P; Zeng, L

    2013-08-01

    This work discusses a new directional probe designed for measurements of fast ion losses and the plasma rotation with a high angular resolution in magnetically confined plasmas. Directional and especially Mach probes are commonly used diagnostics for plasma flow measurements, and their applicability for the fast ion losses detection has been demonstrated. A limitation of static Mach probes is their low angular resolution. At the Tokamak Experiment for Technology Oriented Research, the angular resolution is strongly restricted by the finite number of available measurement channels. In a dynamic plasma, where instabilities can lead to local changes of the field line pitch-angle, plasma flow, or fast ion losses, a low angular resolution makes a precise data analysis difficult and reduces the quality of the measured data. The new probe design, the rotating directional probe, combines the features of early directional probes and Mach probes. It consists of two radially aligned arrays of nine Langmuir probe pins with each array facing opposite directions. During the measurement the probe head rotates along its axis to measure the ion saturation current from all directions. As a result, the rotating directional probe simultaneously provides an angular dependent plasma flow and fast ion losses measurement at different radial positions. Based on the angular dependent data, a precise determination of the current density is made. In addition, the simultaneous measurement of the ion saturation current at different radial positions allows for resolving radially varying field line pitch-angles and identifying the radial dynamic of processes like fast ion losses.

  12. Electron cyclotron emission diagnostics on KSTAR tokamak.

    PubMed

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  13. Electron cyclotron emission diagnostics on KSTAR tokamak

    SciTech Connect

    Jeong, S. H.; Lee, K. D.; Kwon, M.; Kogi, Y.; Kawahata, K.; Nagayama, Y.; Mase, A.

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  14. Origin of Tokamak Density Limit Scalings

    NASA Astrophysics Data System (ADS)

    Gates, D. A.; Delgado-Aparicio, L.

    2012-04-01

    The onset criterion for radiation driven islands [P. H. Rebut and M. Hugon, Plasma Physics and Controlled Nuclear Fusion Research 1984: Proc. 10th Int. Conf. London, 1984, (IAEA, Vienna, 1985), Vol. 2] in combination with a simple cylindrical model of tokamak current channel behavior is consistent with the empirical scaling of the tokamak density limit [M. Greenwald, Nucl. Fusion 28, 2199 (1988)NUFUAU0029-551510.1088/0029-5515/28/12/009]. Many other unexplained phenomena at the density limit are consistent with this novel physics mechanism.

  15. Advanced commercial tokamak study

    SciTech Connect

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.

  16. Completely bootstrapped tokamak

    SciTech Connect

    Weening, R.H. ); Boozer, A.H. )

    1992-01-01

    Numerical simulations of the evolution of large-scale magnetic fields have been developed using a mean-field Ohm's law. The Ohm's law is coupled to a {Delta}{prime} stabilty analysis and a magnetic island growth equation in order to simulate the behavior of tokamak plasmas that are subject to tearing modes. In one set of calculations, the magnetohydrodynamic (MHD)-stable regime of the tokamak is examined via the construction of an {ital l}{sub {ital i}} -{ital q}{sub {ital a}} diagram. The results confirm previous calculations that show that tearing modes introduce a stability boundary into the {ital l}{sub {ital i}} -{ital q}{sub {ital a}} space. In another series of simulations, the interaction between tearing modes and the bootstrap current is investigated. The results indicate that a completely bootstrapped tokamak may be possible, even in the absence of any externally applied loop voltage or current drive.

  17. Physics of Tokamak Plasma Start-up

    NASA Astrophysics Data System (ADS)

    Mueller, Dennis

    2012-10-01

    This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.

  18. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    NASA Astrophysics Data System (ADS)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  19. UCLA Tokamak Program Close Out Report.

    SciTech Connect

    Taylor, Robert John

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  20. A hypothesis of inductive drive to explain the sawtooth measurements of tokamak experiment for technology oriented research (TEXTOR)

    NASA Astrophysics Data System (ADS)

    Chu, T. K.

    2006-07-01

    A hypothesis, based on the current density profile determined from the principle of minimum dissipation of magnetic energy, is applied to explain the measurement of q (0) and current variation in a sawtooth cycle in tokomak experiment for technology oriented research (TEXTOR) [Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1985), Vol. I, p. 193]. A sawtooth oscillation is triggered when the on-axis current density in a configuration with m =0 and n =0 symmetry is driven inductively to a limit.

  1. A hypothesis of inductive drive to explain the sawtooth measurements of tokamak experiment for technology oriented research (TEXTOR)

    SciTech Connect

    Chu, T. K.

    2006-07-15

    A hypothesis, based on the current density profile determined from the principle of minimum dissipation of magnetic energy, is applied to explain the measurement of q(0) and current variation in a sawtooth cycle in tokomak experiment for technology oriented research (TEXTOR) [Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1985), Vol. I, p. 193]. A sawtooth oscillation is triggered when the on-axis current density in a configuration with m=0 and n=0 symmetry is driven inductively to a limit.

  2. Tearing Modes in Tokamaks

    SciTech Connect

    White, R. B.

    2008-05-14

    This lecture gives a basic introduction to magnetic pound elds, magnetic surface destruction, toroidal equilibrium and tearing modes in a tokamak, including the linear and nonlinear development of these modes and their modi pound cation by current drive and bootstrap current, and sawtooth oscillations and disruptions.

  3. High Beta Tokamaks

    SciTech Connect

    Cowley, S.

    1998-11-14

    Perhaps the ideal tokamak would have high {beta} ({beta} {approx}> 1) and classical confinement. Such a tokamak has not been found, and we do not know if one does exist. We have searched for such a possibility, so far without success. In 1990, we obtained analytic equilibrium solutions for large aspect ratio tokamaks at {beta} {approx} {Omicron}(1) [1]. These solutions and the extension at high {beta} poloidal to finite aspect ratio [2] provided a basis for the study of high {beta} tokamaks. We have shown that these configurations can be stable to short scale MHD modes [3], and that they have reduced neoclassical transport [4]. Microinstabilities (such as the {del}T{sub i} mode) seem to be stabilized at high {beta} [5] - this is due to the large local shear [3] and the magnetic well. We have some concerns about modes associated with the compressional branch which may appear at high {beta}. Bill Dorland and Mike Kotschenreuther have studied this issue and our concerns may be unfounded. It is certainly tantalizing, especially given the lowered neoclassical transport values, that these configurations could have no microinstabilities and, one could assume, no anomalous transport. Unfortunately, while this work is encouraging, the key question for high {beta} tokamaks is the stability to large scale kink modes. The MHD {beta} limit (Troyon limit) for kink modes at large aspect ratio is problematically low. There is ample evidence from computations that the limit exists. However, it is not known if stable equilibria exist at much higher {beta}--none have been found. We have explored this question in the asymptotic high {beta} poloidal limit. Unfortunately, we are unable to find stable equilibrium and also unable to show that they don't exist. The results of these calculations will be published when a more definitive answer is found.

  4. Spontaneous generation of rotation in tokamak plasmas

    SciTech Connect

    Parra Diaz, Felix

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  5. Tokamak plasma modelling and atomic processes

    NASA Astrophysics Data System (ADS)

    Kawamura, T.

    1986-06-01

    Topics addressed include: particle control in a tokomak device; ionizing and recombining plasmas; effects of data accuracy on tokamak impurity transport modeling; plasma modeling of tokamaks; and ultraviolet and X-ray spectroscopy of tokamak plasmas.

  6. Tokamak foundation in USSR/Russia 1950-1990

    NASA Astrophysics Data System (ADS)

    Smirnov, V. P.

    2010-01-01

    In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

  7. Improvement of tokamak performance by injection of electrons

    SciTech Connect

    Ono, Masayuki.

    1992-12-01

    Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas.

  8. Tritium catalyzed deuterium tokamaks

    SciTech Connect

    Greenspan, E.; Miley, G.H.; Jung, J.; Gilligan, J.

    1984-04-01

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).

  9. Energy confinement in tokamaks

    SciTech Connect

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston.

  10. TPX tokamak construction management

    SciTech Connect

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-12-31

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly.

  11. Tokamak divertor maps

    NASA Astrophysics Data System (ADS)

    Punjabi, Alkesh; Verma, Arun; Boozer, Allen

    1994-08-01

    A mapping method is developed to investigate the problem of determination and control of heat-deposition patterns on the plates of a tokamak divertor. The deposition pattern is largely determined by the magnetic field lines, which are mathematically equivalent to the trajectories of a single-degree-of-freedom time-dependent Hamiltonian system. Maps are natural tools to study the generic features of such systems. The general theory of maps is presented, and methods for incorporating various features of the magnetic field and particle motion in divertor tokamaks are given. Features of the magnetic field include the profile of the rotational transform, single- versus double-null divertor, reverse map, the effects of naturally occurring low M and N, and externally imposed high-M, high-N perturbations. Particle motion includes radial diffusion, pitch angle and energy scattering, and the electric sheath at the plate. The method is illustrated by calculating the stochastic broadening in a single- null divertor tokamak. Maps provide an efficient, economic and elegant method to study the problem of motion of plasma particles in the stochastic scrape-off layer.

  12. Impurity transport in Tokamaks

    NASA Astrophysics Data System (ADS)

    Amano, T.

    1983-12-01

    Theoretical and experimental efforts directed towards gaining an understanding of impurity behavior in Tokamaks are reviewed. In the Alcator Tokamak experiments, a laser blow-off technique was used to introduce trace amounts of impurities into ohmically heated plasmas. After a series of experiments in which they injected Si, Al, Fe, Mo impurities, an equation representing empirical impurity confinement time was derived. The scaling of this equation was compared with the results of impurity injection experiments on other Tokamaks, FT-I, PDX, TFR, ISX-B. Impurity confinement times in all these cases agree remarkably well, except for the TFR confinement times, which were about a factor of two larger than predicted. In the presence of intense neutral beam injection impurity ions behave differently. Specifically, in the ISX-B experiments, a marked accumulation of impurity ions toward the center of the plasma was observed in the case of counter neutral beam injection. This was interpreted semi-quantitatively by the neoclassical effect of the rotation of the plasma driven by the neutral beam.

  13. Tokamak pump limiters

    NASA Astrophysics Data System (ADS)

    Conn, Robert W.

    1984-12-01

    Experiments with pump limiters on several operating tokamaks have established them as efficient collectors of particles. The gas pressure rise within the chamber behind the limiters has been as high as 50 mTorr when there is no internal chamber pumping. Observations of the plasma power distribution over the front face of these limiter modules yield estimates for the scale length of radial power decay consistent with predictions of relatively simple theory. Interaction of the in-flowing plasma with recycling neutral gas near the limiter deflector plate is predicted to become important when the effective ionization mean free path is comparable to or less than the neutral atom mean path length within the throat structure of the limiter. Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased

  14. The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems

    SciTech Connect

    1995-06-12

    The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development.

  15. Magnetic confinement experiment -- 1: Tokamaks

    SciTech Connect

    Goldston, R.J.

    1994-12-31

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.

  16. Resistive edge mode instability in stellarator and tokamak geometries

    NASA Astrophysics Data System (ADS)

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-01

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  17. Resistive edge mode instability in stellarator and tokamak geometries

    SciTech Connect

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-15

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  18. DIII-D tokamak long range plan. Revision 3

    SciTech Connect

    1992-08-01

    The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998.

  19. DIII-D research operations

    SciTech Connect

    Baker, D.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R D; and collaborative efforts.

  20. Dust Measurements in Tokamaks

    SciTech Connect

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  1. Sawtooth oscillation in tokamaks

    SciTech Connect

    Park, W.; Monticello, D.A.

    1989-03-01

    A three-dimensional nonlinear toroidal full MHD code, MH3D, has been used to study sawtooth oscillations in tokamaks. The profile evolution during the sawtooth crash phase compares well with experiment, but only if neoclassical resistivity is used in the rise phase. (Classical resistivity has been used in most of the previous theoretical sawtooth studies.) With neoclassical resistivity, the q value at the axis drops from 1 to about 0.8 before the crash phase, and then resets to 1 through a Kadomtsev-type complete reconnection process. This ..delta..q/sub 0/ approx. = 0.2 is much larger than ..delta..q/sub o/ approx. = 0.01, which is obtained if classical resistivity is used. The current profile is strongly peaked at the axis with a flat region around the singular surface, and is similar to the Textor profile. To understand this behavior, approximate formulas for the time behavior of current and q values are derived. A functional dependence of sawtooth period scaling is also derived. A semi-empirical scaling is found which fits the experimental data from various tokamaks. Some evidence is presented which indicates that the fast crash time is due to enhanced effective resistivity inside the singular current sheet, generated by, e.g., microinstability and electron parallel viscosity with stochastic fields at the x-point. 16 refs., 5 figs.

  2. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  3. Basic Physics of Tokamak Transport Final Technical Report.

    SciTech Connect

    Sen, Amiya K.

    2014-05-12

    The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficult and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to

  4. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks

    SciTech Connect

    Scharer, J.E.

    1992-01-01

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  5. Texas Experimental Tokamak. Technical progress report, April 1990--April 1993

    SciTech Connect

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.

  6. Electron cyclotron emission imaging in tokamak plasmas

    SciTech Connect

    Munsat, Tobin; Domier, Calvin W.; Kong, Xiangyu; Liang, Tianran; Luhmann, Jr.; Neville C.; Tobias, Benjamin J.; Lee, Woochang; Park, Hyeon K.; Yun, Gunsu; Classen, Ivo. G. J.; Donne, Anthony J. H.

    2010-07-01

    We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the large-aperture optical systems and the linear detector arrays sensitive to millimeter-wavelength radiation. We present the status and recent progress on existing instruments as well as new systems under development for future experiments. We also discuss data analysis techniques relevant to plasma imaging diagnostics and present recent temperature fluctuation results from the tokamak experiment for technology oriented research (TEXTOR).

  7. Tokamak physics experiment: Diagnostic windows study

    SciTech Connect

    Merrigan, M.; Wurden, G.A.

    1995-11-01

    We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented.

  8. Moving Divertor Plates in a Tokamak

    SciTech Connect

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  9. Status of tokamak experiments

    SciTech Connect

    Wolf, G.H.

    1996-03-01

    Plasma-wall interaction, heat removal and ash exhaust have emerged as the dominant problems still to be solved in order to achieve ignition and - even more difficult - to maintain a state of self-sustained thermo-nuclear burn. This is of particular delicacy, since those operational regimes which yield the best energy confinement correspond to an even better particle confinement and confinement of impurities, which then tend to accumulate in the plasma core and to result in disruption or degradation of the tokamak discharge. Therefore, plasma-wall interaction, heat removal and particle exhaust will determine not only the structure and configuration of the plasma edge region, of the wall system and of the materials facing the plasma, but also the final choice of useful confinement regimes. Moreover, the potential effect of powerful {alpha}-particle heating on plasma stability and confinement has to be kept below critical values. For the latter requirement, a final answer can only be obtained in an ITER-type device where ignition and burn will become accessible. 72 refs., 12 figs.

  10. Resistive instabilities in tokamaks

    SciTech Connect

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed.

  11. Fuel retention in tokamaks

    NASA Astrophysics Data System (ADS)

    Loarer, T.

    2009-06-01

    Tritium retention constitutes an outstanding problem for ITER operation and future fusion reactors, particularly for the choice of the first wall materials. In present day tokamaks, fuel retention is evaluated by two complementary methods. The in situ gas balance allows evaluation of how much fuel is retained during a discharge and, typically, up to one day of experiments. Post-mortem analysis is used to determine where the fuel is retained, integrated over an experimental campaign. In all the carbon clad devices, using the two methods, the retention is demonstrated to be very closely related to the carbon net erosion. This results from plasma-wall interaction with ion and charge-exchange fluxes, ELMs and is proportional to the pulse duration. The fuel retention by implantation saturates at high wall temperatures and limits the D/C ratio in the deposited layers but, as far as a carbon source exists, the dominant retention process remains the co-deposition of carbon with deuterium. In full metallic device, in the absence of wall conditioning with boron, co-deposition is strongly reduced and fuel retention below 1% can be achieved. Extrapolation to ITER shows that removing the carbon from the plasma-facing components would increase the number of discharges to 2500 before reaching the maximum tritium limit of 700 g.

  12. Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

    DOE PAGESBeta

    Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; Sandorfi, Andy M.; Smith, S. P.; Wei, Xiangdong

    2015-09-21

    The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here asmore » an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States’ magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress.« less

  13. Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

    SciTech Connect

    Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; Sandorfi, Andy M.; Smith, S. P.; Wei, Xiangdong

    2015-09-21

    The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here as an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States’ magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress.

  14. Tokamak coordinate conventions: COCOS

    NASA Astrophysics Data System (ADS)

    Sauter, O.; Medvedev, S. Yu.

    2013-02-01

    Dealing with electromagnetic fields, in particular current and related magnetic fields, yields "natural" physical vector relations in 3-D. However, when it comes to choosing local coordinate systems, the "usual" right-handed systems are not necessarily the best choices, which means that there are several options being chosen. In the magnetic fusion community such a difficulty exists for the choices of the cylindrical and of the toroidal coordinate systems. In addition many codes depend on knowledge of an equilibrium. In particular, the Grad-Shafranov axisymmetric equilibrium solution for tokamak plasmas, ψ, does not depend on the sign of the plasma current Ip nor that of the magnetic field B0. This often results in ill-defined conventions. Moreover the sign, amplitude and offset of ψ are of less importance, since the free sources in the equation depend on the normalized radial coordinate. The signs of the free sources, dp/dψ and dF2/dψ (p being the pressure, ψ the poloidal magnetic flux and F=RBφ), must be consistent to generate the current density profile. For example, RF and CD calculations (Radio Frequency heating and Current Drive) require an exact sign convention in order to calculate a co- or counter-CD component. It is shown that there are over 16 different coordinate conventions. This paper proposes a unique identifier, the COCOS convention, to distinguish between the 16 most-commonly used options. Given the present worldwide efforts towards code integration, the proposed new index COCOS defining uniquely the COordinate COnventionS required as input by a given code or module is particularly useful. As codes use different conventions, it is useful to allow different sign conventions for equilibrium code input and output, equilibrium being at the core of any calculations in magnetic fusion. Additionally, given two different COCOS conventions, it becomes simple to transform between them. The relevant transformations are described in detail.

  15. Understanding disruptions in tokamaks

    NASA Astrophysics Data System (ADS)

    Zakharov, Leonid

    2011-10-01

    Disruptions in tokamaks are known since 1963 but even now some aspects of them remain a mystery. This talk describes progress made recently in understanding disruptions. A major step forward occurred in 2007 when the importance of galvanic contact of the plasma with the wall in plasma dynamics was pointed out. The toroidal asymmetry of plasma current, observed in JET vertical disruptions, was explained by the theory of the wall touching kink mode. The currents shared by the plasma with the wall and responsible for the asymmetry were identified as generated by the kink mode. Such currents are referred to as Hiro currents. They have shown exceptional consistency with the entire JET disruption data base (more than 5500 cases) and ruled out the long lasting interpretation based on ``halo currents,'' which contradict experiments even in the sign of the measured asymmetry. Accordingly, the sideways forces are understood and their scaling from JET to ITER was justified. Hiro currents provide also a plausible explanation of the current spike at the beginning of the disruptions. The important role of the plasma edge and its interaction with the wall was revealed. Based on this new understanding of disruptions, dedicated experiments on the current spike (J-TEXT, Wuhan, China) and runaway prevention by the repetitive triggering of kink modes (T-10, AUG, Tore Supra) were motivated and are in progress. Accordingly, the need for new, adaptive grid approaches to numerical simulations of disruptions became evident. In addition to the core MHD, simulations of realistic wall geometry, disruption specific plasma edge physics, plasma-wall interaction, and energetic particles need be developed. The first results of simulations of the fast MHD regime, Hiro current generation, and slower plasma decay due to a wall touching kink mode made with the new DSC code are presented. This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  16. Predictive modelling and simulations of internal transport barriers in tokamaks

    NASA Astrophysics Data System (ADS)

    Zhu, Ping

    2001-09-01

    An Internal Transport Barrier (ITB) is a localized region inside a (tokamak) plasma where a steep temperature and/or density gradient forms due to much lower thermal and/or particle transport than in the surrounding regions. Internal transport barriers have now been observed in all large tokamaks after they were first discovered in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) in 1993. While suggesting a promising practical approach to the realization of fusion ignition conditions, this high performance regime poses a great challenge to our understanding of tokamak anomalous transport physics. In this work, the formation and evolution of internal transport barriers in tokamaks are studied through predictive transport modelling and simulations. Neoclassical and anomalous transport of particles, energy, and toroidal momentum are systematically formulated from the ensemble-averaged gyrokinetic equation, for a tokamak plasma with large toroidal flow on the order of the ion thermal speed. This formulation is then used to construct an updated Multi-Mode model (MMM) based on (1)the Weiland fluid model for the drift wave transport, (2)the Scott-Bateman model for drift-Alfvèn mode at the tokamak edge, and (3)poloidal and toroidal momentum transport models by Zhu, Horton and Sugama. The formation of internal transport barriers observed in two optimized shear discharges in the Joint European Torus (JET) and two negative central shear discharges in the Doublet III-D Tokamak (DIII-D) are reproduced in predictive transport simulations that use the updated MultiMode model embedded in the time-dependent one/one and half dimensional transport code BALDUR. The Weiland model for drift modes in the MultiMode model is implemented in combination with either the Hahm-Burrell or the Hamaguchi-Horton flow shear stabilization mechanisms, where the radial electric field is inferred from both the measured toroidal velocity profile and the poloidal velocity profile

  17. Solenoid-free plasma start-up in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  18. Energy losses on tokamak startup

    SciTech Connect

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells.

  19. Fusion product measurements in tokamaks

    SciTech Connect

    Strachan, J.D.

    1985-05-01

    Diagnostic methods and the applications of fusion product measurements in tokamaks are reviewed with emphasis on results from PLT, PDX, and TFTR. Measurements have been made using the 2.5-MeV neutron from the d(d, n)/sup 3/ He reaction, the 3-MeV proton from the d(d, p)t reaction, both the 3.7-MeV alpha and the 14.7-MeV proton from the d(/sup 3/He, p)..cap alpha.. reaction, and the 14-MeV neutron from the d(t, n)..cap alpha.. reaction. The common use of these measurements is the determination of the ion temperature from the magnitude of the d-d neutron emission. For tokamak plasmas, these results are usually in good agreement with the charge exchange ion temperature. Recently, the charged fusion products have been used for high-resolution spectroscopic purposes, and emission profile measurements. Pitch angle resolution of the escaping 3-MeV proton emission has been used to determine the poloidal magnetic field inside the tokamak. Major issues in this field include the expected tritium operation on TFTR where the neutron measurements will determine when tritium will be introduced into the TFTR vessel and provide a measurement of the fusion power multiplication value (Q). The TFTR Q approx. 1 experiments will also provide a chance to measure the confinement of 3.5-MeV alphas in a tokamak.

  20. Development in Diagnostics Application to Control Advanced Tokamak Plasma

    SciTech Connect

    Koide, Y.

    2008-03-12

    For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

  1. Final Technical Report: Global Field Aligned Mesh and Gyrokinetic Field Solver in a Tokamak Edge Geometry

    SciTech Connect

    Cummings, Julian C.

    2013-05-15

    This project was a collaboration between researchers at the California Institute of Technology and the University of California, Irvine to investigate the utility of a global field-aligned mesh and gyrokinetic field solver for simulations of the tokamak plasma edge region. Mesh generation software from UC Irvine was tested with specific tokamak edge magnetic geometry scenarios and the quality of the meshes and the solutions to the gyrokinetic Poisson equation were evaluated.

  2. Bootstrapped tokamak with oscillating field current drive

    SciTech Connect

    Weening, R.H. )

    1993-07-01

    A magnetic helicity conserving mean-field Ohm's law is used to study bootstrapped tokamaks with oscillating field current drive. The Ohm's law leads to the conclusion that the tokamak bootstrap effect can convert the largely alternating current of oscillating field current drive into a direct toroidal plasma current. This plasma current rectification is due to the intrinsically nonlinear nature of the tokamak bootstrap effect, and suggests that it may be possible to maintain the toroidal current of a tokamak reactor by supplementing the bootstrap current with oscillating field current drive. Steady-state tokamak fusion reactors operating with oscillating field current drive could provide an alternative to tokamak reactors operating with external current drive.

  3. Transport of Dust Particles in Tokamak Devices

    SciTech Connect

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  4. Microwave Tokamak Experiment: Overview and status

    SciTech Connect

    Not Available

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs.

  5. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    NASA Astrophysics Data System (ADS)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  6. Comprehensive numerical modelling of tokamaks

    SciTech Connect

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-03

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell.

  7. Neoclassical magnetic microislands in tokamaks

    SciTech Connect

    Kovalishen, E.A.; Mikhailovskii, A.B.; Botov, P.V.; Shirokov, M.S.; Konovalov, S.V.; Tsypin, V.S.; Galvao, R.M.O.

    2005-09-15

    Possibility of existence of neoclassical magnetic microislands (island width smaller than the ion Larmor radius) in a tokamak in the banana regime is shown. The rotation frequency of such islands is found. It is shown that for the case of positive electron temperature gradient, the bootstrap current destabilizes the microislands while the polarization current leads to their stabilization. Maximally possible neoclassical microisland width is estimated.

  8. Gyrosheath near the tokamak edge

    SciTech Connect

    Hazeltine, R.D.; Xiao, H. . Inst. for Fusion Studies); Valanju, P.M. . Fusion Research Center)

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results.

  9. Transport Equations In Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.

    2009-11-01

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for: neoclassical effects on the parallel Ohm's law (trapped particle effects on resistivity, bootstrap current); fluctuation-induced transport; heating, current-drive and flow sources and sinks; small B field non-axisymmetries; magnetic field transients etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed recently using a kinetic-based framework. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales (and constraints they impose) are considered sequentially: compressional Alfv'en waves (Grad-Shafranov equilibrium, ion radial force balance); sound waves (pressure constant along field lines, incompressible flows within a flux surface); and ion collisions (damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on the plasma fluid: 7 ambipolar collision-based ones (classical, neoclassical, etc.) and 8 non-ambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients etc.). The plasma toroidal rotation equation [1] results from setting to zero the net radial current induced by the non-ambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the non-ambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The resultant transport equations will be presented and contrasted with the usual ones. [4pt] [1] J.D. Callen, A.J. Cole, C.C. Hegna, ``Toroidal Rotation In

  10. Magnetic confinement experiment. I: Tokamaks

    SciTech Connect

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.

  11. Instrumentation and controls of an ignited tokamak

    SciTech Connect

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented.

  12. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    SciTech Connect

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  13. Tokamak plasma position dynamics and feedback control

    SciTech Connect

    Burenko, L.; Bailey, J.M.

    1983-01-01

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form.

  14. Steady State Tokamak Equilibria without Current Drive

    SciTech Connect

    Shaing, K.C.; Aydemir, A.Y.; Lin-Liu, Y.R.; Miller, R.L.

    1997-11-01

    Steady state tokamak equilibria without current drive are found. This is made possible by including the potato bootstrap current close to the magnetic axis. Tokamaks with this class of equilibria do not need seed current or current drive, and are intrinsically steady state. {copyright} {ital 1997} {ital The American Physical Society}

  15. Natural current profiles in a tokamak

    SciTech Connect

    Taylor, J.B.

    1990-08-01

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described.

  16. Plasma diagnostics for the compact ignition tokamak

    SciTech Connect

    Medley, S.S.; Young, K.M.

    1988-06-01

    The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.

  17. Transport equations in tokamak plasmas

    SciTech Connect

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-15

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  18. Transport equations in tokamak plasmasa)

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-01

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfvén waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The "mean field" effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  19. Tokamak plasma interaction with limiters

    NASA Astrophysics Data System (ADS)

    Pitcher, Charles Spencer

    1988-08-01

    The importance of plasma purity is discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fueling/recycling and impurity production. The experiments, carried out on the DITE tokomak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behavior; new physical phenomena are presented in all three areas.

  20. Breakdown in the pretext tokamak

    SciTech Connect

    Benesch, J.F.

    1981-06-01

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges.

  1. DIII-D research operations. Annual report, October 1, 1991--September 30, 1992

    SciTech Connect

    Baker, D.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R&D; and collaborative efforts.

  2. Enhancement of confinement in tokamaks

    SciTech Connect

    Furth, H.P.

    1986-05-01

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime.

  3. Superconducting magnet system for the TPX Tokamak

    SciTech Connect

    Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.

    1993-09-15

    The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma.

  4. Analysis of neutral hydrogenic emission spectra in a tokamak

    NASA Astrophysics Data System (ADS)

    Ko, J.; Chung, J.; Jaspers, R. J. E.

    2015-10-01

    Balmer-α radiation by the excitation of thermal and fast neutral hydrogenic particles has been investigated in a magnetically confined fusion device, or tokamak, from the Korea Superconducting Tokamak Advanced Research (KSTAR). From the diagnostic point of view, the emission from thermal neutrals is associated with passive spectroscopy and that from energetic neutrals that are usually injected from the outside of the tokamak to the active spectroscopy. The passive spectroscopic measurement for the thermal Balmer-α emission from deuterium and hydrogen estimates the relative concentration of hydrogen in a deuterium-fueled plasma and therefore, makes a useful tool to monitor the vacuum wall condition. The ratio of hydrogen to deuterium obtained from this measurement qualitatively correlates with the energy confinement of the plasma. The Doppler-shifted Balmer-α components from the fast neutrals features the spectrum of the motional Stark effect (MSE) which is an essential principle for the measurement of the magnetic pitch angle profile. Characterization of this active MSE spectra, especially with multiple neutral beam lines crossing along the observation line of sight, has been done for the guideline of the multi-ion-source heating beam operation and for the optimization of the narrow bandpass filters that are required for the polarimeter-based MSE diagnostic system under construction at KSTAR.

  5. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  6. Control of Dust Inventory in Tokamaks

    SciTech Connect

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Pitcher, C. S.; Taylor, N.; Furlan, J.

    2008-09-07

    Particles with sizes ranging from 100 nm to 100 {mu}m are produced in tokamaks by the interaction of the plasma with the first wall materials and divertor. Dust has not yet been of a major concern in existing tokamaks mainly because their quantities are small and these devices are not nuclear facilities. However, in ITER and in future reactors, they could represent operational and potential safety issues. The aim of this paper is thus to describe the dust creation processes in the tokamak environment. The diagnostics and removal techniques that are needed to be implemented to measure and minimise the dust inventory are also presented. The integration of these techniques into a tokamak environment is also discussed.

  7. Driven-current tokamak (DCT) scoping study

    SciTech Connect

    Reid, R.L.

    1983-01-01

    The present Department of Energy (DOE) plan calls for the construction of an Engineering Test Reactor (ETR) that is to be the last major experimental fusion device prior to the commercialization of fusion power. The plasma driver of the ETR is to be either a long-pulse tokamak or a tandem mirror machine. The possibility of using the Tokamak Fusion Test Reactor (TFTR) facility to consolidate the physics and technology database for the tokamak version of the ETR has been considered. This paper addresses two of the options being considered: (1) a superconducting toroidal field (TF) coil-hydrogen plasma alternative, and (2) a superconducting or hybrid TF coil-high Q alternative. Both options assume essentially steady-state operation through the application of rf current drive. The options are evaluated on the basis of performance and cost determined by application of the Fusion Engineering Design Center (FEDC) Tokamak System Code.

  8. OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS

    SciTech Connect

    LIN-LIU,YR; STAMBAUGH,RD

    2002-11-01

    OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.

  9. D-D tokamak reactor studies

    SciTech Connect

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  10. Computational methods in tokamak transport

    SciTech Connect

    Houlberg, W.A.; Attenberger, S.E.; Lao, L.L.

    1982-06-01

    A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements.

  11. Dust measurements in tokamaks (invited)

    SciTech Connect

    Rudakov, D. L.; Yu, J. H.; Boedo, J. A.; Hollmann, E. M.; Krasheninnikov, S. I.; Moyer, R. A.; Muller, S. H.; Pigarov, A. Yu.; Rosenberg, M.; Smirnov, R. D.; West, W. P.; Boivin, R. L.; Bray, B. D.; Brooks, N. H.; Hyatt, A. W.; Wong, C. P. C.; Roquemore, A. L.; Skinner, C. H.; Solomon, W. M.; Ratynskaia, S.

    2008-10-15

    Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 {mu}m in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C{sub 2} dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  12. A design retrospective of the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Luxon, J. L.

    2002-05-01

    The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and RF heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research programme. An integrated picture of the facility and its capabilities is presented.

  13. Tokamak Diagnostics Using Fusion Products.

    NASA Astrophysics Data System (ADS)

    Heidbrink, William Walter

    Measurements of neutrons and protons produced by the d(d,n)('3)He, d(t,n)(alpha), d(d,p)t, and d(('3)He,p)(alpha) fusion reactions are used to diagnose plasmas in the PLT and PDX tokamaks. An expression for the efficiency of proton detection is derived and confirmed experimentally. The time evolution of the ('3)He density indicates that a scoop limiter may pump ('3)He from the plasma faster than conventional limiters. The confinement of 1.0 MeV tritons and of 0.8 MeV ('3)He ions is studied by measuring the fraction of these fusion-produced ions that burn up in subsequent fusion reactions. In discharges with sawtooth activity and with B(,(phi)) > 2 T, the triton and ('3)He 'burnup' is consistent (within a factor of three) with predictions based on classical theories of ion confinement and slowing down. In discharges with large m = 2 or fishbone instabilities, the ('3)He burnup is less than classically predicted and, in PLT discharges at B(,(phi)) = 1.8 T, the triton burnup is over an order of magnitude smaller than predicted. Expressions for the energy spectrum of ions produced in beam-target fusion reactions are derived. Collimated measurements of the spectrum of 15 MeV protons produced by reactions between energetic ('3)He ions and relatively cold deuterons during fast wave minority heating indicate that the velocity distribution of fast ('3)He ions is peaked perpendicular to the tokamak magnetic field. The ion temperature profile and density of fast deuterons are measured with an array of collimated 3 MeV proton detectors. The fast ions produced by neutral beam injection and by launching lower hybrid waves are concentrated near the magnetic axis. Poloidal field measurements using 3 MeV protons also appear possible. In discharges in which the line radiation from central impurities does not decay, the plasma current profile is broader than in more typical discharges.

  14. Public Data Set: Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup

    DOE Data Explorer

    Hinson, Edward T. [University of Wisconsin-Madison] (ORCID:000000019713140X); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609)

    2016-05-31

    This data set contains openly-documented, machine readable digital research data corresponding to figures published in E.T. Hinson et al., 'Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup,' Physics of Plasmas 23, 052515 (2016).

  15. L to H mode transitions and associated phenomena in divertor tokamaks

    NASA Astrophysics Data System (ADS)

    Punjabi, A.

    1990-09-01

    This is the final report for the research project titled, L to H Mode Transitions and Associated Phenomena in Divertor Tokamaks. The period covered by this project is the fiscal year 1990. This report covers the development of Advanced Two Chamber Model.

  16. Plasma Physics found in JT-60 Tokamak over the Last 20 years

    SciTech Connect

    Kikuchi, M.

    2009-07-07

    This paper summarizes major plasma physics obtained during the period of JT-60 operation for 23 years with special emphasis on research towards steady-state operation of tokamak. Topics included are observation of large bootstrap current fraction, negative shear scenario, demonstration of efficient beam and EC current drive, discovery of current hole, stabilization of the RWM, discovery of ITB (internal transport barrier).

  17. Plasma Physics Regimes in Tokamaks with Li Walls

    SciTech Connect

    L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev

    2003-08-21

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors.

  18. Microtearing modes in tokamak discharges

    NASA Astrophysics Data System (ADS)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  19. Upgrades for the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Duval, Basil; TCV Team

    2013-10-01

    Major upgrades are being implemented on the TCV tokamak to extend its operational domain towards a burning plasma regime. The goals of obtaining high normalized plasma beta and comparable ion and electron temperatures will be achieved with the addition of a 1 MW neutral heating system and 2 MW additional third harmonic EC power. Spatial constraints together with beam occlusion required severe design optimization and the additional of a new large tangential port on the TCV vessel. For EC, the existing vertical launch mirror will be sufficient but new 1MW EC units will be employed with the legacy X3 systems modified for lateral launch. The modifications will not affect TCV's strong RT shaping and EC actuator ranges or the open divertor vacuum chamber that permits access to Snowflake divertor or doublet configurations although some wall protection enhancement is envisaged. TCV can then contribute to disentangling effects of electron-ion coupling, rotation, current and density profile control all as a function of shape in L and H-modes with ITER (or higher) values of plasma beta. Together with fast-ion physics, TCV will also be able to explore heat, particle and momentum transport and turbulence effects in electron-heat dominated discharges for Te/Ti in the (0.02 to 3) range.

  20. Tokamak x ray diagnostic instrumentation

    SciTech Connect

    Hill, K.W.; Beiersdorfer, P.; Bitter, M.; Fredrickson, E.; Von Goeler, S.; Hsuan, H.; Johnson, L.C.; Liew, S.L.; McGuire, K.; Pare, V.

    1987-01-01

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/..delta..E is greater than 10/sup 4/ and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally.

  1. Toroidal Flow in Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Cole, A. J.; Hegna, C. C.

    2007-11-01

    Many effects influence toroidal flow evolution in tokamak plasmas. Momentum sources and radial diffusion due to axisymmetric neoclassical, paleoclassical and anomalous transport are usually considered. In addition, the toroidal flow can be affected by field errors. Small, non-axisymmetric field errors arise from coil irregularities, active control coils and collective plasma magnetic distortions (e.g., NTMs, RWMs). Resonant field errors cause localized electromagnetic torques near rational surfaces in the plasma, which can lock the plasma to the wall leading to magnetic islands and reduced confinement or disruptions. Their penetration into the plasma is limited by flow-shielding effects; but they can be amplified by the plasma response at high beta. Non-resonant field errors cause magnetic pumping and radial banana drifts, and lead to toroidal flow damping over the entire plasma. Many of these processes can also produce momentum pinch and intrinsic flow effects. This poster will seek to present a coherent picture of all these effects and suggest ways they could be tested and distinguished experimentally.

  2. Helicity content and tokamak applications of helicity

    SciTech Connect

    Boozer, A.H.

    1986-05-01

    Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities.

  3. Activation analysis of the compact ignition tokamak

    SciTech Connect

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.

  4. Ripple induced trapped particle loss in tokamaks

    SciTech Connect

    White, R.B.

    1996-05-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks.

  5. Tokamak power systems studies, FY 1985

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  6. MHD stable regime of the tokamak

    SciTech Connect

    Cheng, C.Z.; Furth, H.P.; Boozer, A.H.

    1986-10-01

    A broad family of tokamak current profiles is found to be stable against ideal and resistive MHD kink modes for 1 less than or equal to q(0), with q(a) as low 2. For 0.5 less than or equal to q(0) < and q(a) > 1, current profiles can be found that are unstable only to the m = 1, n = 1 mode. A specific ''optimal'' tokamak profile can be selected from the range of stable solutions, by imposing a common upper limit on dj/dr - corresponding in ohmic equilibrium to a limitation of dT/sub e//dr by anomalous transport.

  7. Optimization of turn position of tokamak inductor

    NASA Astrophysics Data System (ADS)

    Aristov, Yu. A.; Vorobev, G. M.; Kuznetsov, A. V.

    Statement and methods of solution of the problem of optimizing turn position of tokamak induction are considered. Optimization is aimed at determination of inductor turn position, providing the minimal scattering of magnetic field in the region of chamber at any assigned value of volt-seconds. Algorithms of problem solution are described, and results of calculations for STX tokamak are presented. It is shown that development methods can be used for determining optimal position of turns of any coils of poloidal magnetic field, providing the assigned configuration and level of magnetic field.

  8. Rotation of tokamak halo currents

    SciTech Connect

    Boozer, Allen H.

    2012-05-15

    During tokamak disruptions, halo currents, which can be tenths of the total plasma current, can flow at the plasma edge along the magnetic field lines that intercept the chamber walls. Non-axisymmetric halo currents are required to maintain force balance as the plasma kinks when the edge safety factor drops to about two in a vertical displacement event. The plasma quickly assumes a definite toroidal velocity v{sub a}(r) with respect to that of the magnetic kink, v{sub k}, where v{sub a}(r) is set by the radial electric field required for ambipolarity. The plasma velocity, v{sub pl}=v{sub a}+v{sub k}, near the edge is influenced by the interaction with neutrals and with the potential in the halo required for quasi-neutrality on open magnetic field lines, and the plasma velocity in the core is influenced by external error fields. When plasma effects dominate magnetic locking, the magnetic kink should rotate at a diamagnetic speed of either the edge or the core. If the magnetic field lines of the halo plasma intercept the wall at locations of very different electrical conductivity, the toroidal rotation of the halo currents can intermittently stall at wall locations of high conductivity. Such stalling is seen in experiments. The toroidal phase difference between the stalled halo currents and the kink, which is expected to rotate smoothly, must satisfy {delta}{phi}<{+-}{pi}/2. A concern cited by ITER engineers is that the time varying force of the rotating halo could substantially increase the disruption loads on in-vessel components.

  9. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    SciTech Connect

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma.

  10. A need for non-tokamak approaches to magnetic fusion energy

    NASA Astrophysics Data System (ADS)

    Bathke, C. G.; Krakowski, R. A.; Miller, R. L.

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested.

  11. Elementary Processes Underlying Alpha Channeling in Tokamaks

    SciTech Connect

    NM.J. Fisch

    2012-06-15

    Alpha channeling in tokamaks is speculative, but also extraordinarily attractive. Waves that can accomplish this effect have been identified. Key aspects of the theory now enjoy experimental confirmation. This paper will review the elementary processes of wave-particle interactions in plasma that underlie the alpha channeling effect

  12. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1981-04-01

    Ripple transport in tokamaks is discussed for the banana drift collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and the resulting transport coefficients can consequently differ by several orders of magnitude.

  13. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1982-01-01

    Ripple transport in tokamaks is discussed for the ''banana drift'' collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and consequently the resulting transport coefficients can differ by several orders of magnitude.

  14. Microinstabilities in weak density gradient tokamak systems

    SciTech Connect

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.

  15. Toroidal Alfven wave stability in ignited tokamaks

    SciTech Connect

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  16. Fusion product measurements in tokamaks (invited; abstract)

    NASA Astrophysics Data System (ADS)

    Strachan, J. D.

    1985-05-01

    Diagnostic methods and the applications of fusion product measurements in tokamaks are reviewed with emphasis on results from PLT, PDX, and TFTR. Measurements have been made using the 2.5-MeV neutron from the d(d, n)3He reaction, the 3-MeV proton from the d(d, p)t reaction, both the 3.7-MeV alpha and the 14.7-MeV proton from the d(3He, p)α reaction, and the 14-MeV neutron from the d(t, n)α reaction. The common use of these measurements is the determination of the ion temperature from the magnitude of the d-d neutron emission. For tokamak plasmas, these results are usually in good agreement with the charge exchange ion temperature. Recently, the charged fusion products have been used for high-resolution spectroscopic purposes, and emission profile measurements. Pitch angle resolution of the escaping 3-MeV proton emission has been used to determine the poloidal magnetic field inside the tokamak. Major issues in this field include the expected tritium operation on TFTR where the neutron measurements will determine when tritium will be introduced into the TFTR vessel and provide a measurement of the fusion power multiplication value (Q). The TFTR Q˜1 experiments will also provide a chance to measure the confinement of 3.5-MeV alphas in a tokamak.

  17. Analysis of sawtooth relaxation oscillations in tokamaks

    SciTech Connect

    Yamazaki, K.; McGuire, K.; Okabayashi, M.

    1982-07-01

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated.

  18. Diagnostics for neutral-beam-heated tokamaks

    SciTech Connect

    Goldston, R.J.

    1982-12-01

    Diagnostic techniques for neutral-beam-heated tokamak plasmas fall into three categories: (1) magnetic diagnostics for measurements of gross stored energy, (2) profile diagnostics for measurements of stored thermal and beam energy, impurity content and plasma rotation, and (3) fast time resolution diagnostics to study MHD fluctuations and micro-turbulence.

  19. Plasma-gun fueling for tokamak reactors

    SciTech Connect

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment.

  20. Stabilization of tokamak plasma by lithium streams

    SciTech Connect

    L.E. Zakharov

    2000-08-07

    The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.

  1. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  2. Conceptual study of electron ripple injection for tokamak transport control

    SciTech Connect

    Choe, W.; Ono, M.; Chang, C.S.

    1995-08-01

    A non-intrusive method for inducing radial electric field based on electron ripple injection is under development by the Princeton CDX-U group. The radial electric field is known to play an important role in the L-H and H-VH mode transition according to the recent theoretical and experimental research. It is therefore important to develop a non-intrusive tool to control the radial electric field profile in tokamak plasmas. The present technique utilizes externally-applied local magnetic ripple fields to trap electrons at the edge, allowing them to penetrate towards the plasma center via {gradient}B and curvature drifts, causing the flux surfaces to charge up negatively. Electron cyclotron resonance heating is utilized to increase the trapped population and the electron drift velocity by raising the perpendicular energy of trapped electrons. In order to quantify the effects of cyclotron resonance heating on electrons, the temperature anisotropy of resonant electrons in a tokamak plasma is calculated. For the calculation of anisotropic temperatures, energy moments of the bounce-averaged Fokker-Planck equation with a bi-Maxwellian distribution function for heated electrons are solved, assuming a moderate wave power and a constant quasilinear diffusion coefficient. Simulation using a guiding-center orbit model have been performed to understand the behavior of suprathermal electrons in the presence of ripple fields. Examples for CDX-U and ITER parameters are given.

  3. DSC -- Disruption Simulation Code for Tokamaks and ITER applications

    NASA Astrophysics Data System (ADS)

    Galkin, S. A.; Grubert, J. E.; Zakharov, L. E.

    2010-11-01

    Arguably the most important issue facing the further development of magnetic fusion via advanced tokamaks is to predict, avoid, or mitigate disruptions. This recently became the hottest challenging topic in fusion research because of several potentially damaging effects, which could impact the ITER device. To address this issue, two versions of a new 3D adaptive Disruption Simulation Code (DSC) will be developed. The first version will solve the ideal reduced 3D MHD model in the real geometry with a thin conducting wall structure, utilizing the adaptive meshless technique. The second version will solve the resistive reduced 3D MHD model in the real geometry of the conducting structure of the tokamak vessel and will finally be parallelized. The DSC will be calibrated against the JET disruption data and will be capable of predicting the disruption effects in ITER, as well as contributing to the development of the disruption mitigation scheme and suppression of the RE generation. The progress on the first version of the 3D DSC development will be presented.

  4. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  5. Designing tokamaks to withstand electromagnetic disruption loads

    NASA Astrophysics Data System (ADS)

    Crowell, Jeffrey Arnold

    1999-11-01

    Tokamaks, the toroidal plasma confinement devices used to study fusion energy, operate by driving a multi-MA current in the plasma while creating a strong confining magnetic field. In experimental tokamaks under some conditions, the plasma can become unstable, escape its magnetic confines and rapidly cool off. On a time scale of milliseconds, the plasma current decays away in the resulting cold and highly resistive plasma. In these events, called disruptions, the rapid change in plasma current induces large currents in the surrounding conducting structures. The induced currents, flowing in the presence of a strong magnetic field, can apply substantial electromagnetic forces. Some experimental devices, such as the JET facility, have experienced extensive damage from these events. In future power reactors, even greater loads must be absorbed by components also subject to neutron embrittlement. This study models the electromagnetic and structural behavior of conceptual designs of the first generation of power-producing tokamaks to identify the components that are at risk and illuminate design options which mitigate these loads. The problem is a coupled one: the geometry and resistivity of the structure affects the induced currents while the induced currents and resulting loads place demands on the structure. Several new analytical and computational tools for the evaluation of these systems are discussed including a dual-solution technique for taking advantage of the complex electromagnetic symmetries in a typical tokamak design. The finite element method with a differential formulation and an integral method using a Green's function have been applied to 2D and 3D electromagnetic models of tokamaks. The differential formulation was found to be superior in these highly symmetric systems. The most significant design issues arise with the components most proximate to the plasma. Despite toroidal segmentation, damaging electromagnetic loads threaten the first wall and

  6. Sensitivity of magnetic field-line pitch angle measurements to sawtooth events in tokamaks

    NASA Astrophysics Data System (ADS)

    Ko, J.

    2016-11-01

    The sensitivity of the pitch angle profiles measured by the motional Stark effect (MSE) diagnostic to the evolution of the safety factor, q, profiles during the tokamak sawtooth events has been investigated for Korea Superconducting Tokamak Advanced Research (KSTAR). An analytic relation between the tokamak pitch angle, γ, and q estimates that Δγ ˜ 0.1° is required for detecting Δq ˜ 0.05 near the magnetic axis (not at the magnetic axis, though). The pitch angle becomes less sensitive to the same Δq for the middle and outer regions of the plasma (Δγ ˜ 0.5°). At the magnetic axis, it is not straightforward to directly relate the γ sensitivity to Δq since the gradient of γ(R), where R is the major radius of the tokamak, is involved. Many of the MSE data obtained from the 2015 KSTAR campaign, when calibrated carefully, can meet these requirements with the time integration down to 10 ms. The analysis with the measured data shows that the pitch angle profiles and their gradients near the magnetic axis can resolve the change of the q profiles including the central safety factor, q0, during the sawtooth events.

  7. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks. Annual report, November 16, 1991--November 15, 1992

    SciTech Connect

    Scharer, J.E.

    1992-12-31

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  8. Equilibrium system analysis in a tokamak ignition experiment

    SciTech Connect

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  9. Equilibrium system analysis in a tokamak ignition experiment. Final report

    SciTech Connect

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  10. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    NASA Astrophysics Data System (ADS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Garcia, J.; Arimoto, H.; Shoji, T.

    2009-05-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  11. Banana orbits in elliptic tokamaks with hole currents

    NASA Astrophysics Data System (ADS)

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

  12. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    SciTech Connect

    BURRELL,KH

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  13. A low aspect ratio tokamak transmutation system

    NASA Astrophysics Data System (ADS)

    Qiu, L. J.; Wu, Y. C.; Xiao, B. J.; Xu, Q.; Huang, Q. Y.; Wu, B.; Chen, Y. X.; Xu, W. N.; Chen, Y. P.; Liu, X. P.

    2000-03-01

    A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m2 neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude.

  14. The physics of tokamak start-up

    SciTech Connect

    Mueller, D.

    2013-05-15

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  15. Models for impurity effects in tokamaks

    SciTech Connect

    Hogan, J.T.

    1980-03-01

    Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high ..beta.. and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high ..beta.. in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability.

  16. Tritium Retention and Removal in Tokamaks

    SciTech Connect

    Skinner, Charles H.

    2009-02-19

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy. Tritium is an important source term in safety assessments, it is expensive and in short supply. Tritium can be continuously retained in a tokamak by codeposition with eroded carbon or beryllium and JET and TFTR with carbon plasma facing components showed a tritium retention level that would be unacceptable in ITER or future fusion reactors. Asdex-U and Alcator C-mod have shown reduced hydrogenic retention with tungsten clad and molybdenum plasma facing components. Once the tritium inventory approaches the administrative limit, tritium must be removed to permit continued D-T plasma operations. Several candidate techniques are being considered and need to be proven at a relevant speed and efficiency in contemporary tokamaks. Projections for ITER are discussed.

  17. Boundary Plasma Turbulence Simulations for Tokamaks

    SciTech Connect

    Xu, X; Umansky, M; Dudson, B; Snyder, P

    2008-05-15

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.

  18. Properties of dc helicity injected tokamak plasmas

    SciTech Connect

    Darrow, D.S.; Ono, M.; Forest, C.B.; Greene, G.J.; Hwang, Y.S.; Park, H.K. ); Taylor, R.J.; Pribyl, P.A.; Evans, J.D.; Lai, K.F.; Liberati, J.R. )

    1990-06-01

    Several dc helicity injection experiments using an electron beam technique have been conducted on the Current Drive Experiment (CDX) (Phys. Rev. Lett. {bold 59}, 2165 (1987)) and the Continuous Current Tokamak (CCT) (Phys. Rev. Lett. {bold 63}, 2365 (1989)). The data strongly suggest that tokamak plasmas are being formed and maintained by this method. The largest currents driven to date are 1 kA in CDX ({ital q}{sub {ital a}} =5) and 6 kA in CCT ({ital q}{sub {ital a}} =3.5). An initial comparison of discharge properties with helicity theory indicates rough agreement. Current drive energy efficiencies are 9% and 23% of Ohmic efficiency in two cases analyzed. Strong radial electric fields are observed in these plasmas that cause poloidal rotation and, possibly, improved confinement.

  19. Energetics of runaway electrons during tokamak disruptions

    NASA Astrophysics Data System (ADS)

    Riemann, J.; Smith, H. M.; Helander, P.

    2012-01-01

    In a tokamak disruption, a substantial fraction of the plasma current can be converted into runaway electrons. Although these are usually highly relativistic, their total energy is initially much smaller than that of the pre-disruption plasma. However, following a suggestion by Putvinski et al. [Plasma Phys. Controlled Fusion 39, B157 (1997)], it is shown that as the post-disruption plasma drifts toward the first wall, a non-negligible part of the energy contained in the poloidal magnetic field can be converted into kinetic energy of the runaway electrons. This process is simulated numerically, and it is found that in an ITER-like tokamak runaway electrons can gain kinetic energies up to about 70 MJ by this mechanism.

  20. The physics of tokamak start-upa)

    NASA Astrophysics Data System (ADS)

    Mueller, D.

    2013-05-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  1. The Physics of Tokamak Start-up

    SciTech Connect

    D. Mueller

    2012-11-13

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. ITER, the National Spherical Torus eXperiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  2. Spherical Stellarator-Tokamak Hybrid Configurations

    NASA Astrophysics Data System (ADS)

    Hanson, James D.; Yuan, Ying; Gandy, Rex F.; Knowlton, Stephen F.; Doloc, Cristian; Carnevali, Antonino; Hartwell, Gregory

    1996-11-01

    We consider low-aspect ratio stellarator-tokamak hybrid configurations similar to the inclined coils configurations of Moroz(P. E. Moroz, Phys. Plasmas 2), 4269 (1995). and the Small-Aspect Ratio Toroidal Hybrid(D. B. Batchelor et al)., poster at this meeting. (SMARTH) configurations of Batchelor et al. The advantages of these configurations include a current-free q profile which increases with minor radius, (like a tokamak's), and a magnetic divertor structure which does not rotate about the magnetic axis. Our investigations center on configurations suitable to be built as a small, inexpensive exploratory device. Initial work has focused on planar coils (for ease of construction) and small numbers of toroidal coils (for ease of access). Results from field line tracing, equilibrium, and particle orbit studies will be shown.

  3. Rapidly Moving Divertor Plates In A Tokamak

    SciTech Connect

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  4. First Engineering Commissioning of EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Wan, Yuanxi; Li, Jiangang; Weng, Peide; EAST Team

    2006-05-01

    Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak. The first commissioning started on Feb. 1st of 2006 and finished on March 30th of 2006 at the Institute of Plasma Physics, Chinese Academy of Sciences. It consists of leakage testing at both room temperature and low temperature, pumping down, cooling down all coils, current leads, bus bar and the thermal shielding, exciting all the coils, measuring magnetic configuration and warming up the magnets. The electromagnetic, thermal hydraulic and mechanical performance of EAST Toroidal Field (TF) and Poloidal Field (PF) magnets have also been tested. All sub-systems, including pumping system, cryogenic system, PF& TF power supply systems, magnet instrumentation system, quench detection and protection system, water cooling system, data acquisition system, main control system, plasma control system (PCS), interlock and safety system have been successfully tested.

  5. Microtearing modes in spherical and conventional tokamaks

    NASA Astrophysics Data System (ADS)

    Moradi, S.; Pusztai, I.; Guttenfelder, W.; Fülöp, T.; Mollén, A.

    2013-06-01

    The onset and characteristics of microtearing modes (MTM) in the core of spherical (NSTX) and conventional tokamaks (ASDEX Upgrade and JET) are studied through local linear gyrokinetic simulations with GYRO (Candy and Belli 2011 General Atomics Report GA-A26818). For experimentally relevant core plasma parameters in the NSTX and ASDEX Upgrade tokamaks, in agreement with previous works, we find MTMs as the dominant linear instability. Also, for JET-like core parameters considered in our study an MTM is found as the most unstable mode. In all of these plasmas, finite collisionality is needed for MTMs to become unstable and the electron temperature gradient is found to be the fundamental drive. However, a significant difference is observed in the dependence of the linear growth rate of MTMs on electron temperature gradient. While it varies weakly and non-monotonically in JET and ASDEX Upgrade plasmas, in NSTX it increases with the electron temperature gradient.

  6. High beta plasmas in the PBX tokamak

    SciTech Connect

    Bol, K.; Buchenauer, D.; Chance, M.; Couture, P.; Fishman, H.; Fonck, R.; Gammel, G.; Grek, B.; Ida, K.; Itami, K.

    1986-04-01

    Bean-shaped configurations favorable for high ..beta.. discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present ..beta.. limit.

  7. Fast ion orbits in spherical tokamaks

    SciTech Connect

    Solano, E.R.

    1995-07-20

    In a spherical tokamak, the 1/R variation of the toroidal field is extreme, and for a given value of the safety factor a relatively low average toroidal field can be used, together with large plasma current and large plasma minor radius and elongation. The poloidal and toroidal fields are then of similar size. In consequence, the orbits of fast ions depart considerably from the guiding center orbits because of gyromotion in the small magnetic fields in the low field side.

  8. Self-Organized Stationary States of Tokamaks.

    PubMed

    Jardin, S C; Ferraro, N; Krebs, I

    2015-11-20

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  9. Neoclassical tearing modes in a tokamak

    SciTech Connect

    Hahm, T.S.

    1988-12-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation, so that the growth rate of tearing modes driven by ..delta..' is dramatically reduced compared to the usual resistive magnetohydrodynamic values. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed.

  10. Neoclassical tearing modes in a tokamak

    SciTech Connect

    Hahm, T.S.

    1988-08-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation so that the growth rate of tearing modes driven by ..delta..' is dramatically reduced compared to the usual resistive MHD value. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed. 10 refs.

  11. Self-Organized Stationary States of Tokamaks

    SciTech Connect

    Jardin, S. C.; Ferraro, N.; Krebs, I.

    2015-11-01

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  12. Neoclassical transport in high [beta] tokamaks

    SciTech Connect

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high [beta] large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high [beta] large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low [beta] values by a factor ([var epsilon]/q[sup 2][beta])[sup [1/2

  13. Confinement scaling and ignition in tokamaks

    SciTech Connect

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10/sup 15/ cm/sup -3/, high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition.

  14. Tokamak with liquid metal toroidal field coil

    DOEpatents

    Ohkawa, Tihiro; Schaffer, Michael J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  15. ECH on the MTX (Microwave Tokamak Experiment)

    SciTech Connect

    Stallard, B.W.; Byers, J.A.; Hooper, E.B.; Makowski, M.A.; Meassick, S.; Rice, B.W.; Rognlien, T.D.; Verboncoeur, J.

    1989-04-01

    The Microwave Tokamak Experiment (MTX) at LLNL is investigating the heating of high density Tokamak plasmas using an intense pulse FEL. Our first experiments, now beginning, will study the absorption and plasma heating of single FEL pulses (20 ns pulse length and peak power up to 2 GW) at a frequency of 140 GHz. A later phase of experiments also at 140 GHz will study FEL heating at 5 kHz rate for a pulse train up to 50 pulses (35 ns pulse length and peak power up to 4 GW). Future operations are planned at 250 GHz with an average power of 2 MW for a pulse train of 0.5 s. The microwave output of the FEL is transported quasi-optically to the tokamak through a window-less, evacuated pipe of 20 in. diameter, using a six mirror system. Computational modelling of the non-linear absorption for the MTX geometry predicts single-pass absorption of 40% at a density and temperature of 1.8 /times/ 10/sup 20/m/sup /minus/3/ and 1 keV, respectively. To measure plasma microwave absorption and backscatter, diagnostics are available to measure forward and reflected power (parallel wire grid beam-splitter and mirror directional couplers) and power transmitted through the plasma (segmented calorimeter and waveguide detector). Other fast diagnostics include ECE, Thompson scattering, soft x-rays, and fast magnetic probes. 8 refs., 2 figs.

  16. Management and protection system for superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Juszczyk, B.; Wojenski, A.; Zienkiewicz, P.; Kasprowicz, G.; Pozniak, K.; Romaniuk, R.

    2015-09-01

    This paper describes system for a diagnostics of a high-voltage power supply section of tokamaks. System is designed to assure reliability and safety of power supply subsystems. It is divided into two main components: remote and local. Remote part is located near tokamak, whereas local part can be localised away from the tokamak area. The remote side consists of custom, standalone devices. On the other hand, the local device is based on the uTCA.4 architecture. Components are connected with an optic fibre over a link-layer protocol which provides high throughput, low latency and transmission redundancy. All main operations ie. data processing, transmission etc. are performed on the FPGA devices. At the local side there is one device treated as a master device. It implements sort of a routing table which connects consecutive system inputs and outputs. It also provides possibility for some user defined data processing. This document contains general system overview, short description of hardware used in the project and gateware implementation.

  17. Forced Magnetic Reconnection In A Tokamak Plasma

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Hegna, C. C.

    2015-11-01

    The theory of forced magnetic field reconnection induced by an externally imposed resonant magnetic perturbation usually uses a sheared slab or cylindrical magnetic field model and often focuses on the potential time-asymptotic induced magnetic island state. However, tokamak plasmas have significant magnetic geometry and dynamical plasma toroidal rotation screening effects. Also, finite ion Larmor radius (FLR) and banana width (FBW) effects can damp and thus limit the width of a nascent magnetic island. A theory that is more applicable for tokamak plasmas is being developed. This new model of the dynamics of forced magnetic reconnection considers a single helicity magnetic perturbation in the tokamak magnetic field geometry, uses a kinetically-derived collisional parallel electron flow response, and employs a comprehensive dynamical equation for the plasma toroidal rotation frequency. It is being used to explore the dynamics of bifurcation into a magnetically reconnected state in the thin singular layer around the rational surface, evolution into a generalized Rutherford regime where the island width exceeds the singular layer width, and assess the island width limiting effects of FLR and FBW polarization currents. Support by DoE grants DE-FG02-86ER53218, DE-FG02-92ER54139.

  18. Remote feedback stabilization of tokamak instabilities

    SciTech Connect

    Sen, A.K. )

    1994-05-01

    A novel remote suppressor consisting of an injected ion beam has been used for the stabilization of plasma instabilities. A collisionless curvature-driven trapped-particle instability, an [bold E][times][bold B] flute mode and an ion temperature gradient (ITG) instability have been successfully suppressed down to noise levels using this scheme. Furthermore, the first experimental demonstration of a multimode feedback stabilization with a single sensor--suppressor pair has been achieved. Two modes (an [bold E][times][bold B] flute and an ITG mode) were simultaneously stabilized with a simple state-feedback-type method where more state'' information was generated from a single-sensor Langmuir probe by appropriate signal processing. The above experiments may be considered as paradigms for controlling several important tokamak instabilities. First, feedback suppression of edge fluctuations in a tokamak with a suitable form of insulated segmented poloidal limiter sections used as Langmuir-probe-like suppressors is proposed. Other feedback control schemes are proposed for the suppression of electrostatic core fluctuations via appropriately phased ion density input from a modulated neutral beam. Most importantly, a scheme to control major disruptions in tokamaks via feedback suppression of kink (and possibly) tearing modes is discussed. This may be accomplished by using a modulated neutral beam suppressor in a feedback loop, which will supply a momentum input of appropriate phase and amplitude. Simple theoretical models predict modest levels of beam energy, current, and power.

  19. The Spherical Tokamak MEDUSA for Mexico

    NASA Astrophysics Data System (ADS)

    Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

    2011-10-01

    The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14m, a < 0.10m, BT < 0.5T, Ip < 40kA, 3ms pulse) is currently being recomissioned at the Universidad Autónoma de Nuevo León, Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  20. SOL Width Scaling in the MAST Tokamak

    NASA Astrophysics Data System (ADS)

    Ahn, Joon-Wook; Counsell, Glenn; Connor, Jack; Kirk, Andrew

    2002-11-01

    Target heat loads are determined in large part by the upstream SOL heat flux width, Δ_h. Considerable effort has been made in the past to develop analytical and empirical scalings for Δh to allow reliable estimates to be made for the next-step device. The development of scalings for a large spherical tokamak (ST) such as MAST is particularly important both for development of the ST concept and for improving the robustness of scalings derived for conventional tokamaks. A first such scaling has been developed in MAST DND plasmas. The scaling was developed by flux-mapping data from the target Langmuir probe arrays to the mid-plane and fitting to key upstream parameters such as P_SOL, bar ne and q_95. In order to minimise the effects of co-linearity, dedicated campaigns were undertaken to explore the widest possible range of each parameter while keeping the remainder as fixed as possible. Initial results indicate a weak inverse dependence on P_SOL and approximately linear dependence on bar n_e. Scalings derived from consideration of theoretical edge transport models and integration with data from conventional devices is under way. The established scaling laws could be used for the extrapolations to the future machine such as Spherical Tokamak Power Plant (STPP). This work is jointly funded by Euratom and UK Department of Trade and Industry. J-W. Ahn would like to recognise the support of a grant from the British Foreign & Commonwealth Office.

  1. Vlasov tokamak equilibria with shearad toroidal flow and anisotropic pressure

    NASA Astrophysics Data System (ADS)

    Throumoulopoulos, George; Kuiroukidis, Apostolos; Tasso, Henri

    2015-11-01

    By choosing appropriate deformed Maxwellian ion and electron distribution functions depending on the two particle constants of motion, i.e. the energy and toroidal angular momentum, we reduce the Vlasov axisymmetric equilibrium problem for quasineutral plasmas to a transcendental Grad-Shafranov-like equation. This equation is then solved numerically under the Dirichlet boundary condition for an analytically prescribed boundary possessing a lower X-point to construct tokamak equilibria with toroidal sheared ion flow and anisotropic pressure. Depending on the deformation of the distribution functions these steady states can have toroidal current densities either peaked on the magnetic axis or hollow. These two kinds of equilibria may be regarded as a bifurcation in connection with symmetry properties of the distribution functions on the magnetic axis. This work has received funding from (a) the National Programme for the Controlled Thermonuclear Fusion, Hellenic Republic, (b) Euratom research and training programme 2014-2018 under grant agreement No 633053.

  2. Composition And Electrical Properties Of Dust From Tokamak Compass

    SciTech Connect

    Vaverka, J.; Beranek, M.; Pavlu, J.; Richterova, I.; Vysinka, M.; Safrankova, J.; Nemecek, Z.

    2011-11-29

    In spite of the fact that fusion is a subject of the study for many years, there are still a lot of open questions. One of the interesting topics in fusion research is a presence of dust grains in reactors. In the paper, dust grains born in tokamak Compass are studied and compared with samples of a spherical geometry and well known composition. A unique experimental setup was used for investigations of charging properties of such grains and the SEM and EDX spectroscopy was applied for a study of grain composition. We focus on the secondary emission because this process plays a prominent role when a portion of energetic electrons is present in surroundings of a particular grain. It was shown that depending on the grain size and material, energetic electrons charge the grains to positive potentials comparable with the energy of impinging electrons.

  3. Researchers build a secure plasma prison

    SciTech Connect

    Glanz, J.

    1995-07-28

    Research groups at Princeton University`s Tokamak Fusion Test Reactor and at the DIII-D tokamak at General Atomics in San Diego have made a major breakthrough. By tailoring the magnetic fields with unprecedented finesse, they appear to have tamed the plasma instabilities that rattle and tear the fragile magnetic cage, allowing particles to leak out and limiting a tokamak`s performance. In the process they increased the central density of the plasma by as much as threefold and reduced the particle leakage by a factor of 50.

  4. Recent results from the DIII-D tokamak

    SciTech Connect

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ``isoflux control,`` which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles.

  5. Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

    DOE PAGESBeta

    Soukhanovskii, V. A.; Xu, X.

    2015-09-15

    Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

  6. Public Data Set: High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak

    DOE Data Explorer

    Thome, Kathreen E. [University of Wisconsin-Madison] (ORCID:0000000248013922); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bodner, Grant M. [University of Wisconsin-Madison] (ORCID:0000000324979172); Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Kriete, David M. [University of Wisconsin-Madison] (ORCID:0000000236572911); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609); Schlossberg, David J. [University of Wisconsin-Madison] (ORCID:0000000287139448)

    2016-04-27

    This data set contains openly-documented, machine readable digital research data corresponding to figures published in K.E. Thome et al., 'High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak,' Phys. Rev. Lett. 116, 175001 (2016).

  7. D3-D research operations

    NASA Astrophysics Data System (ADS)

    Lahaye, R. J.

    1994-05-01

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

  8. Recent progress on the Compact Ignition Tokamak (CIT)

    SciTech Connect

    Ignat, D.W.

    1987-01-01

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule.

  9. Numerical investigations of plasma parameters in the COMPASS tokamak

    SciTech Connect

    Havlickova, E.; Zagorski, R.; Panek, R.

    2008-09-15

    A numerical investigation of plasma parameters in a diverter configuration of COMPASS tokamak is presented. The plasma parameters in the device are analyzed in the frame of the self-consistent description of the central plasma and edge region. The possibility of achieving high recycling and detached regimes in the boundary layer of the COMPASS tokamak is discussed.

  10. Fokker-Planck/Transport model for neutral beam driven tokamaks

    SciTech Connect

    Killeen, J.; Mirin, A.A.; McCoy, M.G.

    1980-01-01

    The application of nonlinear Fokker-Planck models to the study of beam-driven plasmas is briefly reviewed. This evolution of models has led to a Fokker-Planck/Transport (FPT) model for neutral-beam-driven Tokamaks, which is described in detail. The FPT code has been applied to the PLT, PDX, and TFTR Tokamaks, and some representative results are presented.

  11. A simulation study of a controlled tokamak plasma

    NASA Astrophysics Data System (ADS)

    Fujii, N.; Niwa, Y.

    1980-03-01

    A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.

  12. Effects of neoclassical toroidal viscosity induced by the intrinsic error fields and toroidal field ripple on the toroidal rotation in tokamaks

    NASA Astrophysics Data System (ADS)

    Lee, H. H.; Seol, J.; Ko, W. H.; Terzolo, L.; Aydemir, A. Y.; In, Y.; Ghim, Y.-c.; Lee, S. G.

    2016-08-01

    Effects of neoclassical toroidal viscosity (NTV) induced by intrinsic error fields and toroidal field ripple on cocurrent toroidal rotation in H-mode tokamak plasmas are investigated. It is expected that large NTV torque can be localized at the edge region through the 1/ν-regime in the vicinity of E r ˜ 0 in the cocurrent rotating H-mode plasma. Numerical simulation on toroidal rotation demonstrates that the edge localized NTV torque determined by the intrinsic error fields and toroidal field ripples in the level of most tokamaks can damp the toroidal rotation velocity over the whole region while reducing the toroidal rotation pedestal which is clearly observed in Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. It is found that the NTV torque changes the toroidal rotation gradient in the pedestal region dramatically, but the toroidal rotation profile in the core region responds rigidly without a change in the gradient. On the other hand, it shows that the NTV torque induced by the intrinsic error fields and toroidal field ripple in the level of the KSTAR tokamak, which are expected to be smaller than most tokamaks by at least one order of magnitude, is negligible in determining the toroidal rotation velocity profile. Experimental observation on the toroidal rotation change by the externally applied nonaxisymmetric magnetic fields on KSTAR also suggests that NTV torque arising from nonaxisymmetric magnetic fields can damp the toroidal rotation over the whole region while diminishing the toroidal rotation pedestal.

  13. Tokamak Physics Experiment (TPX) power supply design and development

    SciTech Connect

    Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

    1995-04-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes.

  14. Hybrid Fusion: The Only Viable Development Path for Tokamaks?

    NASA Astrophysics Data System (ADS)

    Manheimer, Wallace

    2009-03-01

    The world needs a great deal of carbon free energy, and soon, for civilization to continue. Fusion's goal is to develop such a carbon free energy source. For the last 4 decades, tokamaks have been the best magnetic fusion has to offer. But what if its development stops short of commercial fusion? This paper introduces `conservative design principles' for tokamaks. These are very simple, are reasonably based in theory, and have always constrained tokamak operation. Assuming they continue to do so, it is unlikely that tokamaks will ever make it as commercial reactors. This is independent of their confinement properties. However because of the large additional gain in hybrid fusion, tokamaks reactors look like they can make it as hybrid fuel producers, and provide large scale power by mid century or shortly thereafter.

  15. First neutral beam injection experiments on KSTAR tokamak.

    PubMed

    Jeong, S H; Chang, D H; Kim, T S; In, S R; Lee, K W; Jin, J T; Chang, D S; Oh, B H; Bae, Y S; Kim, J S; Park, H T; Watanabe, K; Inoue, T; Kashiwagi, M; Dairaku, M; Tobari, H; Hanada, M

    2012-02-01

    The first neutral beam (NB) injection system of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak was partially completed in 2010 with only 1∕3 of its full design capability, and NB heating experiments were carried out during the 2010 KSTAR operation campaign. The ion source is composed of a JAEA bucket plasma generator and a KAERI large multi-aperture accelerator assembly, which is designed to deliver a 1.5 MW, NB power of deuterium at 95 keV. Before the beam injection experiments, discharge, and beam extraction characteristics of the ion source were investigated. The ion source has good beam optics in a broad range of beam perveance. The optimum perveance is 1.1-1.3 μP, and the minimum beam divergence angle measured by the Doppler shift spectroscopy is 0.8°. The ion species ratio is D(+):D(2)(+):D(3)(+) = 75:20:5 at beam current density of 85 mA/cm(2). The arc efficiency is more than 1.0 A∕kW. In the 2010 KSTAR campaign, a deuterium NB power of 0.7-1.5 MW was successfully injected into the KSTAR plasma with a beam energy of 70-90 keV. L-H transitions were observed within a wide range of beam powers relative to a threshold value. The edge pedestal formation in the T(i) and T(e) profiles was verified through CES and electron cyclotron emission diagnostics. In every deuterium NB injection, a burst of D-D neutrons was recorded, and increases in the ion temperature and plasma stored energy were found.

  16. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    SciTech Connect

    Chektybayev, B. Shapovalov, G.; Kolodeshnikov, A.

    2015-05-15

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  17. Observation of ICRF (ion cyclotron range of frequencies) wave-packet propagation in a tokamak plasma

    SciTech Connect

    Greene, G.J.; Gould, R.W.

    1987-11-01

    Experimental observation of ICRF wave-packet propagation in a tokamak plasma is reported. Studies were carried out in the Caltech Research Tokamak in a pure hydrogen plasma and in a regime where fast-wave damping was sufficiently small to permit multiple toroidal transits of the wave-packet. Waves were launched by exciting a small loop antenna with a short burst of rf current and were detected with shielded magnetic probes. Probe scans revealed a large increase in wave-packet amplitude at smaller minor radii, and the packet velocity was found to be independent of radial position. Measurement of the packet transit time yielded direct information about the wave group velocity. Packet velocity was investigated as a function of the fundamental excitation frequency, plasma density, and toroidal magnetic field. Results are compared with the predictions of a cold plasma model which includes a vacuum layer at the edge. 24 refs., 8 figs.

  18. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    NASA Astrophysics Data System (ADS)

    Chektybayev, B.; Shapovalov, G.; Kolodeshnikov, A.

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  19. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

    PubMed

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  20. Development of DEMO-FNS tokamak for fusion and hybrid technologies

    NASA Astrophysics Data System (ADS)

    Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-07-01

    The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

  1. Mathematical modeling plasma transport in tokamaks

    SciTech Connect

    Quiang, Ji

    1995-12-31

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10{sup 20}/m{sup 3} with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  2. Electrostatic analysis of the tokamak edge plasma

    SciTech Connect

    Motley, R.W.

    1981-07-01

    The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in two ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter driving potentials, and (2) it creates zones of back current flow into the limiter. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer provides free energy to drive the edge plasma unstable. Special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer.

  3. Viscosity in the edge of tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Stacey, W. M.

    1993-05-01

    A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the 'short radial gradient scale length' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates.

  4. Magnetic field measurements in tokamak plasmas

    SciTech Connect

    Feldman, U.; Seely, J.F.; Sheeley,Jr., N.R.; Suckewer, S.; Title, A.M.

    1984-11-01

    The measurement of the poloidal magnetic field in a tokamak plasma from the Zeeman splitting and polarization of the magnetic dipole radiation from heavy ions is discussed. When viewed from a direction perpendicular to the toroidal field, the effect of the poloidal field on the circularly polarized radiation is detectable using a photoelectric polarimeter. The Zeeman splittings for a number of magnetic dipole transitions with wavelengths in the range 2300--9300 A are presented. An imaging polarimeter is proposed that can measure the poloidal magnetic field with space and time resolution.

  5. Neoclassical Transport Properties of Tokamak Plasmas

    SciTech Connect

    Weyssow, B.

    2004-03-15

    The classical transport theory is strictly valid for a plasma in a homogeneous and stationary magnetic field. In the '60, experiments have shown that this theory does not apply as a local theory of transport in Tokamaks. It was shown that global geometric characteristics of the confining elements have a strong influence on the transport. Three regimes of collisionality are characteristic of the neoclassical transport theory: the banana regime (the electronic diffusion coefficient increases starting from zero), the plateau regime (the diffusion coefficient is almost independent of the collisionality) and the Pfirsch-Schlueter regime (the electronic diffusion coefficient again increases with the collisionality)

  6. Self-Organized Stationary States of Tokamaks

    DOE PAGESBeta

    Jardin, S. C.; Ferraro, N.; Krebs, I.

    2015-11-17

    We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.

  7. Nonlinear gyrokinetic equations for tokamak microturbulence

    SciTech Connect

    Hahm, T.S.

    1988-05-01

    A nonlinear electrostatic gyrokinetic Vlasov equation, as well as Poisson equation, has been derived in a form suitable for particle simulation studies of tokamak microturbulence and associated anomalous transport. This work differs from the existing nonlinear gyrokinetic theories in toroidal geometry, since the present equations conserve energy while retaining the crucial linear and nonlinear polarization physics. In the derivation, the action-variational Lie perturbation method is utilized in order to preserve the Hamiltonian structure of the original Vlasov-Poisson system. Emphasis is placed on the dominant physics of the collective fluctuations in toroidal geometry, rather than on details of particle orbits. 13 refs.

  8. 3D MHD Simulations of Tokamak Disruptions

    NASA Astrophysics Data System (ADS)

    Woodruff, Simon; Stuber, James

    2014-10-01

    Two disruption scenarios are modeled numerically by use of the CORSICA 2D equilibrium and NIMROD 3D MHD codes. The work follows the simulations of pressure-driven modes in DIII-D and VDEs in ITER. The aim of the work is to provide starting points for simulation of tokamak disruption mitigation techniques currently in the CDR phase for ITER. Pressure-driven instability growth rates previously observed in simulations of DIIID are verified; Halo and Hiro currents produced during vertical displacements are observed in simulations of ITER with implementation of resistive walls in NIMROD. We discuss plans to exercise new code capabilities and validation.

  9. Physics evaluation of compact tokamak ignition experiments

    SciTech Connect

    Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

    1985-01-01

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/S/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs.

  10. Numerical optimization of perturbative coils for tokamaks

    NASA Astrophysics Data System (ADS)

    Lazerson, Samuel; Park, Jong-Kyu; Logan, Nikolas; Boozer, Allen; NSTX-U Research Team

    2014-10-01

    Numerical optimization of coils which apply three dimensional (3D) perturbative fields to tokamaks is presented. The application of perturbative 3D magnetic fields in tokamaks is now commonplace for control of error fields, resistive wall modes, resonant field drive, and neoclassical toroidal viscosity (NTV) torques. The design of such systems has focused on control of toroidal mode number, with coil shapes based on simple window-pane designs. In this work, a numerical optimization suite based on the STELLOPT 3D equilibrium optimization code is presented. The new code, IPECOPT, replaces the VMEC equilibrium code with the IPEC perturbed equilibrium code, and targets NTV torque by coupling to the PENT code. Fixed boundary optimizations of the 3D fields for the NSTX-U experiment are underway. Initial results suggest NTV torques can be driven by normal field spectrums which are not pitch-resonant with the magnetic field lines. Work has focused on driving core torque with n = 1 and edge torques with n = 3 fields. Optimizations of the coil currents for the planned NSTX-U NCC coils highlight the code's free boundary capability. This manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the U.S. Department of Energy.

  11. Constrained ripple optimization of Tokamak bundle divertors

    SciTech Connect

    Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.

  12. Plasma engineering analysis of Tennessee Tokamak

    SciTech Connect

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Shannon, T.E.; Uckan, N.A.

    1983-01-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses dueterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. Detailed analyses are performed in the areas of (1) transport simulation using the 1-1/2-D WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control sytems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

  13. Applications of fast wave in spherical tokamaks

    SciTech Connect

    Chiu, S.C.; Chan, V.S.; Lin-Liu, Y.R.; Miller, R.L.; Prater, R.; Politzer, P.

    1997-04-01

    In spherical tokamaks (ST), the magnetic field strength varies over a wide range across the plasma, and at high betas it deviates significantly from the 1/R dependence of conventional tokamaks. This, together with the high density expected in ST, poses challenging problems for RF heating and current drive. In this paper, the authors investigate the various possible applications of fast waves (FW) in ST. The adjoint technique of calculating current drive is implemented in the raytracing code CURRAY. The applicability of high harmonic and subharmonic FW to steady state ST is considered. They find that high harmonic FW tends to be totally absorbed before reaching the core and may be considered a candidate for off axis current drive while the subharmonic FW tends to be absorbed mainly in the core region and may be considered for central current drive. A difficult problem is the maintenance of current at the startup stage. In the bootstrap ramp-up scenario, the current ramp-up is mainly provided by the bootstrap current. Under this condition, the role of rf becomes mainly the sustainment of plasma through electron heating. Using a slab full-wave code SEMAL, the authors find that the ion-ion-hybrid mode conversion scheme is a promising candidate. The effect of possible existence of edge Alfven resonance and high harmonic cyclotron resonance is investigated and regimes of minimization of edge heating identified.

  14. Applications of fast wave in spherical tokamaks

    SciTech Connect

    Chiu, S.C.; Chan, V.S.; Lin-Liu, Y.R.; Miller, R.L.; Prater, R.; Politzer, P.

    1997-04-01

    In spherical tokamaks (ST), the magnetic field strength varies over a wide range across the plasma, and at high betas it deviates significantly from the 1/R dependence of conventional tokamaks. This, together with the high density expected in ST, poses challenging problems for RF heating and current drive. In this paper, we investigate the various possible applications of fast waves (FW) in ST. The adjoint technique of calculating current drive is implemented in the raytracing code CURRAY. The applicability of high harmonic and subharmonic FW to steady state ST is considered. We find that high harmonic FW tends to be totally absorbed before reaching the core and may be considered a candidate for off axis current drive while the subharmonic FW tends to be absorbed mainly in the core region and may be considered for central current drive. A difficult problem is the maintenance of current at the startup stage. In the bootstrap ramp-up scenario, the current ramp-up is mainly provided by the bootstrap current. Under this condition, the role of rf becomes mainly the sustainment of plasma through electron heating. Using a slab full-wave code SEMAL, we find that the ion-ion-hybrid mode conversion scheme is a promising candidate. The effect of possible existence of edge Alfv{acute e}n resonance and high harmonic cyclotron resonance is investigated and regimes of minimization of edge heating identified. {copyright} {ital 1997 American Institute of Physics.}

  15. Electrostatic Dust Detection and Removal in Tokamaks

    NASA Astrophysics Data System (ADS)

    Hensley, R.; Skinner, C. H.; Roquemore, A. L.

    2006-10-01

    The inventory of in-vessel dust particles in next-step tokamaks will increase with the rise in stored energy and pulse duration. Dust levels will need to be measured and controlled for safety reasons and to avoid plasma contamination. A novel electrostatic dust detector has been developed with a sensitivity appropriate for the carbon dust levels expected in next-step devices.^23 Higher sensitivity is desired for real-time measurements in contemporary tokamaks that have less dust. We report on results from a larger area, more sensitive detector. A 2 x 2 circuit board has two interlocking combs of copper traces spaced by 25 microns and biased at 30-50 V. The carbon test dust is delivered to the circuit board by a mesh tray vibrated at 60 Hz. The impinging dust creates a short circuit and the resulting current pulse is recorded. We will present results on the dust detection sensitivity and dust removal efficiency of these new detectors in three environments: air, vacuum, and inert gas. ^2 C. Voinier et al., J. Nucl. Mater. 346 (2005) 266-271. ^3 C. Parker et al., PPPL Report, PPPL-4169.

  16. Modular pump limiter systems for large tokamaks

    NASA Astrophysics Data System (ADS)

    Uckan, T.; Klepper, C. C.; Mioduszewski, P. K.; McGrath, R. T.

    1987-09-01

    Long-pulse (greater than 10-s) operation of large tokamaks with high-power (greater than 10-MW) heating and extensive external fueling will require correspondingly efficient particle exhaust for density control. A pump limiter can provide the needed exhaust capability by removing a small percentage of the particles, which would otherwise be recycled. Single pump limiter modules have been operated successfully on ISX-B, PDX, TEXTOR, and PLT. An axisymmetric pump limiter is now being installed and will be studied in TEXTOR. A third type of pump limiter is a system that consists of several modules and exhibits performance different from that of a single module. To take advantage of the flexibility of a modular pump limiter system in a high-power, long-pulse device, the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The design parameters for the modules are then determined from the system requirements for particle and power removal. Design criteria and parameters are presented, and the impact on module design of the state of the art in engineering technolgy is discussed. The relationship between modules is considered from the standpoint of flux coverage and shadowing effects. The results are applied to the Tore Supra tokamak. A preliminary conceptual design for the Tore Supra pump limiter system is discussed, and the design parameters of the limiter modules are presented.

  17. Modular pump limiter systems for large tokamaks

    SciTech Connect

    Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.; McGrath, R.T.

    1987-09-01

    Long-pulse (>10-s) operation of large tokamaks with high-power (>10-MW) heating and extensive external fueling will require correspondingly efficient particle exhaust for density control. A pump limiter can provide the needed exhaust capability by removing a small percentage of the particles, which would otherwise be recycled. Single pump limiter modules have been operated successfully on ISX-B, PDX, TEXTOR, and PLT. An axisymmetric pump limiter is now being installed and will be studied in TEXTOR. A third type of pump limiter is a system that consists of several modules and exhibits performance different from that of a single module. To take advantage of the flexibility of a modular pump limiter system in a high-power, long-pulse device, the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The design parameters for the modules are then determined from the system requirements for particle and power removal. Design criteria and parameters are presented, and the impact on module design of the state of the art in engineering technology is discussed. The relationship between modules are considered from the standpoint of flux coverage and shadowing effects. The results are applied to the Tore Supra tokamak. A preliminary conceptual design for the Tore Supra pump limiter system is discussed, and the design parameters of the limiter modules are presented. 21 refs., 12 figs.

  18. Neoclassical theory inside transport barriers in tokamaks

    NASA Astrophysics Data System (ADS)

    Shaing, K. C.; Hsu, C. T.

    2012-02-01

    Inside the transport barriers in tokamaks, ion energy losses sometimes are smaller than the value predicted by the standard neoclassical theory. This improvement can be understood in terms of the orbit squeezing theory in addition to the sonic poloidal E ×B Mach number Up,m that pushes the tips of the trapped particles to the higher energy. In general, Up,m also includes the poloidal component of the parallel mass flow speed. These physics mechanisms are the corner stones for the transition theory of the low confinement mode (L-mode) to the high confinement mode (H-mode) in tokamaks. Here, detailed transport fluxes in the banana regime are presented using the parallel viscous forces calculated earlier. It is found, as expected, that effects of orbit squeezing and the sonic Up,m reduce the ion heat conductivity. The former reduces it by a factor of |S|3/2 and the later by a factor of R(Up ,m2)exp(-Up ,m2) with R(Up ,m2), a rational function. Here, S is the orbit squeezing factor.

  19. /sup 3/He functions in tokamak-pumped laser systems

    SciTech Connect

    Jassby, D.L.

    1986-10-01

    /sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.

  20. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    NASA Astrophysics Data System (ADS)

    Zheng, Linjin; Horton, W.; Miura, H.; Shi, T. H.; Wang, H. Q.

    2016-08-01

    Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew-Goldburger-Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

  1. Study of Plasma Interaction with Titanium Coated Ferritic Steel in IR-T1 Tokamak

    SciTech Connect

    Ghoranneviss, M.; Talebitaher, A.; Arvin, R.; Mohammadi, S.; Nikmohamadi, A.; Milani, M.; Salem, M. K.; Sari, A. H.; Yousefi, M. R.; Shokouhi, A.; Khorshid, P.; Saboohi, S.

    2008-04-07

    Studies of plasma interaction with titanium coated ferritic steel is performed on IR-T1 tokamak. Titanium coating is one of the candidates for the plasma facing materials in a tokomak. Titaniumization is carried out by a sputtering method. Some of the samples were baked (3 hours at 460 deg. C) before sputtering. Atomic Force Microscopy (AFM) analyses before and after discharge in r/a = l .04 carried out. The samples (with distinctive titanium layers) were placed at different depths inside the vacuum vessel of the IR-T1 tokamak in the SOL region. A comparison of the titanium coated steel with bare ferritic steel exposed to plasma tokamak and glow discharges is made in this research. Depth of impurity penetration and retention, and the surface roughness are measured by using surface analysis methods. Rutherford backscattering method is used to measure the content of nitrogen, oxygen and titanium, before and after discharges. The result is shown a change in roughness with respect to position of samples.

  2. Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

    NASA Astrophysics Data System (ADS)

    Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; Holcomb, C. T.; Lao, L. L.; McKee, G. R.; Meneghini, O.; Staebler, G. M.; Grierson, B. A.; Qian, J. P.; Solomon, W. M.; Turnbull, A. D.; Holland, C.; Guo, W. F.; Ding, S. Y.; Pan, C. K.; Xu, G. S.; Wan, B. N.

    2016-06-01

    Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN , despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

  3. Study of Plasma Interaction with Titanium Coated Ferritic Steel in IR-T1 Tokamak

    NASA Astrophysics Data System (ADS)

    Ghoranneviss, M.; Khorshid, P.; Saboohi, S.; Talebitaher, A.; Arvin, R.; Mohammadi, S.; Nikmohamadi, A.; Milani, M.; Salem, M. K.; Sari, A. H.; Yousefi, M. R.; Shokouhi, A.

    2008-04-01

    Studies of plasma interaction with titanium coated ferritic steel is performed on IR-T1 tokamak. Titanium coating is one of the candidates for the plasma facing materials in a tokomak. Titaniumization is carried out by a sputtering method. Some of the samples were baked (3 hours at 460 °C) before sputtering. Atomic Force Microscopy (AFM) analyses before and after discharge in r/a = l .04 carried out. The samples (with distinctive titanium layers) were placed at different depths inside the vacuum vessel of the IR-T1 tokamak in the SOL region. A comparison of the titanium coated steel with bare ferritic steel exposed to plasma tokamak and glow discharges is made in this research. Depth of impurity penetration and retention, and the surface roughness are measured by using surface analysis methods. Rutherford backscattering method is used to measure the content of nitrogen, oxygen and titanium, before and after discharges. The result is shown a change in roughness with respect to position of samples.

  4. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak.

    PubMed

    Ren, J; Zuo, G Z; Hu, J S; Sun, Z; Yang, Q X; Li, J G; Zakharov, L E; Xie, H; Chen, Z X

    2015-02-01

    A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak-both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST. PMID:25725839

  5. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak.

    PubMed

    Ren, J; Zuo, G Z; Hu, J S; Sun, Z; Yang, Q X; Li, J G; Zakharov, L E; Xie, H; Chen, Z X

    2015-02-01

    A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak-both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  6. Quiescent double barrier high-confinement mode plasmas in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Burrell, K. H.; Austin, M. E.; Brennan, D. P.; DeBoo, J. C.; Doyle, E. J.; Fenzi, C.; Fuchs, C.; Gohil, P.; Greenfield, C. M.; Groebner, R. J.; Lao, L. L.; Luce, T. C.; Makowski, M. A.; McKee, G. R.; Moyer, R. A.; Petty, C. C.; Porkolab, M.; Rettig, C. L.; Rhodes, T. L.; Rost, J. C.; Stallard, B. W.; Strait, E. J.; Synakowski, E. J.; Wade, M. R.; Watkins, J. G.; West, W. P.

    2001-05-01

    High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 energy confinement times τE), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved βNH89=7 for up to 5 times the τE of 150 ms. The βNH89 values of 7 substantially exceed the value of 4 routinely achieved in the standard ELMing H mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H mode is possible because of the presence of an edge magnetohydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.

  7. ECH by FEL and gyrotron sources on the Microwave Tokamak Experiment (MTX) tokamak

    SciTech Connect

    Stallard, B.W.; Turner, W.C.; Allen, S.L.; Byers, J.A.; Felker, B.; Fenstermacher, M.E.; Ferguson, S.W.; Hooper, E.G.; Thomassen, K.I.; Throop, A.L. ); Makowski, M.A. )

    1990-08-09

    The Microwave Tokamak Experiment (MTX) at LLNL is studying the physics of intense pulse ECH is a high-density tokamak plasma using a microwave FEL. Related technology development includes the FEL, a windowless quasi-optical transmission system, and other microwave components. Initial plasma experiments have been carried out at 140 GHz with single rf pulses generated using the ETA-II accelerator and the ELF wiggler. Peak power levels up to 0.2 GW and pulse durations up to 10 ns were achieved for injection into the plasma using as untapered wiggler. FEL pulses were transmitted over 33 m from the FEL to MTX using six mirrors mounted in a 50-cm-diam evacuated pipe. Measurements of the microwave beam and transmission through the plasma were carried out. For future rapid pulse experiments at high average power (4 GW peak power, 5kHz pulse rate, and {bar P} > 0.5 MW) using the IMP wiggler with tapered magnetic field, a gyrotron (140 GHz, 400 kW cw or up to 1 MW short pulse) is being installed to drive the FEL input or to directly heat the tokamak plasma at full gyrotron power. Quasi-optic techniques will be used to couple the gyrotron power. For direct plasma heating, the gyrotron will couple into the existing mirror transport system. Using both sources of rf generation, experiments are planned to investigate intense pulse absorption and tokamak physics, such as the ECH of a pellet-fueled plasma and plasma control using localized heating. 12 refs., 9 figs.

  8. Realizing steady-state tokamak operation for fusion energy

    NASA Astrophysics Data System (ADS)

    Luce, T. C.

    2011-03-01

    Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

  9. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  10. Neutral beam injector performance on the PLT and PDX tokamaks

    SciTech Connect

    Schilling, G.; Ashcroft, D.L.; Eubank, H.P.; Grisham, L.R.; Kozub, T.A.; Kugel, H.W.; Rossmassler, J.; Williams, M.D.

    1981-02-01

    An overall injector system description is presented first, and this will be followed by a detailed discussion of those problems unique to multiple injector operation on the tokamaks, i.e., power transmission, conditioning, reliability, and failures.

  11. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    NASA Astrophysics Data System (ADS)

    Giunta, A. S.; Henderson, S.; O'Mullane, M.; Harrison, J.; Doyle, J. G.; Summers, H. P.

    2016-09-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  12. Compact Ignition Tokamak Program: status of FEDC studies

    SciTech Connect

    Flanagan, C.A.

    1985-01-01

    Viewgraphs on the Compact Ignition Tokamak Program comprise the report. The technical areas discussed are the mechanical configuration status, magnet analysis, stress analysis, cooling between burns, TF coil joint, and facility/device layout options. (WRF)

  13. An emerging understanding of H-mode discharges in tokamaks

    SciTech Connect

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the [upsilon][sub E][sup [yields

  14. Operating tokamaks with steady-state toroidal current

    SciTech Connect

    Fisch, N.J.

    1981-04-01

    Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements.

  15. Neoclassical diffusion of heavy impurities in a rotating tokamak plasma

    SciTech Connect

    Wong, K.L.; Cheng, C.Z.

    1987-08-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle simulation is carried out and the results offer a qualitative explanation for some experimental data from the Tokamak Test Reactor (TFTR). 13 refs., 2 figs.

  16. Fast ion effects on magnetic instabilities in the PDX tokamak

    NASA Astrophysics Data System (ADS)

    Buchenauer, D. A.

    Modification and excitation of nondisruptive magnetic instabilities due to near perpendicular neutral beam injection on the PDX tokamak were made to determine the importance of these instabilities at low q. The instabilities consisted of resistive MHD modes, beam driven ideal MHD modes, and beam driven ion cyclotron modes. Evidence of enhanced transport is presented for several of these instabilities as well as comparison of the experimental results with theory. Possible consequences for reactor type tokamaks and high power auxiliary heating systems are discussed.

  17. 'Snowflake' H Mode in a Tokamak Plasma

    SciTech Connect

    Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.

    2010-10-08

    An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy ({Delta}W{sub ELM}/W{sub p}) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.

  18. Anisotropic pressure tokamak equilibrium and stability considerations

    SciTech Connect

    Salberta, E.R.; Grimm, R.C.; Johnson, J.L.; Manickam, J.; Tang, W.M.

    1987-02-01

    Investigation of the effect of pressure anisotropy on tokamak equilibrium and stability is made with an MHD model. Realistic perpendicular and parallel pressure distributions, P/sub perpendicular/(psi,B) and P/sub parallel/(psi,B), are obtained by solving a one-dimensional Fokker-Planck equation for neutral beam injection to find a distribution function f(E, v/sub parallel//v) at the position of minimum field on each magnetic surface and then using invariance of the magnetic moment to determine its value at each point on the surface. The shift of the surfaces of constant perpendicular and parallel pressure from the flux surfaces depends strongly on the angle of injection. This shift explains the observed increase or decrease in the stability conditions. Estimates of the stabilizing effect of hot trapped ions indicates that a large fraction must be nonresonant and thus decoupled from the bad curvature before it becomes important.

  19. Transport Bifurcation in a Rotating Tokamak Plasma

    SciTech Connect

    Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

    2010-11-19

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  20. Cooldown of the Compact Ignition Tokamak

    SciTech Connect

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  1. Fast tomographic methods for the tokamak ISTTOK

    SciTech Connect

    Carvalho, P. J.; Coelho, R.; Neto, A.; Pereira, T.; Silva, C.; Fernandes, H.; Gori, S.; Toussaint, U. v.

    2008-04-07

    The achievement of long duration, alternating current discharges on the tokamak IST-TOK requires a real-time plasma position control system. The plasma position determination based on magnetic probes system has been found to be inadequate during the current inversion due to the reduced plasma current. A tomography diagnostic has been therefore installed to supply the required feedback to the control system. Several tomographic methods are available for soft X-ray or bolo-metric tomography, among which the Cormack and Neural networks methods stand out due to their inherent speed of up to 1000 reconstructions per second, with currently available technology. This paper discusses the application of these algorithms on fusion devices while comparing performance and reliability of the results. It has been found that although the Cormack based inversion proved to be faster, the neural networks reconstruction has fewer artifacts and is more accurate.

  2. Nonlinear lower hybrid modeling in tokamak plasmas

    SciTech Connect

    Napoli, F.; Schettini, G.; Castaldo, C.; Cesario, R.

    2014-02-12

    We present here new results concerning the nonlinear mechanism underlying the observed spectral broadening produced by parametric instabilities occurring at the edge of tokamak plasmas in present day LHCD (lower hybrid current drive) experiments. Low frequency (LF) ion-sound evanescent modes (quasi-modes) are the main parametric decay channel which drives a nonlinear mode coupling of lower hybrid (LH) waves. The spectrum of the LF fluctuations is calculated here considering the beating of the launched LH wave at the radiofrequency (RF) operating line frequency (pump wave) with the noisy background of the RF power generator. This spectrum is calculated in the frame of the kinetic theory, following a perturbative approach. Numerical solutions of the nonlinear LH wave equation show the evolution of the nonlinear mode coupling in condition of a finite depletion of the pump power. The role of the presence of heavy ions in a Deuterium plasma in mitigating the nonlinear effects is analyzed.

  3. Tearing mode analysis in tokamaks, revisited

    SciTech Connect

    Nishimura, Y.; Callen, J.D.; Hegna, C.C.

    1998-12-01

    A new {Delta}{sup {prime}} shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio ({epsilon}{le}0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of the finite pressure term. Numerical results compare favorably with Furth {ital et al.} [H. P. Furth {ital et al.}, Phys. Fluids {bold 16}, 1054 (1973)] results. The effects of finite pressure, which are shown to decrease {Delta}{sup {prime}}, are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric elements, stabilizes the tearing mode significantly, even in a low-{beta} regime before the toroidal magnetic curvature effects come into play. {copyright} {ital 1998 American Institute of Physics.}

  4. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  5. Plasma rotation in the PDX tokamak

    SciTech Connect

    Brau, K.; Bitter, M.; Goldston, R.J.; Manos, D.; McGuire, K.; Suckewer, S.

    1983-06-01

    Toroidal and poloidal rotation has been measured in the Poloidal Divertor Experiment (PDX) tokamak in ohmic- and neutral-beam-heated plasmas in a variety of discharge conditions and in both circular and diverted configurations. Rotation velocities were deduced from Doppler shifts of magnetic dipole (M1) lines and lines of optically allowed transitions in the visible and uv regions, from K/sub ..cap alpha../ emission, and also from an array of magnetic pickup loops. Poloidal and toroidal rotation velocities in ohmically heated discharges were unusually less than 3 x 10/sup 5/ cm/sec. Near the plasma edge the toroidal-rotation velocity varies with poloidal angle both before and during neutral-beam injection. No systematic poloidal rotation was observed during neutral-beam injection centered about or displaced 10 cm from the horizontal midplane, which implies that the poloidal damping time tau/sub theta/ < 0.5 tau/sub ii/, consistent with theoretical estimates.

  6. Numerical simulation of fueling in tokamaks

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.; Milora, S.L.

    1982-04-01

    We describe the numerical simulation of fueling and particle transport in both present and future tokamak plasmas. Models for pellet ablation and plasma density behavior after pellet injection are compared with experimental results in ISX and PDX plasmas and then extended to fusion reactor conditions. The role of fast ion ablation due to intense neutral beam injection and fusion alphas is examined along with pellet size and velocity considerations. In plasmas with high pumping efficiency (which may be obtained with divertor operation), pellet injection can significantly reduce fueling rates while maintaining more flexibility in control of the density profile than afforded by gas puffing. When fueling is dominated by gas puffing or high recycle from the walls or limiter, control of the fueling and density profiles is reduced and particle fluxes to the wall increase.

  7. Numerical simulation of fueling in tokamaks

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.; Milora, S.L.

    1981-01-01

    We describe the numerical simulation of fueling and particle transport in both present and future tokamak plasmas. Models for pellet ablation and plasma density behavior after pellet injection are compared with experimental results in ISX and PDX plasmas and then extended to fusion reactor conditions. The role of fast ion ablation due to intense neutral beam injection and fusion alphas is examined along with pellet size and velocity considerations. In plasmas with high pumping efficiency (which may be obtained with divertor operation), pellet injection can significantly reduce fuel handling requirements and interaction of the plasma with the chamber walls while maintaining more flexibility in control of the density profile than afforded by gas puffing. When fueling is dominated by gas puffing or high recycle from the walls or limiter, control of the fueling and density profiles is reduced while plasma/wall interactions increase.

  8. Control of Asymmetric Magnetic Perturbations in Tokamaks

    SciTech Connect

    Park, Jong-kyu; Schaffer, Michael J.; Menard, Jonathan E.; Boozer, Allen H.

    2007-10-03

    The sensitivity of tokamak plasmas to very small deviations from the axisymmetry of the magnetic field |δ→(over)Β/→(over)Β|≈ 10–4 is well known. What was not understood until very recently is the importance of the perturbation to the plasma equilibrium in assessing the effects of externally produced asymmetries in the magnetic field, even far from a stability limit. DIII-D and NSTX experiments find that when the deleterious effects of asymmetries are mitigated, the external asymmetric field was often made stronger and with an increased interaction with the magnetic field of the unperturbed equilibrium fields. This paper explains these counter intuitive results. The explanation using ideal perturbed equilibria has important implications for the control of field errors in all toroidal plasmas.

  9. Vertically stabilized elongated cross-section tokamak

    DOEpatents

    Sheffield, George V.

    1977-01-01

    This invention provides a vertically stabilized, non-circular (minor) cross-section, toroidal plasma column characterized by an external separatrix. To this end, a specific poloidal coil means is added outside a toroidal plasma column containing an endless plasma current in a tokamak to produce a rectangular cross-section plasma column along the equilibrium axis of the plasma column. By elongating the spacing between the poloidal coil means the plasma cross-section is vertically elongated, while maintaining vertical stability, efficiently to increase the poloidal flux in linear proportion to the plasma cross-section height to achieve a much greater plasma volume than could be achieved with the heretofore known round cross-section plasma columns. Also, vertical stability is enhanced over an elliptical cross-section plasma column, and poloidal magnetic divertors are achieved.

  10. Trail-A Tokamak RAIL Gun Limiter

    SciTech Connect

    Yu, W.S; Fillo, J.A.; Powell, J.R.; Usher, J.L.

    1984-09-01

    An attractive new limiter concept is investigated. The Tokamak RAIl Gun Limiter (TRAIL) system directs a stream of moderate velocity pellets (100 to 200 m/s) through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an electromagnetic mass accelerator. Heat fluxes of about30000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements ( about0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state of the art (several kilometres per second). Accelerators injecting pellets at about1 km/s can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  11. TRAIL: a tokamak rail gun limiter

    SciTech Connect

    Yu, W S; Powell, J R; Usher, J L

    1980-01-01

    An attractive new limiter concept is investigated. The TRAIL (Tokamak Rail Gun Limiter) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled after cooling, to the injector of an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state of the art (several Km/sec). Accelerators injecting pellets at approx. 1 Km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  12. TRAIL: a tokamak rail gun limiter

    SciTech Connect

    Yu, W.S.; Powell, J.R.; Usher, J.L.

    1980-01-01

    An attractive new limiter concept is investigated. The TRAIL (Tokamak Rail Gun Limiter) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  13. Dust Studies in DIII-D Tokamak

    SciTech Connect

    Rudakov, D L; West, W P; Groth, M; Yu, J H; Boedo, J A; Bray, B D; Brooks, N H; Fenstermacher, M E; Hollmann, E M; Hyatt, A W; Krasheninnikov, S I; Lasnier, C J; Moyer, R A; Pigarov, A Y; Smirnov, R; Solomon, W M; Wong, C C

    2008-04-15

    Studies of submicron dust using Mie scattering from Nd:YAG lasers and video data of micron to sub-millimeter sized dust on DIII-D tokamak have provided the first data of dust sources and transport during tokamak discharges. During normal operation on DIII-D dust observation rates are low, a few events per discharge or less. The net carbon content of the dust corresponds to a carbon atom density a few orders of magnitude below the core impurity density. Statistical analysis of Mie data collected over months of operation reveal correlation of increased dust rate with increased heating power and impulsive wall loading due to edge localized modes (ELMs) and disruptions. Generation of significant amounts of dust by disruptions is confirmed by the camera data. However, dust production by disruptions alone is insufficient to account for estimated in-vessel dust inventory in DIII-D. After an extended entry vent, thousands of dust particles are observed by cameras in the first 2-3 plasma discharges. Individual particles moving at velocities up to {approx}300 m/s, breakup of larger particles into pieces, and collisions of particles with walls are observed. After {approx}70 discharges, dust levels are reduced to a few events per discharge. In order to calibrate diagnostics and benchmark modeling, milligram amounts of micron-sized carbon dust have been injected into DIII-D discharges, leading to the core carbon density increase by a factor of 2-3. Following injection, dust trajectories in the divertor are mostly in the toroidal direction, consistent with the ion drag force. Dust from the injection is observed in the outboard midplane by a fast framing camera. The observed trajectories and velocities of the dust particles are in qualitative agreement with modeling by the 3D DustT code.

  14. Dust Studies in DIII-D Tokamak

    SciTech Connect

    Rudakov, D. L.; Yu, J. H.; Boedo, J. A.; Hollmann, E. M.; Krasheninnikov, S. I.; Moyer, R. A.; Pigarov, A. Yu.; Smirnov, R.; West, W. P.; Bray, B. D.; Brooks, N. H.; Hyatt, A. W.; Wong, C. P. C.; Groth, M.; Fenstermacher, M. E.; Lasnier, C. J.; Solomon, W. M.

    2008-09-07

    Studies of submicron dust using Mie scattering from Nd:YAG lasers and video data of micron to sub-millimeter sized dust on DIII-D tokamak have provided the first data of dust sources and transport during tokamak discharges. During normal operation on DIII-D dust observation rates are low, a few events per discharge or less. The net carbon content of the dust corresponds to a carbon atom density a few orders of magnitude below the core impurity density. Statistical analysis of Mie data collected over months of operation reveal correlation of increased dust rate with increased heating power and impulsive wall loading due to edge localized modes (ELMs) and disruptions. Generation of significant amounts of dust by disruptions is confirmed by the camera data. However, dust production by disruptions alone is insufficient to account for estimated in-vessel dust inventory in DIII-D. After an extended entry vent, thousands of dust particles are observed by cameras in the first 2-3 plasma discharges. Individual particles moving at velocities up to {approx}300 m/s, breakup of larger particles into pieces, and collisions of particles with walls are observed. After {approx}70 discharges, dust levels are reduced to a few events per discharge. In order to calibrate diagnostics and benchmark modeling, milligram amounts of micron-sized carbon dust have been injected into DIII-D discharges, leading to the core carbon density increase by a factor of 2-3. Following injection, dust trajectories in the divertor are mostly in the toroidal direction, consistent with the ion drag force. Dust from the injection is observed in the outboard midplane by a fast framing camera. The observed trajectories and velocities of the dust particles are in qualitative agreement with modeling by the 3D DustT code.

  15. Self-organized stationary states of tokamaks

    NASA Astrophysics Data System (ADS)

    Jardin, Stephen

    2015-11-01

    We report here on a nonlinear mechanism that forms and maintains a self-organized stationary (sawtooth free) state in tokamaks. This process was discovered by way of extensive long-time simulations using the M3D-C1 3D extended MHD code in which new physics diagnostics have been added. It is well known that most high-performance modes of tokamak operation undergo ``sawtooth'' cycles, in which the peaking of the toroidal current density triggers a periodic core instability which redistributes the current density. However, certain modes of operation are known, such as the ``hybrid'' mode in DIII-D, ASDEX-U, JT-60U and JET, and the long-lived modes in NSTX and MAST, which do not experience this cycle of instability. Empirically, it is observed that these modes maintain a non-axisymmetric equilibrium which somehow limits the peaking of the toroidal current density. The physical mechanism responsible for this has not previously been understood, but is often referred to as ``flux-pumping,'' in which poloidal flux is redistributed in order to maintain q0 >1. In this talk, we show that in long-time simulations of inductively driven plasmas, a steady-state magnetic equilibrium may be obtained in which the condition q0 >1 is maintained by a dynamo driven by a stationary marginal core interchange mode. This interchange mode, unstable because of the pressure gradient in the ultra-low shear region in the center region, causes a (1,1) perturbation in both the electrostatic potential and the magnetic field, which nonlinearly cause a (0,0) component in the loop voltage that acts to sustain the configuration. This hybrid mode may be a preferred mode of operation for ITER. We present parameter scans that indicate when this sawtooth-free operation can be expected.

  16. The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

    SciTech Connect

    Kessel, C. E.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Tillack, M. S.; Najmabadi, F.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; El-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Radar, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.

    2014-03-05

    Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a {beta}N{sup total} of 5.75, H{sub 98} of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m{sup 2}. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a {beta}N{sup total} of 2.5, H{sub 98} of 1.25, n/n{sub Gr} of 1.3, and peak divertor heat flux of 10 MW/m{sup 2}. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m{sup 2}. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.

  17. The ARIES Advanced and Conservative Tokamak Power Plant Study

    SciTech Connect

    Kessel, C. E; Tillak, M. S; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; EL-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Rader, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.

    2015-12-22

    Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.

  18. The ARIES Advanced and Conservative Tokamak Power Plant Study

    DOE PAGESBeta

    Kessel, C. E; Tillak, M. S; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; et al

    2015-12-22

    Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, anmore » n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

  19. Diagnostic of fusion neutrons on JET tokamak using diamond detector

    NASA Astrophysics Data System (ADS)

    Nemtsev, G.; Amosov, V.; Marchenko, N.; Meshchaninov, S.; Rodionov, R.; Popovichev, S.; JET EFDA contributors

    2014-08-01

    In 2011-2012, an experimental campaign with a significant yield of fusion neutrons was carried out on the JET tokamak. During this campaign the facility was equipped with two diamond detectors based on natural and artificial CVD diamond. These detectors were designed and manufactured in State Research Center of Russian Federation TRINITI. The detectors measure the flux of fast neutrons with energies above 0.2 MeV. They have been installed in the torus hall and the distance from the center of plasma was about 3 m. For some of the JET pulses in this experiment, the neutron flux density corresponded to the operational conditions in collimator channels of ITER Vertical Neutron Camera. The main objective of diamond monitors was the measurement of total fast neutron flux at the detector location and the estimation of the JET total neutron yield. The detectors operate as threshold counters. Additionally a spectrometric measurement channel has been configured that allowed us to distinguish various energy components of the neutron spectrum. In this paper we describe the neutron signal measuring and calibration procedure of the diamond detector. Fluxes of DD and DT neutrons at the detector location were measured. It is shown that the signals of total neutron yield measured by the diamond detector correlate with signals measured by the main JET neutron diagnostic based on fission chambers with high accuracy. This experiment can be considered as a successful test of diamond detectors in ITER-like conditions.

  20. Spectroscopic Measurements on the Lithium Tokamak eXperiment

    NASA Astrophysics Data System (ADS)

    Granstedt, E. M.; Kaita, R.; Majeski, R.; Gray, T. K.; Maingi, R.; Tritz, K.; Soukhanovskii, V. A.

    2011-10-01

    The Lithium Tokamak eXperiment (LTX) is a spherical torus designed to investigate the very low-recycling, lithium wall regime for magnetically confined plasmas. Since lithium surfaces primarily influence plasma performance through their effect on wall recycling, comprehensive measurements of hydrogen fluxes from the wall are necessary. Three instruments measure Lyman- α emission around most of the poloidal cross-section: two arrays view the inboard shell and outboard shell, and a single diode views a molybdenum limiter. These measurements will be used with a neutral transport code to calculate recycling and the fueling profile. Lithium wall conditioning also affects plasma performance through modifying wall impurity fluxes. A visible survey spectrometer and filterscope measurements of lithium, carbon, and oxygen emission lines are used to quantify the fluxes of light impurities ejected from the walls. Trends in the core penetration of these light impurities are measured by an XUV spectrometer, which is also used to examine high-Z impurity emission. Finally, an AXUV array is used as a radiometer to quantify the radiation emission profile. Supported by US DOE contracts DE-AC02-09CH11466, DE-AC52-07NA27344 and an NSF Graduate Research Fellowship.

  1. Diagnostic of fusion neutrons on JET tokamak using diamond detector

    SciTech Connect

    Nemtsev, G.; Amosov, V.; Marchenko, N.; Meshchaninov, S.; Rodionov, R.; Popovichev, S.; Collaboration: JET EFDA Conbributors

    2014-08-21

    In 2011-2012, an experimental campaign with a significant yield of fusion neutrons was carried out on the JET tokamak. During this campaign the facility was equipped with two diamond detectors based on natural and artificial CVD diamond. These detectors were designed and manufactured in State Research Center of Russian Federation TRINITI. The detectors measure the flux of fast neutrons with energies above 0.2 MeV. They have been installed in the torus hall and the distance from the center of plasma was about 3 m. For some of the JET pulses in this experiment, the neutron flux density corresponded to the operational conditions in collimator channels of ITER Vertical Neutron Camera. The main objective of diamond monitors was the measurement of total fast neutron flux at the detector location and the estimation of the JET total neutron yield. The detectors operate as threshold counters. Additionally a spectrometric measurement channel has been configured that allowed us to distinguish various energy components of the neutron spectrum. In this paper we describe the neutron signal measuring and calibration procedure of the diamond detector. Fluxes of DD and DT neutrons at the detector location were measured. It is shown that the signals of total neutron yield measured by the diamond detector correlate with signals measured by the main JET neutron diagnostic based on fission chambers with high accuracy. This experiment can be considered as a successful test of diamond detectors in ITER-like conditions.

  2. Polarimetric spectra analysis for tokamak pitch angle measurements

    NASA Astrophysics Data System (ADS)

    Ko, J.; Chung, J.; Lange, A. G. G.; de Bock, M. F. M.

    2013-10-01

    Measurements of the internal magnetic field structures using conventional polarimetric approaches are considered extremely challenging in fusion-reactor environments whereas the information on current density profiles is essential to establish steady-state and advance operation scenarios in such reactor-relevant devices. Therefore, on ITER a hybrid system is proposed for the current density measurements that uses both polarimetry and spectral measurements. The spectrum-based approaches have been tested in the Korea Superconducting Tokamak Advanced Research (KSTAR) during the past two plasma campaigns. As such, KSTAR is a test-bed for the proposed ITER hybrid system. Measurements in the plasma core are based on the motional Stark effect (MSE) spectrum of the neutral beam emission. For the edge profiles, the Zeeman effect (ZE) acting on the lithium emission spectrum of the newly installed (2013) Lithium-beam-diagnostic is exploited. The neutral beam emission spectra, complicated by the multi-ion-source beam injection, are successfully fitted making use of the data provided by the Atomic Data and Analysis Structure (ADAS) database package. This way pitch angle profiles could be retrieved from the beam emission spectra. With the same spectrometer/CCD hardware as on MSE, but with a different wavelength range and different lines of sight, the first ZE spectrum measurements have been made. The Zeeman splitting comparable to and greater than the instrumental broadening has been routinely detected at high toroidal field operations ( ~ 3 Tesla).

  3. EBT: an alternate concept to tokamaks and mirrors

    SciTech Connect

    Glowienka, J.C.

    1980-01-01

    The ELMO Bumpy Torus (EBT) is a hybrid magnetic trap formed by a series of toroidally connected simple mirrors. It differs from a tokamak, the present main-line approach, in that plasma stability and heating are obtained in a current-free geometry by the application of steady-state, high power, electron cyclotron resonance heating (ECH) producing a steady-state plasma. The primary motivation for EBT confinement research is the potential for a steady-state, highly accessible reactor with high ..beta... In the present EBT-I/S device, electron confinement has been observed to agree with the predictions of theory. The major emphasis of the experimental program is on the further scaling of plasma parameters in the EBT-I/S machine with ECH frequency (10.6, 18, and 28 GHz), resonant magnetic field (0.3, 0.6, and 1 T), and heating power (30, 60, and 200 kW). In addition, substantial efforts are under way or planned in the areas of ion cyclotron heating, neutral beam heating, plasma-wall interactions, impurity control, synchrotron radiation, and divertors. Recently, EBT has been selected as the first alternative concept to be advanced to the proof-of-principle stage; this entails a major device scale-up to allow a reasonable extrapolation to a DT-burning facility. The status and future plans of the EBT program, in particular the proof-of-principle experiment (EBT-P), are discussed.

  4. Remote maintenance concepts for the Compact Ignition Tokamak

    SciTech Connect

    Davis, F.C.; Hager, E.R.

    1988-01-01

    Because deuterium-tritium fuel will be used in the Compact Ignition Tokamak (CIT), remote handling technology is needed to carry out some maintenance operations on the machine. In keeping with the compact, low-cost nature of CIT, remote maintenance is provided only for systems with the highest probability of failure. Remote operations include removing, repairing (if feasible), and replacing such components as thermal protection tiles on the first wall, radio-frequency (rf) heating modules, and diagnostic modules. For maintenance inside the vacuum vessel, major pieces of equipment under development include an articulated boom manipulator with servomanipulators, an inspection manipulator, and special tooling. For maintenance outside the cryostat, remote equipment includes a bridge-mounted manipulator system, equipment for decontamination and hot cell activities, and for handling and packaging solid radioactive waste. The conceptual design phase of the CIT project is nearing completion; research and development activities in support of the project include demonstrations of remote maintenance operations on full-size partial mock-ups. 9 figs.

  5. SODA: The reduced database for the TdeV tokamak

    SciTech Connect

    Cote, A.; Michaud, D.; Caumartin, J.; de Villers, P.; Gauthier, Y.; Gauvreau, J.; Larsen, J.

    1997-01-01

    SODA which stands for Syst{grave e}me d{close_quote}Organisation des Donn{acute e}es et d{close_quote}Analyse, is a general database for TdeV. SODA has the following goals: to produce a database of a reduced set of physical data; to ensure that these data are validated; to record all the parameters relevant to tokamak operation and experiments; to facilitate the retrieval of data using given selection criteria; and to improve data accessibility and analysis. The relational database ORACLE{trademark} has been chosen to provide flexibility and to accommodate the increasing expectations of the TdeV researchers. In-house expertise allows custom-made tables and centralized data management. In the process of creating SODA several new interfaces for the scientific coordinator, machine operator, and diagnosticians have been added to provide a better definition of the experiment for the archiving system. The database includes the more relevant machine and diagnostic parameters, plasma perturbations (rf, biasing, gas{hor_ellipsis}), mean and standard deviation of physical signals, plasma profiles, and code results (equilibrium{hor_ellipsis}) for selected time windows in a discharge. Users of the X-window interface of SODA are not required to know the database structure or the SQL language. SODA has been operating successfully for over a year and its capabilities are continuously expanding. {copyright} {ital 1997 American Institute of Physics.}

  6. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    SciTech Connect

    BURRELL,HK

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, they have made significant progress in developing the building blocks needed for AT operation: (1) they have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. The authors have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiated power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  7. Maryland Controlled Fusion Research Program. Progress report, November 1, 1993--October 31, 1994

    SciTech Connect

    Antonsen, T.M. Jr.; Drake, J.F.; Finn, J.M.; Guzdar, P.N.; Hassam, A.; Liu, C.S.; Ott, E.

    1994-07-01

    The theoretical research activity at the University of Maryland supported by the Department of Energy focusses on two major tasks. They are: A. Transport, generation of shear flow and L-H transitions in Tokamaks; and B. MHD stability of Tokamaks. Titles of specific research projects under these fields carried out in this period include: Poloidal rotation of tokamak plasmas at super-poloidal-sonic speeds; Rotation of plasmas in neoclassical regimes; New unstable branch of drift resistive ballooning modes in tokamaks; 3D fluid simulations of the drift resistive ballooning modes - twisted coordinate system; Disintegration of ion banana orbits in tokamak edge plamsas; Fast reconnection in high temperature plasmas; Dynamics of sawtooth collapse in tokamak plasmas; Effect of trapped particles on MHD modes; MHD stability of high beta toroidal equilibria.

  8. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    NASA Astrophysics Data System (ADS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G. P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-04-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  9. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    SciTech Connect

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-12-31

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  10. Progress toward long-pulse high-performance Advanced Tokamak discharges on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Wade, M. R.; Luce, T. C.; Politzer, P. A.; Ferron, J. R.; Allen, S. L.; Austin, M. E.; Baker, D. R.; Bray, B.; Brennen, D. P.; Burrell, K. H.; Casper, T. A.; Chu, M. S.; DeBoo, J. C.; Doyle, E. J.; Garofalo, A. M.; Gohil, P.; Gorelov, I. A.; Greenfield, C. M.; Groebner, R. J.; Heidbrink, W. W.; Hsieh, C.-L.; Hyatt, A. W.; Jayakumar, R.; Kinsey, J. E.; La Haye, R. J.; Lao, L. L.; Lasnier, C. J.; Lazarus, E. A.; Leonard, A. W.; Lin-Liu, Y. R.; Lohr, J.; Mahdavi, M. A.; Makowski, M. A.; Murakami, M.; Petty, C. C.; Pinsker, R. I.; Prater, R.; Rettig, C. L.; Rhodes, T. L.; Rice, B. W.; Strait, E. J.; Taylor, T. S.; Thomas, D. M.; Turnbull, A. D.; Watkins, J. G.; West, W. P.; Wong, K.-L.

    2001-05-01

    Significant progress has been made in obtaining high-performance discharges for many energy confinement times in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. Normalized performance (measured by the product of βNH89 and indicative of the proximity to both conventional β limits and energy confinement quality, respectively) ˜10 has been sustained for >5 τE with qmin>1.5. These edge localized modes (ELMing) H-mode discharges have β˜5%, which is limited by the onset of resistive wall modes slightly above the ideal no-wall n=1 limit, with approximately 75% of the current driven noninductively. The remaining Ohmic current is localized near the half-radius. The DIII-D electron cyclotron heating system is being upgraded to replace this inductively driven current with localized electron cyclotron current drive (ECCD). Density control, which is required for effective ECCD, has been successfully demonstrated in long-pulse high-performance ELMing H-mode discharges with βNH89˜7 for up to 6.3 s. In plasma shapes compatible with good density control in the present divertor configuration, the achieved βN is somewhat less than that in the high βNH89=10 discharges.

  11. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    SciTech Connect

    1995-09-22

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided.

  12. Electromagnetic Torque in Tokamaks with Toroidal Asymmetries

    NASA Astrophysics Data System (ADS)

    Logan, Nikolas Christopher

    Toroidal rotation and rotation shear strongly influences stability and confinement in tokamaks. Breaking of the toroidal symmetry by fields orders of magnitude smaller than the axisymmetric field can, however, produce electromagnetic torques that significantly affect the plasma rotation, stability and confinement. These electromagnetic torques are the study of this thesis. There are two typical types of electromagnetic torques in tokamaks: 1) "resonant torques" for which a plasma current defined by a single toroidal and single poloidal harmonic interact with external currents and 2) "nonresonant torques" for which the global plasma response to nonaxisymmetric fields is phase shifted by kinetic effects that drive the rotation towards a neoclassical offset. This work describes the diagnostics and analysis necessary to evaluate the torque by measuring the rate of momentum transfer per unit area in the vacuum region between the plasma and external currents using localized magnetic sensors to measure the Maxwell stress. These measurements provide model independent quantification of both the resonant and nonresonant electromagnetic torques, enabling direct verification of theoretical models. Measured values of the nonresonant torque are shown to agree well with the perturbed equilibrium nonambipolar transport (PENT) code calculation of torque from cross field transport in nonaxisymmetric equilibria. A combined neoclassical toroidal viscosity (NTV) theory, valid across a wide range of kinetic regimes, is fully implemented for the first time in general aspect ratio and shaped plasmas. The code captures pitch angle resonances, reproducing previously inaccessible collisionality limits in the model. The complete treatment of the model enables benchmarking to the hybrid kinetic MHD stability codes MARS-K and MISK, confirming the energy-torque equivalency principle in perturbed equilibria. Experimental validations of PENT results confirm the torque applied by nonaxisymmetric

  13. Magnetic control of magnetohydrodynamic instabilities in tokamaks

    SciTech Connect

    Strait, E. J.

    2015-02-15

    Externally applied, non-axisymmetric magnetic fields form the basis of several relatively simple and direct methods to control magnetohydrodynamic (MHD) instabilities in a tokamak, and most present and planned tokamaks now include a set of non-axisymmetric control coils for application of fields with low toroidal mode numbers. Non-axisymmetric applied fields are routinely used to compensate small asymmetries (δB/B∼10{sup −3} to 10{sup −4}) of the nominally axisymmetric field, which otherwise can lead to instabilities through braking of plasma rotation and through direct stimulus of tearing modes or kink modes. This compensation may be feedback-controlled, based on the magnetic response of the plasma to the external fields. Non-axisymmetric fields are used for direct magnetic stabilization of the resistive wall mode—a kink instability with a growth rate slow enough that feedback control is practical. Saturated magnetic islands are also manipulated directly with non-axisymmetric fields, in order to unlock them from the wall and spin them to aid stabilization, or position them for suppression by localized current drive. Several recent scientific advances form the foundation of these developments in the control of instabilities. Most fundamental is the understanding that stable kink modes play a crucial role in the coupling of non-axisymmetric fields to the plasma, determining which field configurations couple most strongly, how the coupling depends on plasma conditions, and whether external asymmetries are amplified by the plasma. A major advance for the physics of high-beta plasmas (β = plasma pressure/magnetic field pressure) has been the understanding that drift-kinetic resonances can stabilize the resistive wall mode at pressures well above the ideal-MHD stability limit, but also that such discharges can be very sensitive to external asymmetries. The common physics of stable kink modes has brought significant unification to the topics of static error

  14. Power supplies and quench protection for the Tokamak Physics Experiment

    SciTech Connect

    Neumeyer, C.L.

    1994-07-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.

  15. Tokamak power system studies at ANL

    SciTech Connect

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-06-01

    The following features, in particular, have been examined: (a) large aspect ratio (A approx. = 6), which may ease maintenance; (b) high beta (..beta.. greater than or equal to 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I approx. = 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors.

  16. Thermo-Oxidation of Tokamak Carbon Dust

    SciTech Connect

    J.W. Davis; B.W.N. Fitzpatrick; J.P. Sharpe; A.A. Haasz

    2008-04-01

    The oxidation of dust and flakes collected from the DIII-D tokamak, and various commercial dust specimens, has been measured at 350 ºC and 2.0 kPa O2 pressure. Following an initial small mass loss, most of the commercial dust specimens showed very little effect due to O2 exposure. Similarly, dust collected from underneath DIII-D tiles, which is thought to comprise largely Grafoil™ particulates, also showed little susceptibility to oxidation at this temperature. However, oxidation of the dust collected from tile surfaces has led to ~ 18% mass loss after 8 hours; thereafter, little change in mass was observed. This suggests that the surface dust includes some components of different composition and/or structure – possibly fragments of codeposited layers. The oxidation of codeposit flakes scraped form DIII-D upper divertor tiles showed an initial 25% loss in mass due to heating in vacuum, and the gradual loss of 30-38% mass during the subsequent 24 hours exposure to O2. This behavior is significantly different from that observed for the oxidation of thinner DIII-D codeposit specimens which were still adhered to tile surfaces, and this is thought to be related to the low deuterium content (D/C ~ 0.03 – 0.04) of the flakes.

  17. Continuous tokamak operation with an internal transformer

    SciTech Connect

    Singer, C.E.; Mikkelsen, D.R.

    1982-10-01

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia timescales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.

  18. Carbonization of the DIII-D tokamak

    SciTech Connect

    Jackson, G.L.; Lippmann, S.; Petrie, T.W.; DeBoo, J.C.; Ferron, J.R.; Schissel, D.P.; Taylor, T.S. ); Winters, J. . Inst. fuer Plasmaphysik); Hill, D.N. )

    1990-12-01

    The DIII-D tokamak has been carbonized by the application of a carbon film covering all plasma facing surfaces. Carbonization was done in order to reduce the metal impurity influx and central metal accumulation especially during beam heated D{sup O} {yields} D{sup +} H-mode discharges. After carbonization, nickel impurity line radiation was reduced by a factor of 10 during the ohmic phase of the discharge and up to a factor of 30 during the H-mode phase. The reduction of metal impurities also produced a reduction of total radiated power and allowed high current operation. After carbonization, the highest plasma current in a double null divertor, and the highest stored energy ever achieved on DIII-D were observed, 3 MA and 3.6 MJ respectively. A toroidal beta, {beta}{sub T} = 5.1%, at full field, 2.1 T, was obtained. On the first day after carbonization, H-mode density profiles were more peaked than pre-carbonization discharges. 20 refs., 6 figs.

  19. Tokamak L/H mode transition

    SciTech Connect

    Tsui, K. H.; Navia, C. E.

    2012-01-15

    Through the non field-aligned rotational tokamak equilibrium of a divergence-free plasma flow with a pair of transformed plasma variables w-vector{sub *}=({mu}{rho}){sup 1/2}{nu}-vector and {mu}p{sub *}=({mu}p+w{sub *}{sup 2}/2)[K. H. Tsui, Phys. Plasmas 18, 072502 (2011)], a preliminary understanding of the L/H equilibrium transition is proposed through a feedback cycle, where the higher plasma flux due to external drives enters the rotational Grad-Shafranov equation through the velocity dependent poloidal plasma {beta} to generate the H equilibrium. This H rotational mode has the characteristics of higher normal electric field and plasma pressure. Coupled to the transport properties of E-vector x B-vector drift transport barrier leading to a higher plasma pressure, this makes the H mode a self-sustained equilibrium. The higher plasma {beta} then feeds back to the equilibrium and completes the feedback loop.

  20. Pellet imaging techniques in the ASDEX tokamak

    SciTech Connect

    Wurden, G.A. ); Buechl, K.; Hofmann, J.; Lang, R.; Loch, R.; Rudyj, A.; Sandmann, W. )

    1990-11-01

    As part of a USDOE/ASDEX collaboration, a detailed examination of pellet ablation in ASDEX with a variety of diagnostics has allowed a better understanding of a number of features of hydrogen ice pellet ablation in a plasma. In particular, fast-gated photos with an intensified Xybion CCD video camera allow {ital in} {ital situ} velocity measurements of the pellet as it penetrates the plasma. With time resolution of typically 100 ns and exposures every 50 {mu}s, the evolution of each pellet in a multipellet ASDEX tokamak plasma discharge can be followed. When the pellet cloud track has striations, the light intensity profile through the cloud is hollow (dark near the pellet), whereas at the beginning or near the end of the pellet trajectory the track is typically smooth (without striations) and has a gaussian-peaked light emission profile. New, single pellet Stark broadened {ital D}{sub {alpha}}, {ital D}{sub {beta}}, and {ital D}{sub {gamma}} spectra, obtained with a tangentially viewing scanning mirror/spectrometer with Reticon array readout, are consistent with cloud densities of 2{times}10{sup 17} cm{sup {minus}3} or higher in the regions of strongest light emission. A spatially resolved array of {ital D}{sub {alpha}} detectors shows that the light variations during the pellet ablation are not caused solely by a modulation of the incoming energy flux as the pellet crosses rational {ital q} surfaces, but instead are a result of dynamic, nonstationary, ablation process.

  1. Divertor design for the Tokamak Physics Experiment

    SciTech Connect

    Hill, D.N.; Braams, B.; Brooks, J.N.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

  2. Physics issues of high bootstrap current tokamaks

    NASA Astrophysics Data System (ADS)

    Ozeki, T.; Azumi, M.; Ishii, Y.; Kishimoto, Y.; Fu, G. Y.; Fujita, T.; Rewoldt, G.; Kikuchi, M.; Kamada, Y.; Kimura, H.; Kusama, Y.; Saigusa, M.; Ide, S.; Shirai, H.

    1997-05-01

    Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of qmin is close to the region of high $\</p>
      </li>

      <li>
      <p><a target=Thomson scattering on the Texas Experimental Tokamak

    SciTech Connect

    Porter, J.L. Jr.

    1985-01-01

    A Thomson scattering diagnostic was constructed on the Texas Experimental Tokamak (TEXT) which is capable of measuring local electron temperatures in the range 10 eV less than or equal to T/sub e/ less than or equal to 2000 eV for densities greater than or equal to 5 x 10/sup 12/ cm/sup -3/. This diagnostic has been used to study the edge region of the plasma, the scaling of the central electron temperature and Z/sub effective/ over a wide range of discharge conditions, and the evolution of the electron temperature profile during the plasma formation and during a fast current ramp initiated well into the discharge. Current diffusion was modeled during the current rise portion of the standard TEXT discharge and during discharges in which the plasma current was rapidly increased after a steady-state discharge had been established by using the measured electron temperature profiles as input to a computer code which solves the one dimensional poloidal magnetic field diffusion equation. Because of the large differences found between the predicted and measured values of the loop voltage and the quantity ..beta../sub p/ + 1/sub i//2 during the initial current rise portion of the discharge it is concluded that the plasma current penetrates faster than can be explained by classical resistive current diffusion during this portion of the discharge.

  3. Tearing mode analysis in tokamaks, revisited

    SciTech Connect

    Nishimura, Y.; Callen, J.D.; Hegna, C.C.

    1997-12-01

    A new {Delta}{prime} shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio ({epsilon} {le} 0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of finite pressure term. Numerical results compare favorably with Furth et al. results. The effects of finite pressure, which are shown to decrease {Delta}{prime}, are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric element stabilizes the tearing mode significantly, even in a low {beta} regime before the toroidal magnetic curvature effects come into play. Double tearing modes in toroidal geometries are examined as well. Furthermore, m {ge} 2 tearing mode stability criteria are compared with three dimensional initial value MHD simulation by the FAR code.

  4. Fast bolometric measurements on the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Furno, I.; Weisen, H.; Mlynar, J.; Pitts, R. A.; Llobet, X.; Marmillod, Ph.; Pochon, G. P.

    1999-12-01

    The design and first results are presented from a bolometric diagnostic with high temporal resolution recently installed on the TCV tokamak. The system consists of two pinhole cameras viewing the plasma from above and below at the same toroidal location. Each camera is equipped with an AXUV-16ELO linear array of 16 p-n junction photodiodes, characterized by a flat spectral sensitivity from ultraviolet to x-ray energies, a high temporal response (<0.5 μs), and insensitivity to low-energy neutral particles emitted by the plasma. This high temporal resolution allows the study of transient phenomena such as fast magnetohydrodynamic (MHD) activity hitherto inaccessible with standard bolometry. In the case of purely electromagnetic radiation, good agreement has been found when comparing results from the new diagnostic with those from a standard metal foil bolometer system. This comparison has also revealed that the contribution of neutrals to the foil bolometer measurements can be extremely important under certain operating conditions, precluding the application of tomographic techniques for reconstruction of the radiation distribution.

  5. Neoclassical transport in high {beta} tokamaks

    SciTech Connect

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high {beta} large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high {beta} large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low {beta} values by a factor ({var_epsilon}/q{sup 2}{beta}){sup {1/2}} II. This factor is the ratio of plasma volume in the boundary layer to the volume in the core. The fraction of trapped particles on a given flux surface (f{sub t}) is also reduced by this factor so that {approximately} {sub ({var_epsilon}}/q{sup 2}{beta}){sup {1/2}}. Special attention is given to the current equation, since this is thought to be relevant at low 3 and therefore may also be relevant at high {beta}. The bootstrap current term is found to exceed the actual current by a factor of the square root of the aspect ratio.

  6. Controlling sawtooth oscillations in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Chapman, I. T.

    2011-01-01

    The sawtooth instability in tokamak plasmas results in a periodic reorganization of the core plasma. A typical sawtooth cycle consists of a quiescent period, during which the plasma density and temperature increase, followed by the growth of a helical magnetic perturbation, which in turn is followed by a rapid collapse of the central pressure. The stabilizing effects of fusion-born α particles are likely to lead to long sawtooth periods in burning plasmas. However, sawteeth with long quiescent periods have been observed to result in the early triggering of neo-classical tearing modes (NTMs) at low plasma pressure, which can, in turn, significantly degrade confinement. Consequently, recent experiments have identified various methods to deliberately control sawtooth oscillations in an attempt to avoid seeding NTMs whilst retaining the benefits of small, frequent sawteeth, such as the prevention of core impurity accumulation. Sawtooth control actuators include current drive schemes, such as electron cyclotron current drive, and tailoring the fast ion population in the plasma using neutral beam injection or ion cyclotron resonance heating.

  7. Multiple time scale methods in tokamak magnetohydrodynamics

    SciTech Connect

    Jardin, S.C.

    1984-01-01

    Several methods are discussed for integrating the magnetohydrodynamic (MHD) equations in tokamak systems on other than the fastest time scale. The dynamical grid method for simulating ideal MHD instabilities utilizes a natural nonorthogonal time-dependent coordinate transformation based on the magnetic field lines. The coordinate transformation is chosen to be free of the fast time scale motion itself, and to yield a relatively simple scalar equation for the total pressure, P = p + B/sup 2//2..mu../sub 0/, which can be integrated implicitly to average over the fast time scale oscillations. Two methods are described for the resistive time scale. The zero-mass method uses a reduced set of two-fluid transport equations obtained by expanding in the inverse magnetic Reynolds number, and in the small ratio of perpendicular to parallel mobilities and thermal conductivities. The momentum equation becomes a constraint equation that forces the pressure and magnetic fields and currents to remain in force balance equilibrium as they evolve. The large mass method artificially scales up the ion mass and viscosity, thereby reducing the severe time scale disparity between wavelike and diffusionlike phenomena, but not changing the resistive time scale behavior. Other methods addressing the intermediate time scales are discussed.

  8. Zonal flows in tokamaks with anisotropic pressure

    NASA Astrophysics Data System (ADS)

    Ren, Haijun

    2014-04-01

    Zonal flows (ZFs) in a tokamak plasma with anisotropic pressure are investigated. The dynamics of perpendicular and parallel pressures are determined by the Chew-Goldberger-Low double equations and low-β condition is adopted, where β is the ratio of plasma pressure to the magnetic field pressure. The dispersion relation is analytically derived and illustrates two branches of ZFs. The low frequency zonal flow (LFZF) branch becomes unstable when χ, the ratio of the perpendicular pressure to the parallel one, is greater than a threshold value χc, which is about 3.8. In the stable region, its frequency increases first and then decreases with increasing χ. For χ = 1, the frequency of LFZF agrees well with the experimental observation. For the instability, the growth rate of LFZF increases with χ. The geodesic acoustic mode branch is shown to be always stable with a frequency increasing with χ. The safety factor is shown to diminish the frequencies of both branches or the growth rate of LFZF.

  9. Impurities in the Lithium Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Boyle, D. P.; Bell, R. E.; Kaita, R.; Majeski, R.; Biewer, T. M.; Gray, T. K.; Tritz, K.; Widmann, K.

    2014-10-01

    The Lithium Tokamak Experiment (LTX) is designed to study the low-recycling regime through the use of close-fitting, lithium-coated, heatable shell quadrants surrounding the plasma volume. Lithium coatings can getter and bury impurities, but they can also become covered by impurity compounds. Liquefied coatings can both dissolve impurity compounds and bring them to the surface, while sputtering and evaporation rates increase strongly with temperature. Here, we use spectroscopic measurements to assess the effects of varying wall conditions on plasma impurities, mainly Li, C, and O. A passive Doppler spectroscopy system measures toroidal and poloidal impurity profiles using fixed-wavelength and variable-wavelength visible spectrometers. In addition, survey and high-resolution extreme ultraviolet spectrometers detect emission from higher charge states. Preliminary results show that fresh Li coatings generally reduced C and O emission. C emission decreased sharply following the first solid Li coatings. Inverted toroidal profiles in a discharge with solid Li coatings show peaked Li III emissivity and temperature profiles. Recently, experiments with fresh liquid coatings led to especially strong O reduction. Results from these and additional experiments will be presented. Supported by US DOE Contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

  10. Recent progress in the tokamak edge modeling

    NASA Astrophysics Data System (ADS)

    Petravic, M.; Heifetz, D.; Heifetz, S.; Post, D.

    1984-12-01

    Tokamak edge modeling, with a particular emphasis on divertors, was reviewed in detail in 1982. At that time the emphasis was on the qualitative behavior of the scrape-off plasma and the atomic processes involved in the neutral-plasma interaction. While no detailed comparisons with the experiments were available, the data nevertheless showed all the basic features of the cool high-density regime predicted by the models. The two most important modeling developments of 1983 were the introduction of accurate magnetic geometries and the inclusion of impurity transport in the plasma equations. This made possible detailed comparisons with the PDX and ASDEX experiments which on the one hand showed remarkable agreement while on the other hand pointed to new areas of uncertainty, i.e., the plasma-wall and neutral-wall interactions. In another development, the scrape-off models are beginning to be linked to the main plasma transport in order to provide better boundary conditions for the main plasma models, and in particular to model limiters. The fully two-dimensional plasma flow models should be particularly useful in this area.

  11. Natural fueling of a tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Wan, Weigang; Parker, Scott E.; Chen, Yang; Perkins, Francis W.

    2010-04-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward toward the core. This mechanism is due to the quasineutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection is augmented by an inward pinch of cold DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM [Y. Chen and S. E. Parker, J. Comput. Phys. 220, 839 (2007)] and is analyzed using quasilinear theory. Profiles similar to those used for conservative International Thermonuclear Experimental Reactor [R. Aymar et al., Nucl. Fusion 41, 1301 (2001)] transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rate and energy transport. Natural fueling requires a two-component plasma and ion-ion and charge-exchange collisions set limits on this favorable effect.

  12. The dynamic mutation characteristics of thermonuclear reaction in Tokamak.

    PubMed

    Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

    2014-01-01

    The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z 2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given.

  13. Driven magnetic reconnection in the COMPASS-C tokamak

    SciTech Connect

    Morris, A.W.; Carolan, P.G.; Fitzpatrick, R.; Hender, T.C.; Todd, T.N. , Abingdon, Oxon )

    1992-02-01

    The question of the influence of nonaxisymmetric field perturbations on tokamaks is investigated. Recent experiments in the COMPASS-C tokamak (in {ital Proceedings} {ital of} {ital the} 15{ital th} {ital Symposium} {ital on} {ital Fusion} {ital Technology}, Utrecht (North-Holland, Amsterdam, 1989), Vol. 1, p. 361) with externally applied helical fields reveal that magnetic islands do not appear until the applied field exceeds a certain value, when plasma rotation and confinement are affected. A new resistive magnetohydrodynamic model including plasma rotation now provides an explanation of this threshold, and is quantitatively consistent with experimental results in Ohmic plasmas. The results indicate the tolerable error fields in future tokamaks. The effects of perturbations with various poloidal and toroidal mode numbers have been studied.

  14. Bootstrap currents in radio-frequency-driven tokamak equilibria

    SciTech Connect

    Hsiao, Ming-Yuan; Ehst, D.A.; Evans, K. Jr.

    1988-05-01

    Interest in the bootstrap current arising from neoclassical transport in tokamaks has increased recently in view of certain experimental observations. In this study, the bootstrap current is calculated for a number of rf current-driven tokamaks. Two-dimensional, self-consistent, steady-state tokamak MHD equilibria are obtained by including both the transport-driven bootstrap current and the externally driven rf current. The self-consistency is acomplished by iterating between two-dimensional MHD equilibrium calculations and the current calculations (including bootstrap and rf ray-tracing). Calculations for other reactor parameters of interest are also carried out. It is found that for reactor-grade plasmas, the bootstrap current contribution to the toroidal current is, in general, important. An approxiamte scaling law for GAMMA, based on parametric survey performed, is also obtained. 16 refs., 8 figs., 1 tab.

  15. A method for determining poloidal coil configurations for tokamak devices

    SciTech Connect

    Evans, K. Jr.

    1990-12-01

    This paper presents a method for obtaining the locations and currents of the poloidal coil systems for a tokamak, given an desirable magnetohydrodynamic equilibrium for the device. The method involves a simultaneous minimization of the match to the desired poloidal field and the stored energy in the coils, subject to the constraints necessary to achieve decoupling of the equilibrium and inductive-current-drive (ohmic-heating) systems and to achieve a given coupling of the current-drive system with the plasma. A compendium of mutual and self-inductance formulas as they apply to tokamak systems is presented, as well as examples of how the method has been used in the design of several tokamaks. Finally, a user manual for a computer code that implements this method is provided. 14 refs., 11 figs., 1 tab.

  16. Probing spherical tokamak plasmas using charged fusion products

    NASA Astrophysics Data System (ADS)

    Boeglin, Werner U.; Perez, Ramona V.; Darrow, Douglass S.; Cecconello, Marco; Klimek, Iwona; Allan, Scott Y.; Akers, Rob J.; Jones, Owen M.; Keeling, David L.; McClements, Ken G.; Scannell, Rory

    2015-11-01

    The detection of charged fusion products, such as protons and tritons resulting from D(d,p)t reactions, can be used to determine the fusion reaction rate profile in large spherical tokamak plasmas with neutral beam heating. The time resolution of a diagnostic of this type makes it possible to study the slowly-varying beam density profile, as well as rapid changes resulting from MHD instabilities. A 4-channel prototype proton detector (PD) was installed and operated on the MAST spherical tokamak in August/September 2013, and a new 6-channel system for the NSTX-U spherical tokamak is under construction. PD and neutron camera measurements obtained on MAST will be compared with TRANSP calculations, and the design of the new NSTX-U system will be presented, together with the first results from this diagnostic, if available. Supported in part by DOE DE-SC0001157.

  17. The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

    PubMed Central

    Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

    2014-01-01

    The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given. PMID:24892099

  18. Stochastic modeling of plasma mode forecasting in tokamak

    NASA Astrophysics Data System (ADS)

    Saadat, Sh.; Salem, M.; Ghoranneviss, M.; Khorshid, P.

    2012-04-01

    The structure of magnetohydrodynamic (MHD) modes has always been an interesting study in tokamaks. The mode number of tokamak plasma is the most important parameter, which plays a vital role in MHD instabilities. If it could be predicted, then the time of exerting external fields, such as feedback fields and Resonance Helical Field, could be obtained. Autoregressive Integrated Moving Average (ARIMA) and Seasonal Autoregressive Integrated Moving Average are useful models to predict stochastic processes. In this paper, we suggest using ARIMA model to forecast mode number. The ARIMA model shows correct mode number (m = 4) about 0.5 ms in IR-T1 tokamak and equations of Mirnov coil fluctuations are obtained. It is found that the recursive estimates of the ARIMA model parameters change as the plasma mode changes. A discriminator function has been proposed to determine plasma mode based on the recursive estimates of model parameters.

  19. Plasma Shape and Current Density Profile Control in Advanced Tokamak Operating Scenarios

    NASA Astrophysics Data System (ADS)

    Shi, Wenyu

    The need for new sources of energy is expected to become a critical problem within the next few decades. Nuclear fusion has sufficient energy density to potentially supply the world population with its increasing energy demands. The tokamak is a magnetic confinement device used to achieve controlled fusion reactions. Experimental fusion technology has now reached a level where tokamaks are able to produce about as much energy as is expended in heating the fusion fuel. The next step towards the realization of a nuclear fusion tokamak power plant is ITER, which will be capable of exploring advanced tokamak (AT) modes, characterized by a high fusion gain and plasma stability. The extreme requirements of the advanced modes motivates researchers to improve the modeling of the plasma response as well as the design of feedback controllers. This dissertation focuses on several magnetic and kinetic control problems, including the plasma current, position and shape control, and data-driven and first-principles-driven modeling and control of plasma current density profile and the normalized plasma pressure ratio betaN. The plasma is confined within the vacuum vessel by an external electromagnetic field, produced primarily by toroidal and poloidal field coils. The outermost closed plasma surface or plasma boundary is referred to as the shape of the plasma. A central characteristic of AT plasma regimes is an extreme elongated shape. The equilibrium among the electromagnetic forces acting on an elongated plasma is unstable. Moreover, the tokamak performance is improved if the plasma is located in close proximity to the torus wall, which guarantees an efficient use of available volume. As a consequence, feedback control of the plasma position and shape is necessary. In this dissertation, an Hinfinity-based, multi-input-multi-output (MIMO) controller for the National Spherical Torus Experiment (NSTX) is developed, which is used to control the plasma position, shape, and X

  1. Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Lucia, Matthew James

    The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance

  2. Two phase liquid helium flow testing to simulate the operation of a cryocondensation pump in the D3-D tokamak

    NASA Astrophysics Data System (ADS)

    Laughon, G. J.; Baxi, C. B.; Campbell, G. L.; Mahdavi, M. A.; Makariou, C. C.; Smith, J. P.; Schaffer, M. J.; Schaubel, K. M.; Menon, M. M.

    1994-06-01

    A liquid helium-cooled cryocondensation pump has been installed in the D3-D tokamak fusion energy research experiment at General Atomics. The pump is located within the tokamak vacuum chamber beneath the divertor baffle plates and is utilized for plasma density and contamination control. Two-phase helium flows through the pump at 5 to 10 g/s utilizing the heat transfer and constant temperature characteristics of boiling liquid . helium. The pump is designed for a pumping speed of 32,000 1/s. Extensive testing was performed with a prototypical pump test fixture. Several pump geometries (simple tube, coaxial flow plug, and coaxial slotted insert) were tested, in an iterative process, to determine which was the most satisfactory for stable cryocondensation pumping. Results from the different tests illustrating the temperature distribution and flow characteristics for each configuration are presented.

  3. Two phase liquid helium flow testing to simulate the operation of a cryocondensation pump in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Laughon, G. J.; Baxi, C. B.; Campbell, G. L.; Mahdavi, M. A.; Makariou, C. C.; Menon, M. M.; Smith, J. P.; Schaffer, M. J.; Schaubel, K. M.

    A liquid helium-cooled cryocondensation pump has been installed in the DIII=D tokamak fusion energy research experiment at General Atomics. The pump is located within the tokamak vacuum chamber beneath the divertor baffle plates and is utilized for plasma density and contamination control. Two-phase helium flows through the pump at 5 to 10 g/s utilizing the beat transfer and constant temperature characteristics of boiling liquid helium. The pump is designed for a pumping speed of 32,0001/s. Extensive testing was performed with a prototypical pump test fixture. Several pump geometries (simple tube, coaxial flow plug, and coaxial slotted insert) were tested, in an iterative process, to determine which was the most satisfactory for stable cryocondensation pumping. Results from the different tests illustrating the temperature distribution and flow characteristics for each configuration are presented.

  4. Fractal structure of films deposited in a tokamak

    NASA Astrophysics Data System (ADS)

    Budaev, V. P.; Khimchenko, L. N.

    2007-04-01

    The surface of amorphous films deposited in the T-10 tokamak was studied in a scanning tunnel microscope. The surface relief on a scale from 10 nm to 100 μm showed a stochastic surface topography and revealed a hierarchy of grains. The observed variety of irregular structures of the films was studied within the framework of the concept of scale invariance using the methods of fractal geometry and statistical physics. The experimental probability density distribution functions of the surface height variations are close in shape to the Cauchy distribution. The stochastic surface topography of the films is characterized by a Hurst parameter of H = 0.68-0.85, which is evidence of a nontrivial self-similarity of the film structure. The fractal character and porous structure of deposited irregular films must be considered as an important issue related to the accumulation of tritium in the ITER project. The process of film growth on the surface of tokamak components exposed to plasma has been treated within the framework of the general concept of inhomogeneous surface growth. A strong turbulence of the edge plasma in tokamaks can give rise to fluctuations in the incident flux of particles, which leads to the growth of fractal films with grain dimensions ranging from nano-to micrometer scale. The shape of the surface of some films found in the T-10 tokamak has been interpreted using a model of diffusion-limited aggregation (DLA). The growth of films according to the discrete DLA model was simulated using statistics of fluctuations observed in a turbulent edge plasma of the T-10 tokamak. The modified DLA model reproduces well the main features of the surface of some films deposited in tokamaks.

  5. A Midsize Tokamak As Fast Track To Burning Plasmas

    SciTech Connect

    E. Mazzucato

    2010-07-14

    This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.

  6. Accessibility of second regions of stability in tokamaks

    SciTech Connect

    Manickam, J.

    1985-12-01

    Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs.

  7. Resistive demountable toroidal-field coils for tokamak reactors

    SciTech Connect

    Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

    1981-07-01

    Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments.

  8. Adaptive grid finite element model of the tokamak scrapeoff layer

    SciTech Connect

    Kuprat, A.P.; Glasser, A.H.

    1995-07-01

    The authors discuss unstructured grids for application to transport in the tokamak edge SOL. They have developed a new metric with which to judge element elongation and resolution requirements. Using this method, the authors apply a standard moving finite element technique to advance the SOL equations while inserting/deleting dynamically nodes that violate an elongation criterion. In a tokamak plasma, this method achieves a more uniform accuracy, and results in highly stretched triangular finite elements, except near separatrix X-point where transport is more isotropic.

  9. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    SciTech Connect

    Hively, L.M.

    1980-01-01

    High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-..beta.., non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated.

  10. Shear flow destabilization of a slowly rotating tokamak

    NASA Astrophysics Data System (ADS)

    Chu, M. S.

    1998-01-01

    The Kelvin-Helmholtz destabilizing effect of shear in toroidal rotation on ideal magnetohydrodynamic localized interchange is studied in a tokamak with a general geometry. The method of maximizing the growth rate given by Frieman and Rotenberg is utilized. An explicit stability criterion is given for a slowly rotating tokamak with a non-negligible shearing rate in its rotation profile. It is found that rotation shear can weaken the stabilizing effect of the magnetic field shear and also allow the coupling of the sound wave to the shear Alfvén wave which destabilizes the plasma.

  11. What is the fate of runaway positrons in tokamaks?

    SciTech Connect

    Liu, Jian; Qin, Hong; Fisch, Nathaniel J.; Teng, Qian; Wang, Xiaogang

    2014-06-19

    In this study, massive runaway positrons are generated by runaway electrons in tokamaks. The fate of these positrons encodes valuable information about the runaway dynamics. The phase space dynamics of a runaway position is investigated using a Lagrangian that incorporates the tokamak geometry, loop voltage, radiation and collisional effects. It is found numerically that runaway positrons will drift out of the plasma to annihilate on the first wall, with an in-plasma annihilation possibility less than 0.1%. The dynamics of runaway positrons provides signatures that can be observed as diagnostic tools.

  12. Current Status and Future Technical Challenges for Tokamak Magnets

    SciTech Connect

    Martovetsky, N; Minervini, J; Okuno, K; Salpiero, E; Filatov, O

    2002-11-11

    Magnet technology for fusion in the last decade has been focusing mostly on the development of magnets for tokamaks--the most advanced fusion concept at the moment. The largest and the most complex tokamak under development is ITER. To demonstrate adequate design approaches to large magnets for ITER and to develop industrial capabilities, two large model coils and three insert coils, all using full-scale conductor, were built and tested by the international collaboration during 1994-2002. The status of the magnet technology and directions of future developments are discussed in this paper.

  13. Design and Analysis of the Thermal Shield of EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Xie, Han; Liao, Ziying

    2008-04-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

  14. Linear and nonlinear kinetic-stability studies in tokamaks

    SciTech Connect

    Tang, W.M.; Chance, M.S.; Chen, L.; Krommes, J.A.; Lee, W.W.; Rewoldt, G.

    1982-09-01

    This paper presents results of theoretical investigations on important linear kinetic properties of low frequency instabilities in toroidal systems and on nonlinear processes which could significantly influence their impact on anomalous transport. Analytical and numerical methods and also particle simulations have been employed to carry out these studies. In particular, the following subjects are considered: (1) linear stability analysis of kinetic instabilities for realistic tokamak equilibria and the application of such calculations to the PDX and PLT tokamak experiments including the influence of a hot beam-ion component; (2) determination of nonlinearly saturated, statistically steady states of three interacting drift modes; and (3) gyrokinetic particle simulation of drift instabilities.

  15. Steady state self-induced current in tokamak

    SciTech Connect

    Gott, Yu. V.; Yurchenko, E. I.

    2009-11-15

    A model, which may make it possible to self-consistently calculate the self-driven current in tokamaks taking into account asymmetry and bootstrap currents, is presented. It is shown that the described self-driven current can provide steady-state tokamak operation without the seed current produced with the help of additional methods. The total self-consistent, self-driven current does not depend on magnetic field magnitude and is proportional to the square root from plasma pressure. The experimental data obtained in the National Spherical Torus Experiment are satisfactorily described by this model.

  16. Simulation of EAST vertical displacement events by tokamak simulation code

    NASA Astrophysics Data System (ADS)

    Qiu, Qinglai; Xiao, Bingjia; Guo, Yong; Liu, Lei; Xing, Zhe; Humphreys, D. A.

    2016-10-01

    Vertical instability is a potentially serious hazard for elongated plasma. In this paper, the tokamak simulation code (TSC) is used to simulate vertical displacement events (VDE) on the experimental advanced superconducting tokamak (EAST). Key parameters from simulations, including plasma current, plasma shape and position, flux contours and magnetic measurements match experimental data well. The growth rates simulated by TSC are in good agreement with TokSys results. In addition to modeling the free drift, an EAST fast vertical control model enables TSC to simulate the course of VDE recovery. The trajectories of the plasma current center and control currents on internal coils (IC) fit experimental data well.

  17. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    SciTech Connect

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  18. Magnetic flux reconstruction methods for shaped tokamaks

    SciTech Connect

    Tsui, Chi-Wa

    1993-12-01

    The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p` and FF` functions). The author treats the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green`s function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perception neural network as an interface, and the volume integration of plasma current density using Green`s functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising.

  19. Transport-driven toroidal rotation in the tokamak edge

    NASA Astrophysics Data System (ADS)

    Stoltzfus-Dueck, T.

    2011-10-01

    The edge of H-mode tokamak plasmas without external momentum input almost always rotates toroidally in the co-current direction, which has prompted a theoretical search for non-diffusive momentum transport mechanisms. In contrast to these efforts, the present work treats a model drift-kinetic ion equation for the pedestal and SOL containing only parallel free streaming, magnetic drifts, and spatially inhomogeneous but purely diffusive transport. The solution demonstrates that passing-ion orbits and spatially inhomogeneous diffusion interact to cause a variation of the orbit-averaged diffusivities that depends on the sign of v∥, typically resulting in preferential transport of counter-current ions. If the plasma at the boundary with the core is allowed to rotate toroidally to annihilate toroidal momentum transport, the resulting pedestal-top rotation reaches experimentally relevant values and exhibits several features in qualitative agreement with experiment. It is almost always in the co-current direction, with a rate that is proportional to Ti|ped-top /BpolLTe for small qρi /LTe , thus inversely proportional to Ip in accord with Rice scaling. It is independent of the toroidal velocity and its radial gradient, representing a residual stress. The Ti|ped-top /BpolLTe scaling implies co-current spin-up at the transition to H-mode, as Ti increases and the gradient of Te steepens. Untested predictions of the model include a sensitivity of the rotation to the major-radial position of the X-point, with a more inboard X-point leading to stronger co-current rotation. Beyond intrinsic rotation predictions, comparison of heat and momentum transport reveals that neutral beam injection must be significantly unbalanced in the counter-current direction to cause zero toroidal rotation at the pedestal top. Work supported by a research fellowship from the Alexander von Humboldt Foundation.

  20. HPGe well-type detectors for neutron activation measurements on the Frascati Tokamak Upgrade tokamak

    SciTech Connect

    Bertalot, L.; Damiani, M.; Esposito, B.; Lagamba, L.; Podda, S.; Batistoni, P.; De Felice, P.; Biagini, R.

    1997-01-01

    We describe an improvement of the neutron activation system in operation on the Frascati Tokamak Upgrade (FTU) tokamak for the measurement of the total neutron yield. A HPGe well-type detector (200 cm{sup 3} active volume) is used to detect the photoemission from neutron activated samples ({sup 115m}In336.2 keV {gamma} rays from DD neutrons on indium for FTU). Due to their high geometrical efficiency, HPGe well-type detectors are particularly suited to the FTU low-level activity measurements. A particular effort has been devoted to the calibration of the measuring system. In particular, a multi-{gamma} calibration source (59{endash}1332 keV energy range) with a density of 7.31 g/cm{sup 3} consisting of a stack of indium foils has been prepared. This assures that the shape and volume of the calibration source are the same as those of the samples used in the actual measurements. The full-energy-peak efficiency at the {sup 115m}In336.2 keV line is 0.197 with an overall uncertainty of 2{percent} (1{sigma}). For a better characterization of the detector response as a function of the sample density, a further calibration source with the same geometry has been prepared in a gel aqueous solution (density {approximately}1 g/cm{sup 3}). The calibration curves for the well-type detector at the two different density values are compared. {copyright} {ital 1997 American Institute of Physics.}

  1. Modelling multi-ion plasma gun simulations of Tokamak disruptions

    SciTech Connect

    Ehst, D.A.

    1995-08-01

    The effect of impurity ions in plasma gun ablation tests of various targets is considered. Inclusion of reasonable amounts of impurity ({approximately}10%) is adequate to explain observed energy transmission and erosion measurements. The gun tests and the computer code calculations are relevant to the parameter range expected for major disruptions on large tokamaks.

  2. Stochasticity and the m = 1 mode in tokamaks. [Sawtooth oscillations

    SciTech Connect

    Izzo, R.; Monticello, D.A.; Stodiek, W.; Park, W.

    1986-05-01

    It has recently been proposed that stochasticity resulting from toroidal coupling could lead to a saturation of the m = 1 internal mode in tokamaks. We present results from the nonlinear evolution of the m = 1 mode with toroidal coupling that show that stochasticity is not enough to cause saturation of the m = 1 mode.

  3. Loop-voltage tomography in tokamaks using transient synchrotron radiation

    SciTech Connect

    Fisch, N.J.; Kritz, A.H. . Plasma Physics Lab.; Hunter Coll., New York, NY . Dept. of Physics)

    1989-07-01

    The loop voltage in tokamaks is particularly difficult to measure anywhere but at the plasma periphery. A brief, deliberate, perturbation of hot plasma electrons, however, produces a transient radiation response that is sensitive to this voltage. We investigate how such a radiation response can be used to diagnose the loop voltage. 24 refs., 6 figs.

  4. Tokamak Scenario Trajectory Optimization Using Fast Integrated Simulations

    NASA Astrophysics Data System (ADS)

    Urban, Jakub; Artaud, Jean-François; Vahala, Linda; Vahala, George

    2015-11-01

    We employ a fast integrated tokamak simulator, METIS, for optimizing tokamak discharge trajectories. METIS is based on scaling laws and simplified transport equations, validated on existing experiments and capable of simulating a full tokamak discharge in about 1 minute. Rapid free-boundary equilibrium post-processing using FREEBIE provides estimates of PF coil currents or forces. We employ several optimization strategies for optimizing key trajectories, such as Ip or heating power, of a model ITER hybrid discharge. Local and global algorithms with single or multiple objective functions show how to reach optimum performance, stationarity or minimum flux consumption. We constrain fundamental operation parameters, such as ramp-up rate, PF coils currents and forces or heating power. As an example, we demonstrate the benefit of current over-shoot for hybrid mode, consistent with previous results. This particular optimization took less than 2 hours on a single PC. Overall, we have established a powerful approach for rapid, non-linear tokamak scenario optimization, including operational constraints, pertinent to existing and future devices design and operation.

  5. The impact of improved physics on commercial tokamak reactors

    SciTech Connect

    Galambos, J.D.; Perkins, L.J.; Haney, S.; Mandrekas, J.

    1994-01-01

    Improvements in the confinement and beta capability of tokamak devices have long been a goal of the fusion program. We examine the impact of improvements in present day confinement and beta capabilities on commercial tokamak reactors. We characterize confinement with the achievable enhancement factor (H) over the ITER89 Power scaling confinement time, and beta by the Troyon coefficient g. A surprisingly narrow range of plasma confinement and beta are found to be useful in minimizing the cost of electricity for a tokamak reactor. Improvements in only one of these quantities is not useful beyond some point, without accompanying improvements in the other. For the plasma beta limited by a Troyon coefficient (g) near 4.3 (%mT/MA), confinement levels characterized by H factor enhancements of only 2 are useful for our nominal steady-state driven tokamak. These confinement levels are similar to those observed in present day experiments. If the permissible Troyon beta coefficient is near 6, the useful H factor confinement range increases to 2.5, still close to present day confinement levels. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement, relative to the steady-state cases. For a Troyon beta limit coefficient g near 4.3, H factors up to 2.5 are useful, and for g near 6, H factors up to 3 are useful.

  6. Confinement of high energy trapped particles in tokamaks

    SciTech Connect

    Goldston, R.J.; White, R.B.; Boozer, A.H.

    1981-04-01

    The banana orbits of high energy trapped particles in tokamaks are found to diffuse rapidly in the radial direction if the toroidal ripple exceeds a low critical value. During this diffusion the energy, the magnetic moment, and the value of the magnetic field strength at the banana tips are conserved.

  7. Confinement of high-energy trapped particles in tokamaks

    SciTech Connect

    Goldston, R.J.; White, R.B.; Boozer, A.H.

    1981-08-31

    The banana orbits of high-energy trapped particles in tokamaks are found to diffuse rapidly in the radial direction if the toroidal ripple exceeds a low critical value. During this diffusion the energy, the magnetic moment, and the value of the magnetic field strength at the banana tips are conserved.

  8. 2-D Imaging of Electron Temperature in Tokamak Plasmas

    SciTech Connect

    T. Munsat; E. Mazzucato; H. Park; C.W. Domier; M. Johnson; N.C. Luhmann Jr.; J. Wang; Z. Xia; I.G.J. Classen; A.J.H. Donne; M.J. van de Pol

    2004-07-08

    By taking advantage of recent developments in millimeter wave imaging technology, an Electron Cyclotron Emission Imaging (ECEI) instrument, capable of simultaneously measuring 128 channels of localized electron temperature over a 2-D map in the poloidal plane, has been developed for the TEXTOR tokamak. Data from the new instrument, detailing the MHD activity associated with a sawtooth crash, is presented.

  9. Dynamic diagnostics of the error fields in tokamaks

    NASA Astrophysics Data System (ADS)

    Pustovitov, V. D.

    2007-07-01

    The error field diagnostics based on magnetic measurements outside the plasma is discussed. The analysed methods rely on measuring the plasma dynamic response to the finite-amplitude external magnetic perturbations, which are the error fields and the pre-programmed probing pulses. Such pulses can be created by the coils designed for static error field correction and for stabilization of the resistive wall modes, the technique developed and applied in several tokamaks, including DIII-D and JET. Here analysis is based on the theory predictions for the resonant field amplification (RFA). To achieve the desired level of the error field correction in tokamaks, the diagnostics must be sensitive to signals of several Gauss. Therefore, part of the measurements should be performed near the plasma stability boundary, where the RFA effect is stronger. While the proximity to the marginal stability is important, the absolute values of plasma parameters are not. This means that the necessary measurements can be done in the diagnostic discharges with parameters below the nominal operating regimes, with the stability boundary intentionally lowered. The estimates for ITER are presented. The discussed diagnostics can be tested in dedicated experiments in existing tokamaks. The diagnostics can be considered as an extension of the 'active MHD spectroscopy' used recently in the DIII-D tokamak and the EXTRAP T2R reversed field pinch.

  10. Gamma ray imager on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Pace, D. C.; Cooper, C. M.; Taussig, D.; Eidietis, N. W.; Hollmann, E. M.; Riso, V.; Van Zeeland, M. A.; Watkins, M.

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

  11. Gamma ray imager on the DIII-D tokamak.

    PubMed

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons. PMID:27131674

  12. Conceptual Design of Alborz Tokamak Poloidal Coils System

    NASA Astrophysics Data System (ADS)

    Mardani, M.; Amrollahi, R.

    2013-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

  13. Energetic-particle stabilization of ballooning modes in Tokamaks

    NASA Astrophysics Data System (ADS)

    Rosenbluth, M. N.; Tsai, S. T.; van Dam, J. W.; Engguist, M. G.

    1983-07-01

    Introduction of an anisotropic, highly energetic trapped-particle species into a Tokamak may allow direct stable access to the high-beta regime of second stability. Under certain conditions, the mode at marginal stability acquires a real frequency close to the precessional drift frequency of the energetic particles, perhaps correlating with recent fishbone observations on PDX.

  14. Energetic Particle Stabilization of Ballooning Modes in Tokamaks

    NASA Astrophysics Data System (ADS)

    Rosenbluth, M. N.; Tsai, S. T.; van Dam, J. W.; Engquist, M. G.

    1983-11-01

    Introduction of an anisctropic, highly energetic trapped-particle species into a tokamak may allow direct stable access to the high-beta regime of second stability. Under certain conditions, the mode at marginal stability acquires a real frequency close to the precessional drift frequency of the energetic particles, perhaps correlating with recent "fishbone" observations on PDX.

  15. Energetic particle stabilization of ballooning modes in tokamaks

    SciTech Connect

    Rosenbluth, M.N.; Tsai, S.T.; Van Dam, J.W.; Engquist, M.G.

    1983-11-21

    Introduction of an anisotropic, highly energetic trapped-particle species into a tokamak may allow direct stable access to the high-beta regime of second stability. Under certain conditions, the mode at marginal stability acquires a real frequency close to the precessional drift frequency of the energetic particles, perhaps correlating with recent ''fishbone'' observations on PDX.

  16. Observation of finite-. beta. MHD phenomena in tokamaks

    SciTech Connect

    McGuire, K.M.

    1984-09-01

    Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.

  17. Gamma ray imager on the DIII-D tokamak.

    PubMed

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

  18. Bootstrap current close to magnetic axis in tokamaks

    SciTech Connect

    Shaing, K.C.; Hazeltine, R.D.

    1996-12-01

    It is shown that the bootstrap current density close to the magnetic axis in tokamaks does not vanish in simple electron-ion plasmas because the fraction of the trapped particles is finite. The magnitude of the current density could be comparable to that in the outer core region. This may reduce or even eliminate the need of the seed current.

  19. Disruption avoidance through active magnetic feedback in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Paccagnella, Roberto; Zanca, Paolo; Yanovskiy, Vadim; Finotti, Claudio; Manduchi, Gabriele; Piron, Chiara; Carraro, Lorella; Franz, Paolo; RFX Team

    2014-10-01

    Disruptions avoidance and mitigation is a fundamental need for a fusion relevant tokamak. In this paper a new experimental approach for disruption avoidance using active magnetic feedback is presented. This scheme has been implemented and tested on the RFX-mod device operating as a circular tokamak. RFX-mod has a very complete system designed for active mode control that has been proved successful for the stabilization of the Resistive Wall Modes (RWMs). In particular the current driven 2/1 mode, unstable when the edge safety factor, qa, is around (or even less than) 2, has been shown to be fully and robustly stabilized. However, at values of qa (qa > 3), the control of the tearing 2/1 mode has been proved difficult. These results suggested the idea to prevent disruptions by suddenly lowering qa to values around 2 where the tearing 2/1 is converted to a RWM. Contrary to the universally accepted idea that the tokamaks should disrupt at low qa, we demonstrate that in presence of a well designed active control system, tokamak plasmas can be driven to low qa actively stabilized states avoiding plasma disruption with practically no loss of the plasma internal energy.

  20. Electron-cyclotron-heating experiments in tokamaks and stellarators

    SciTech Connect

    England, A.C.

    1983-01-01

    This paper reviews the application of high-frequency microwave radiation to plasma heating near the electron-cyclotron frequency in tokamaks and stellarators. Successful plasma heating by microwave power has been demonstrated in numerous experiments. Predicted future technological developments and current theoretical understanding suggest that a vigorous program in plasma heating will continue to yield promising results.

  1. Recent Progress on Spherical Torus Research

    SciTech Connect

    Ono, Masayuki; Kaita, Robert

    2014-01-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  2. UCLA program in reactor studies: The ARIES tokamak reactor study. Progress report, December 1, 1990--November 30, 1991

    SciTech Connect

    Not Available

    1991-12-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ``modest`` extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  3. Scrape-off layer plasma modeling for the DIII-D tokamak

    SciTech Connect

    Porter, G.D.; Rognlien, T.D.; Allen, S.L.

    1994-09-01

    The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing.

  4. Nonlinear ω*-stabilization of the m=1 mode in tokamaks

    NASA Astrophysics Data System (ADS)

    Rogers, B.; Zakharov, L.

    1995-09-01

    Earlier studies of sawtooth oscillations in Tokamak Fusion Test Reactor (TFTR) supershots [Levinton et al., Phys. Rev. Lett. 72, 2895 (1994); Zakharov et al., ``Onset and stabilization of sawtooth oscillations in tokamaks,'' in Plasma Physics and Controlled Nuclear Fusion Research, Proceedings of the 15th International Conference, Seville, 1994 (International Atomic Energy Agency, Vienna, in press)] have found an apparent contradiction between conventional linear theory and experiment: even in sawtooth-free discharges, the theory typically predicts instability due to a nearly ideal m=1 mode. Here, the nonlinear evolution of such mode is analyzed using numerical simulations of a two-fluid magnetohydrodynamic (MHD) model. We find the mode saturates nonlinearly at a small amplitude provided the ion and electron drift-frequencies ω*i,e are somewhat above the linear stability threshold of the collisionless m=1 reconnecting mode. The comparison of the simulation results to m=1 mode activity in TFTR suggests additional, stabilizing effects outside the present model are also important.

  5. Physics of radiation-driven islands near the tokamak density limit

    NASA Astrophysics Data System (ADS)

    Gates, D. A.; Delgado-Aparicio, L.; White, R. B.

    2013-06-01

    In previous work (Gates and Delgado-Aparicio 2012 Phys. Rev. Lett. 108 165004), the onset criterion for radiation-driven islands (Rebut et al 1985 Proc. 10th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research 1984 (London, UK, 1984) vol 2 (Vienna: IAEA) p 197) in combination with a simple cylindrical model of tokamak current channel behaviour was shown to be consistent with the empirical scaling of the tokamak density limit (Greenwald et al 1988 Nucl. Fusion 28 2199). A number of the unexplained phenomena at the density limit are consistent with this novel physics mechanism. In this work, a more formal theoretical underpinning, consistent with cylindrical tearing mode theory, is developed for the onset criteria of these modes. The appropriate derivation of the radiation-driven addition to the modified Rutherford equation (MRE) is discussed. Additionally, the ordering of the terms in the MRE is examined in a regime near the density limit. It is hoped that, given the apparent success of this simple model in explaining the observed global scalings, it will lead to a more comprehensive analysis of the possibility that radiation-driven islands are the physics mechanism responsible for the density limit. In particular, with modern diagnostic capabilities detailed measurements of current densities, electron densities and impurity concentrations at rational surfaces should be possible, enabling verification of the concepts described above.

  6. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Xie, H.; Chen, Z. X.; Zakharov, L. E.

    2015-02-15

    A program involving the extensive and systematic use of lithium (Li) as a “first,” or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  7. Geodesic acoustic modes in toroidally rotating tokamaks with an arbitrary β

    SciTech Connect

    Ren, Haijun; Li, Ding; Chu, Paul K

    2013-07-15

    Theoretical research on the geodesic acoustic mode (GAM) induced by the equilibrium toroidal rotation flow (ETRF) in the tokamak plasmas with an arbitrary β is performed by using the ideal magnetohydrodynamic model, where β is the ratio of the plasma pressure and magnetic field pressure. Two equations determining the poloidal displacement ξ{sub θ} and the divergence of the Lagrangian perturbation are obtained and suitable for arbitrary cross-section tokamaks with large-aspect-ratios. The dispersion relations are then derived for two different coupling patterns by assuming ξ{sub ±2}=0 and ξ{sub ±4}=0, respectively, where ξ{sub m}=(1/2π)∫ξ{sub θ}e{sup imθ}dθ with θ being the poloidal angle under the circular cross-section condition. In both patterns, the ETRF will increase the frequencies of the GAMs but β can decrease them. The GAM for ξ{sub ±2}=0 has a larger frequency than GAM for ξ{sub ±4}=0.

  8. Performance of a New Ion Source for KSTAR Tokamak Plasma Heating

    NASA Astrophysics Data System (ADS)

    Tae-Seong, Kim; Seung, Ho Jeong; Doo, Hee Chang; Kwang, Won Lee; Sang-Ryul, In

    2014-06-01

    In the experimental campaign of 2010 and 2011 on KSTAR, the NBI-1 system was equipped with one prototype ion source and operated successfully, providing a neutral beam power of 0.7-1.6 MW to the tokamak plasma. The new ion source planned for the 2012 KSTAR campaign had a much more advanced performance compared with the previous one. The target performance of the new ion source was to provide a neutral deuterium beam of 2 MW to the tokamak plasma. The ion source was newly designed, fabricated, and assembled in 2011. The new ion source was then conditioned up to 64 A/100 keV over a 2-hour beam extraction and performance tested at the NB test stand (NBTS) at the Korea Atomic Energy Research Institute (KAERI) in 2012. The measured optimum perveance at which the beam divergence is a minimum was about 2.5 μP, and the minimum beam divergent angle was under 1.0° at 60 keV. These results indicate that the 2.0 MW neutral beam power at 100 keV required for the heating of plasma in KSTAR can be delivered by the installation of the new ion source in the KSTAR NBI-1 system.

  9. Analytic model for coaxial helicity injection in tokamak plasmas

    SciTech Connect

    Weening, R. H.

    2011-12-15

    Using a partial differential equation for the time evolution of the mean-field poloidal magnetic flux that incorporates resistivity {eta} and hyper-resistivity {Lambda} terms, an exact analytic solution is obtained for steady-state coaxial helicity injection (CHI) in force-free large aspect ratio tokamaks. The analytic mean-field Ohm's law model allows for calculation of the tokamak CHI current drive efficiency and the plasma inductances at arbitrary levels of magnetic fluctuations, or dynamo activity. The results of the mean-field model suggest that CHI approaching Ohmic efficiency is only possible in tokamaks when the size of the effective current drive boundary layer, {delta}{identical_to}({Lambda}/{eta}){sup 1/2}, becomes greater than half the size of the plasma, {delta}>a/2, with a the plasma minor radius. The electron thermal diffusivity due to magnetic fluctuation induced transport is obtained from the expression {chi}{sub e}={Lambda}/{mu}{sub 0}d{sub e}{sup 2}, with {mu}{sub 0} the permeability of free space and d{sub e} the electron skin depth, which for typical tokamak fusion plasma parameters is on the order of a millimeter. Thus, the ratio of the energy confinement time to the resistive diffusion time in a tokamak plasma driven by steady-state CHI approaching Ohmic efficiency is shown to be constrained by the relation {tau}{sub E}/{tau}{sub {eta}}<(d{sub e}/a){sup 2}{approx_equal}10{sup -6}. The mean-field model suggests that steady-state CHI can be viewed most simply as a boundary layer of stochastically wandering magnetic field lines.

  10. LIDAR Thomson scattering for advanced tokamaks. Final report

    SciTech Connect

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-03-18

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured.

  11. High power heating of magnetic reconnection in merging tokamak experimentsa)

    NASA Astrophysics Data System (ADS)

    Ono, Y.; Tanabe, H.; Yamada, T.; Gi, K.; Watanabe, T.; , T., Ii; Gryaznevich, M.; Scannell, R.; Conway, N.; Crowley, B.; Michael, C.

    2015-05-01

    Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R˜105. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magnetic reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field Brec2 ˜ Bp2. The guide toroidal field Bt does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field Bt, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with Bp > 0.4 T will enables us to heat the plasma to the alpha heating regime: Ti > 5 keV without using any additional heating facility.

  12. On steady poloidal and toroidal flows in tokamak plasmas

    SciTech Connect

    McClements, K. G.

    2010-08-15

    The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B{sub {theta}/}B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B{sub {theta}/}B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

  13. Status of and prospects for advanced tokamak regimes from multi-machine comparisons using the 'International Tokamak Physics Activity' database

    NASA Astrophysics Data System (ADS)

    Litaudon, X.; Barbato, E.; Bécoulet, A.; Doyle, E. J.; Fujita, T.; Gohil, P.; Imbeaux, F.; Sauter, O.; Sips, G.; ITPA Group on Transport; Internal ITB Physics; Connor, J. W.; Doyle, E. J.; Esipchuk, Yu; Fujita, T.; Fukuda, T.; Gohil, P.; Kinsey, J.; Kirneva, N.; Lebedev, S.; Litaudon, X.; Mukhovatov, V.; Rice, J.; Synakowski, E.; Toi, K.; Unterberg, B.; Vershkov, V.; Wakatani, M.; International ITB Database Working Group; Aniel, T.; Baranov, Yu F.; Barbato, E.; Bécoulet, A.; Behn, R.; Bourdelle, C.; Bracco, G.; Budny, R. V.; Buratti, P.; Doyle, E. J.; Esipchuk, Yu; Esposito, B.; Ide, S.; Field, A. R.; Fujita, T.; Fukuda, T.; Gohil, P.; Gormezano, C.; Greenfield, C.; Greenwald, M.; Hahm, T. S.; Hoang, G. T.; Hobirk, J.; Hogeweij, D.; Ide, S.; Isayama, A.; Imbeaux, F.; Joffrin, E.; Kamada, Y.; Kinsey, J.; Kirneva, N.; Litaudon, X.; Luce, T. C.; Murakami, M.; Parail, V.; Peng, Y.-K. M.; Ryter, F.; Sakamoto, Y.; Shirai, H.; Sips, G.; Suzuki, T.; Synakowski, E.; Takenaga, H.; Takizuka, T.; Tala, T.; Wade, M. R.; Weiland, J.

    2004-05-01

    Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H\\times \\beta _{\\rm N}/q^{2}_{95} , is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters. A preliminary version of this study was presented in the 29th EPS Conf. on Plasma Phys. and Control. Fusion (Montreux, Switzerland, 17 21 June 2002) [1].

  14. Fueling studies on the lithium tokamak experiment

    NASA Astrophysics Data System (ADS)

    Lundberg, Daniel Patrick

    Lithium plasma facing components reduce the flux of "recycled" particles entering the plasma edge from the plasma facing components. This results in increased external fueling requirements and provides the opportunity to control the magnitude and distribution of the incoming particle flux. It has been predicted that the plasma density profile will then be determined by the deposition profile of the external fueling, rather than dominated by the recycled particle flux. A series of experiments on the Lithium Tokamak Experiment demonstrate that lithium wall coatings facilitate control of the neutral and plasma particle inventories. With fresh lithium coatings and careful gas injection programming, over 90% of the injected particle inventory can be absorbed in the lithium wall during a discharge. Furthermore, dramatic changes in the fueling requirements and plasma parameters were observed when lithium coatings were applied. This is largely due to the elimination of water as an impurity on the plasma facing components. A Molecular Cluster Injector (MCI) was developed for the fueling of LTX plasmas. The MCI uses a supersonic nozzle, cooled to liquid nitrogen temperatures, to create the conditions necessary for molecular cluster formation. It has been predicted that molecular clusters will penetrate deeper into plasmas than gas-phase molecules via a reduced ionization cross-section and by improving the collimation of the neutral jet. Using an electron beam diagnostic, the densities of the cryogenic MCI are measured to be an order of magnitude higher than in the room-temperature jets formed with the same valve pressure. This indicates increased collimation relative to what would be expected from ideal gas dynamics alone. A systematic study of the fueling efficiencies achieved with the LTX fueling systems is presented. The fueling efficiency of the Supersonic Gas Injector (SGI) is demonstrated to be strongly dependent on the distance between the nozzle and plasma edge. The

  15. Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER

    NASA Astrophysics Data System (ADS)

    Razumova, K. A.; Andreev, V. F.; Kislov, A. Ya.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Yu. D.; Shafranov, T. V.; T-10 Team; Donné, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; Jaspers, R.; TEXTOR Team; Kantor, M.; Walsh, M.

    2009-06-01

    The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile has a universal shape for a wide class of tokamaks and regimes, if the normalized radius ρ = r/(IpR/κB)1/2 is used. The consequences of this phenomenon for low aspect ratio tokamaks, the optimal deposition of additional heating, fast velocity of heat/cold pulse propagation and the possibility of obtaining a peaked density profile in ITER are discussed.

  16. Development of magnetohydrodynamic modes during sawteeth in tokamak plasmas

    SciTech Connect

    Firpo, M.-C.; Ettoumi, W.; Farengo, R.; Ferrari, H. E.; García-Martínez, P. L.; Lifschitz, A. F.

    2013-07-15

    A dynamical analysis applied to a reduced resistive magnetohydrodynamics model is shown to explain the chronology of the nonlinear destabilization of modes observed in tokamak sawteeth. A special emphasis is put on the nonlinear self-consistent perturbation of the axisymmetric m = n = 0 mode that manifests through the q-profile evolution. For the very low fusion-relevant resistivity values, the q-profile is shown to remain almost unchanged on the early nonlinear timescale within the central tokamak region, which supports a partial reconnection scenario. Within the resistive region, indications for a local flattening or even a local reversed-shear of the q-profile are given. The impact of this ingredient in the occurrence of the sawtooth crash is discussed.

  17. Collisionless microtearing modes in hot tokamaks: Effect of trapped electrons

    SciTech Connect

    Swamy, Aditya K.; Ganesh, R.; Brunner, S.; Vaclavik, J.; Villard, L.

    2015-07-15

    Collisionless microtearing modes have recently been found linearly unstable in sharp temperature gradient regions of large aspect ratio tokamaks. The magnetic drift resonance of passing electrons has been found to be sufficient to destabilise these modes above a threshold plasma β. A global gyrokinetic study, including both passing electrons as well as trapped electrons, shows that the non-adiabatic contribution of the trapped electrons provides a resonant destabilization, especially at large toroidal mode numbers, for a given aspect ratio. The global 2D mode structures show important changes to the destabilising electrostatic potential. The β threshold for the onset of the instability is found to be generally downshifted by the inclusion of trapped electrons. A scan in the aspect ratio of the tokamak configuration, from medium to large but finite values, clearly indicates a significant destabilizing contribution from trapped electrons at small aspect ratio, with a diminishing role at larger aspect ratios.

  18. Tooling concepts for ITER tokamak assembly and remote disassembly

    SciTech Connect

    Oikawa, A.; Puhn, F.; Helary, J.L.; Shaw, R.; Friend, M.; Piec, Z.; Tachikawa, N.; Acks, M.; Basile, A.

    1995-12-31

    Since ITER has many of the characteristics of a full-scale tokamak reactor, its provisions for assembly and replaceability are relevant to a future fusion power plant. The performance of ITER is dependent on tight tolerances, mainly for the magnets and plasma facing components. The magnetic field must be highly uniform in the toroidal direction to ensure good plasma energy and particle confinement. Alignment of the plasma facing surface of the first wall and divertor target plates is required to avoid large local heat loads on the plasma facing components and, as a consequence, their erosion and contamination of the plasma with impurities. Because of the large and heavy components the major challenge of the ITER tokamak assembly is to hold such tight tolerances using guide tools, adjustable interfaces, accurate measuring tools, and specific procedures to compensate for deformation and fabrication tolerances. The assembly tooling plan also includes verification of the essential remote handling operations.

  19. Cryogenic requirements for the JT-60SA Tokamak

    NASA Astrophysics Data System (ADS)

    Michel, Frederic; Hitz, D.; Hoa, Christine; Lamaison, Valerie; Kamiya, Koji; Roussel, Pascal; Wanner, Manfred; Yoshida, Kiyoshi

    2012-06-01

    The superconducting tokamak JT-60SA is part of the Broader Approach Programmeagreed between Japan and Europe. CEA is in charge of the cryogenic system procurementincluding the Warm Compression Station, the gas storages, the Refrigerator Cold Box andthe Auxiliary Cold Box (ACB) which has to be installed on the JAEA Naka site in 2016.This paper summarizes the updated cryogenic requirements for the tokamak JT-60SAcryogenic system. The cryogenic system has a refrigeration capacity of about 9 kW equivalent at 4.5K, to supply cryopump panels at 3.7 K, superconducting magnets and cold structures at 4.4 K, HTS current leads at 50 K, and thermal shields at 80 K. This paper presents the static and variable heat loads of the different cooling loops and the results of the rmohydraulic calculations to derive the transient heat loads at the interface between the magnet system cooling loops and the Auxiliary Cold Box.

  20. Recent Results of IRAN-T1 Tokamak

    SciTech Connect

    Dorranian, D.; Ghoranneviss, M.; Salem, M. K.; Mahmoodi D, M.; Arvin, R.; Talebitaher, Alireza; Abhari, Ali; Khorshid, P.; Hojabri, A.

    2006-12-04

    In this article after introducing the IR-T1 tokamak and its diagnostic systems a brief discussion on the range of grossly stable operating conditions of its plasma by Hugill diagram is presented. Hard disruption instability is studied experimentally in the next part, which confirms that MHD behavior in small tokamaks can be characterized by a single parameter q(a), safety factor at plasma edge. Finally the characteristics of the new regime of IR-T1 are reported. By our new model of triggering different fields (toroidal, ohmic and vertical), the plasma duration time is increased up to 35 ms with Ip of about 25 kA. By modifying capacitance and charging voltage of ohmic and vertical fields the spike oscillations which was appeared in the plasma behavior is taken out. The role of cleaning the vacuum chamber and using heavier gas for glow discharge and the effect of base pressure is described in detail.

  1. 3D MHD disruptions simulations of tokamaks plasmas

    NASA Astrophysics Data System (ADS)

    Paccagnella, Roberto; Strauss, Hank; Breslau, Joshua

    2008-11-01

    Tokamaks Vertical Displacement Events (VDEs) and disruptions simulations in toroidal geometry by means of a single fluid visco-resistive magneto-hydro-dynamic (MHD) model are presented in this paper. The plasma model, implemented in the M3D code [1], is completed with the presence of a 2D homogeneous wall with finite resistivity. This allows the study of the relatively slowly growing magneto-hydro-dynamical perturbation, the resistive wall mode (RWM), which is, in this work, the main drive of the disruptions. Amplitudes and asymmetries of the halo currents pattern at the wall are also calculated and comparisons with tokamak experimental databases and predictions for ITER are given. [1] W. Park, E.V. Belova, G.Y. Fu, X.Z. Tang, H.R. Strauss, L.E. Sugiyama, Phys. Plasmas 6 (1999) 1796.

  2. [alpha]-particle transport-driven current in tokamaks

    SciTech Connect

    Heikkinen, J.A. ); Sipilae, S.K. )

    1995-03-01

    It is shown that the radial transport of fusion-born energetic [alpha] particles, induced by electrostatic waves traveling in one poloidal direction, is directly connected to a net momentum of [alpha] particles in the toroidal direction in tokamaks. Because the momentum change is almost independent of toroidal velocity, the energy required for the momentum generation remains small on an [alpha]-particle population sustained by an isotropic time-independent source. By numerical toroidal Monte Carlo calculations it is shown that the current carried by [alpha] particles in the presence of intense well penetrated waves can reach several mega-amperes in reactor-sized tokamaks. The current obtained can greatly exceed the neoclassical bootstrap current of the [alpha] particles.

  3. Advances in Dust Detection and Removal for Tokamaks

    NASA Astrophysics Data System (ADS)

    Campos, A.; Skinner, C. H.; Roquemore, A. L.; Leisure, J. O. V.; Wagner, S.

    2008-11-01

    Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. An electrostatic dust detector[1] developed in the laboratory is being applied to NSTX. In the tokamak environment, large particles or fibres can fall on the grid potentially causing a permanent short. We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments with varying nozzle designs, backing pressures, puff durations, and exit flow orientations have obtained an optimal configuration that effectively removes particles from a 25 cm^2 area. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tripolar grid of fine interdigitated traces has been designed that generates an electrostatic travelling wave for conveying dust particles to a ``drain.'' First trials have shown particle motion in optical microscope images. [1] C. H. Skinner et al., J. Nucl. Mater., 376 (2008) 29.

  4. Multilayer mirror based line emission tomography for spherical Tokamaks

    SciTech Connect

    Stutman, D.; Hwang, Y.S.; Menard, J.; Choe, W.; Ono, M.; Finkenthal, M.; May, M.J.; Regan, S.P.; Soukhanovskii, V.; Moos, H.W.

    1997-01-01

    Due to their highly shaped plasma and possible poloidal asymmetry in impurity concentration, spherical Tokamaks will require tomographic reconstruction of local emissivities to assess impurity content and transport. To collect in an effective manner the data required for such reconstruction, we develop arrays of high throughput {open_quotes}mini-monochromators{close_quotes} using extreme ultraviolet multilayer mirrors as dispersive elements and filtered surface barrier diodes as detectors. We discuss monochromator optimization and show that by working at near normal incidence throughput and spectral resolution are simultaneously maximized. A system proposed for tomographic reconstruction of CV and CVI resonance emission at 33.7 and 40.5 {Angstrom} respectively, achieves 0.9 {Angstrom} spectral resolution, 2 cm spatial resolution, and 0.2 ms temporal resolution, together with good sensitivity and background rejection. Preliminary results obtained from CDX-U low aspect ratio tokamak are also presented. {copyright} {ital 1997 American Institute of Physics.}

  5. Estimation of Electron Temperature on Glass Spherical Tokamak (GLAST)

    NASA Astrophysics Data System (ADS)

    Hussain, S.; Sadiq, M.; Shah, S. I. W.; GLAST Team

    2015-03-01

    Glass Spherical Tokamak (GLAST) is a small spherical tokamak indigenously developed in Pakistan with an insulating vacuum vessel. A commercially available 2.45 GHz magnetron is used as pre-ionization source for plasma current startup. Different diagnostic systems like Rogowski coils, magnetic probes, flux loops, Langmuir probe, fast imaging and emission spectroscopy are installed on the device. The plasma temperature inside of GLAST, at the time of maxima of plasma current, is estimated by taking into account the Spitzer resistivity calculations with some experimentally determined plasma parameters. The plasma resistance is calculated by using Ohm's law with plasma current and loop voltage as experimentally determined inputs. The plasma resistivity is then determined by using length and area of the plasma column. Finally, the average plasma electron temperature is predicted to be 12.65eV for taking neon (Ne) as a working gas.

  6. Pseudo-MHD ballooning modes in tokamak plasmas

    SciTech Connect

    Callen, J.D.; Hegna, C.C.

    1996-08-01

    The MHD description of a plasma is extended to allow electrons to have both fluid-like and adiabatic-regime responses within an instability eigenmode. In the resultant {open_quotes}pseudo-MHD{close_quotes} model, magnetic field line bending is reduced in the adiabatic electron regime. This makes possible a new class of ballooning-type, long parallel extent, MHD-like instabilities in tokamak plasmas for {alpha} > s{sup 2}(2 {sup 7/3}/9) (r{sub p}/R{sub 0}) or-d{radical}{Beta}/dr > (2{sup 1/6} /3)(s/ R{sub 0q}), which is well below the ideal-MHD stability boundary. The marginally stable pressure profile is similar in both magnitude and shape to that observed in ohmically heated tokamak plasmas.

  7. Shape Optimization for DIII-D Advanced Tokamak Plasmas

    SciTech Connect

    C.E. Kesse; J.R. Ferron; C.M. Greenfield; J.E. Menard; T.S. Taylor

    2003-07-30

    The advanced tokamak program on DIII-D is targeting the full integration of high-beta and high-bootstrap/noninductive current fraction for long-pulse lengths and the high confinement consistent with these features. Central to achieving these simultaneously is access to the highest ideal beta limits possible to maximize the headroom for experimental operation with RWM control. A study of the ideal-MHD stability is done for plasmas modeled after DIII-D advanced tokamak plasmas, varying the plasma elongation, triangularity, and outboard squareness. The highest beta(sub)N limits reach 6-7 for the n=1 kink mode for all elongation, outer squareness values, and plasma triangularity equals 0.8.

  8. Equilibrium analysis of tokamak discharges with toroidal variation

    SciTech Connect

    Zwingmann, W.; Becoulet, M.; Moreau, Ph.; Nardon, E.

    2006-11-30

    Tokamaks provide a field structure that is almost axisymmetric around the torus axis. There are however always small toroidal variations due to the limited number of toroidal field coils, the magnetic field ripple. On the other hand, non-axisymmetric external fields are applied on purpose to ergodise the field structure close to the separatrix, to control the heat and particle transport across the plasma boundary. We present a perturbation method to calculate the magnetic field of tokamak discharges with with weak toroidal variation. The method is applied for the equilibrium reconstruction of Tore Supra discharges with toroidal ripple. The perturbation method does not rely on a flux surface representation and can therefore be applied to structures with magnetic islands. We obtain the plasma response to the field of ergodising external coils, as proposed for the ITER device.

  9. Maintenance concept development for the Compact Ignition Tokamak

    SciTech Connect

    Macdonald, D.

    1988-01-01

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs.

  10. On the Production of Relativistic Runaway Electrons in Damavand Tokamak

    NASA Astrophysics Data System (ADS)

    Moslehi-Fard, Mahmoud

    2013-02-01

    Experimental observations in Damavand tokamak show that hard X-ray is produced by either disruption with I p < 20 kA or by shots with I p > 20 kA. Hard X-ray also persists from the initiation of plasma discharge to the end. Occurrence of multiple spikes in hard X-ray during the discharge is evident. The propagation of hard X-ray is attributed to runaway electrons. We observe runaway electrons in two regimes with different characteristics. Regime (RADI) is similar to the observations of other Tokamak during disruption on that the plasma current is reduced abruptly and interpreted by Dreicer theory. In the regime of RADII, hard X-ray and subsequently runaway electrons are observed from starting of plasma discharge which provides the condition that the most of runaway electrons contain the toroidal plasma current. Runaway electron beam excites whistler waves and scattered electrons in velocity space and prevent growing the runaway electrons beam.

  11. Commissioning of heating neutral beams for COMPASS-D tokamak

    SciTech Connect

    Deichuli, P.; Davydenko, V.; Belov, V.; Gorbovsky, A.; Dranichnikov, A.; Ivanov, A.; Sorokin, A.; Mishagin, V.; Abdrashitov, A.; Kolmogorov, V.; Kondakov, A.

    2012-02-15

    Two neutral beam injectors have been developed for plasma heating on COMPASS-D tokamak (Institute of Plasma Physics, Prague). The 4-electrodes multihole ion-optical system with beam focusing was chosen to provide the low divergence 300 kW power in both deuterium and hydrogen atoms. The accelerating voltage is 40 kV at extracted ion current up to 15 A. The power supply system provides the continuous and modulated mode of the beam injection at a maximal pulse length 300 ms. The optimal arrangement of the cryopanels and the beam duct elements provides sufficiently short-length beamline which reduces the beam losses. The evolution of the impurities and molecular fraction content is studied in the process of the high voltage conditioning of the newly made ion sources. Two injectors of the same type have been successfully tested and are ready for operation at tokamak in IPP, Prague.

  12. Drift-wave fluctuation in an inviscid tokamak plasma

    NASA Astrophysics Data System (ADS)

    Yang, Jian-Rong; Mao, Jie-Jian; Tang, Xiao-Yan

    2013-11-01

    In order to describe the characterization of resistive drift-wave fluctuation in a tokamak plasma, a coupled inviscid two-dimensional Hasegawa—Wakatani model is investigated. Two groups of new analytic solutions with and without phase shift between the fluctuant density and the fluctuant potential are obtained by using the special function transformation method. It is demonstrated that the fluctuant potential shares similar spatio—temporal variations with the density. It is found from the solutions without phase shift that the effect of the diffusion and adiabaticity on the fluctuant density is quite complex, and that the fluctuation may be controlled through the adiabaticity and diffusion. By using the typical parameters in the quasi-adiabatic regime in the solutions with phase shift, it is shown that the density gradient becomes larger as the contours become dense toward the plasma edge and the contours have irregular structures, which reveal the nonuniform distribution in the tokamak edge.

  13. Testing of low Z coated limiters in tokamak fusion devices

    SciTech Connect

    Whitely, J.B.; Mullendore, A.W.; Langley, R.A.

    1980-01-01

    Extensive testing on a laboratory scale has been used to select those coatings most suitable for this environment. From this testing which included pulsed electron beam heating, low energy ion bombardment and arcing, chemical vapor deposited coating of TiB/sub 2/ and TiC on Poco graphite substrates have been selected and tested as limiters in ISX. Both limiter materials gave clean, stable, reproducible tokamak discharges the first day of operation. After one weeks exposure, the TiC limiter showed only superficial damage with no coating failure. The TiB/sub 2/ limiter had some small areas of coating failure. TiC coated graphite limiters have also been briefly tested in the tokamaks Alcator and PDX with favorable results.

  14. Internal disruptions and RHF in IR-T1 tokamak

    SciTech Connect

    Ghoranneviss, M.; Masnavi, M.; Khademian, A.

    1996-12-31

    Sawtooth oscillations are observed on IR-T1 Tokamak during low ql discharge with a disruption time of about 30--60 {micro}s. The q = 1 singular surface occurs at radius 3--3.5 cm and inverted Sawtooth from chords outside this radius. The superimposed (m = 1) oscillation with a frequency of about 19 kHz {approx} 25 kHz, according to the tokamak discharges parameters, preceding the Sawtooth oscillation. One major effect the Sawtooth oscillation is to flatten the temperature and density profiles approximately out to a mixing radius rm = {radical}2 rs,. Furthermore, by applying RHFs (L = 2 and L = 3), the Sawtooth behavior is modified. The magnitude of the weak RHFs used in the experiments did not exceed 1% of Bp. Results showed that the weak RHFs magnetic perturbation would change the MHD instabilities and the Sawtooth behavior, as well as plasma, confinement.

  15. Three-dimensional analysis of tokamaks and stellarators

    PubMed Central

    Garabedian, Paul R.

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  16. Three-dimensional equilibria in axially symmetric tokamaks

    PubMed Central

    Garabedian, Paul R.

    2006-01-01

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  17. Comparison of a radial fractional transport model with tokamak experiments

    SciTech Connect

    Kullberg, A. Morales, G. J.; Maggs, J. E.

    2014-03-15

    A radial fractional transport model [Kullberg et al., Phys. Rev. E 87, 052115 (2013)], that correctly incorporates the geometric effects of the domain near the origin and removes the singular behavior at the outer boundary, is compared to results of off-axis heating experiments performed in the Rijnhuizen Tokamak Project (RTP), ASDEX Upgrade, JET, and DIII-D tokamak devices. This comparative study provides an initial assessment of the presence of fractional transport phenomena in magnetic confinement experiments. It is found that the nonlocal radial model is robust in describing the steady-state temperature profiles from RTP, but for the propagation of heat waves in ASDEX Upgrade, JET, and DIII-D the model is not clearly superior to predictions based on Fick's law. However, this comparative study does indicate that the order of the fractional derivative, α, is likely a function of radial position in the devices surveyed.

  18. Controls of Magnetic Islands by Pellet Injection in Tokamaks

    SciTech Connect

    Shaing, K. C.; Houlberg, Wayne A; Peng, Yueng Kay Martin

    2007-01-01

    The appearance of magnetic islands in tokamaks degrades plasma confinement. It is therefore important to control or eliminate the growth of the islands to improve the performance of a tokamak. A theory is developed to control magnetic islands using the localized pressure gradient driven bootstrap current by injecting pellets at the O-point of the island to create a peaked plasma pressure profile inside the island. This localized bootstrap current replenishes the missing equilibrium bootstrap current density that causes the island to grow in the first place. It is shown that the effect of the localized bootstrap current tends to reduce or eliminate the original drive for the growth of the island in the island evolution equation. The theory is also valid for the localized bootstrap current created by localized heating, but with much less effectiveness. A possibility of eliminating the island by controlling the equilibrium profiles is also discussed.

  19. On the bootstrap current in stellarators and tokamaks

    SciTech Connect

    Helander, P.; Geiger, J.; Maassberg, H.

    2011-09-15

    The expression for the long-mean-free-path limit of the bootstrap current in stellarators is rederived in such a way that the expansion procedure is identical to that used in the corresponding calculation for a tokamak. In addition, the first correction due to finite collisionality is calculated and shown to vanish in quasi-isodynamic configurations without net current. This correction, which is proportional to the square root of the collisionality, is found to compare well with a numerical solution of the first-order drift kinetic equation in spherical tokamak geometry. Numerically, it appears that there is a similar correction in general stellarator geometry, which however depends on the strength of the radial electric field.

  20. Control of magnetic islands by pellet injection in tokamaks

    SciTech Connect

    Shaing, K. C.; Houlberg, W. A.; Peng, M.

    2007-07-15

    The appearance of magnetic islands in tokamaks degrades plasma confinement. It is therefore important to control or eliminate the growth of the islands to improve the performance of a tokamak. A theory is developed to control magnetic islands using the localized pressure gradient driven bootstrap current by injecting pellets at the O-point of the island to create a peaked plasma pressure profile inside the island. This localized bootstrap current replenishes the missing equilibrium bootstrap current density that causes the island to grow in the first place. It is shown that the effect of the localized bootstrap current tends to reduce or eliminate the original drive for the growth of the island in the island evolution equation. The theory is also valid for the localized bootstrap current created by localized heating, but with much less effectiveness. A possibility of eliminating the island by controlling the equilibrium profiles is also discussed.

  1. Control of magnetic islands by pellet injection in tokamaks

    SciTech Connect

    Shaing, K. C.; Rome, James A; Peng, Yueng Kay Martin

    2007-01-01

    The appearance of magnetic islands in tokamaks degrades plasma confinement. It is therefore important to control or eliminate the growth of the islands to improve the performance of a tokamak. A theory is developed to control magnetic islands using the localized pressure gradient driven bootstrap current by injecting pellets at the O-point of the island to create a peaked plasma pressure profile inside the island. This localized bootstrap current replenishes the missing equilibrium bootstrap current density that causes the island to grow in the first place. It is shown that the effect of the localized bootstrap current tends to reduce or eliminate the original drive for the growth of the island in the island evolution equation. The theory is also valid for the localized bootstrap current created by localized heating, but with much less effectiveness. A possibility of eliminating the island by controlling the equilibrium profiles is also discussed. (c) 2007 American Institute of Physics.

  2. Equilibrium calculations for plasma control in CIT (Compact Ignition Tokamak)

    SciTech Connect

    Strickler, D.J.; Peng, Y-K.M. . Fusion Engineering Design Center); Pomphrey, N.; Jardin, S.C. . Plasma Physics Lab.)

    1990-01-01

    The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs.

  3. Gyrokinetic Simulation of Global Turbulent Transport Properties in Tokamak Experiments

    SciTech Connect

    Wang, W.X.; Lin, Z.; Tang, W.M.; Lee, W.W.; Ethier, S.; Lewandowski, J.L.V.; Rewoldt, G.; Hahm, T.S.; Manickam, J.

    2006-01-01

    A general geometry gyro-kinetic model for particle simulation of plasma turbulence in tokamak experiments is described. It incorporates the comprehensive influence of noncircular cross section, realistic plasma profiles, plasma rotation, neoclassical (equilibrium) electric fields, and Coulomb collisions. An interesting result of global turbulence development in a shaped tokamak plasma is presented with regard to nonlinear turbulence spreading into the linearly stable region. The mutual interaction between turbulence and zonal flows in collisionless plasmas is studied with a focus on identifying possible nonlinear saturation mechanisms for zonal flows. A bursting temporal behavior with a period longer than the geodesic acoustic oscillation period is observed even in a collisionless system. Our simulation results suggest that the zonal flows can drive turbulence. However, this process is too weak to be an effective zonal flow saturation mechanism.

  4. On the economic prospects of nuclear fusion with tokamaks

    NASA Astrophysics Data System (ADS)

    Pfirsch, D.; Schmitter, K. H.

    1987-12-01

    A method of cost and construction energy estimation for tokamak fusion power stations conforming to the present stage of fusion development is described. The method is based on first-wall heat load constraints rather than Beta limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled Environmental Impact and Economic Prospects of Nuclear Fusion (1986). It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is derived when it is properly applied for cost estimation of advanced gas-cooled and Magnox reactors, the two examples presented by the European study to disprove it.

  5. Molecular emission in the edge plasma of T-10 tokamak

    SciTech Connect

    Zimin, A. M.; Krupin, V. A.; Troynov, V. I.; Klyuchnikov, L. A.

    2015-12-15

    The experiments on recording molecular emission in the edge plasma of the T-10 tokamak are described. To obtain reliable spectra with sufficient spectral, temporal, and spatial resolution, the optical circuit is optimized for various experimental conditions. Typical spectra measured in two sections of the tokamak are shown. It is shown that, upon varying the parameters of the discharge, the molecular spectrum not only changes significantly in intensity but also undergoes a qualitative change in the rotational and vibrational structure. For a detailed analysis, we use the Fulcher-α system (d{sup 3}Π{sub u}–a{sup 3}Σ{sub g}{sup +}) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X{sup 1}Σ{sub g}{sup +} and upper excited state d{sup 3}Π{sub u} are estimated by the measured spectra.

  6. Development on JET of advanced tokamak operations for ITER

    NASA Astrophysics Data System (ADS)

    Tuccillo, A. A.; Crisanti, F.; Litaudon, X.; Baranov, Yu. F.; Becoulet, A.; Becoulet, M.; Bertalot, L.; Castaldo, C.; Challis, C. D.; Cesario, R.; DeBaar, M. R.; de Vries, P. C.; Esposito, B.; Frigione, D.; Garzotti, L.; Giovannozzi, E.; Giroud, C.; Gorini, G.; Gormezano, C.; Hawkes, N. C.; Hobirk, J.; Imbeaux, F.; Joffrin, E.; Lomas, P. J.; Mailloux, J.; Mantica, P.; Mantsinen, M. J.; Mazon, D.; Moreau, D.; Murari, A.; Pericoli-Ridolfini, V.; Rimini, F.; Sips, A. C. C.; Sozzi, C.; Tudisco, O.; Van Eester, D.; Zastrow, K.-D.; work-programme contributors, JET-EFDA

    2006-02-01

    Recent research on advanced tokamak in JET has focused on scenarios with both monotonic and reversed shear q-profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide internal transport barriers (ITBs), r/a ~ 0.7, are formed at ITER relevant triangularity δ ~ 0.45 and moderate plasma current, IP = 1.5-2.5 MA, with ne/nG ~ 60% when ELMs are moderated by Ne injection. At higher current (IP <= 3.5 MA, δ ~ 0.25) wide ITBs sitting at r/a >= 0.5, in the positive shear region, have been developed. Generally MHD events terminate these barriers otherwise limited in strength by power availability. ITBs with core density close to Greenwald value, Te ~ Ti and low toroidal rotation (4 times lower than standard ITBs) are obtained in plasma target preformed by opportune timing of lower hybrid current drive (LHCD), pellet injection and a small amount of NBI power. Wide ITBs, r/a ~ 0.6, of moderate strength, can be sustained without impurities accumulation for a time close to neoclassical resistive time in 3 T/1.8 MA discharges that exhibit reversed magnetic shear profiles and type-III ELMy edge. These discharges have been extended to the maximum duration allowed by JET subsystems (20 s) bringing to the record of injected energy in a JET discharge: E ~ 330 MJ. Portability of ITB physics has been addressed through dedicated similarity experiments. The ITB is identified as a layer of reduced diffusivity studying the propagation of the heat wave generated by modulating the ICRF mode conversion (MC) electron heating. Impressive results, QDT ~ 0.25, are obtained in these deuterium discharges with 3He minority when the MC layer is located in the core. The ion behaviour has been investigated in pure LHCD electron ITBs optimizing the 3He minority concentration for direct ion heating. Preliminary results of particle transport, studied via injection of a trace of tritium and an Ar-Ne mixture, will be presented.

  7. (Injection of compact toroids for tokamak fueling and current drive)

    SciTech Connect

    Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.

    1991-01-01

    The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.

  8. Note for the Mirnov signal analysis in tokamaks

    SciTech Connect

    Kikuchi, M.

    1985-05-01

    The relation between Mirnov coil signals and the current perturbation on the rational surface is examined analytically by using the approximate Green's function for the case of large aspect ratio circular tokamaks. Satellite island formation, phase modulation effect due to the poloidal variation of the field line pitch, and the shift effect of the plasma column with respect to the center of the vacuum chamber are examined. The detectability of these effects from Mirnov coil signals is discussed for TFTR.

  9. Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport

    SciTech Connect

    Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.

    2011-03-18

    The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.

  10. Problems with the concept of plasma equilibrium in tokamaks

    SciTech Connect

    Carreras, B.A.

    1992-06-01

    The equilibrium condition for a magnetically confined plasma in normally formulated in terms of macroscopic equations. In these equations, the plasma pressure is assumed to be a function of the magnetic flux with continuous derivatives. However, in three- dimensional systems this is not necessarily the case. Here, we look at the case of an intrinsically three-dimensional realistic tokamak, and we discuss the possible interconnection between the equilibrium and anomalous transport.

  11. FEM (Free Electron Maser) for tokamak: Final report

    SciTech Connect

    Not Available

    1987-01-01

    This paper studies the feasibility of a microwave source for heating a tokamak reactor. The free electron maser (FEM) shows great promise for being this source. The topics covered in this paper are microwave generation with FEM, efficiency enhancement, parameter scaling, space charge scaling, beam energy spread and efficiency scaling, electron beam line with energy recovery, achromatic bend, multi-stage depressed voltage electron beam collector, and development plans. 12 refs., 10 figs., 5 tabs. (LSP)

  12. Thermally excited proton spin-flip laser emission in tokamaks

    SciTech Connect

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser.

  13. Multi-field plasma sandpile model in tokamaks and applications

    NASA Astrophysics Data System (ADS)

    Peng, X. D.; Xu, J. Q.

    2016-08-01

    A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

  14. Drift-Wave Instabilities and Transport in Non - Tokamak Geometry

    NASA Astrophysics Data System (ADS)

    Hua, Daniel Duc

    Motivated by experimental scaling laws that suggest an improvement in the confinement time of fusion plasmas in tokamaks with elongated cross section, we search theoretically for favorable dependence on elongation for drift-wave instabilities, which may be responsible for anomalous transport in tokamaks. First, using thermodynamic methods, we derive upper bounds on thermal diffusivities for drift-wave instabilities in tokamaks but find no elongation dependence to lowest order. Also, compared with experimentally inferred ion thermal diffusivities from the DIIID tokamak, the thermodynamic bounds are as much as 100 times bigger in the plasma core. Second, utilizing a simulation code to calculate linear growth rates, we obtain mixing-length estimates of ion thermal diffusivities for a specific drift wave, the ion-temperature-gradient (ITG) mode, which becomes unstable only if the temperature gradient exceeds a finite threshold value (whereas the thermodynamic constraints allow instability for any value). We find that the simulation growth rates and the diffusivities estimated from them do decrease for increasing elongation, due to finite Larmor radius effects (which do not explicitly appear in the thermodynamic constraints). Compared with the experimentally inferred diffusivities, the simulation diffusivities are similar near the edge but are 10 times bigger in the core. However, a small adjustment in the temperature profile, within experimental and theoretical uncertainties, would produce good agreement everywhere. Therefore, we suggest that for the DIIID experiments studied, the plasma is actually very close to the ITG instability threshold in the core and farther away from threshold near the edge, but not far enough to induce the full thermodynamic level of diffusivities. This conjecture is supported by model transport calculations that reproduce the experimental diffusivity profile fairly well.

  15. Computational model of surface ablation from tokamak disruptions

    SciTech Connect

    Ehst, D.; Hassanein, A.

    1993-10-01

    Energy transfer to material surfaces is dominated by photon radiation through low temperature plasma vapors if tokamak disruptions are due to low kinetic energy particles ( < 100 eV). Simple models of radiation transport are derived and incorporated into a fast-running computer routine to model this process. The results of simulations are in fair agreement with plasma gun erosion tests on several metal targets.

  16. Viscous damping of toroidal angular momentum in tokamaks

    SciTech Connect

    Stacey, W. M.

    2014-09-15

    The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various “neoclassical toroidal viscosity” mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.

  17. Imaging charge exchange recombination spectroscopy on the TEXTOR tokamak

    NASA Astrophysics Data System (ADS)

    Howard, J.; Jaspers, R.; Lischtschenko, O.; Delabie, E.; Chung, J.

    2010-12-01

    We describe the application of a simple spatial-heterodyne coherence-imaging filter for 2D Doppler imaging of charge exchange recombination (CXR) emission from a heating beam in the TEXTOR tokamak. Results obtained by the CXR imaging system are found to be consistent with measurements obtained using a standard multi-channel spectrometer-based system. We describe the system, indicate possible enhancements and future applications for imaging CXRS.

  18. Edge region hydrogen line emission in the PDX tokamak

    SciTech Connect

    McNeill, D.H.; Bell, M.G.; Grek, B.; LeBlanc, B.

    1984-02-01

    Measurements of the H/sub ..cap alpha../ line shape and of the spatial distribution of the H/sub ..cap alpha../ emissivity in the PDX tokamak are interpreted in terms of molecular dissociation reactions at the plasma edge. The influx of molecules is shown to be roughly proportional to the edge emission intensity over a wide range of temperatures. The H/sub 2/ particle lifetime is estimated for various types of discharges in PDX.

  19. Edge region hydrogen line emission in the PDX tokamak

    SciTech Connect

    McNeill, D.H.; Bell, M.G.; Grek, B.; LeBlanc, B.

    1984-04-01

    Measurements of the H/sub ..cap alpha../ line shape and of the spatial distribution of H/sub ..cap alpha../ emissivity in the PDX tokamak are interpreted in terms of molecular dissociation reactions at the plasma edge. The in-flux of molecules is shown to be roughly proportional to the edge emission intensity over a wide range of temperatures. The H/sub 2/ particle lifetime is estimated for various types of discharges in PDX.

  20. Stochastic Acceleration of Dust Particles in Tokamak Edge Plasmas

    SciTech Connect

    Marmolino, C.; De Angelis, U.; Ivlev, A. V.; Morfill, G. E.

    2008-10-15

    Stochastic heating of dust particles resulting from dust charge fluctuations is considered in the conditions of the scrape-off-layer (SOL) in tokamak plasmas. It is shown that kinetic energies corresponding to velocities of {approx_equal}Km/s can be reached in times of order {approx_equal}1 ms by micron-size dust particles interacting with a background of stochastically heated nano-size dust particles.

  1. A moving finite element model of the tokamak scrapeoff layer

    SciTech Connect

    Glasser, A.H.; Kuprat, A.P.

    1993-10-01

    Most numerical simulations of the tokamak scrapeoff layer use a mapping to flux coordinates and a piecewise equidistributed grid in those coordinates to resolve the multiple length scales and anisotropy characteristic of this problem. We have developed an alternative numerical method using simple cylindrical coordinates with a complex adaptive grid scheme. It is based on an understructured grid of traingles which move adaptively, aligning themselves with the magnetic field and concentrating in regions of sharp gradients.

  2. Key Aspects of EBW Heating and Current Drive in Tokamaks

    NASA Astrophysics Data System (ADS)

    Urban, Jakub; Decker, Joan; Preinhaelter, Josef; Taylor, Gary; Vahala, Linda; Vahala, George

    2010-11-01

    Electron Bernstein wave (EBW) heating and current drive is modeled by coupled mode conversion, ray-tracing (AMR) and Fokker-Planck (LUKE) codes. Deposition and current drive profiles are determined for EBW with various injection parameters under realistic spherical tokamak conditions. There parameters are varied to investigate the robustness of the applied scenarios. The importance of relativistic corrections to EBW absorption is considered. The differences between various relativistic models are explored.

  3. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  4. Dust-Particle Transport in Tokamak Edge Plasmas

    SciTech Connect

    Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D

    2005-09-12

    Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.

  5. Variational Symplectic Orbit Code in 3-D Tokamak Geometry

    NASA Astrophysics Data System (ADS)

    Ellison, Charles; Qin, Hong; Tang, William M.

    2011-10-01

    Since advanced tokamak experiments - including ITER - are long-pulse systems, it is important to develop accurate numerical methods to track plasma dynamics over an extended temporal period. When attempting to model the motion of individual particles, standard integrators (e.g. 4th order Runge-Kutta) discretize the differential equations of motion - but do not possess desired properties such as energy conservation. The variational symplectic integrator adopts instead a different approach via minimizing the action of the guiding center motion to determine iteration rules. Consequently, the Lagrangian symplectic structure is conserved, and the numerical energy error is bounded by a small number for all time-steps. In previous work, the theoretical basis for this method was introduced, but the implementation was for 2-D geometry. To address realistic experimental scenarios, the variational symplectic integrator has been implemented for 3-D tokamak geometry for the first time. Sample results will be presented and compared with those from standard Runge-Kutta-based 3-D tokamak orbit codes. This work was supported by the DOE contract # DE-AC02-09CH11466 and the DOE FES Fellowship.

  6. Performance Projections For The Lithium Tokamak Experiment (LTX)

    SciTech Connect

    Majeski, R.; Berzak, L.; Gray, T.; Kaita, R.; Kozub, T.; Levinton, F.; Lundberg, D. P.; Manickam, J.; Pereverzev, G. V.; Snieckus, K.; Soukhanovskii, V.; Spaleta, J.; Stotler, D.; Strickler, T.; Timberlake, J.; Yoo, J.; Zakharov, L.

    2009-06-17

    Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

  7. Performance projections for the lithium tokamak experiment (LTX)

    NASA Astrophysics Data System (ADS)

    Majeski, R.; Berzak, L.; Gray, T.; Kaita, R.; Kozub, T.; Levinton, F.; Lundberg, D. P.; Manickam, J.; Pereverzev, G. V.; Snieckus, K.; Soukhanovskii, V.; Spaleta, J.; Stotler, D.; Strickler, T.; Timberlake, J.; Yoo, J.; Zakharov, L.

    2009-05-01

    Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

  8. Currents induced in tokamaks by electron cyclotron heating

    SciTech Connect

    Eldridge, O. C.

    1980-10-01

    Generation of a plasma current is predicted in association with strong electron cyclotron heating in tokamaks or in any plasma with transverse magnetic field gradients. The current predicted in present-day tokamaks is of the order of one-quarter ampere per watt, which is large enough to be detected in heating experiments in progress. The current scales linearly with electron temperature and heating power and inversely with density and major radius. The mechanism depends on the Doppler shift for electrons streaming along magnetic field lines. Electrons streaming toward the source of radiation are resonant at a larger magnetic field. When the interaction is strong, radiation incident from the high field side is absorbed before reaching the cold electron resonant surface, and, so, a unidirectional population of electrons is heated. The anisotropic electron distribution gains momentum by collisions with ions. For small tokamaks the extraordinary wave should be launched for current drive, but for reactors the ordinary wave produces a sufficiently strong interaction.

  9. Economic analyses of alpha channeling in tokamak power plants.

    SciTech Connect

    Ehst, D.A.

    1998-09-17

    The hot-ion-mode of operation [1] has long been thought to offer optimized performance for long-pulse or steady-state magnetic fusion power plants. This concept was revived in recent years when theoretical considerations suggested that nonthermal fusion alpha particles could be made to channel their power density preferentially to the fuel ions [2,3]. This so-called anomalous alpha particle slowing down can create plasmas with fuel ion temperate T{sub i} somewhat larger than the electron temperature T{sub e}, which puts more of the beta-limited plasma pressure into the useful fuel species (rather than non-reacting electrons). As we show here, this perceived benefit may be negligible or nonexistent for tokamaks with steady state current drive. It has likewise been argued [2,3] that alpha channeling could be arranged such that little or no external power would be needed to generate the steady state toroidal current. Under optimistic assumptions we show that such alpha-channeling current drive would moderately improve the economic performance of a first stability tokamak like ARIES-I [4], however a reversed-shear (advanced equilibrium) tokamak would likely not benefit since traditional radio-wave (rf) electron-heating current drive power would already be quite small.

  10. Modeling of Anomalous Transport in Tokamaks with FACETS code

    NASA Astrophysics Data System (ADS)

    Pankin, A. Y.; Batemann, G.; Kritz, A.; Rafiq, T.; Vadlamani, S.; Hakim, A.; Kruger, S.; Miah, M.; Rognlien, T.

    2009-05-01

    The FACETS code, a whole-device integrated modeling code that self-consistently computes plasma profiles for the plasma core and edge in tokamaks, has been recently developed as a part of the SciDAC project for core-edge simulations. A choice of transport models is available in FACETS through the FMCFM interface [1]. Transport models included in FMCFM have specific ranges of applicability, which can limit their use to parts of the plasma. In particular, the GLF23 transport model does not include the resistive ballooning effects that can be important in the tokamak pedestal region and GLF23 typically under-predicts the anomalous fluxes near the magnetic axis [2]. The TGLF and GYRO transport models have similar limitations [3]. A combination of transport models that covers the entire discharge domain is studied using FACETS in a realistic tokamak geometry. Effective diffusivities computed with the FMCFM transport models are extended to the region near the separatrix to be used in the UEDGE code within FACETS. 1. S. Vadlamani et al. (2009) %First time-dependent transport simulations using GYRO and NCLASS within FACETS (this meeting).2. T. Rafiq et al. (2009) %Simulation of electron thermal transport in H-mode discharges Submitted to Phys. Plasmas.3. C. Holland et al. (2008) %Validation of gyrokinetic transport simulations using %DIII-D core turbulence measurements Proc. of IAEA FEC (Switzerland, 2008)

  11. Operation of a tokamak reactor in the radiative improved mode

    NASA Astrophysics Data System (ADS)

    Morozov, D. Kh.; Mavrin, A. A.

    2016-03-01

    The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ E with an increase in the introduced power has been analyzed.

  12. Impurity control and vacuum pumping system design and analysis for next-generation tokamaks

    SciTech Connect

    Haines, J.R.

    1985-01-01

    Impurity control system design and performance studies were performed in support of the Tokamak Fusion Core Experiment (TFCX) preconceptual design. Efforts concentrated on the pumped limiter and vacuum pumping system design configuration, thermal/mechanical and erosion lifetime performance of the limiter protective surface, and helium ash removal performance. Analysis results indicate that the limiter/vacuum pumping system design provides marginally adequate helium ash removal. Difficulties in providing adequate helium ash removal for more compact or higher fusion-power-density devices are addressed. Erosion, primarily by disruption-induced vaporization and/or melting, limits the protective surface lifetime to about one calendar year or only about 60 full power hours of operation. In addition to evaluating impurity control system performance for nominal TFCX conditions, these studies attempt to focus on the key plasma physics and engineering design issues that should be addressed in future research and development programs.

  13. Study of Total Detection Efficiency (η) of Charge Exchange Analyzer (CXA) in TVD Tokamak

    NASA Astrophysics Data System (ADS)

    Kazemi, M.; Habibi, M.; Tafreshi, M.

    2013-02-01

    Diagnostic devices detection efficiency consist of many parameters. Contribution of each part in main detection efficiency need to study all aspect of devices and experimentally researches. In this paper experimental results and theoretical study merged to figure out the total detection efficiency of CXA. Most important primary parameters has been recognized (stripping foil thickness, incident energy, etc.) and effect of these parameters in related part studied. For CXA of TVD tokamak total detection efficiency presented and shown for 3 keV this parameter equal to .09 and stripping efficiency is equal 1.1 approximately. Choosing appropriate system for stripping of neutral beam is the most important factor in CXA diagnostic devices specially for high energy beams.

  14. Data acquisition and processing system of the electron cyclotron emission imaging system of the KSTAR tokamak

    SciTech Connect

    Kim, J. B.; Lee, W.; Yun, G. S.; Park, H. K.; Domier, C. W.; Luhmann, N. C. Jr.

    2010-10-15

    A new innovative electron cyclotron emission imaging (ECEI) diagnostic system for the Korean Superconducting Tokamak Advanced Research (KSTAR) produces a large amount of data. The design of the data acquisition and processing system of the ECEI diagnostic system should consider covering the large data production and flow. The system design is based on the layered structure scalable to the future extension to accommodate increasing data demands. Software architecture that allows a web-based monitoring of the operation status, remote experiment, and data analysis is discussed. The operating software will help machine operators and users validate the acquired data promptly, prepare next discharge, and enhance the experiment performance and data analysis in a distributed environment.

  15. Application of an array processor to the analysis of magnetic data for the Doublet III tokamak

    SciTech Connect

    Wang, T.S.; Saito, M.T.

    1980-08-01

    Discussed herein is a fast computational technique employing the Floating Point Systems AP-190L array processor to analyze magnetic data for the Doublet III tokamak, a fusion research device. Interpretation of the experimental data requires the repeated solution of a free-boundary nonlinear partial differential equation, which describes the magnetohydrodynamic (MHD) equilibrium of the plasma. For this particular application, we have found that the array processor is only 1.4 and 3.5 times slower than the CDC-7600 and CRAY computers, respectively. The overhead on the host DEC-10 computer was kept to a minimum by chaining the complete Poisson solver and free-boundary algorithm into one single-load module using the vector function chainer (VFC). A simple time-sharing scheme for using the MHD code is also discussed.

  16. The rapid inward diffusion of cold ions in tokamaks and their effect on ion transport

    NASA Astrophysics Data System (ADS)

    Ware, A. A.

    1990-06-01

    The observed increase with density of the density asymmetry caused by the centrifugal force of toroidal motion in the PDX tokamak [Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1981), Vol. 1, p. 665], which is contrary to conventional theory, is explained by the presence of an excess of low-energy ions with 10%-15% concentration. The prime source being recycling, it is shown that low-energy ions undergo rapid inward diffusion (too rapid to thermalize with the outward diffusing energetic ions) because of the combined effects of large νPA, electrostatic diffusion, and negative Er and ∂Ti/∂r. The presence of the low-energy ions alters dramatically the predictions of neoclassical theory and many hydrogen and impurity ion transport phenomena now have simple explanations.

  17. Static and Dynamic Mechanical Analyses for the Vacuum Vessel of EAST Superconducting Tokamak Device

    NASA Astrophysics Data System (ADS)

    Song, Yuntao; Yao, Damao; Du, Shijun; Wu, Songtao; Weng, Peide

    2006-03-01

    EAST (experimental advanced superconducting tokamak) is an advanced steady-state plasma physics experimental device, which is being constructed as the Chinese National Nuclear Fusion Research Project. During the plasma operation the vacuum vessel as one of the key component will withstand the electromagnetic force due to the plasma disruption, the Halo current and the toroidal field coil quench, the pressure of boride water and the thermal load due to 250 oC baking by pressurized nitrogen gas. In this paper a report of the static and dynamic mechanical analyses of the vacuum vessel is made. Firstly the applied loads on the vacuum vessel were given and the static stress distribution under the gravitational loads, the pressure loads, the electromagnetic loads and thermal loads were investigated. Then a series of primary dynamic, buckling and fatigue life analyses were performed to predict the structure's dynamic behavior. A seismic analysis was also conducted.

  18. The rapid inward diffusion of cold ions in tokamaks and their effect on ion transport

    SciTech Connect

    Ware, A.A. )

    1990-06-01

    The observed increase with density of the density asymmetry caused by the centrifugal force of toroidal motion in the PDX tokamak ({ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Nuclear} {ital Fusion} {ital Research} (IAEA, Vienna, 1981), Vol. 1, p. 665), which is contrary to conventional theory, is explained by the presence of an excess of low-energy ions with 10%--15% concentration. The prime source being recycling, it is shown that low-energy ions undergo rapid inward diffusion (too rapid to thermalize with the outward diffusing energetic ions) because of the combined effects of large {nu}{sub PA}, electrostatic diffusion, and negative {ital E}{sub {ital r}} and {partial derivative}{ital T}{sub {ital i}}/{partial derivative}{ital r}. The presence of the low-energy ions alters dramatically the predictions of neoclassical theory and many hydrogen and impurity ion transport phenomena now have simple explanations.

  19. Imaging polarimeter/interferometer arrays for tokamak measurements. Technical progress report FY 84. [Imaging polarimeters and interferometer arrays for tokamaks

    SciTech Connect

    Not Available

    1984-01-01

    The Task IIIB program has continued to make significant strides during the last year. Laboratory test studies continued in our development efforts on imaging polarimeter and interferometer arrays in support of the tokamak measurements carried out under Task IIIA. This work ensures that the system optics and resolution are completely understood prior to attempting actual tokamak measurements. New microbolometer designs and fabrication techniques increased their sensitivity by over an order of magnitude compared with the previous devices. In addition, the development of sensitive monolithic integrated Schottky diode detector arrays has shown rapid progress. Heterodyne noise temperature of less than 9000/sup 0/K have already been achieved at 94 GHz with extension into the submillimeter region anticipated during the coming year.

  20. Far infrared tangential interferometry/polarimetry on the National Spherical Tokamak Experiment

    SciTech Connect

    Park, H.K.; Domier, C.W.; Geck, W.R.; Luhmann, N.C. Jr.

    1999-01-01

    Measurement of the core B{sub T}(r,t) value is essential in the National Spherical Tokamak Experiment (NSTX), since the effects of paramagnetism and diamagnetism in the NSTX are expected to be considerably greater than that in higher aspect ratio tokamaks. Therefore, without independent B{sub T}(r,t) measurement, plasma parameters dependent upon B{sub T} such as the {ital q} profile and local {beta} value cannot be evaluated. Tangential interferometer/polarimeter systems (eight channels) [H. Park, L. Guttadora, C. Domier, W. R. Geck, and N. C. Luhman, Jr., First and Second NSTX Research Forums, Princeton, NJ, 1997 (unpublished)] for the NSTX will provide temporally and radially resolved toroidal field profile [B{sub T}(r,t)] and two-dimensional electron density profile [n{sub e}(r,t)] data. The outcome of the proposed system is extremely important to the study of confinement, heating, and stability of the NSTX plasmas. The research task is largely based on utilizing existing hardware from the TFTR multichannel infrared interferometer system [D. K. Mansfield, H. K. Park, L. C. Johnson, H. Anderson, S. Foote, B. Clifton, and C. H. Ma, Appl. Opt. {bold 26}, 4469 (1987) and H. K. Park, D. K. Mansfield, and C. L. Johnson, Proceedings of the 3rd International Symposium on Laser-Aided Plasma Diagnostic, Los Angeles, CA, 28{endash}30 Oct. 1987 (unpublished), pp. 96{endash}104] which will be reconfigured into a tangential system for NSTX, and to develop the additional hardware required to complete the system. {copyright} {ital 1999 American Institute of Physics.}

  1. High power heating of magnetic reconnection in merging tokamak experiments

    SciTech Connect

    Ono, Y.; Tanabe, H.; Gi, K.; Watanabe, T.; Ii, T.; Yamada, T.; Gryaznevich, M.; Scannell, R.; Conway, N.; Crowley, B.; Michael, C.

    2015-05-15

    Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R∼10{sup 5}. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magnetic reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field B{sub rec}{sup 2}  ∼  B{sub p}{sup 2}. The guide toroidal field B{sub t} does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field B{sub t}, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with B{sub p} > 0.4 T will enables us to heat the plasma to the alpha heating regime: T{sub i} > 5 keV without using any additional heating facility.

  2. The STOR-1M Tokamak: Experiments on Current Reversal and Fast Current Ramping.

    NASA Astrophysics Data System (ADS)

    Wolfe, Sean William

    1987-09-01

    Experiments on the STOR-1M tokamak have addressed the following problems in magnetic confinement fusion research: (1) the attainment of quasi-continuous operation in a fusion reactor by inductively driving a current which alternates polarity; (2) plasma heating by inducing turbulence to enhance the Ohmic dissipation; and, (3) the stability of tokamak plasmas carrying large currents. STOR-1M plasmas have major and minor radii of 22 cm and 3.5 cm, respectively, and are produced with hydrogen at a pressure of 0.9 mTorr. Typical discharges carry a current of 5 kA with a toroidal field of 1 T. Around the current peak, electrons at a density of 1 x 10('13) cm(' -3) are heated to a temperature of 80 eV. The ions, with an effective charge number of 2, reach a temperature of 30 eV. Input energy is confined for almost 0.1 ms at the current peak, and the total discharge length is usually 4.5 ms. To simulate the current reversal phase in an ac tokamak reactor, a sinusoidal plasma current has been sustained for one cycle. Peak currents of 8 kA and electron densities of 1.8 x 10('13) cm('-3) have been attained. The electron density at the reversal is always at least 2 x 10('12) cm('-3). The unexpected equilibrium when the toroidal current goes through zero may be due to vertical plasma currents closing through the limiter or chamber walls. To induce turbulence for plasma heating, an electric field pulse of amplitude up to 360 V/m and width 20 (mu)s drives up to 10 kA of current on top of a normal discharge. After the pulse, electron temperatures of 300 eV and ion temperatures of 200 eV have been recorded. About 200 (mu)s after the pulse, the electron density and temperature reach a peak, implying that containment of energy is enhanced. The safety factor at the plasma surface during the pulse can be as low as 1.5. Disruptive behaviour, in the form of current interruption and loop voltage spikes, is observed when the safety factor is between 1.85 and 2.1. Outside this range, the

  3. ADX: a high field, high power density, advanced divertor and RF tokamak

    NASA Astrophysics Data System (ADS)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  4. Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

    NASA Astrophysics Data System (ADS)

    Ren, Q.

    2015-11-01

    Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

  5. DIII-D research operations. Annual report, October 1, 1992--September 30, 1993

    SciTech Connect

    La Haye, R.J.

    1994-05-01

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

  6. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    SciTech Connect

    Simonen, T.C.; Baker, D.

    1993-01-01

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies.

  7. Summary of the IEA Workshop on Alpha Physics and Tritium Issues in Large Tokamaks

    SciTech Connect

    Cheng, C.Z.; Stratton, B.; Zweben, S.J.; Pitcher, C.S.

    1993-11-01

    A brief summary is presented of the talks given during this meeting, which was held at PPPL and sponsored by the IEA (International Energy Agency) as part of the Large Tokamak collaboration. These talks are summarized into four sessions: tritium issues in large tokamaks, alpha particle simulation experiments, alpha particle theory, and alpha particle diagnostics.

  8. TSC simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

    SciTech Connect

    Jardin, S.C.; Schmidt, J.A.

    1997-03-01

    The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrapeoff layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer significant advantages over the more conventional method of controlling this instability when applied in a reactor environment.

  9. Characteristics of solid-target charge-exchange analyzers for energetic ion diagnostics on tokamaks

    SciTech Connect

    Beiersdorfer, P.; Roquemore, A.L.; Kaita, R.

    1987-05-01

    Compact electrostatic charge-exchange analyzers have been constructed for installation in areas of high magnetic fields and restricted access near tokamak fusion devices. The analyzers employed carbon stripping foils, and have been calibrated for proton energies between 1 and 70 keV. They have been successfully used to study charge-exchange losses in auxiliary-heated tokamak plasmas.

  10. Measurement of the hydrogen recombination coefficient in the TEXT tokamak as a function of outgassing and power radiated during tokamak discharges

    SciTech Connect

    Langley, R.A.; Rowan, W.L.; Bravinec, R.V.; Nelin, K.

    1986-01-01

    The global recombination rate coefficient k/sub r/ for hydrogen has been measured in the TEXT tokamak vacuum for various surface conditions. An attempt was made to correlate the measured values of k/sub r/ with RGA data taken prior to each k/sub r/ measurement and with the power radiated during tokamak discharges produced after each k/sub r/ measurement. The results show that: k/sub r/ increases during a series of tokamak discharges, k/sub r/ is relatively insensitive to power radiated during tokamak discharges, k/sub r/ increases with the RGA measurements of mass 28 and 40 but not with mass 18. In addition, it was found that the RGA mass 18 (H/sub 2/O) signal decreased as glow discharge experiments with hydrogen were performed.

  11. Measurement of the hydrogen recombination coefficient in the TEXT tokamak as a function of outgassing and power radiated during tokamak discharges

    SciTech Connect

    Langley, R.A.; Rowan, W.L.; Bravenec, R.V.; Nelin, K.

    1986-10-01

    The global recombination rate coefficient k/sub r/ for hydrogen has been measured in the TEXT tokamak vacuum vessel for various surface conditions. An attempt was made to correlate the measured values of k/sub r/ with residual gas analyzer (RGA) data taken before each measurement of k/sub r/ and with the power radiated during tokamak discharges produced after each measurement of k/sub r/. The results show that k/sub r/ increases during a series of tokamak discharges, k/sub r/ is relatively insensitive to power radiated during tokamak discharges, and k/sub r/ increases with the RGA measurements of mass 28 and 40 but not with those of mass 18. In addition, it was found that the mass 18 (H/sub 2/O) signal decreases as glow discharge experiments with hydrogen were performed.

  12. Recent results from the DIII-D tokamak and implications for future devices

    SciTech Connect

    Luxon, J.L.

    1995-02-01

    Improvements to the DIII-D tokamak have led to significant new research results and enhanced performance. These results provide important inputs to the design of next generation divertor systems including the upgrade of the DIII-D divertor. The use of graphite for the plasma facing components and careful wall preparation has enabled the routine achievement of regimes of enhanced energy confinement. In elongated discharges, triangularity has been found to be important in attaining good discharge performance as measured by the product of the normalized plasma pressure and the energy confinement time, {beta}{tau}{sub E} This constrains the design of the divertor configuration (X-point location). Active pumping of the divertor region using an in-situ toroidal cryogenic pump has demonstrated control of the plasma density in H-mode discharges and allowed the dependence of confinement on plasma density and current to be separately determined. Helium removal from the plasma edge sufficient to achieve effective ash removal in reactor discharges has also been demonstrated using this pumping configuration. The reduction of the heat flux to the divertor plates has been demonstrated using two different techniques to increase the radiation in the boundary regions of the plasma and thus reduce the heat flux to the divertor plates; deuterium gas injection has been used to create a strongly radiating localized zone near the X-point, and impurity (neon) injection to enhance the radiation from the plasma mantle. Precise shaping of the plasma current profile has been found to be important in achieving enhanced tokamak performance. Transiently shaped current profiles have been used to demonstrate regimes of plasmas with high beta and good confinement. Control of the current profile also is important to sustaining the plasma in the Very High (VH)-mode of energy confinement.

  13. A charged fusion product diagnostic for a spherical tokamak

    NASA Astrophysics Data System (ADS)

    Perez, Ramona Leticia Valenzuela

    Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are

  14. Dynamic modeling of transport and positional control of tokamaks

    SciTech Connect

    Jardin, S.C.; Pomphrey, N.; DeLucia, J.

    1985-10-01

    We describe here a numerical model of a free boundary axisymmetric tokamak plasma and its associated control systems. The plasma is modeled with a hybrid method using two-dimensional velocity and flux functions with surface-averaged MHD equations describing the evolution of the adiabatic invariants. Equations are solved for the external circuits and for the effects of eddy currents in nearby conductors. The method is verified by application to several test problems and used to simulate the formation of a bean-shaped plasma in the PBX experiment.

  15. Eikonal waves, caustics and mode conversion in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Jaun, A.; Tracy, E. R.; Kaufman, A. N.

    2007-01-01

    Ray optics is used to model the propagation of short electromagnetic plasma waves in toroidal geometry. The new RAYCON code evolves each ray independently in phase space, together with its amplitude, phase and focusing tensor to describe the transport of power along the ray. Particular emphasis is laid on caustics and mode conversion layers, where a linear phenomenon splits a single incoming ray into two. The complete mode conversion algorithm is described and tested for the first time, using the two space dimensions that are relevant in a tokamak. Applications are shown, using a cold plasma model to account for mode conversion at the ion-hybrid resonance in the Joint European Torus.

  16. Radiation−condensation instability in tokamaks with mixed impurities

    SciTech Connect

    Morozov, D. Kh.; Pshenov, A. A.

    2015-08-15

    Radiation−condensation instability (RCI) is one of the possible mechanisms behind the formation of microfaceted asymmetric radiation from the edge (MARFE) of a tokamak. It has been previously shown by the authors that RCI in carbon-seeded plasma can be stabilized using neon injection. Recently, beryllium- and tungsten-seeded plasmas became a subject of great interest. Therefore, in the present paper, RCI stability analysis of the edge plasma seeded with beryllium, tungsten, nitrogen, and carbon is performed. The influence of neutral hydrogen fluxes from the wall on the marginal stability limit is studied as well.

  17. Anomalous transport and confinement scaling studies in tokamaks

    SciTech Connect

    Tang, W.M.; Cheng, C.Z.; Krommes, J.A.; Lee, W.W.; Oberman, C.R.; Perkins, F.W.; Rewoldt, G.; Smith, R.; Bonoli, P.; Coppi, B.

    1984-09-01

    In addressing the general issue of anomalous energy transport, this paper reports on results of theoretical studies concerning: (1) the characteristics and relative strength of the dominant kinetic instabilities likely to be present under realistic tokamak operating conditions; (2) specific nonlinear processes relevant to the saturation and transport properties of drift-type instabilities; (3) the construction of semiempirical models for electron thermal transport and the scaling trends inferred from them; and (4) the application of specific anomalous transport models to simulate recent large-scale confinement experiments (TFTR and JET) and current drive experiments.

  18. Microwave Imaging Reflectometry for the Visualization of Turbulence in Tokamaks

    SciTech Connect

    E. Mazzucato

    1999-12-16

    Understanding the mechanism of anomalous transport in magnetically confined plasmas requires the use of sophisticated diagnostic tools for the measurement of short-scale turbulent fluctuations. This paper describes the conceptual design of an experimental technique for the global visualization of density fluctuations in tokamaks. The proposed method is based on microwave reflectometry and consists in using a large diameter probing beam, collecting the reflected waves with a large aperture antenna, and forming an image of the reflecting plasma layer onto a 2D array of microwave receivers. Based on results from a series of numerical simulations, the theoretical feasibility conditions of the proposed method are discussed.

  19. Resonant magnetic perturbations and edge ergodization on the COMPASS tokamak

    SciTech Connect

    Cahyna, P.; Fuchs, V.; Krlin, L.

    2008-09-15

    Results of calculations of resonant magnetic perturbation spectra on the COMPASS tokamak are presented. Spectra of the perturbations are calculated from the vacuum field of the perturbation coils. Ergodization is then estimated by applying the criterion of overlap of the resulting islands and verified by field line tracing. Results show that for the chosen configuration of perturbation coils an ergodic layer appears in the pedestal region. The ability to form an ergodic layer is similar to the theoretical results for the ELM suppression experiment at DIII-D; thus, a comparable effect on ELMs can be expected.

  20. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  1. Formation and Stability of Impurity "snakes" in Tokamak Plasmas

    SciTech Connect

    L. Delgado-Aparicio, et. al.

    2013-01-28

    New observations of the formation and dynamics of long-lived impurity-induced helical "snake" modes in tokamak plasmas have recently been carried-out on Alcator C-Mod. The snakes form as an asymmetry in the impurity ion density that undergoes a seamless transition from a small helically displaced density to a large crescent-shaped helical structure inside q < 1, with a regularly sawtoothing core. The observations show that the conditions for the formation and persistence of a snake cannot be explained by plasma pressure alone. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature

  2. Impurity Line Emissions in VUV Region of TCABR Tokamak

    SciTech Connect

    Machida, M.; Daltrini, A. M.; Severo, J. H. F.; Nascimento, I. C.; Sanada, E. K.; Elizondo, J. I.; Kuznetsov, Y. K.; Galvao, R. M. O.

    2008-04-07

    Spectral emissions in the vacuum ultraviolet region from 50 nm to 320 nm have been measured on TCABR tokamak using an one meter VUV spectrometer and a MCP coupled to a CCD detector. Among the 98 emissions classified, 37 are from first order diffraction, 29 are from second order, 24 are from third order, 7 from fourth order, and one from fifth order diffraction. Main impurity lines are OII to OVII, CII to CIV, NIII to N V, FVII, besides working gas plasma hydrogen Lyman lines.

  3. The tokamak density limit: A thermo-resistive disruption mechanism

    NASA Astrophysics Data System (ADS)

    Gates, David

    2015-11-01

    A magnetic island growth mechanism based on radiative cooling of the internal island flux surfaces is shown to produce the correct physical scaling to explain one of the long standing mysteries of tokamak physics - the empirical Greenwald density limit. In this presentation we will review the phenomenology of the density limit and the correlation between the Greenwald limit and the onset threshold for radiation-driven tearing modes. The behavior of magnetic islands with a 3D electron temperature distribution which is consistent with a large ratio of radial to parallel heat conductivity - and a corresponding 3D resistivity profile - is examined for islands with near-zero net heating in the island interior. The effect of varying impurity mix on the local island onset threshold is consistent with the established experimental scalings for tokamaks at the density limit. A simple analytic theory is developed which reveals the effect of heating and cooling in the island interior as well as the effect of island asymmetry. It is shown that a new term accounting for the thermal effects of island asymmetry is a crucial addition to the Modified Rutherford Equation. The resultant model exhibits a robust onset of a rapidly growing tearing mode - consistent with the disruption mechanism observed at the density limit in tokamaks. Additionally, a fully non-linear 3D cylindrical calculation is performed that simulates the effect of net island heating / cooling by raising / suppressing the temperature in the core of the island. In both the analytic theory and the numerical simulation a sudden threshold for explosive growth is found to be due to the interaction between three distinct thermal non-linearities, which affect the island resistivity, thereby modifying the growth dynamics. Expanding on the model presented, we speculate that the mechanism described may be applicable to a much wider range of tokamak disruptions than just those near the Greenwald limit. This work is supported

  4. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    SciTech Connect

    Motloch, C.G.; Bonney, R.F.; Levine, J.D.; McKenzie-Carter, M.A.; Masson, L.S.; Commander, J.C.

    1995-04-01

    This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR.

  5. Halo current diagnostic system of experimental advanced superconducting tokamak

    SciTech Connect

    Chen, D. L.; Shen, B.; Sun, Y.; Qian, J. P. Wang, Y.; Xiao, B. J.; Granetz, R. S.

    2015-10-15

    The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

  6. Turbulence studies in Tokamak boundary plasmas with realistic divertor geometry

    SciTech Connect

    Xu, X.Q.

    1998-10-14

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT [1] and the linearized shooting code BAL[2] to study turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant, resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters.

  7. Experimental study of the principles governing tokamak transport

    NASA Astrophysics Data System (ADS)

    Wagner, F.; Gruber, O.; Lackner, K.; Murmann, H. D.; Speth, E.; Becker, G.; Bosch, H. S.; Brocken, H.; Cattanei, G.; Dorst, D.; Eberhagen, A.; Elsner, A.; Erckmann, V.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Grieger, G.; Grigull, P.; Haas, G.; Hacker, H.; Hartfuss, H. J.; Jäckel, H.; Jaenicke, R.; Janeschitz, G.; Junker, J.; Karger, F.; Kasparek, W.; Keilhacker, M.; Kick, M.; Klüber, O.; Kornherr, M.; Kroiss, H.; Kuehner, M.; Lenoci, M.; Lisitano, G.; Maassberg, M.; Mahn, C.; Marlier, S.; Mayer, H. M.; McCormick, K.; Meisel, D.; Mertens, V.; Müller, E. R.; Müller, .; Müller, G.; Niedermeyer, H.; Ohlendorf, W.; Poschenrieder, W.; Rapp, H.; Rau, F.; Renner, H.; Riedler, H.; Ringler, H.; Sardei, F.; Schüller, P. G.; Schwörer, K.; Siller, G.; Söldner, F.; Steuer, K.-H.; Thumm, M.; Tutter, M.; Vollmer, O.; Weller, A.; Wilhelm, R.; Wobig, H.; Würsching, E.; Zippe, M.

    1986-05-01

    Both in ohmically and beam-heated L-mode discharges of ASDEX, the electron-temperature (Te) profile shape can be varied over a wide range by the choice of the safety factor qa. The power-deposition profile, on the contrary, has no effect on the Te-profile shape. In current-free WVII-A stellarator plasmas, no such invariance property is found. An independent constraint seems to fix the current distribution j(r) of the tokamak, which defines the conditions of electron heat transport.

  8. Tokamak transmutation of (nuclear) waste (TTW): Parametric studies

    SciTech Connect

    Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.

    1994-06-01

    Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses.

  9. Measuring the escaping beam ions from a tokamak plasma

    SciTech Connect

    Buchenauer, D.; Heidbrink, W.W.; Roquemore, L.; McGuire, K.

    1987-12-01

    A new technique using a silicon surface barrier (SSB) diode has been developed for measuring the escaping fast ion flux from a tokamak plasma. Calibration of the detector with an ion beam showed that at a fixed energy the diode's output current varied linearly with the incident deuteron flux. The diode was mounted inside the PDX vacuum vessel with collimating apertures designed to admit the spiraling orbits of 50-keV deuterons expelled from the plasma by MHD instabilities. Results from PDX indicated that relative measurements of the escaping fast ion flux due to several plasma instabilities could be made.

  10. GAM observation in the TUMAN-3M tokamak

    NASA Astrophysics Data System (ADS)

    Bulanin, V. V.; Askinazi, L. G.; Belokurov, A. A.; Kornev, V. A.; Lebedev, V.; Petrov, A. V.; Tukachinsky, A. S.; Vildjunas, M. I.; Wagner, F.; Yashin, A. Yu

    2016-04-01

    Results of an experimental study of geodesic acoustic modes (GAM) in the TUMAN-3M tokamak are reported. With Doppler backscattering (DBS) the basic properties of the GAM such as frequency, conditions for the GAM existence and the GAM radial location have been identified. The two-frequency Doppler reflectometer system was employed to reveal an interplay between low frequency sheared poloidal rotation, ambient turbulence level and the GAM intensity. Bicoherence analysis of the DBS data evidences the presence of a nonlinear interaction between the GAM and plasma turbulence.

  11. Nonlinear tearing instabilities in tokamaks with locally flattened current profiles

    SciTech Connect

    Reiman, A.H.

    1988-07-01

    Nonlinear tearing stability is evaluated for current profiles which are linearly stabilized by flattening the current in the neighborhood of the rational surface. When marginally stable to the linear instability, these profiles remain unstable in the presence of a small but finite island. The growth of the island saturated only when the island reaches the width it would have attained in the absence of flattening. Implications are discused for proposed methods of tearing mode stabilization and for theories of the tokamak sawtooth oscillation. 19 refs., 1 fig.

  12. Tritium Experience in Large Tokamaks: Application to ITER

    SciTech Connect

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R{ampersand}D avenues to address them. This paper summarizes the presentations, discussions, and recommendations.

  13. Halo current diagnostic system of experimental advanced superconducting tokamak.

    PubMed

    Chen, D L; Shen, B; Granetz, R S; Sun, Y; Qian, J P; Wang, Y; Xiao, B J

    2015-10-01

    The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

  14. Ambipolarity in a tokamak with magnetic field ripple

    NASA Astrophysics Data System (ADS)

    Hazeltine, R. D.

    2016-08-01

    In view of the recognized importance of electrostatic fields regarding turbulent transport, the radial electric field in a tokamak with magnetic field ripple is reconsidered. Terms in the ambipolarity condition involving the radial derivative of the field are derived from an extended drift-kinetic equation, including effects of second order in the gyroradius. Such terms are of interest in part because of their known importance in rotational relaxation equations for the axisymmetric case. The electric field is found to satisfy a nonlinear differential equation that is universal in a certain sense, and that implies spatial relaxation of the potential to its conventionally predicted value.

  15. Nonlinear saturation of ballooning modes in tokamaks and stellarators

    PubMed Central

    Bauer, F.; Garabedian, P.; Betancourt, O.

    1988-01-01

    The spectral code BETAS computes plasma equilibrium in a toroidal magnetic field B = [unk]s × [unk]Ψ with remarkable accuracy because the finite difference scheme employed in the radial direction allows for discontinuities of the flux function Ψ across the nested surfaces s = const. Instability of higher modes in stellarators like the Heliotron E can be detected in roughly an hour on the best supercomputers by calculating bifurcated equilibria that are defined over just one field period. The method has been validated by comparing results about nonlinear saturation of ballooning modes in tokamaks with numerical data from the PEST code. PMID:16593984

  16. Enhancement of the Bootstrap Current in a Tokamak Pedestal

    SciTech Connect

    Kagan, Grigory; Catto, Peter J.

    2010-07-23

    The strong radial electric field in a subsonic tokamak pedestal modifies the neoclassical ion parallel flow velocity, as well as the radial ion heat flux. Existing experimental evidence of the resulting alteration in the poloidal flow of a trace impurity is discussed. We then demonstrate that the modified parallel ion flow can noticeably enhance the pedestal bootstrap current when the background ions are in the banana regime. Only the coefficient of the ion temperature gradient drive term is affected. The revised expression for the pedestal bootstrap current is presented. The prescription for inserting the modification into any existing banana regime bootstrap current expression is given.

  17. Liquid nitrogen cooling considerations of the Compact Ignition Tokamak

    SciTech Connect

    Dabiri, A.E.

    1986-10-01

    A simple model was developed to estimate the cooldown time between pulses of toroidal field (TF) coils of the Compact Ignition Tokamak (CIT) using liquid nitrogen. Good agreement was obtained between the analysis results and those measured in the early fusion experimental devices. A cooldown time of about 1 h would reduce the TF coil temperature to about 80 K. An R and D experimental program is required to determine the actual cooldown time between pulses, an issue in the conceptual design of the CIT.

  18. Carborane films: Applications to first-wall problems in tokamaks

    SciTech Connect

    Doyle, B.L.; Walsh, D.S.; Wampler, W.R.; Hays, A.K. ); Dylla, H.F.; Manos, D.M.; Kilpatrick, S.J. . Plasma Physics Lab.)

    1990-01-01

    RF plasma-assisted CVD and sputter deposition of amorphous boron-carbon layers similar to those being used in the TFTR tokamak at the Princeton Plasma Physics Laboratory have been performed. The initial stoichiometry has been determined using Rutherford backscattering spectrometry and elastic recoil detection. Films have also been implanted with deuterium in order to determine H-isotope pumping capacity. These studies, in addition to characterizations made of layers collected on probes in TFTR, have been used to optimize the boronization parameters and to better understand the effects of boronization on TFTR. 10 refs., 3 figs.

  19. Ion plateau transport near the tokamak magnetic axis

    SciTech Connect

    Shaing, K.C.; Hazeltine, R.D.

    1998-04-01

    Conventional neoclassical transport theory does not pertain near the magnetic axis, where orbital variation of the minor radius and the poloidal field markedly change the nature of guiding-center trajectories. Instead of the conventional tokamak banana-shaped trajectories, near-axis orbits, called potato orbits, are radially wider and lead to distinctive kinetic considerations. Here it is shown that there is a plateau regime for the near-axis case; the corresponding potato-plateau ion thermal conductivity is computed. {copyright} {ital 1998 American Institute of Physics.}

  20. Poloidal flow damping with potato orbits in tokamaks

    SciTech Connect

    Shaing, K.C.

    2005-10-01

    The poloidal flow damping rate in the vicinity of the magnetic axis in tokamaks is calculated using the time-dependent plasma viscosity. It is found that the damping rate is of the order of {nu}{sub ii}/f{sub t}{sup 2}, where {nu}{sub ii} is the ion-ion collision frequency, and f{sub t} is the fraction of the trapped potatoes. The corresponding neoclassical polarization or inertia enhancement factor is [1+({sigma}{sub p}q{sup 2}/f{sub t})], where {sigma}{sub p} is a numerical number of the order of unity, and q is the safety factor.