Pre-Results of the Real-Time ODIN Validation on MARTe Using Plasma Linearized Model in FTU Tokamak

NASA Astrophysics Data System (ADS)

Sadeghi, Yahya; Boncagni, Luca

2012-06-01

MARTe is a modular framework for real-time control aspects. At present time there are several MARTe systems under development at Frascati Tokamak Upgrade (Boncagni et al. in First steps in the FTU migration towards a modular and distributed real time control architecture based on MARTe and RTNet, 2010) such as the LH power percentage system, the gas puffing control system, the real-time ODIN plasma equilibrium reconstruction system and the position/current feedback control system (in a design phase) (Boncagni et al. in J Fusion Eng Design). The real-time reconstruction of magnetic flux in FTU tokamak is an important issue to estimate some quantities that can be use to control the plasma. This paper addresses the validation of real-time implementation of that task on MARTe.

Establishment and Assessment of Plasma Disruption and Warning Databases from EAST

NASA Astrophysics Data System (ADS)

Wang, Bo; Robert, Granetz; Xiao, Bingjia; Li, Jiangang; Yang, Fei; Li, Junjun; Chen, Dalong

2016-12-01

Disruption database and disruption warning database of the EAST tokamak had been established by a disruption research group. The disruption database, based on Structured Query Language (SQL), comprises 41 disruption parameters, which include current quench characteristics, EFIT equilibrium characteristics, kinetic parameters, halo currents, and vertical motion. Presently most disruption databases are based on plasma experiments of non-superconducting tokamak devices. The purposes of the EAST database are to find disruption characteristics and disruption statistics to the fully superconducting tokamak EAST, to elucidate the physics underlying tokamak disruptions, to explore the influence of disruption on superconducting magnets and to extrapolate toward future burning plasma devices. In order to quantitatively assess the usefulness of various plasma parameters for predicting disruptions, a similar SQL database to Alcator C-Mod for EAST has been created by compiling values for a number of proposed disruption-relevant parameters sampled from all plasma discharges in the 2015 campaign. The detailed statistic results and analysis of two databases on the EAST tokamak are presented. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2014GB103000)

Modeling of Resistive Wall Modes in Tokamak and Reversed Field Pinch Configurations of KTX

NASA Astrophysics Data System (ADS)

Han, Rui; Zhu, Ping; Bai, Wei; Lan, Tao; Liu, Wandong

2016-10-01

Resistive wall mode is believed to be one of the leading causes for macroscopic degradation of plasma confinement in tokamaks and reversed field pinches (RFP). In this study, we evaluate the linear RWM instability of Keda Torus eXperiment (KTX) in both tokamak and RFP configurations. For the tokamak configuration, the extended MHD code NIMROD is employed for calculating the dependence of the RWM growth rate on the position and conductivity of the vacuum wall for a model tokamak equilibrium of KTX in the large aspect-ratio approximation. For the RFP configuration, the standard formulation of dispersion relation for RWM based on the MHD energy principle has been evaluated for a cylindrical α- Θ model of KTX plasma equilibrium, in an effort to investigate the effects of thin wall on the RWM in KTX. Full MHD calculations of RWM in the RFP configuration of KTX using the NIMROD code are also being developed. Supported by National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002, 2015GB101004, 2011GB106000, and 2011GB106003.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Skinner, C.H.; Brooks, J.N.

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 themore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stencel, J.R.; Finley, V.L.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory for CY90. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The PPPL has engaged in fusion energy research sincemore » 1951 and in 1990 had one of its two large tokamak devices in operation: namely, the Tokamak Fusion Test Reactor. The Princeton Beta Experiment-Modification is undergoing new modifications and upgrades for future operation. A new machine, the Burning Plasma Experiment -- formerly called the Compact Ignition Tokamak -- is under conceptual design, and it is awaiting the approval of its draft Environmental Assessment report by DOE Headquarters. This report is required under the National Environmental Policy Act. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. 59 refs., 39 figs., 45 tabs.« less

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seokho H; Berry, Jan

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclearmore » pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.« less

Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the control and data acquisition systems for MEDUSA-CR device which are based on National Instruments (NI) software (LabView) and hardware on loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the energy, vacuum, gas fueling, and security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Alternate fusion fuels workshop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

Phase space effects on fast ion distribution function modeling in tokamaks

NASA Astrophysics Data System (ADS)

Podestà, M.; Gorelenkova, M.; Fredrickson, E. D.; Gorelenkov, N. N.; White, R. B.

2016-05-01

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.

Phase space effects on fast ion distribution function modeling in tokamaks

DOE Data Explorer

White, R. B. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Podesta, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gorelenkova, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Fredrickson, E. D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gorelenkov, N. N. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.

Relaunch of the Interactive Plasma Physics Educational Experience (IPPEX)

NASA Astrophysics Data System (ADS)

Dominguez, A.; Rusaitis, L.; Zwicker, A.; Stotler, D. P.

2015-11-01

In the late 1990's PPPL's Science Education Department developed an innovative online site called the Interactive Plasma Physics Educational Experience (IPPEX). It featured (among other modules) two Java based applications which simulated tokamak physics: A steady state tokamak (SST) and a time dependent tokamak (TDT). The physics underlying the SST and the TDT are based on the ASPECT code which is a global power balance code developed to evaluate the performance of fusion reactor designs. We have relaunched the IPPEX site with updated modules and functionalities: The site itself is now dynamic on all platforms. The graphic design of the site has been modified to current standards. The virtual tokamak programming has been redone in Javascript, taking advantage of the speed and compactness of the code. The GUI of the tokamak has been completely redesigned, including more intuitive representations of changes in the plasma, e.g., particles moving along magnetic field lines. The use of GPU accelerated computation provides accurate and smooth visual representations of the plasma. We will present the current version of IPPEX as well near term plans of incorporating real time NSTX-U data into the simulation.

Fast Time Response Electromagnetic Disruption Mitigation Concept

DOE PAGES

Raman, R.; Jarboe, T.; Jernigan, Thomas C.; ...

2015-09-28

An important and urgent issue for ITER is predicting and controlling disruptions. Tokamaks and spherical tokamaks have the potential to disrupt. Methods to rapidly quench the discharge after an impending disruption is detected are essential to protect the vessel and internal components. The warning time for the onset of some disruptions in tokamaks could be <10 ms, which poses stringent requirements on the disruption mitigation system for reactor systems. In this proposed method, a cylindrical boron nitride projectile containing a radiative payload composed of boron, boron nitride, or beryllium particulate matter and weighing similar to 15 g is accelerated tomore » velocities on the order of 1 to 2 km/s in <2 ms in a linear rail gun accelerator. A partially fragmented capsule is then injected into the tokamak discharge in the 3- to 6-ms timescale, where the radiative payload is dispersed. The device referred to as an electromagnetic particle injector has the potential to meet the short warning timescales for which a reactor disruption mitigation system must be built. The system is fully electromagnetic, with no mechanical moving parts, which ensures high reliability after a period of long standby.« less

Understanding L-H transition in tokamak fusion plasmas

NASA Astrophysics Data System (ADS)

Xu, Guosheng; Wu, Xingquan

2017-03-01

This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.

Fusion energy division annual progress report, period ending December 31, 1980

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-11-01

The ORNL Program encompasses most aspects of magnetic fusion research including research on two magnetic confinement programs (tokamaks and ELMO bumpy tori); the development of the essential technologies for plasma heating, fueling, superconducting magnets, and materials; the development of diagnostics; the development of atomic physics and radiation effect data bases; the assessment of the environmental impact of magnetic fusion; the physics and engineering of present-generation devices; and the design of future devices. The integration of all of these activities into one program is a major factor in the success of each activity. An excellent example of this integration is themore » extremely successful application of neutral injection heating systems developed at ORNL to tokamaks both in the Fusion Energy Division and at Princeton Plasma Physics Laboratory (PPPL). The goal of the ORNL Fusion Program is to maintain this balance between plasma confinement, technology, and engineering activities.« less

SOVRaD - A Digest of Recent Soviet R and D Articles. Volume 2, Number 6, 1976

DTIC Science & Technology

1976-06-01

6 Laser- Powered Rocket Model 1 High- Power CO2 Laser Radiation Effect in SF6 1 Tests With 9-Beam Laser Fusion Systems 1 Focusing Optics For...Boundary Layer 6 Deformation Theory of Artif.cial Muscles . 6 Dolphin Swimming Stereophotogrammetry 7 Stable Spark Gap for High- Power Pulsers 7...8 Resume of Soviet Tokamak Program .............. 9 First Measurements of Tokamak-10 Plasma , . . 10 Electrochemical Power Generation 11

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tierney, Brian; Dart, Eli; Tierney, Brian

The Energy Sciences Network (ESnet) is the primary provider of network connectivity for the U.S. Department of Energy Office of Science, the single largest supporter of basic research in the physical sciences in the United States of America. In support of the Office of Science programs, ESnet regularly updates and refreshes its understanding of the networking requirements of the instruments, facilities, scientists, and science programs that it serves. This focus has helped ESnet to be a highly successful enabler of scientific discovery for over 20 years. In March 2008, ESnet and the Fusion Energy Sciences (FES) Program Office of themore » DOE Office of Science organized a workshop to characterize the networking requirements of the science programs funded by the FES Program Office. Most sites that conduct data-intensive activities (the Tokamaks at GA and MIT, the supercomputer centers at NERSC and ORNL) show a need for on the order of 10 Gbps of network bandwidth for FES-related work within 5 years. PPPL reported a need for 8 times that (80 Gbps) in that time frame. Estimates for the 5-10 year time period are up to 160 Mbps for large simulations. Bandwidth requirements for ITER range from 10 to 80 Gbps. In terms of science process and collaboration structure, it is clear that the proposed Fusion Simulation Project (FSP) has the potential to significantly impact the data movement patterns and therefore the network requirements for U.S. fusion science. As the FSP is defined over the next two years, these changes will become clearer. Also, there is a clear and present unmet need for better network connectivity between U.S. FES sites and two Asian fusion experiments--the EAST Tokamak in China and the KSTAR Tokamak in South Korea. In addition to achieving its goal of collecting and characterizing the network requirements of the science endeavors funded by the FES Program Office, the workshop emphasized that there is a need for research into better ways of conducting remote collaboration with the control room of a Tokamak running an experiment. This is especially important since the current plans for ITER assume that this problem will be solved.« less

Remote experimental site concept development

NASA Astrophysics Data System (ADS)

Casper, Thomas A.; Meyer, William; Butner, David

1995-01-01

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel area allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.

A Research Program of Spherical Tokamak in China

NASA Astrophysics Data System (ADS)

He, Ye-xi

2002-08-01

The mission of this program is to explore the spherical torus plasma with a SUNIST spherical tokamak. Main experiments in the start phase will be involved with breakdown and plasma current set-up with a mode of saving volt-second and without ohmic heating system, equilibrium and instability, current driving, heating and profile modification. The SUNIST is a university-scale conceptual spherical tokamak, with R = 0.3 m, A 1.3, Ip ~ 50 kA, BT < 0.15 T, and PRF = 100 kW. The only peculiarity of SUNIST is that there is a toroidal insulating break along the outer wall of vacuum vessel. The expected that advantages of this arrangement are helpful not only for saving flux swing, but also for having a deep understanding of what will influence the discharge startup and globe performances of plasma under different conditions of strong vessel eddy and ECR power assistance. Of course, the vessel structure of cross seal will be at a great risk of controlling vacuum quality, although we have achieved positive results on simulation test and vacuum vessel test.

Energetic-ion-driven global instabilities in stellarator/helical plasmas and comparison with tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toi, K.; Ogawa, K.; Isobe, M.

2011-01-01

Comprehensive understanding of energetic-ion-driven global instabilities such as Alfven eigenmodes (AEs) and their impact on energetic ions and bulk plasma is crucially important for tokamak and stellarator/helical plasmas and in the future for deuterium-tritium (DT) burning plasma experiments. Various types of global modes and their associated enhanced energetic ion transport are commonly observed in toroidal plasmas. Toroidicity-induced AEs and ellipticity-induced AEs, whose gaps are generated through poloidal mode coupling, are observed in both tokamak and stellarator/helical plasmas. Global AEs and reversed shear AEs, where toroidal couplings are not as dominant were also observed in those plasmas. Helicity induced AEs thatmore » exist only in 3D plasmas are observed in the large helical device (LHD) and Wendelstein 7 Advanced Stellarator plasmas. In addition, the geodesic acoustic mode that comes from plasma compressibility is destabilized by energetic ions in both tokamak and LHD plasmas. Nonlinear interaction of these modes and their influence on the confinement of the bulk plasma as well as energetic ions are observed in both plasmas. In this paper, the similarities and differences in these instabilities and their consequences for tokamak and stellarator/helical plasmas are summarized through comparison with the data sets obtained in LHD. In particular, this paper focuses on the differences caused by the rotational transform profile and the 2D or 3D geometrical structure of the plasma equilibrium. Important issues left for future study are listed.« less

The LTX- β Research Program

NASA Astrophysics Data System (ADS)

Majeski, R.; Bell, R. E.; Boyle, D. P.; Hughes, P. E.; Kaita, R.; Kozub, T.; Merino, E.; Zhang, X.; Biewer, T. M.; Canik, J. M.; Elliott, D. B.; Reinke, M. L.; Bialek, J.; Hansen, C.; Jarboe, T.; Kubota, S.; Rhodes, T.; Dorf, M. A.; Rognlien, T.; Scotti, F.; Soukhanovskii, V. A.; Koel, B. E.; Donovan, D.; Maan, A.

2017-10-01

LTX- β, the upgrade to the Lithium Tokamak Experiment, approximately doubles the toroidal field (to 3.4 kG) and plasma current (to 150 - 175 kA) of LTX. Neutral beam injection at 20 kV, 30 A will be added in February 2018, with systems provided by Tri-Alpha Energy. A 9.3 GHz, 100 kW, short-pulse (5-10 msec) source will be available in summer 2018 for electron Bernstein wave heating. New lithium evaporation sources will allow between-shots recoating of the walls. Upgrades to the diagnostic set are intended to strengthen the research program in the critical areas of equilibrium, core transport, scrape-off layer physics, and plasma-material interactions. The LTX- β research program will combine the capability for gradient-free temperature profiles, to stabilize ion and electron temperature gradient-driven modes, with approaches to stabilization of ∇n-driven modes, such as the trapped electron mode (TEM). Candidate stabilization mechanisms for the TEM include sheared flow stabilization, which can be tested on LTX- β. The goal will be to minimize anomalous transport in a low aspect ratio tokamak, which would lead to a very compact, tokamak-based fusion core. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

Pivotal issues on relativistic electrons in ITER

NASA Astrophysics Data System (ADS)

Boozer, Allen H.

2018-03-01

The transfer of the plasma current from thermal to relativistic electrons is a threat to ITER achieving its mission. This danger is significantly greater in the nuclear than in the non-nuclear phase of ITER operations. Two issues are pivotal. The first is the extent and duration of magnetic surface breaking in conjunction with the thermal quenches. The second is the exponential sensitivity of the current transfer to three quantities: (1) the poloidal flux change required to e-fold the number of relativistic electrons, (2) the time τa after the beginning of the thermal quench before the accelerating electric field exceeds the Connor-Hastie field for runaway, and (3) the duration of the period τ_op in which magnetic surfaces remain open. Adequate knowledge does not exist to devise a reliable strategy for the protection of ITER. Uncertainties are sufficiently large that a transfer of neither a negligible nor the full plasma current to relativistic electrons can be ruled out during the non-nuclear phase of ITER. Tritium decay can provide a sufficiently strong seed for a dangerous relativistic-electron current even if τa and τ_op are sufficiently long to avoid relativistic electrons during non-nuclear operations. The breakup of magnetic surfaces that is associated with thermal quenches occurs on a time scale associated with fast magnetic reconnection, which means reconnection at an Alfvénic rather than a resistive rate. Alfvénic reconnection is well beyond the capabilities of existing computational tools for tokamaks, but its effects can be studied using its property of conserving magnetic helicity. Although the dangers to ITER from relativistic electrons have been known for twenty years, the critical issues have not been defined with sufficient precision to formulate an effective research program. Studies are particularly needed on plasma behavior in existing tokamaks during thermal quenches, behavior which could be clarified using methods developed here.

CXSFIT Code Application to Process Charge-Exchange Recombination Spectroscopy Data at the T-10 Tokamak

NASA Astrophysics Data System (ADS)

Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.

2017-12-01

The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.

User's manual for COAST 4: a code for costing and sizing tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

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.

Cutoff Probe for Tokamak SOL Measurement

NASA Astrophysics Data System (ADS)

Na, Byung-Keun; You, Kwang-Ho; Kim, Dae-Woong; You, Shin-Jae; Kim, Jung-Hyung; Chang, Hong-Young

2013-09-01

Since a cutoff probe was developed, there have been a lot of improvements in methodology and analysis for low temperature plasmas. However, in order to apply the cutoff probe to the Tokamak scrape-off layer (SOL), three important issues should be solved - speed, thermal protection, and short-distance (a few mm) wave propagation in magnetized plasmas. In this presentation, the improvement of cutoff probe for Tokamak is shown. The above problems can be solved using the following methods: (a) the cutoff probe can be used with short impulse of a few nano-seconds for speed improvement. (b) Ceramic covers were used for thermal protection. (c) In magnetized plasmas, the cutoff peak can be analyzed using circuit modeling and CST Microwave studio. To verify the proposed methods, the cutoff probe was applied to a Helicon plasma, and the results were compared to laser Thomson scattering results. Based on the result in the Helicon plasma, the cutoff probe will be applied to far-SOL region at the KSTAR 2013 campaign, and SOL region at the later campaign.

Prospects for Attractive Fusion Power

NASA Astrophysics Data System (ADS)

Najmabadi, Farrokh

2006-10-01

During the past ten years, the ARIES Team, a national team involving universities, national laboratories, and industry, has studied a variety of magnetic fusion power plants (tokamaks, stellarators, ST, and RFP). In this paper, we present the top-level requirements and goals for commercial fusion power plants developed with consultation with US utilities and industry. We will review several ARIES designs and discuss the candidate options for physics operation regime as well engineering design of various components (e.g., choice of structural material, coolant, breeder). For each option, we will discuss (1) the potential to satisfy the requirements and goals, and (2) the critical R&D needs. In particular, we will discuss fusion R&D issues which are similar to those of advanced fission systems. For tokamaks, our results indicate that dramatic improvement over first-stability operation can be obtained through either utilization of high-field magnets (e.g., high-temperature superconductors) or operation in advanced-tokamak modes (e.g., reversed-shear). In particular, if full benefits of reversed-shear operation are realized, as is assumed in ARIES-AT, tokamak power plants will have a cost of electricity competitive with other sources of electricity. Emerging technologies such as advanced Baryon cycle, high-temperature superconductor, and advanced manufacturing techniques can improve the cost and attractiveness of fusion plants.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Voorhees, D.R.; Rossmassler, R.L.; Zimmer, G.

The tritium analytical system at TFTR is used to determine the purity of tritium bearing gas streams in order to provide inventory and accountability measurements. The system includes a quadrupole mass spectrometer (QMS) and beta scintillator originally configured at Monsanto Mound Research Laboratory. The system was commissioned and tested in 1992 and is used daily for analysis of calibration standards, incoming tritium shipments, gases evolved from uranium storage beds and effluent gases from the tokamak. The instruments are controlled by a personal computer with customized software written with a graphical programming system designed for data acquisition and control. A discussionmore » of the instrumentation, control systems, system parameters, procedural methods, algorithms, and operational issues will be presented. Measurements of gas holding tanks and tritiated water waste streams using ion chamber instrumentation are discussed elsewhere. 7 refs., 3 figs.« less

High Adjustable Shear Map for a Single-null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Whitten, Michaelangelo; Lam, Maria; Punjabi, Alkesh

1996-11-01

An explicit map that has an adjustable shear s is x1 = x0 - k y0 [ ( 1 - y0 ) ( 1 + s y0 ) + s x ^21 ] y1 = y0 + k x1 [ 1 + s ( x^21 + y^20 ) ] Tokamak shear corresponds to negative s. Thus we can construct maps for variable shear for a single-null divertor tokamak (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 3322, 69 (1992) ^, (Punjabi A, Verma A and Boozer A, J Plasma Phys), 52, 91 (1994). Here we present the results from an initial study of this map. This work is supported by US DOE OFES. Michelangelo Whitten is a HU CFRT Summer Fusion High School Workshop scholar from Bowie High School in El Paso, Texas. He is supported by NASA SHARP Plus Program.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, W., E-mail: woochanglee@unist.ac.kr; Lee, D. J.; Park, H. K.

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{sup −1}.more » The upper limit corresponds to the normalized wavenumber k{sub θ}ρ{sub e} 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.« less

Non-Solenoidal Startup Research Directions on the Pegasus Toroidal Experiment

NASA Astrophysics Data System (ADS)

Fonck, R. J.; Bongard, M. W.; Lewicki, B. T.; Reusch, J. A.; Winz, G. R.

2017-10-01

The Pegasus research program has been focused on developing a physical understanding and predictive models for non-solenoidal tokamak plasma startup using Local Helicity Injection (LHI). LHI employs strong localized electron currents injected along magnetic field lines in the plasma edge that relax through magnetic turbulence to form a tokamak-like plasma. Pending approval, the Pegasus program will address a broader, more comprehensive examination of non-solenoidal tokamak startup techniques. New capabilities may include: increasing the toroidal field to 0.6 T to support critical scaling tests to near-NSTX-U field levels; deploying internal plasma diagnostics; installing a coaxial helicity injection (CHI) capability in the upper divertor region; and deploying a modest (200-400 kW) electron cyclotron RF capability. These efforts will address scaling of relevant physics to higher BT, separate and comparative studies of helicity injection techniques, efficiency of handoff to consequent current sustainment techniques, and the use of ECH to synergistically improve the target plasma for consequent bootstrap and neutral beam current drive sustainment. This has an ultimate goal of validating techniques to produce a 1 MA target plasma in NSTX-U and beyond. Work supported by US DOE Grant DE-FG02-96ER54375.

Dynamic optimization of open-loop input signals for ramp-up current profiles in tokamak plasmas

NASA Astrophysics Data System (ADS)

Ren, Zhigang; Xu, Chao; Lin, Qun; Loxton, Ryan; Teo, Kok Lay

2016-03-01

Establishing a good current spatial profile in tokamak fusion reactors is crucial to effective steady-state operation. The evolution of the current spatial profile is related to the evolution of the poloidal magnetic flux, which can be modeled in the normalized cylindrical coordinates using a parabolic partial differential equation (PDE) called the magnetic diffusion equation. In this paper, we consider the dynamic optimization problem of attaining the best possible current spatial profile during the ramp-up phase of the tokamak. We first use the Galerkin method to obtain a finite-dimensional ordinary differential equation (ODE) model based on the original magnetic diffusion PDE. Then, we combine the control parameterization method with a novel time-scaling transformation to obtain an approximate optimal parameter selection problem, which can be solved using gradient-based optimization techniques such as sequential quadratic programming (SQP). This control parameterization approach involves approximating the tokamak input signals by piecewise-linear functions whose slopes and break-points are decision variables to be optimized. We show that the gradient of the objective function with respect to the decision variables can be computed by solving an auxiliary dynamic system governing the state sensitivity matrix. Finally, we conclude the paper with simulation results for an example problem based on experimental data from the DIII-D tokamak in San Diego, California.

Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

NASA Astrophysics Data System (ADS)

Snyder, Philip

2017-10-01

High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finley, V.L.; Wiezcorek, M.A.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY93. The report is prepared to provide the U.S. Department of Energy (DOE) and the public with information on the level of radioactive and non-radioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1993. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The Princeton Plasmamore » Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1993, PPPL had both of its two large tokamak devices in operation; the Tokamak Fusion Test Reactor (TFTR) and the Princeton Beta Experiment-Modification (PBX-M). PBX-M completed its modifications and upgrades and resumed operation in November 1991. TFTR began the deuterium-tritium (D-T) experiments in December 1993 and set new records by producing over six million watts of energy. The engineering design phase of the Tokamak Physics Experiment (TPX), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL`s next machine, began in 1993 with the planned start up set for the year 2001. In 1993, the Environmental Assessment (EA) for the TFRR Shutdown and Removal (S&R) and TPX was prepared for submittal to the regulatory agencies.« less

Tokamak Operation with Safety Factor q 95 < 2 via Control of MHD Stability

DOE PAGES

Piovesan, Paolo; Hanson, Jeremy M.; Martin, Piero; ...

2014-07-24

Magnetic feedback control of the resistive-wall mode has enabled DIII-D to access stable operation at safety factor q95 = 1:9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at given toroidal magnetic eld. In tokamaks with a divertor, the limit occurs at q95 = 2, as con rmed in DIII-D. Since the energy con cement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a wholemore » new high-current regime not accessible before. This result brings signi cant possible bene ts in terms of fusion performance, but it also extends resistive wall mode physics and its control to conditions never explored before. In present experiments, q95 < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.« less

Merging-compression formation of high temperature tokamak plasma

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Sykes, A.

2017-07-01

Merging-compression is a solenoid-free plasma formation method used in spherical tokamaks (STs). Two plasma rings are formed and merged via magnetic reconnection into one plasma ring that then is radially compressed to form the ST configuration. Plasma currents of several hundred kA and plasma temperatures in the keV-range have been produced using this method, however until recently there was no full understanding of the merging-compression formation physics. In this paper we explain in detail, for the first time, all stages of the merging-compression plasma formation. This method will be used to create ST plasmas in the compact (R ~ 0.4-0.6 m) high field, high current (3 T/2 MA) ST40 tokamak. Moderate extrapolation from the available experimental data suggests the possibility of achieving plasma current ~2 MA, and 10 keV range temperatures at densities ~1-5  ×  1020 m-3, bringing ST40 plasmas into a burning plasma (alpha particle heating) relevant conditions directly from the plasma formation. Issues connected with this approach for ST40 and future ST reactors are discussed

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lazarus, E; Peng, Yueng Kay Martin

Oak Ridge National Laboratory (ORNL) proposes to build the Spherical Torus Experiment (STX), a very low aspect ratio toroidal confinement device. This proposal concentrates on tokamak operation of the experiment; however, it can in principle be operated as a pinch or reversed-field pinch as well. As a tokamak, the spherical torus confines a plasma that is characterized by high toroidal beta, low poloidal beta, large natural elongation, high plasma current for a given edge q, and strong paramagnetism. These features combine to offer the possibility of a compact, low-field fusion device. The figure below shows that when compared to amore » conventional tokamak the spherical torus represents a major change in geometry. The primary goals of the experiment will be to demonstrate a capability for high beta (20%) in the first stability regime, to extend our knowledge of tokamak confinement scaling, and to test oscillating-field current drive. The experiment will operate in the high-beta, collisionless regime, which is achieved in STX at low temperatures because of the geometry. At a minimum, operation of STX will help to resolve fundamental questions regarding the scaling of beta and confinement in tokamaks. Complete success in this program would have a significant impact on toroidal fusion research in that it would demonstrate solutions to the problems of beta and steady-state operation in the tokamak. The proposed device has a major radius of 0.45 m, a toroidai field of 0.5 T, a plasma current of 900 kA, and heating by neutral beam injection. We estimate 30 months for design, construction, and assembly. The budget estimate, including contingency and escalation, is $6.8 million.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neilson, G. H.; Heitzenroeder, P.; Lyon, J.

Stellarators use 3D plasma and magnetic field shaping to produce a steady-state disruption-free magnetic confinement configuration. Compact stellarators have additional attractive properties — quasi-symmetric magnetic fields and low aspect ratio. The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a lowripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabricationmore » of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start.« less

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.

Synchrotron radiation intensity and energy of runaway electrons in EAST tokamak

NASA Astrophysics Data System (ADS)

Zhang, YK; Zhou, RJ; Hu, LQ; Chen, MW; Chao, Y.; EAST team

2018-05-01

Not Available Project supported by the National Natural Science Foundation of China (Grant Nos. 11775263 and 11405219), the JSPS-NRF-NSFC A3 Foresight Program in the Field of Plasma Physics, China (Grant No. 11261140328), and the National Magnetic Confnement Fusion Science Program of China (Grant No. 2015GB102004).

Effects of Equilibrium Toroidal Flow on Locked Mode and Plasma Response in a Tokamak

NASA Astrophysics Data System (ADS)

Zhu, Ping; Huang, Wenlong; Yan, Xingting

2016-10-01

It is widely believed that plasma flow plays significant roles in regulating the processes of mode locking and plasma response in a tokamak in presence of external resonant magnetic perturbations (RMPs). Recently a common analytic relation for both locked mode and plasma response has been developed based on the steady-state solution to the coupled dynamic system of magnetic island evolution and torque balance. The analytic relation predicts the size of the magnetic island of a locked mode or a static nonlinear plasma response for a given RMP amplitude, and rigorously proves a screening effect of the equilibrium toroidal flow. To test the theory, we solve for the locked mode and the nonlinear plasma response in presence of RMP for a circular-shaped limiter tokamak equilibrium with constant toroidal flow, using the initial-value, full MHD simulation code NIMROD. The comparison between the simulation results and the theory prediction, in terms of the quantitative screening effects of equilibrium toroidal flow, will be reported and discussed. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Heat Transfer Issues in Finite Element Analysis of Bounding Accidents in PPCS Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pampin, R.; Karditsas, P.J.

2005-05-15

Modelling of temperature excursions in structures of conceptual power plants during hypothetical worst-case accidents has been performed within the European Power Plant Conceptual Study (PPCS). A new, 3D finite elements (FE) based tool, coupling the different calculations to the same tokamak geometry, has been extensively used to conduct the neutron transport, activation and thermal analyses for all PPCS plant models. During a total loss of cooling, the usual assumption for the bounding accident, passive removal of the decay heat from activated materials depends on conduction and radiation heat exchange between components. This paper presents and discusses results obtained during themore » PPCS bounding accident thermal analyses, examining the following issues: (a) radiation heat exchange between the inner surfaces of the tokamak, (b) the presence of air within the cryostat volume, and the heat flow arising from the circulation pattern provided by temperature differences between various parts, and (c) the thermal conductivity of pebble beds, and its degradation due to exposure to neutron irradiation, affecting the heat transfer capability and thermal response of a blanket based on these components.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koniges, A.E.; Craddock, G.G.; Schnack, D.D.

The purpose of the workshop was to assemble workers, both within and outside of the fusion-related computations areas, for discussion regarding the issues of dynamically adaptive gridding. There were three invited talks related to adaptive gridding application experiences in various related fields of computational fluid dynamics (CFD), and nine short talks reporting on the progress of adaptive techniques in the specific areas of scrape-off-layer (SOL) modeling and magnetohydrodynamic (MHD) stability. Adaptive mesh methods have been successful in a number of diverse fields of CFD for over a decade. The method involves dynamic refinement of computed field profiles in a waymore » that disperses uniformly the numerical errors associated with discrete approximations. Because the process optimizes computational effort, adaptive mesh methods can be used to study otherwise the intractable physical problems that involve complex boundary shapes or multiple spatial/temporal scales. Recent results indicate that these adaptive techniques will be required for tokamak fluid-based simulations involving the diverted tokamak SOL modeling and MHD simulations problems related to the highest priority ITER relevant issues.Individual papers are indexed separately on the energy data bases.« less

Improving spatial and spectral resolution of TCV Thomson scattering

NASA Astrophysics Data System (ADS)

Hawke, J.; Andrebe, Y.; Bertizzolo, R.; Blanchard, P.; Chavan, R.; Decker, J.; Duval, B.; Lavanchy, P.; Llobet, X.; Marlétaz, B.; Marmillod, P.; Pochon, G.; Toussaint, M.

2017-12-01

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

Fusion programs in applied plasma physics

NASA Astrophysics Data System (ADS)

1992-07-01

The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (1) Applied Plasma Physics Theory Program, (2) Alpha Particle Diagnostic, (3) Edge and Current Density Diagnostic, and (4) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (RF). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described.

Current status of final design and R&D for ITER blanket shield blocks in Korea

NASA Astrophysics Data System (ADS)

Ha, M. S.; Kim, S. W.; Jung, H. C.; Hwang, H. S.; Heo, Y. G.; Kim, D. H.; Ahn, H. J.; Lee, H. G.; Jung, K. J.

2015-07-01

The main function of the ITER blanket shield block (SB) is to provide nuclear shielding and support the first wall (FW) panel. It needs to accommodate all the components located on the vacuum vessel (in particular the in-vessel coils, blanket manifolds and the diagnostics). The conceptual, preliminary and final design reviews have been completed in the framework of the Blanket Integrated Product Team. The Korean Domestic Agency has successfully completed not only the final design activities, including thermo-hydraulic and thermo-mechanical analyses for SBs #2, #6, #8 and #16, but also the SB full scale prototype (FSP) pre-qualification program prior to issuing of the procurement agreement. SBs #2 and #6 are located at the in-board region of the tokamak. The pressure drop was less than 0.3 MPa and fully satisfied the design criteria. The thermo-mechanical stresses were also allowable even though the peak stresses occurred at nearby radial slit end holes, and their fatigue lives were evaluated over many more than 30 000 cycles. SB #8 is one of the most difficult modules to design, since this module will endure severe thermal loading not only from nuclear heating but also from plasma heat flux at uncovered regions by the FW. In order to resolve this design issue, the neutral beam shine-through module concept was applied to the FW uncovered region and it has been successfully verified as a possible design solution. SB #16 is located at the out-board central region of the tokamak. This module is under much higher nuclear loading than other modules and is covered by an enhanced heat flux FW panel. In the early design stage, many cooling headers on the front region were inserted to mitigate peak stresses near the access hole and radial slit end hole. However, the cooling headers on the front region needed to be removed in order to reduce the risk from cover welding during manufacturing. A few cooling headers now remain after efforts through several iterations to remove them and to optimize the cooling channels. The SB #8 FSP was manufactured and tested in accordance with the pre-qualification program based on the preliminary design, and related R&D activities were implemented to resolve the fabrication issues. This paper provides the current status of the final design and relevant R&D activities of the blanket SB.

In-vessel tritium retention and removal in ITER

NASA Astrophysics Data System (ADS)

Federici, G.; Anderl, R. A.; Andrew, P.; Brooks, J. N.; Causey, R. A.; Coad, J. P.; Cowgill, D.; Doerner, R. P.; Haasz, A. A.; Janeschitz, G.; Jacob, W.; Longhurst, G. R.; Nygren, R.; Peacock, A.; Pick, M. A.; Philipps, V.; Roth, J.; Skinner, C. H.; Wampler, W. R.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER - and more generically in any other next-step experimental fusion facility fuelled with tritium - the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Formation of the internal transport barrier in KSTAR

NASA Astrophysics Data System (ADS)

Chung, J.; Kim, H. S.; Jeon, Y. M.; Kim, J.; Choi, M. J.; Ko, J.; Lee, K. D.; Lee, H. H.; Yi, S.; Kwon, J. M.; Hahn, S.-H.; Ko, W. H.; Lee, J. H.; Yoon, S. W.

2018-01-01

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles.

Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, Rajesh

1992-08-01

The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensionalmore » (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles.« less

Enhanced erosion of tungsten plasma-facing components subject to simultaneous heat pulses and deuterium plasma

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Doerner, R.; Tynan, G.

2009-04-01

When an ELM occurs in tokamaks, up to 30% of the pedestal energy can be deposited on the wall of the tokamak causing heating and material loss due to sublimation/evaporation and melt layer splashing of plasma-facing components (PFCs) and expansion of the ejected material into the plasma. A short-pulse laser system capable of reproducing the thermal load of an ELM heat pulse has been integrated into the existing PFC research program in PISCES, a laboratory facility capable of reproducing plasma-materials interactions expected during normal operation of large tokamaks. An Nd:YAG laser capable of delivering up to 1 J of energy over a 7 ns pulsewidth is used for the experiments. Laser heat pulse only, H +/D + plasma only, and laser plus plasma experiments were conducted and initial results indicate enhanced erosion of tungsten exposed to simultaneous plasma and heat pulses, as compared to exposure to separate plasma-only or heat pulse-only conditions.

Optimal control of a coupled partial and ordinary differential equations system for the assimilation of polarimetry Stokes vector measurements in tokamak free-boundary equilibrium reconstruction with application to ITER

NASA Astrophysics Data System (ADS)

Faugeras, Blaise; Blum, Jacques; Heumann, Holger; Boulbe, Cédric

2017-08-01

The modelization of polarimetry Faraday rotation measurements commonly used in tokamak plasma equilibrium reconstruction codes is an approximation to the Stokes model. This approximation is not valid for the foreseen ITER scenarios where high current and electron density plasma regimes are expected. In this work a method enabling the consistent resolution of the inverse equilibrium reconstruction problem in the framework of non-linear free-boundary equilibrium coupled to the Stokes model equation for polarimetry is provided. Using optimal control theory we derive the optimality system for this inverse problem. A sequential quadratic programming (SQP) method is proposed for its numerical resolution. Numerical experiments with noisy synthetic measurements in the ITER tokamak configuration for two test cases, the second of which is an H-mode plasma, show that the method is efficient and that the accuracy of the identification of the unknown profile functions is improved compared to the use of classical Faraday measurements.

Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks

NASA Astrophysics Data System (ADS)

Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

2017-10-01

The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.

The QUASAR facility

NASA Astrophysics Data System (ADS)

Gates, David

2013-10-01

The QUAsi-Axisymmetric Research (QUASAR) stellarator is a new facility which can solve two critical problems for fusion, disruptions and steady-state, and which provides new insights into the role of magnetic symmetry in plasma confinement. If constructed it will be the only quasi-axisymmetric stellarator in the world. The innovative principle of quasi-axisymmetry (QA) will be used in QUASAR to study how ``tokamak-like'' systems can be made: 1) Disruption-free, 2) Steady-state with low recirculating power, while preserving or improving upon features of axisymmetric tokamaks, such as 1) Stable at high pressure simultaneous with 2) High confinement (similar to tokamaks), and 3) Scalable to a compact reactor Stellarator research is critical to fusion research in order to establish the physics basis for a magnetic confinement device that can operate efficiently in steady-state, without disruptions at reactor-relevant parameters. The two large stellarator experiments - LHD in Japan and W7-X under construction in Germany are pioneering facilities capable of developing 3D physics understanding at large scale and for very long pulses. The QUASAR design is unique in being QA and optimized for confinement, stability, and moderate aspect ratio (4.5). It projects to a reactor with a major radius of ~8 m similar to advanced tokamak concepts. It is striking that (a) the EU DEMO is a pulsed (~2.5 hour) tokamak with major R ~ 9 m and (b) the ITER physics scenarios do not presume steady-state behavior. Accordingly, QUASAR fills a critical gap in the world stellarator program. This work supported by DoE Contract No. DEAC02-76CH03073.

Current Drive

DOE Office of Scientific and Technical Information (OSTI.GOV)

Faulconer, D.W

2004-03-15

Certain devices aimed at magnetic confinement of thermonuclear plasma rely on the steady flow of an electric current in the plasma. In view of the dominant place it occupies in both the world magnetic-confinement fusion effort and the author's own activity, the tokamak toroidal configuration is selected as prototype for discussing the question of how such a current can be maintained. Tokamaks require a stationary toroidal plasma current, this being traditionally provided by a pulsed magnetic induction which drives the plasma ring as the secondary of a transformer. Since this mechanism is essentially transient, and steady-state fusion reactor operation hasmore » manifold advantages, significant effort is now devoted to developing alternate steady-state means of generating toroidal current. These methods are classed under the global heading of 'noninductive current drive' or simply 'current drive', generally, though not exclusively, employing the injection of waves and/or toroidally directed particle beams. In what follows we highlight the physical mechanisms underlying surprisingly various approaches to driving current in a tokamak, downplaying a number of practical and technical issues. When a significant data base exists for a given method, its experimental current drive efficiency and future prospects are detailed.« less

Numerical modeling of lower hybrid current drive in fully non-inductive plasma start-up experiments on TST-2

NASA Astrophysics Data System (ADS)

Tsujii, N.; Takase, Y.; Ejiri, A.; Shinya, T.; Togashi, H.; Yajima, S.; Yamazaki, H.; Moeller, C. P.; Roidl, B.; Sonehara, M.; Takahashi, W.; Toida, K.; Yoshida, Y.

2017-12-01

Non-inductive plasma start-up is a critical issue for spherical tokamaks since there is not enough room to provide neutron shielding for the center solenoid. Start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Because of the low magnetic field of a spherical tokamak, the plasma density needs to be kept at a very low value during the plasma current ramp-up so that the plasma core remains accessible to the LH waves. However, we have found that higher density was required to sustain larger plasma current. The achievable plasma current was limited by the maximum operational toroidal field of TST-2. The existence of an optimum density for LH current drive and its toroidal field dependence is explained through a numerical simulation based on a ray tracing code and a Fokker-Planck solver. In order to access higher density at the same magnetic field, a top-launch antenna was recently installed in addition to the existing outboard-launch antenna. Increase in the density limit was observed when the power was launched from the top antenna, consistently with the numerical predictions.

Toroidal rotation induced by 4.6 GHz lower hybrid current drive on EAST tokamak

NASA Astrophysics Data System (ADS)

Yin, Xiang-Hui; Chen, Jun; Hu, Rui-Ji; Li, Ying-Ying; Wang, Fu-Di; Fu, Jia; Ding, Bo-Jiang; Wang, Mao; Liu, Fu-Kun; Zang, Qing; Shi, Yue-Jiang; Lyu, Bo; Wan, Bao-Nian; EAST Team

2017-10-01

Not Available Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2013GB112004 and 2015GB103002), the National Natural Science Foundation of China (Grant Nos. 11405212 and 11261140328), and the Major Program of Development Foundation of Hefei Center for Physical Science and Technology China (Grant No. 2016FXZY008).

Thermonuclear Power Engineering: 60 Years of Research. What Comes Next?

NASA Astrophysics Data System (ADS)

Strelkov, V. S.

2017-12-01

This paper summarizes results of more than half a century of research of high-temperature plasmas heated to a temperature of more than 100 million degrees (104 eV) and magnetically insulated from the walls. The energy of light-element fusion can be used for electric power generation or as a source of fissionable fuel production (development of a fusion neutron source—FNS). The main results of studies of tokamak plasmas which were obtained in the Soviet Union with the greatest degree of thermal plasma isolation among all other types of devices are presented. As a result, research programs of other countries were redirected to tokamaks. Later, on the basis of the analysis of numerous experiments, the international fusion community gradually came to an opinion that it is possible to build a tokamak (ITER) with Q > 1 (where Q is the ratio of the fusion power to the external power injected into the plasma). The ITER program objective is to achieve Q = 1-10 for a discharge time of up to 1000 s. The implementation of this goal does not solve the problem of a steadystate operation. The solution to this problem is a reliable first wall and current generation. This is a task of the next fusion power plant construction stage, called DEMO. Comparison of DEMO and FNS parameters shows that, at this development stage, the operating parameters and conditions of these devices are identical.

Overview of the EUROfusion Medium Size Tokamak scientific program

NASA Astrophysics Data System (ADS)

Bernert, Matthias; Bolzonella, Tommaso; Coda, Stefano; Hakola, Antti; Meyer, Hendrik; Eurofusion Mst1 Team; Tcv Team; Mast-U Team; ASDEX Upgrade Team

2017-10-01

Under the EUROfusion MST1 program, coordinated experiments are conducted at three European medium sized tokamaks (ASDEX Upgrade, TCV and MAST-U). It complements the JET program for preparing a safe and efficient operation for ITER and DEMO. Work under MST1 benefits from cross-machine comparisons but also makes use of the unique capabilities of each device. For the 2017/2018 campaign 25 topic areas were defined targeting three main objectives: 1) Development towards an edge and wall compatible H-mode scenario with small or no ELMs. 2) Investigation of disruptions in order to achieve better predictions and improve avoidance or mitigation schemes. 3) Exploring conventional and alternative divertor configurations for future high P/R scenarios. This contribution will give an overview of the work done under MST1 exemplified by the highlight results for each top objective from the last campaigns, such as evaluation of natural small ELM scenarios, runaway mitigation and control, assessment of detachment in alternative divertor configurations and highly radiative scenarios. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57, 102014''.

Tokamak power reactor ignition and time dependent fractional power operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« less

Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

2018-05-01

Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Christopher; Nygren, R. E.; Chrobak, C P.

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels ofmore » isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.« less

Closeout Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Kenneth W. Gentle

2004-11-29

We are conducting a collaborative research program on two tokamaks, HT-7 and EAST (Experimental Advanced Superconducting Tokamak, formerly HT-7U), with the Institute of Plasma Physics of the Chinese Academy of Sciences (CASIPP) located in Hefei, PRC. The work that we planned for this year included conducting transport experiments on HT-7, completing plans for expansion of the HT-7 diagnostic set, and reaching an agreement on how UT-FRC can best participate in experiments on HT-7U. These goals were accomplished as summarized in the next section. Note that the experimental portion of the work is still underway. The experimental campaign for HT-7 beganmore » just a few weeks before this report was compiled.« less

Bench Test of the Vibration Compensation Interferometer for EAST Tokamak

NASA Astrophysics Data System (ADS)

Li, Gongshun; Yang, Yao; Liu, Haiqing; Jie, Yinxian; Zou, Zhiyong; Wang, Zhengxing; Zeng, Long; Wei, Xuechao; Li, Weiming; Lan, Ting; Zhu, Xiang; Liu, Yukai; Gao, Xiang

2016-02-01

A visible laser-based vibration compensation interferometer has recently been designed for the EAST tokamak and the bench test has been finished. The system was optimized for its installation on EAST. The value of the final optical power before the detectors without plasma has been calculated from the component bench test result, which is quite close to the measured value. A nanometer level displacement (of the order of the laser's wavelength) has been clearly measured by a modulation of piezoelectric ceramic unit, proving the system's capability. supported by the National Magnetic Confinement Fusion Program of China (Nos. 2014GB106002, 2014GB106003, 2014GB106004) and National Natural Science Foundation of China (Nos. 11105184, 11375237, 11505238)

Plasma anisotropy and the radial particle flux in a rippled tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hazeltine, R. D.

We show that an often used relation between the radial particle flux and the divergence of the gyrotropic stress is an algebraic identity, unrelated to momentum conservation. Our calculation is completely general with regard to toroidal geometry and plasma collisionality. The result bears on the role of anisotropy in momentum relaxation and also clarifies certain methodological issues.

US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

McGrath, R.T.; Yamashina, T.

This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition.

EDITORIAL: Message from the Editor

NASA Astrophysics Data System (ADS)

Schüller, F. C.

2005-01-01

The group of 25 articles published in this special issue of Nuclear Fusion aims to monitor the progress made with experiments on fusion physics that have been conducted worldwide up to the end of 2004. These articles are based on overview reports from the various experimental teams presented at the Fusion Energy Conference (FEC 2004). This conference was organized by the IAEA together with the Portuguese host organization CFN-IST and was held in Vilamoura, Portugal, in early November 2004. The overviews presented at the conference have been rewritten and extended for the purpose of this special issue and submitted to the standard double-referee peer-review of Nuclear Fusion. Most teams have made use of this opportunity. Therefore this issue, which also includes four conference summaries, presents a reasonably complete picture of the progress made since FEC 2002 in Lyon. The articles are placed in the following sequence: Conference summaries Theory of magnetic confinement Experimental confinement, plasma-material interactions and innovative concepts Experiments on stability, energetic particles, waves and current drive Inertial confinement fusion Tokamaks Performance: JT-60U, JET, DIII-D, ASDEX-U, C-MOD Steady state/long pulse operation: Tore Supra, HT-7, TRIAM Spherical tokamaks: MAST, NSTX Tritium experiments: JET Diagnostics and heating methods: JET (diagnostics), T-10 (ECRH and diagnostics) and FTU (LHH + ECRH) New devices: HL-2A Small devices Alternative magnetic confinement concepts Stellarators: LHD, TJ-II Reversed field pinches: MST Inertial confinement Direct drive Heavy ion beam fusion Readers will also notice the supplementary issue of the journal (volume 45, issue 10A). This extra issue contains the 15-year overview report on progress in fusion research as written by the International Fusion Research Council (IFRC) under the editorial responsibility of the IFRC. Both issues together will give the interested reader a state-of-the-art picture of the progress in nuclear fusion research.

Two Strategic Decisions Facing Fusion

NASA Astrophysics Data System (ADS)

Baldwin, D. E.

1998-06-01

Two strategic decisions facing the U.S. fusion program are described. The first decision deals with the role and rationale of the tokamak within the U. S. fusion program, and it underlies the debate over our continuing role in the evolving ITER collaboration (mid-1998). The second decision concerns how to include Inertial Fusion Energy (IFE) as a viable part of the national effort to harness fusion energy.

The ITER project construction status

NASA Astrophysics Data System (ADS)

Motojima, O.

2015-10-01

The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

Localized Electron Heating by Strong Guide-Field Magnetic Reconnection

NASA Astrophysics Data System (ADS)

Guo, Xuehan; Sugawara, Takumichi; Inomoto, Michiaki; Yamasaki, Kotaro; Ono, Yasushi; UTST Team

2015-11-01

Localized electron heating of magnetic reconnection was studied under strong guide-field (typically Bt 15Bp) using two merging spherical tokamak plasmas in Univ. Tokyo Spherical Tokamak (UTST) experiment. Our new slide-type two-dimensional Thomson scattering system documented for the first time the electron heating localized around the X-point. The region of high electron temperature, which is perpendicular to the magnetic field, was found to have a round shape with radius of 2 [cm]. Also, it was localized around the X-point and does not agree with that of energy dissipation term Et .jt . When we include a guide-field effect term Bt / (Bp + αBt) for Et .jt where α =√{ (vin2 +vout2) /v∥2 } , the energy dissipation area becomes localized around the X-point, suggesting that the electrons are accelerated by the reconnection electric field parallel to the magnetic field and thermalized around the X-point. This work was supported by JSPS A3 Foresight Program ``Innovative Tokamak Plasma Startup and Current Drive in Spherical Torus,'' a Grant-in-Aid from the Japan Society for the Promotion of Science (JSPS) Fellows 15J03758.

Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

NASA Astrophysics Data System (ADS)

Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

2017-10-01

Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

Numerical Solution of the Electron Heat Transport Equation and Physics-Constrained Modeling of the Thermal Conductivity via Sequential Quadratic Programming Optimization in Nuclear Fusion Plasmas

NASA Astrophysics Data System (ADS)

Paloma, Cynthia S.

The plasma electron temperature (Te) plays a critical role in a tokamak nu- clear fusion reactor since temperatures on the order of 108K are required to achieve fusion conditions. Many plasma properties in a tokamak nuclear fusion reactor are modeled by partial differential equations (PDE's) because they depend not only on time but also on space. In particular, the dynamics of the electron temperature is governed by a PDE referred to as the Electron Heat Transport Equation (EHTE). In this work, a numerical method is developed to solve the EHTE based on a custom finite-difference technique. The solution of the EHTE is compared to temperature profiles obtained by using TRANSP, a sophisticated plasma transport code, for specific discharges from the DIII-D tokamak, located at the DIII-D National Fusion Facility in San Diego, CA. The thermal conductivity (also called thermal diffusivity) of the electrons (Xe) is a plasma parameter that plays a critical role in the EHTE since it indicates how the electron temperature diffusion varies across the minor effective radius of the tokamak. TRANSP approximates Xe through a curve-fitting technique to match experimentally measured electron temperature profiles. While complex physics-based model have been proposed for Xe, there is a lack of a simple mathematical model for the thermal diffusivity that could be used for control design. In this work, a model for Xe is proposed based on a scaling law involving key plasma variables such as the electron temperature (Te), the electron density (ne), and the safety factor (q). An optimization algorithm is developed based on the Sequential Quadratic Programming (SQP) technique to optimize the scaling factors appearing in the proposed model so that the predicted electron temperature and magnetic flux profiles match predefined target profiles in the best possible way. A simulation study summarizing the outcomes of the optimization procedure is presented to illustrate the potential of the proposed modeling method.

Design of tangential multi-energy SXR cameras for tokamak plasmas

NASA Astrophysics Data System (ADS)

Yamazaki, H.; Delgado-Aparicio, L. F.; Pablant, N.; Hill, K.; Bitter, M.; Takase, Y.; Ono, M.; Stratton, B.

2017-10-01

A new synthetic diagnostic capability has been built to study the response of tangential multi-energy soft x-ray pin-hole cameras for arbitrary plasma densities (ne , D), temperature (Te) and ion concentrations (nZ). For tokamaks and future facilities to operate safely in a high-pressure long-pulse discharge, it is imperative to address key issues associated with impurity sources, core transport and high-Z impurity accumulation. Multi-energy soft xray imaging provides a unique opportunity for measuring, simultaneously, a variety of important plasma properties (e.g. Te, nZ and ΔZeff). These systems are designed to sample the continuum- and line-emission from low- to high-Z impurities (e.g. C, O, Al, Si, Ar, Ca, Fe, Ni and Mo) in multiple energy-ranges. These x-ray cameras will be installed in the MST-RFP, as well as NSTX-U and DIII-D tokamaks, measuring the radial structure of the photon emissivity with a radial resolution below 1 cm at a 500 Hz frame rate and a photon-energy resolution of 500 eV. The layout and response expected for the new systems will be shown for different plasma conditions and impurity concentrations. The effect of toroidal rotation driving poloidal asymmetries in the core radiation is also addressed for the case of NSTX-U.

Fusion Studies in Japan

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi

2016-05-01

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

Tomographic capabilities of the new GEM based SXR diagnostic of WEST

NASA Astrophysics Data System (ADS)

Jardin, A.; Mazon, D.; O'Mullane, M.; Mlynar, J.; Loffelmann, V.; Imrisek, M.; Chernyshova, M.; Czarski, T.; Kasprowicz, G.; Wojenski, A.; Bourdelle, C.; Malard, P.

2016-07-01

The tokamak WEST (Tungsten Environment in Steady-State Tokamak) will start operating by the end of 2016 as a test bed for the ITER divertor components in long pulse operation. In this context, radiative cooling of heavy impurities like tungsten (W) in the Soft X-ray (SXR) range [0.1 keV; 20 keV] is a critical issue for the plasma core performances. Thus reliable tools are required to monitor the local impurity density and avoid W accumulation. The WEST SXR diagnostic will be equipped with two new GEM (Gas Electron Multiplier) based poloidal cameras allowing to perform 2D tomographic reconstructions in tunable energy bands. In this paper tomographic capabilities of the Minimum Fisher Information (MFI) algorithm developed for Tore Supra and upgraded for WEST are investigated, in particular through a set of emissivity phantoms and the standard WEST scenario including reconstruction errors, influence of noise as well as computational time.

Self-regulated oscillation of transport and topology of magnetic islands in toroidal plasmas

DOE PAGES

Ida, K.; Kobayashi, T.; Evans, T. E.; ...

2015-11-04

The coupling between the transport and magnetic topology is an important issue because the structure of magnetic islands, embedded in a toroidal equilibrium field, depends on the nature of the transport at the edge of the islands. Measurements of modulated heat pulse propagation in the DIII-D tokamak have revealed the existence of self-regulated oscillations in the radial energy transport into magnetic islands that are indicative of bifurcations in the island structure and transport near the q = 2 surface. Large amplitude heat pulses are seen in one state followed by small amplitude pulses later in the discharge resulting in amore » repeating cycle of island states. These two states are interpreted as a bifurcation of magnetic island with high and low heat pulse accessibility. In conclusion, this report describes the discovery of a bifurcation in the coupled dynamics between the transport and topology of magnetic islands in tokamak plasmas.« less

A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ren, J.; Zuo, G. Z.; Hu, J. S.

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 thinmore » 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.« less

Current drive by spheromak injection into a tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, M.R.; Bellan, P.M.

1990-04-30

We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma ({ital {bar n}}{sub 3} increases by a factor of 6) then becomes hollow, suggestive of an interchange instability.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Podestà, M., E-mail: mpodesta@pppl.gov; Gorelenkova, M.; Fredrickson, E. D.

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions.more » The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.« less

Development of FullWave : Hot Plasma RF Simulation Tool

NASA Astrophysics Data System (ADS)

Svidzinski, Vladimir; Kim, Jin-Soo; Spencer, J. Andrew; Zhao, Liangji; Galkin, Sergei

2017-10-01

Full wave simulation tool, modeling RF fields in hot inhomogeneous magnetized plasma, is being developed. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated in configuration space without limiting approximations by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. This approach allows for better resolution of plasma resonances, antenna structures and complex boundaries. The formulation of FullWave and preliminary results will be presented: construction of the finite differences for approximation of derivatives on adaptive cloud of computational points; model and results of nonlocal conductivity kernel calculation in tokamak geometry; results of 2-D full wave simulations in the cold plasma model in tokamak geometry using the formulated approach; results of self-consistent calculations of hot plasma dielectric response and RF fields in 1-D mirror magnetic field; preliminary results of self-consistent simulations of 2-D RF fields in tokamak using the calculated hot plasma conductivity kernel; development of iterative solver for wave equations. Work is supported by the U.S. DOE SBIR program.

Overview of Current Drive Experiment-Upgrade (CDX-U)

NASA Astrophysics Data System (ADS)

Hwang, Y. S.; Choe, W.; Stutman, D.; Lo, E.; Menard, J.; Ono, M.; Jones, T. G.; Armstrong, R.

1996-11-01

The CDX-U tokamak is a spherical tokamak (ST) facility with R ≈ 32 cm, R/a >= 1.4, and B_TF ≈ 1 kG. With an OH power supply of 60 mV-S capability, experiments were conducted with Ip up to ~ 100 kA and q(a) >= 3.5. The ST plasma performance has been studied along with various MHD-related activities. By appropriate discharge programing, it was possible to obtain MHD-quiescent discharges with a factor of 2 - 3 improvement in the electron energy confinement. Recently, the outer vacuum vessel was replaced with a toroidally continuous stainless steel chamber to accomodate the fast wave antenna. With the newly installed antenna, preliminary heating experiments using high harmonic fast waves have been pursued. The success of fast wave heating is a crucial element for achieving high beta plasmas in ST devices such as NSTX. Also, preliminary electron ripple injection (ERI) experiments were performed in CDX-U to examine the feasibility of this technique for improving ST tokamak confinement. To support the ST physics investigation, various novel plasma profile diagnostics such as the multi-pass Thomson scattering, soft x-ray tomography, and tangential-phase-contrast-imaging systems are under development on CDX-U.

Understanding and predicting the dynamics of tokamak discharges during startup and rampdown

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, G. L.; Politzer, P. A.; Humphreys, D. A.

Understanding the dynamics of plasma startup and termination is important for present tokamaks and for predictive modeling of future burning plasma devices such as ITER. We report on experiments in the DIII-D tokamak that explore the plasma startup and rampdown phases and on the benchmarking of transport models. Key issues have been examined such as plasma initiation and burnthrough with limited inductive voltage and achieving flattop and maximum burn within the technical limits of coil systems and their actuators while maintaining the desired q profile. Successful rampdown requires scenarios consistent with technical limits, including controlled H-L transitions, while avoiding verticalmore » instabilities, additional Ohmic transformer flux consumption, and density limit disruptions. Discharges were typically initiated with an inductive electric field typical of ITER, 0.3 V/m, most with second harmonic electron cyclotron assist. A fast framing camera was used during breakdown and burnthrough of low Z impurity charge states to study the formation physics. An improved 'large aperture' ITER startup scenario was developed, and aperture reduction in rampdown was found to be essential to avoid instabilities. Current evolution using neoclassical conductivity in the CORSICA code agrees with rampup experiments, but the prediction of the temperature and internal inductance evolution using the Coppi-Tang model for electron energy transport is not yet accurate enough to allow extrapolation to future devices.« less

Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-01-01

This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

ALCBEAM - Neutral beam formation and propagation code for beam-based plasma diagnostics

NASA Astrophysics Data System (ADS)

Bespamyatnov, I. O.; Rowan, W. L.; Liao, K. T.

2012-03-01

ALCBEAM is a new three-dimensional neutral beam formation and propagation code. It was developed to support the beam-based diagnostics installed on the Alcator C-Mod tokamak. The purpose of the code is to provide reliable estimates of the local beam equilibrium parameters: such as beam energy fractions, density profiles and excitation populations. The code effectively unifies the ion beam formation, extraction and neutralization processes with beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. This paper describes the physical processes interpreted and utilized by the code, along with exploited computational methods. The description is concluded by an example simulation of beam penetration into plasma of Alcator C-Mod. The code is successfully being used in Alcator C-Mod tokamak and expected to be valuable in the support of beam-based diagnostics in most other tokamak environments. Program summaryProgram title: ALCBEAM Catalogue identifier: AEKU_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKU_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 66 459 No. of bytes in distributed program, including test data, etc.: 7 841 051 Distribution format: tar.gz Programming language: IDL Computer: Workstation, PC Operating system: Linux RAM: 1 GB Classification: 19.2 Nature of problem: Neutral beams are commonly used to heat and/or diagnose high-temperature magnetically-confined laboratory plasmas. An accurate neutral beam characterization is required for beam-based measurements of plasma properties. Beam parameters such as density distribution, energy composition, and atomic excited populations of the beam atoms need to be known. Solution method: A neutral beam is initially formed as an ion beam which is extracted from the ion source by high voltage applied to the extraction and accelerating grids. The current distribution of a single beamlet emitted from a single pore of IOS depends on the shape of the plasma boundary in the emission region. Total beam extracted by IOS is calculated at every point of 3D mesh as sum of all contributions from each grid pore. The code effectively unifies the ion beam formation, extraction and neutralization processes with neutral beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. Running time: 10 min for a standard run.

In-situ Testing of the EHT High Gain and Frequency Ultra-Stable Integrators

NASA Astrophysics Data System (ADS)

Miller, Kenneth; Ziemba, Timothy; Prager, James; Slobodov, Ilia; Lotz, Dan

2014-10-01

Eagle Harbor Technologies (EHT) has developed a long-pulse integrator that exceeds the ITER specification for integration error and pulse duration. During the Phase I program, EHT improved the RPPL short-pulse integrators, added a fast digital reset, and demonstrated that the new integrators exceed the ITER integration error and pulse duration requirements. In Phase II, EHT developed Field Programmable Gate Array (FPGA) software that allows for integrator control and real-time signal digitization and processing. In the second year of Phase II, the EHT integrator will be tested at a validation platform experiment (HIT-SI) and tokamak (DIII-D). In the Phase IIB program, EHT will continue development of the EHT integrator to reduce overall cost per channel. EHT will test lower cost components, move to surface mount components, and add an onboard Field Programmable Gate Array and data acquisition to produce a stand-alone system with lower cost per channel and increased the channel density. EHT will test the Phase IIB integrator at a validation platform experiment (HIT-SI) and tokamak (DIII-D). Work supported by the DOE under Contract Number (DE-SC0006281).

INTRODUCTION: Award of the 2004 Hannes Alfvén Prize of the European Physical Society to J W Connor, R J Hastie and J B Taylor

NASA Astrophysics Data System (ADS)

Lister, Jo, Dr

2004-12-01

Jack Connor, Jim Hastie and Bryan Taylor The Hannes Alfvén Prize of the European Physical Society for Outstanding Contributions to Plasma Physics (2004) has been awarded to Jack Connor, Jim Hastie and Bryan Taylor `for their seminal contributions to a wide range of issues of fundamental importance to the success of magnetic confinement fusion, including: the development of gyro-kinetic theory; the prediction of the bootstrap current; dimensionless scaling laws; pressure-limiting instabilities, and micro-stability and transport theory'. Jack Connor, Jim Hastie and Bryan Taylor form one of the most successful teams of theoretical physicists in the history of magnetic confinement fusion. They have made important contributions individually, but their greatest discoveries have mostly been accomplished jointly, either in pairs or as a team involving all three. Their early work, in the 1960s, included the development of the gyro-kinetic theory for fine-scale plasma instabilities, which today forms the basis of the most advanced turbulence simulation codes in tokamak and stellarator research. The theoretical prediction of the bootstrap current, made in 1970-71 was not confirmed experimentally for over a decade but is now regarded as crucial to the success of the tokamak as a steady-state fusion power source. Their work on collisional transport also included the prediction of impurity ion accumulation, which is observed in internal transport barriers and is a key concern for long-pulse tokamak operation. The relativistic threshold for runaway electrons, identified in 1975, forms the basis of the most recent tokamak disruption mitigation schemes. In the late 1970s, the team developed the theory for ballooning instabilities, which provided an important ingredient in the `Troyon-Sykes' β-limit—an expression that is still used as a guide to the performance of tokamaks and in the design of ITER. Ballooning mode theory has also contributed to the understanding of instabilities in space plasmas such as magnetospheres and the solar corona. Finally, coming right up to date, the ballooning mode is thought to be a key ingredient in edge-localized modes (ELMs), which are a main issue for ITER, and ballooning stability is an important feature of modern stellarators. In the late 1970s and through the 1980s, the concept of dimensionless scaling laws was introduced and developed (following work by Kadomtsev), enabling scalings for transport coefficients to be derived without tackling all the details of the plasma turbulence. The same ideas are still used today to provide various constraints on confinement scaling laws, for example, on which the ITER design is largely based. The linear theory of toroidal drift waves was also developed by the team during this period, and into the 1990s. Key results on the role of shear damping in toroidal geometry, the identification of modes with extended radial correlation lengths, and the role of flow shear in reducing these correlation lengths (and hence transport) were deduced. All of these are key ideas that are often components in theoretical models for tokamak confinement and the generation of transport barriers. This laudation can only address a small number of the areas in which this formidable team of theoretical plasma physicists have made great contributions to our understanding of magnetically confined plasmas. It is appropriate and timely that their contributions are recognized as they approach the end of their careers.

Effects of Density and Impurity on Edge Localized Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Zhu, Ping

2017-10-01

Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Bartlit, John R.; Causey, Rion A.; Haines, John R.

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10(exp 19) ions/((sq cm)(s)) and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment.

High heat flux issues for plasma-facing components in fusion reactors

NASA Astrophysics Data System (ADS)

Watson, Robert D.

1993-02-01

Plasma facing components in tokamak fusion reactors are faced with a number of difficult high heat flux issues. These components include: first wall armor tiles, pumped limiters, diverter plates, rf antennae structure, and diagnostic probes. Peak heat fluxes are 15 - 30 MW/m2 for diverter plates, which will operate for 100 - 1000 seconds in future tokamaks. Disruption heat fluxes can approach 100,000 MW/m2 for 0.1 ms. Diverter plates are water-cooled heat sinks with armor tiles brazed on to the plasma facing side. Heat sink materials include OFHC, GlidcopTM, TZM, Mo-41Re, and niobium alloys. Armor tile materials include: carbon fiber composites, beryllium, silicon carbide, tungsten, and molybdenum. Tile thickness range from 2 - 10 mm, and heat sinks are 1 - 3 mm. A twisted tape insert is used to enhance heat transfer and increase the burnout safety margin from critical heat flux limits to 50 - 60 MW/m2 with water at 10 m/s and 4 MPa. Tests using rastered electron beams have shown thermal fatigue failures from cracks at the brazed interface between tiles and the heat sink after only 1000 cycles at 10 - 15 MW/m2. These fatigue lifetimes need to be increased an order of magnitude to meet future requirements. Other critical issues for plasma facing components include: surface erosion from sputtering and disruption erosion, eddy current forces and runaway electron impact from disruptions, neutron damage, tritium retention and release, remote maintenance of radioactive components, corrosion-erosion, and loss-of-coolant accidents.

Overview of the TCV tokamak program: scientific progress and facility upgrades

NASA Astrophysics Data System (ADS)

Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.; Bin, W.; Blanchard, P.; Blanken, T. C.; Boedo, J. A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F. H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I. T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B. P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J. P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.-Ph.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D. L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H. B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I. G.; Molina Cabrera, P. A.; Moret, J.-M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A. H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M. Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vijvers, W. A. J.; Vlahos, L.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.; the EUROfusion MST1 Team

2017-10-01

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

The generalized accessibility and spectral gap of lower hybrid waves in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, Hironori

1994-03-01

The generalized accessibility of lower hybrid waves, primarily in the current drive regime of tokamak plasmas, which may include shifting, either upward or downward, of the parallel refractive index (n{sub {parallel}}), is investigated, based upon a cold plasma dispersion relation and various geometrical constraint (G.C.) relations imposed on the behavior of n{sub {parallel}}. It is shown that n{sub {parallel}} upshifting can be bounded and insufficient to bridge a large spectral gap to cause wave damping, depending upon whether the G.C. relation allows the oblique resonance to occur. The traditional n{sub {parallel}} upshifting mechanism caused by the pitch angle of magneticmore » field lines is shown to lead to contradictions with experimental observations. An upshifting mechanism brought about by the density gradient along field lines is proposed, which is not inconsistent with experimental observations, and provides plausible explanations to some unresolved issues of lower hybrid wave theory, including generation of {open_quote}seed electrons.{close_quote}« less

Modeling of Steady-state Scenarios for the Fusion Nuclear Science Facility, Advanced Tokamak Approach

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Chan, V. S.; Prater, R.; Smith, S. P.; St. John, H. E.; Meneghini, O.

2013-10-01

A Fusion National Science Facility (FNSF) would complement ITER in addressing the community identified science and technology gaps to a commercially attractive DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. Steady-state plasma operation is highly desirable to address the requirements for fusion nuclear technology testing [1]. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues with a more compact device, and the direct relevance to an attractive target power plant. Key features of AT are fully noninductive current drive, strong plasma cross section shaping, internal profiles consistent with high bootstrap fraction, and operation at high beta, typically above the free boundary limit, βN > 3 . Work supported by GA IR&D funding, DE-FC02-04ER54698, and DE-FG02-95ER43309.

Modeling of the control of the driven current profile in ICRF MCCD on EAST plasma

NASA Astrophysics Data System (ADS)

Yin, L.; Yang, C.; Gong, X. Y.; Lu, X. Q.; Cao, J. J.; Wu, Z. Y.; Chen, Y.; Du, D.

2018-05-01

Control of the current profile is a crucial issue for improved confinement and the inhibition of instability in advanced tokamak operation. Using typical discharge data for the Experimental Advanced Superconducting Tokamak, numerical simulations of driven-current profile control in mode conversion current drive (MCCD) in the ion cyclotron range of frequencies were performed employing a full-wave method and Ehst-Karney efficiency formula. Results indicate that the driven current profile in MCCD can be effectively modified by shifting the mode conversion layer. The peak of the driven current can be located at an aimed position in the normalized minor radius range (-0.60 ≤r/a≤0) by changing the radiofrequency and the minority-ion concentration. The efficiency of the off-axis MCCD can reach 233 kA/MW through optimization, and the mode converted ion cyclotron wave plays an important role in such scenarios. The effects of electron temperature and plasma density on the driven current profile are also investigated.

Structured Cable for High-Current Coils of Tokamaks

NASA Astrophysics Data System (ADS)

Benson, Christopher; McIntyre, Peter; Sattarov, Akhdiyor; Mann, Thomas

2011-10-01

The 45 kA superconducting cable for the ITER central solenoid coil has yielded questionable results in two recent tests. In both cases the cable Tc increased after cycling only a fraction of the design life, indicating degradation due to fatigue and fracture among the superconducting strands. The Accelerator Research Lab at Texas A&M University is developing a design for a Nb3Sn structured cable suitable for such tokamak coils. The superconductor is configured in 6 sub-cables, and each subcable is supported within a channel of a central support structure within a high-strength armor sheath. The structured cable addresses two issues that are thought to compromise opposition at high current. The strands are supported without cross-overs (which produce stress concentration); and armor sheath and core structure bypass stress through the coil and among subcables so that the stress within each subcable is only what is produced directly upon it. Details of the design and plans for development will be presented.

Final Technical Report for the Center for Momentum Transport and Flow Organization (CMTFO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forest, Cary B.; Tynan, George R.

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physicsmore » issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tacholine physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.« less

Center for Momentum Transport and Flow Organization (CMTFO). Final technical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tynan, George R.; Diamond, P. H.; Ji, H.

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physicsmore » issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tachocline physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.« less

Protection of tokamak plasma facing components by a capillary porous system with lithium

NASA Astrophysics Data System (ADS)

Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

2015-08-01

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

ELM mitigation techniques

NASA Astrophysics Data System (ADS)

Evans, T. E.

2013-07-01

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ˜30 MJ which is expected to result in an energy density of 10-15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20-30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

Area of Stochastic Scrape-Off Layer for a Single-Null Divertor Tokamak Using Simple Map

NASA Astrophysics Data System (ADS)

Fisher, Tiffany; Verma, Arun; Punjabi, Alkesh

1996-11-01

The magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The Simple map is characterized by a single parameter k representing the toroidal asymmetry. The width of the stochastic scrape-off layer and its area varies with the map parameter k. We calculate the area of the stochastic scrape-off layer for different k's and obtain a parametric expression for the area in terms of k and y _LastGoodSurface(k). This work is supported by US DOE OFES. Tiffany Fisher is a HU CFRT Summer Fusion High school Workshop Scholar from New Bern High School in North Carolina. She is supported by NASA SHARP Plus Program.

Phase space effects on fast ion distribution function modeling in tokamaks

DOE PAGES

Podesta, M.; Gorelenkova, M.; Fredrickson, E. D.; ...

2016-04-14

Here, integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities,ad-hocmodels can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. Themore » kick model implemented in the tokamaktransport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.« less

Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, Anne

The tokamak is a type of toroidal device used to confine a fusion plasma using large magnetic fields. Tokamaks and stellarators the leading devices for confining plasmas for fusion, and the capability to predict performance in these magnetically confined plasmas is essential for developing a sustainable fusion energy source. The magnetic configuration of tokamaks and stellarators does not exist in Nature, yet, the fundamental processes governing transport in fusion plasmas are universal – turbulence and instabilities, driven by inhomogeneity and asymmetry in the plasma, conspire to transport heat and particles across magnetic field lines and can play critical roles inmore » impurity confinement and generation of intrinsic rotation. Turbulence exists in all plasmas, and in neutral fluids as well. The study of turbulence is essential to developing a fundamental understanding of the nature of the fourth state of matter, plasmas. Experimental studies of turbulence in tokamaks date back to early scattering observations from the late 1970s. Since that time, great advances in turbulence diagnostics have been made, all of which have significantly enhanced our knowledge and understanding of turbulence in tokamaks. Through comparisons with advanced gyrokinetic theory and turbulent-transport models a great deal of evidence exists to implicate turbulent-driven transport as an important mechanism determining transport in all channels: heat, particle and momentum However, prediction and control of turbulent-driven transport remains elusive. Key to development of predictive transport models for magnetically confined fusion plasmas is validation of the nonlinear gyrokinetic transport model, which describes transport due to turbulence. Validation of gyrokinetic codes must include detailed and quantitative comparisons with measured turbulence characteristics, in addition to comparisons with inferred transport levels and equilibrium profiles. For this reason, advanced plasma diagnostics for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for testing and validating predictive models for the transport of heat and particles in fusion plasmas due to turbulence. Once validated, the models are used to predict performance in ITER and other burning plasmas, such as the MIT ARC design. Most recently, data from the newly developed, so-called “CECE diagnostic” [Cima 1995, White 2008] and “nT phase angle measurements” [Haese 1999, White 2010] ]will be combined with data from density fluctuation diagnostics at ASDEX Upgrade to support a long-term program of physics research in turbulence and transport that will allow for more stringent testing and validation of gyrokinetic turbulent-transport codes. This work directly impacts the development of predictive transport models in the U.S. FES program, such as TGLF, developed by General Atomics, which are used to predict performance in ITER and other burning plasma devices as part of advancing the development of fusion energy sciences.« less

Advanced divertor configurations with large flux expansion

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; McLean, A.; Menard, J. E.; Paul, S. F.; Podesta, M.; Raman, R.; Ryutov, D. D.; Scotti, F.; Kaita, R.; Maingi, R.; Mueller, D. M.; Roquemore, A. L.; Reimerdes, H.; Canal, G. P.; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

2013-07-01

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3-7 MW/m2 to 0.5-1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L-H power threshold, enhanced stability of the peeling-ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2-3) Type I ELM frequency and slightly increased (20-30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma-material interface issues.

Enabling High Fidelity Measurements of Energy and Pitch Angle for Escaping Energetic Ions with a Fast Ion Loss Detector

NASA Astrophysics Data System (ADS)

Chaban, R.; Pace, D. C.; Marcy, G. R.; Taussig, D.

2016-10-01

Energetic ion losses must be minimized in burning plasmas to maintain fusion power, and existing tokamaks provide access to energetic ion parameter regimes that are relevant to burning machines. A new Fast Ion Loss Detector (FILD) probe on the DIII-D tokamak has been optimized to resolve beam ion losses across a range of 30 - 90 keV in energy and 40° to 80° in pitch angle, thereby providing valuable measurements during many different experiments. The FILD is a magnetic spectrometer; once inserted into the tokamak, the magnetic field allows energetic ions to pass through a collimating aperture and strike a scintillator plate that is imaged by a wide view camera and narrow view photomultiplier tubes (PMTs). The design involves calculating scintillator strike patterns while varying probe geometry. Calculated scintillator patterns are then used to design an optical system that allows adjustment of the focus regions for the 1 MS/s resolved PMTs. A synthetic diagnostic will be used to determine the energy and pitch angle resolution that can be attained in DIII-D experiments. Work supported in part by US DOE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Studies of runaway electrons via Cherenkov effect in tokamaks

NASA Astrophysics Data System (ADS)

Zebrowski, J.; Jakubowski, L.; Rabinski, M.; Sadowski, M. J.; Jakubowski, M. J.; Kwiatkowski, R.; Malinowski, K.; Mirowski, R.; Mlynar, J.; Ficker, O.; Weinzettl, V.; Causa, F.; COMPASS; FTU Teams

2018-01-01

The paper concerns measurements of runaway electrons (REs) which are generated during discharges in tokamaks. The control of REs is an important task in experimental studies within the ITER-physics program. The NCBJ team proposed to study REs by means of Cherenkov-type detectors several years ago. The Cherenkov radiation, induced by REs in appropriate radiators, makes it possible to identify fast electron beams and to determine their spatial- and temporal-characteristics. The results of recent experimental studies of REs, performed in two tokamaks - COMPASS in Prague and FTU in Frascati, are summarized and discussed in this paper. Examples of the electron-induced signals, as recorded at different experimental conditions and scenarios, are presented. Measurements performed with a three-channel Cherenkov-probe in COMPASS showed that the first fast electron peaks can be observed already during the current ramp-up phase. A strong dependence of RE-signals on the radial position of the Cherenkov probe was observed. The most distinct electron peaks were recorded during the plasma disruption. The Cherenkov signals confirmed the appearance of post-disruptive RE beams in circular-plasma discharges with massive Ar-puffing. During experiments at FTU a clear correlation between the Cherenkov detector signals and the rotation of magnetic islands was identified.

Overview of MST Research

NASA Astrophysics Data System (ADS)

Chapman, B. E.

2017-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with ohmic heating and minimal external magnetization, (2) predictive capability in toroidal confinement physics, and (3) basic plasma physics is summarized. Validation of key plasma models is a program priority, which is enhanced by programmable power supplies (PPS) to maximize inductive capability. The existing PPS enables access to very low plasma current, down to Ip =0.02 MA. This greatly expands the Lundquist number range S =104 -108 and allows nonlinear, 3D MHD computation using NIMROD and DEBS with dimensionless parameters that overlap those of MST plasmas. A new, second PPS will allow simultaneous PPS control of the Bp and Bt circuits. The PPS also enables MST tokamak operation, thus far focused on disruptions and RMP suppression of runaway electrons. Gyrokinetic modeling with GENE predicts unstable TEM in improved-confinement RFP plasmas. Measured fluctuations have TEM properties including a density-gradient threshold larger than for tokamak plasmas. Turbulent energization of an electron tail occurs during sawtooth reconnection. Probe measurements hint that drift waves are also excited via the turbulent cascade in standard RFP plasmas. Exploration of basic plasma science frontiers in MST RFP and tokamak plasmas is proposed as part of WiPPL, a basic science user facility. Work supported by USDoE.

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition.

NASA Astrophysics Data System (ADS)

Sharma, Uttam; Chauhan, Sachin S.; Sharma, Jayshree; Sanyasi, A. K.; Ghosh, J.; Choudhary, K. K.; Ghosh, S. K.

2016-10-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS.

Numerical studies of edge localized instabilities in tokamaks

NASA Astrophysics Data System (ADS)

Wilson, H. R.; Snyder, P. B.; Huysmans, G. T. A.; Miller, R. L.

2002-04-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code.

Stability at high performance in the MAST spherical tokamak

NASA Astrophysics Data System (ADS)

Buttery, R. J.; Akers, R.; Arends, E.; Conway, N. J.; Counsell, G. F.; Cunningham, G.; Gimblett, C. G.; Gryaznevich, M.; Hastie, R. J.; Hole, M. J.; Lehane, I.; Martin, R.; Patel, A.; Pinfold, T.; Sauter, O.; Taylor, D.; Turri, G.; Valovic, M.; Walsh, M. J.; Wilson, H. R.; MAST Team

2004-09-01

The development of reliable H-modes on MAST, together with advances in heating power and a range of high spatial resolution diagnostics, has provided a platform to enable MAST to address some of the most important issues of tokamak stability. In particular the high bgr potential of the spherical tokamak is highlighted with stable operation at bgrN ~ 5-6, bgrT ~ 16% and bgrp up to ~2. Magnetic diagnostic evaluation of the global bgr parameters is independently confirmed by kinetic profile data. Calculations indicate that the bgrN values are in the vicinity of no-wall stability limits. Studies of neoclassical tearing modes (NTMs) have been extended to explore their effects and develop avoidance strategies. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs—by avoiding large sawteeth a much higher bgrN value has been reached. The significance of NTMs is confirmed, with large islands observed using the 300 point Thomson scattering diagnostic, and locking of large n = 1 modes frequently leading to disruptions, which become more rapid at low q95. The role of error fields has been explored. H-mode plasmas are also limited by edge localized modes (ELMs), with confinement degraded as the ELM frequency rises. However, in contrast to the conventional tokamak, the ELMs in high performing regimes on MAST (HIPB98Y2 ~ 1) appear to be type III in nature. Modelling using the ELITE code, which incorporates finite n corrections, identifies instability to peeling modes, consistent with a type III interpretation. It also shows considerable scope to raise pressure gradients before ballooning type modes (perhaps associated with type I ELMs) occur. The calculations show that narrow pedestals can support much stronger pressure gradients than might be expected from simple n = infin ballooning calculations. Finally sawteeth are shown to degrade confinement by ~10-15% in particular cases examined. They are observed not to remove the q = 1 surface in the cases where snakes are present—various physics models of the sawteeth are now being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER.

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.

Advanced Tokamak Stability Theory

NASA Astrophysics Data System (ADS)

Zheng, Linjin

2015-03-01

The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

Plasma wall interaction, a key issue on the way to a steady state burning fusion device

NASA Astrophysics Data System (ADS)

Philipps, V.

2006-04-01

The International Tokamak Experimental Reactor (ITER), the first burning fusion plasma experiment based on the tokamak principle, is ready for construction. It is based on many years of fusion research resulting in a robust design in most of the areas. Present day fusion research concentrates on the remaining critical issues which are, to a large extent, connected with processes of plasma wall interaction. This is mainly due to extended duty cycle and the increase of the plasma stored energy in comparison with present-day machines. Critical topics are the lifetime of the plasma facing components (PFC) and the long-term tritium retention. These processes are controlled mainly by material erosion, both during steady state operation and transient power losses (disruptions and edge localized modes (ELMs)) and short- and long-range material migration and re-deposition. The extrapolation from present-day 'full carbon wall' devices suggests that the long-term tritium retention in a burning fusion device would be unacceptably high under these conditions allowing for only an unacceptable limited number of pulses in a D T mixture. As a consequence of this, research activities have been strengthened to understand in more detail the underlying processes of material erosion and re-deposition, to develop methods to remove retained tritium from the PFCs and remote areas of a fusion device and to explore these processes and the plasma performance in more detail with metallic PFC, such as beryllium (Be) and tungsten (W), which are foreseen for the ITER experiment. This paper outlines the main physical mechanisms leading to material erosion, migration and re-deposition and the associated fuel retention. It addresses the experimental database in these areas and describes the further research strategies that will be needed to tackle critical issues.

Overview of the recent DiMES and MiMES experiments in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Wong, C. P. C.; Litnovsky, A.; Wampler, W. R.; Boedo, J. A.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hollmann, E. M.; Jacob, W.; Krasheninnikov, S. I.; Krieger, K.; Lasnier, C. J.; Leonard, A. W.; McLean, A. G.; Marot, M.; Moyer, R. A.; Petrie, T. W.; Philipps, V.; Smirnov, R. D.; Stangeby, P. C.; Watkins, J. G.; West, W. P.; Yu, J. H.

2009-12-01

Divertor and midplane material evaluation systems (DiMES and MiMES) in the DIII-D tokamak are used to address a variety of plasma-material interaction (PMI) issues relevant to ITER. Among the topics studied are carbon erosion and re-deposition, hydrogenic retention in the gaps between plasma-facing components (PFCs), deterioration of diagnostic mirrors from carbon deposition and techniques to mitigate that deposition, and dynamics and transport of dust. An overview of the recent experimental results is presented.

Damping Rate Measurements of Medium n Alfv'en Eigenmodes in JET

NASA Astrophysics Data System (ADS)

Klein, Alexander; Testa, Duccio; Snipes, Joseph; Fasoli, Ambrogio; Carfantan, Hervé

2007-11-01

Alfv'en Eigenmodes (AE's) with mode numbers 5 < n < 20 are expected to be unstable in burning tokamaks and may lead to loss of fast particle confinement. The active MHD spectroscopy program at JET has already provided a wealth of information about low n (n <= 2) AE's in the past decade, but a recently installed array of four antennas is capable of driving higher mode numbered (n < 100, 30 < f < 350 kHz) perturbations. In the latest JET campaign, the damping rates for several types of AE's were measured parasitically in a wide range of tokamak scenarios. We review the active MHD diagnostic and present the first measurements of medium-n AE stability on JET, then describe future plans for the active MHD spectroscopy project. The data analysis involves a novel method for resolving multiple AE's that exist at identical frequencies, which uses techniques based on the SparSpec code.

New Technique of AC drive in Tokamak using Permanent Magnets

NASA Astrophysics Data System (ADS)

Matteucci, Jackson; Zolfaghari, Ali

2013-10-01

This study investigates a new technique of capturing the rotational energy of alternating permanent magnets in order to inductively drive an alternating current in tokamak devices. The use of rotational motion bypasses many of the pitfalls seen in typical inductive and non-inductive current drives. Three specific designs are presented and assessed in the following criteria: the profile of the current generated, the RMS loop voltage generated as compared to the RMS power required to maintain it, the system's feasibility from an engineering perspective. All of the analysis has been done under ideal E&M conditions using the Maxwell 3D program. Preliminary results indicate that it is possible to produce an over 99% purely toroidal current with a RMS d Φ/dt of over 150 Tm2/s, driven by 20 MW or less of rotational power. The proposed mechanism demonstrates several key advantages including an efficient mechanical drive system, the generation of pure toroidal currents, and the potential for a quasi-steady state fusion reactor. The following quantities are presented for various driving frequencies and magnet strengths: plasma current generated, loop voltage, torque and power required. This project has been supported by DOE Funding under the SULI program.

Neoclassical Toroidal Viscosity Torque Induced by Plasma Response in a Low- β Tokamak with Edge Pedestal

NASA Astrophysics Data System (ADS)

Yan, Xingting; Zhu, Ping; Sun, Youwen

2016-10-01

The characteristic profile and magnitude are predicted in theory for the neoclassical toroidal viscosity (NTV) torque induced by the plasma response to the resonant magnetic perturbation (RMP) in a tokamak with an edge pedestal, using the newly developed module coupling the NIMROD and the NTVTOK codes. For a low β equilibrium, the NTV torque is mainly induced by the dominant toroidal mode of plasma response. The NTV torque profile is radially localized and peaked, which is determined by profiles of both the equilibrium temperature and the plasma response fields. In general, the peak of NTV torque profile is found to trace the pedestal location. The magnitude of NTV torque is extremely sensitive to the β of pedestal top; for a given plasma response, the peak value of NTV torque can increase by three orders of magnitude, when the pedestal β increases by only one order of magnitude. This suggests a more significant role of NTV torque in higher plasma β regimes. Supported by the National Magnetic Confinement Fusion Program of China under Grant Nos. 2014GB124002 and 2015GB101004, and the 100 Talent Program of the Chinese Academy of Sciences.

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.

Foreword to Special Issue: Papers from the 54th Annual Meeting of the APS Division of Plasma Physics, Providence, Rhode Island, USA, 2012

NASA Astrophysics Data System (ADS)

Skiff, Fred; Davidson, Ronald C.

2013-05-01

Each year, the annual meeting of the APS Division of Plasma Physics (DPP) brings together a broad representation of the many active subfields of plasma physics and enjoys an audience that is equally diverse. The meeting was well attended and largely went as planned despite the interventions of hurricane Sandy which caused the city of Providence to shut-down during the first day of the conference. The meeting began on Monday morning with a review of the physics of cosmic rays, 2012 being the 100th year since their discovery, which illustrated the central importance of plasma physics to astrophysical problems. Subsequent reviews covered the importance of tokamak plasma boundaries, progress towards ignition on the National Ignition Facility (NIF), and magnetized plasma turbulence. The Maxwell prize address, by Professor Liu Chen, covered the field of nonlinear Alfvén wave physics. Tutorial lectures were presented on the verification of gyrokinetics, new capabilities in laboratory astrophysics, magnetic flux compression, and tokamak plasma start-up.

Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

NASA Astrophysics Data System (ADS)

Labate, C.; Di Gironimo, G.; Renno, F.

2015-09-01

Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

Investigation of Plasma Surface Interactions with the PISCES ELM Laser System

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Baldwin, M.; Hanna, J.; Doerner, R.; Lynch, T.; Palmer, T.; Tynan, G. R.

2007-11-01

When an ELM occurs in tokamaks, up to 30% of the pedestal energy can be deposited on the wall of the tokamak causing heating & material loss due to sublimation, evaporation and melt splashing of plasma facing components (PFCs) and expansion of the ejected material into the plasma. We have explored heat pulses using an electrical power circuit to draw electrons from the plasma to heat samples ohmically. This system is limited in power to ˜250kJ/m^2 at the minimum pulse width of 10ms and depletes the plasma column, complicating spectroscopy. We have completed calculations that indicate that a pulsed laser system can be used to simulate the heat pulse of ELMs. We are integrating laser systems into the existing PFC research program in PISCES, a laboratory facility capable of reproducing plasma-materials interactions expected during normal operation of large tokamaks. Two Nd:YAG lasers capable of delivering up to 50J of energy over various pulsewidths are used for the experiments. Laser heat pulse only, H+/D+ plasma only, and laser+plasma experiments were conducted and initial results indicate that metals behave very differently while exposed to plasma and simultaneous heat pulses. We will also discuss initial results for carbon PFCs and material transport into the plasma. Supported by US DoE grant DE-FG02-07ER-54912.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Last Good Surface

NASA Astrophysics Data System (ADS)

Phan, Huong; Ali, Halima; Punjabi, Alkesh

2000-10-01

The Simple Map^1 is a representation of the magnetic field inside a single-null divertor tokamak. It is given by the equations: X_n+1=X_n kYn (1-Y_n), Y_n+1= Y_n+kX_n+1. These equations mimic the motion of the magnetic field lines in a single-null divertor tokamak. The fixed stable point is (0,0) and the unstable fixed oint is (0,1). k is fixed at 0.60. In our work, the starting values of Y in the map is kept in the interval of 0 to 1, and the starting value of X is 0. Using the successive bifurcation method, we first run these equations for 10^6 iterations to find the approximate value of Y when chaos occurs. We examine the neighborhood of this Y value to find the exact value of Y for the last good surface. We call this value Y_lgs. We find Y_lgs to be 0.997135768 for k=0.60 and X=0. This work is supported by US DOE OFES. Ms. Huong Phan is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew P. Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

A generalized plasma dispersion function for electron damping in tokamak plasmas

DOE PAGES

Berry, L. A.; Jaeger, E. F.; Phillips, C. K.; ...

2016-10-14

Radio frequency wave propagation in finite temperature, magnetized plasmas exhibits a wide range of physics phenomena. The plasma response is nonlocal in space and time, and numerous modes are possible with the potential for mode conversions and transformations. Additionally, diffraction effects are important due to finite wavelength and finite-size wave launchers. Multidimensional simulations are required to describe these phenomena, but even with this complexity, the fundamental plasma response is assumed to be the uniform plasma response with the assumption that the local plasma current for a Fourier mode can be described by the Stix conductivity. But, for plasmas with non-uniformmore » magnetic fields, the wave vector itself is nonlocal. When resolved into components perpendicular (k ) and parallel (k ||) to the magnetic field, locality of the parallel component can easily be violated when the wavelength is large. The impact of this inconsistency is that estimates of the wave damping can be incorrect (typically low) due to unresolved resonances. For the case of ion cyclotron damping, this issue has already been addressed by including the effect of parallel magnetic field gradients. In this case, a modified plasma response (Z function) allows resonance broadening even when k || = 0, and this improves the convergence and accuracy of wave simulations. In our paper, we extend this formalism to include electron damping and find improved convergence and accuracy for parameters where electron damping is dominant, such as high harmonic fast wave heating in the NSTX-U tokamak, and helicon wave launch for off-axis current drive in the DIII-D tokamak.« less

Dust measurements in tokamaks (invited).

PubMed

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; Fenstermacher, M E; Groth, M; Lasnier, C J; McLean, A G; Stangeby, P C

2008-10-01

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 microm 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(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

Compact fusion energy based on the spherical tokamak

NASA Astrophysics Data System (ADS)

Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

2018-01-01

Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

HBT-EP Program: MHD Dynamics and Active Control through 3D Fields and Currents

NASA Astrophysics Data System (ADS)

Navratil, G. A.; Bialek, J.; Brooks, J. W.; Byrne, P. J.; Desanto, S.; Levesque, J. P.; Mauel, M. E.; Stewart, I. G.; Hansen, C. J.

2017-10-01

The HBT-EP active mode control research program aims to: (i) advance understanding of the effects of 3D shaping on advanced tokamak fusion performance, (ii) resolve important MHD issues associated with disruptions, and (iii) measure and mitigate the effects of 3D scrape-off layer (SOL) currents through active and passive control of the plasma edge and conducting boundary structures. Comparison of kink mode structure and RMP response in circular versus diverted plasmas shows good agreement with DCON modeling. SOL current measurements have been used to study SOL current dynamics and current-sharing with the vacuum vessel wall during kink-mode growth and disruptions. A multi-chord extreme UV/soft X-ray array is being installed to provide detailed internal mode structure information. Internal local electrodes were used to apply local bias voltage at two radial locations to study the effect of rotation profile on MHD mode rotation and stability and radial current flow through the SOL. A GPU-based low latency control system using 96 inputs and 64 outputs to apply magnetic perturbations for active control of kink modes is extended to directly control the SOL currents for kink-mode control. An extensive array of SOL current monitors and edge drive electrodes are being installed for pioneering studies of helical edge current control. Supported by U.S. DOE Grant DE-FG02-86ER53222.

3D toroidal physics: testing the boundaries of symmetry breaking

NASA Astrophysics Data System (ADS)

Spong, Don

2014-10-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to lead to a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D ELM-suppression fields to stellarators with more dominant 3D field structures. There is considerable interest in the development of unified physics models for the full range of 3D effects. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. Fortunately, significant progress is underway in theory, computation and plasma diagnostics on many issues such as magnetic surface quality, plasma screening vs. amplification of 3D perturbations, 3D transport, influence on edge pedestal structures, MHD stability effects, modification of fast ion-driven instabilities, prediction of energetic particle heat loads on plasma-facing materials, effects of 3D fields on turbulence, and magnetic coil design. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with future fusion reactors. The development of models to address 3D physics and progress in these areas will be described. This work is supported both by the US Department of Energy under Contract DE-AC05-00OR22725 with UT-Battelle, LLC and under the US DOE SciDAC GSEP Center.

Tokamak und Stellarator - zwei Wege zur Fusionsenergie: Fusionsforschung

NASA Astrophysics Data System (ADS)

Milch, Isabella

2006-07-01

Im Laufe der Fusionsforschung haben sich zwei Bautypen für ein zukünftiges Kraftwerk als besonders aussichtsreich erwiesen: Tokamak und Stellarator. Mit dem geplanten Tokamak-Experimentalreaktor ITER steht die internationale Fusionsforschung vor der Demonstration eines Energie liefernden Plasmas. Parallel soll die in Greifswald entstehende Forschungsanlage Wendelstein 7-X die Kraftwerkstauglichkeit des alternativen Bauprinzips der Stellaratoren zeigen.

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

NASA Astrophysics Data System (ADS)

Pace, D. C.; Austin, M. E.; Bardoczi, L.; Collins, C. S.; Crowley, B.; Davis, E.; Du, X.; Ferron, J.; Grierson, B. A.; Heidbrink, W. W.; Holcomb, C. T.; McKee, G. R.; Pawley, C.; Petty, C. C.; Podestà, M.; Rauch, J.; Scoville, J. T.; Spong, D. A.; Thome, K. E.; Van Zeeland, M. A.; Varela, J.; Victor, B.

2018-05-01

An engineering upgrade to the neutral beam system at the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic ( E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2 MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities and results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Austin, Max E.; Bardoczi, Laszlo; Collins, Cami S.

Here, an engineering upgrade to the neutral beam system at the DIII-D tokamak enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic (E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities andmore » results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.« less

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

DOE PAGES

Austin, Max E.; Bardoczi, Laszlo; Collins, Cami S.; ...

2018-04-20

Here, an engineering upgrade to the neutral beam system at the DIII-D tokamak enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic (E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities andmore » results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.« less

The Simple Map for a Single-null Divertor Tokamak: How to Look for Self-Similarity in Chaos

NASA Astrophysics Data System (ADS)

Nguyen, Christina; Ali, Halima; Punjabi, Alkesh

2000-10-01

The movement of magnetic field lines inside a single-null divertor tokamak can be described by the Simple Map^1. The Simple Map in the Poincaré Surface of Section is given by the equations: X_1=X_0-KY_0(1-Y_0) and Y_1=Y_0+KX_1. In these equations, K remains constant at 0.60. However, the values for X0 and Y0 are changed. These values are changed so that we can zoom into chaos. Chaos lies between the region (0,0.997) and (0,1). In chaos, there lies order. As we zoom into chaos, we again find chaos and order that looks like the original good surfaces and chaos. This phenomenon is called self-similarity. Self-similarity can occur for an infinite number of times if one magnifies into the chaotic region. For this work, we write a program in a computer language called Fortran 77 and Gnuplot. This work is supported by US DOE OFES. Ms. Christina Nguyen is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

The Helicity Injected Torus (HIT) Program

NASA Astrophysics Data System (ADS)

Jarboe, T. R.; Gu, P.; Hamp, W.; Izzo, V.; Jewell, P.; Liptac, J.; McCollam, K. J.; Nelson, B. A.; Raman, R.; Redd, A. J.; Shumlak, U.; Sieck, P. E.; Smith, R. J.; Jain, K. K.; Nagata, M.; Uyama, T.

2000-10-01

The purpose of the Helicity Injected Torus (HIT) program is to develop current drive techniques for low-aspect-ratio toroidal plasmas. The present HIT-II spherical tokamak experiment is capable of both Coaxial Helicity Injection (CHI) and transformer action current drive. The HIT-II device itself is modestly sized (major radius R = 0.3 m, minor radius a = 0.2 m, with an on-axis magnetic field of up to Bo = 0.5 T), but has demonstrated toroidal plasma currents of up to 200 kA, using either CHI or transformer drive. An overview of ongoing research on HIT-II plasmas, including recent results, will be presented. An electron-locking model has been developed for helicity injection current drive; a description of this model will be presented, as well as comparisons to experimental results from the HIT and HIT-II devices. Empirical results from both the HIT program and past spheromak research, buttressed by theoretical developments, have led to the design of the upcoming HIT-SI (Helicity Injected Torus with Steady Inductive helicity injection) device (T.R. Jarboe, Fusion Technology 36, p. 85, 1999). HIT-SI will be able to form a high-beta spheromak, a low aspect ratio RFP or a spherical tokamak using constant inductive helicity injection. The HIT-SI design and construction progress will be presented.

Review of the magnetic fusion program by the 1986 ERAB Fusion Panel

NASA Astrophysics Data System (ADS)

Davidson, Ronald C.

1987-09-01

The 1986 ERAB Fusion Panel finds that fusion energy continues to be an attractive energy source with great potential for the future, and that the magnetic fusion program continues to make substantial technical progress. In addition, fusion research advances plasma physics, a sophisticated and useful branch of applied science, as well as technologies important to industry and defense. These factors fully justify the substantial expenditures by the Department of Energy in fusion research and development (R&D). The Panel endorses the overall program direction, strategy, and plans, and recognizes the importance and timeliness of proceeding with a burning plasma experiment, such as the proposed Compact Ignition Tokamak (CIT) experiment.

Simulations of the effects of density and temperature profile on SMBI penetration depth based on the HL-2A tokamak configuration

NASA Astrophysics Data System (ADS)

Wu, Xueke; Li, Huidong; Wang, Zhanhui; Feng, Hao; Zhou, Yulin

2017-06-01

Not Available Project supported by the National Natural Science Foundation for Young Scientists of China (Grant No. 11605143), the Undergraduate Training Programs for Innovation and Entrepreneurship of Sichuan Province, China (Grant No. 05020732), the National Natural Science Foundation of China (Grant No. 11575055), the Fund from the Department of Education in Sichuan Province of China (Grant No. 15ZB0129), the China National Magnetic Confinement Fusion Science Program (Grant No. 2013GB107001), the National ITER Program of China (Contract No. 2014GB113000), and the Funds of the Youth Innovation Team of Science and Technology in Sichuan Province of China (Grant No. 2014TD0023).

What`s fair is fair

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nachtrieb, R.; Freidberg, J.P.

The newly elucidated strategy for the magnetic fusion program set forth by the Department of Energy calls for increased emphasis on alternate concepts. This strategy is motivated by the recognition that in spite of its many attractive features, a tokamak tends to be a low power density device, ultimately translating into large and corresponding expensive reactor. ITER, as it is currently envisaged, is a good example of a large, expensive, plain vanilla tokamak. In its defense, ITER rightly claims that its base design is very conservative in order to minimize the risk of failure. In order to increase power densitymore » and reduce cost there are two qualitatively different approaches that one can follow: discover advanced modes of tokamak operation or develop near alternate concepts. To decide which path to follow is a difficult task because of the uncertainties involved in making accurate comparisons between different concepts at different stages of development. One area, however, that most would agree is meaningful is ideal MHD stability. For any given concept to be credible as a reactor, it must at least be stable against macroscopic ideal MHD modes. The TPX design, for instance, goes to considerable trouble to obtain stability against external kinks: a close fitting metallic cage, rotation to stabilize the resistive wall version of the external kink, and, if all else fails, feedback. For credibility any other advanced tokamak or alternate concept should be held to the same standards of ideal MHD stability. As a first step in addressing this requirement we have investigated the stability of the RFP since it can be simply and accurately modeled as a straight cylinder. The RFP is well known to have good stability at high P against internal modes but is very unstable to external modes. We have developed a linear stability code which treats the plasma as an ideal compressible fluid, and includes longitudinal flow and a resistive wall.« less

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.

Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment

NASA Astrophysics Data System (ADS)

Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.

2018-05-01

A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.

Study of high field side/low field side asymmetry in the electron temperature profile with electron cyclotron emission

NASA Astrophysics Data System (ADS)

Gugliada, V. R.; Austin, M. E.; Brookman, M. W.

2017-10-01

Electron cyclotron emission (ECE) provides high resolution measurements of electron temperature profiles (Te(R , t)) in tokamaks. Calibration accuracy of this data can be improved using a sawtooth averaging technique. This improved calibration will then be utilized to determine the symmetry of Te profiles by comparing low field side (LFS) and high field side (HFS) measurements. Although Te is considered constant on flux surfaces, cases have been observed in which there are pronounced asymmetries about the magnetic axis, particularly with increased pressure. Trends in LFS/HFS overlap are examined as functions of plasma pressure, MHD mode presence, heating techniques, and other discharge conditions. This research will provide information on the accuracy of the current two-dimensional mapping of flux surfaces in the tokamak. Findings can be used to generate higher quality EFITs and inform ECE calibration. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER549698.

Recent progress of the Laser-driven Ion-beam Trace Probe

NASA Astrophysics Data System (ADS)

Yang, Xiaoyi; Xiao, Chijie; Chen, Yihang; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

2017-10-01

The Laser-driven Ion-beam Trace Probe (LITP) is a new method to diagnose the poloidal magnetic field and radial electric field in tokamaks. Recently significant progresses have been made as follows. 1) The experimental system has been set up on the PKU Plasma Test (PPT) linear device and begun to validate the principle of LITP, including the ion source, the ion detector and the poloidal magnetic field cable. Preliminary experimental results matched the theoretical prediction well. 2) The reconstruction principle has been improved including the nonlinear effect. 3) Tomography methods have been applied in the reconstruction codes. Now the laser-driven ion-beam accelerator has been setup on the PPT device, and further test of LITP will start soon. After that a prototype of LITP system will be designed and setup on the HL-2A tokamak device. This work was supported by the CHINA MOST under 2012YQ030142, ITER-CHINA program 2015GB120001 and National Natural Science Foundation of China under 11575014 and 11375053.

The scientific prototype, the only reasonable next step for the American MFE program; or why FESAC will never solicit my advice again

NASA Astrophysics Data System (ADS)

Manheimer, Wallace

2014-10-01

The scientific prototype is a tokamak which builds on what has been accomplished in TFTR, JET and JT-60. Instead of attempting to advance the plasma parameters, or investigate a new confinement configuration, it takes the tokamak plasma parameters already achieved (or actually nearly already achieved), Q about 1 and run it at steady state or high duty cycle in a DT plasma. It is very much a nuclear device requiring all of the safeguards of any nuclear device. It is an important step forward for either pure fusion or fusion breeding, and it is difficult to see how fusion can advance very far with out the knowledge the scientific prototype would provide. The poster will be divided into two parts. The first part examines options other than the scientific prototype and shows why they should be rejected. The second part explains the scientific prototype in somewhat more detail.

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.

Measurement of the ratio of hydrogen to deuterium at the KSTAR 2009 experimental campaign

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwak, Jong-Gu; Wang, Son Jong; Kim, Sun Ho

The control of the ratio of hydrogen to the deuterium is one of the very important issues for ion cyclotron range of frequency (ICRF) minority heating as well as the plasma wall interaction in the tokamak. The ratio of hydrogen to deuterium during the tokamak shot was deduced from the emission spectroscopy measurements during the KSTAR 2009 experimental campaign. Graphite tiles were used for the plasma facing components (PFCs) at KSTAR and its surface area exposed to the plasma was about 11 m{sup 2}. The data showed that it remained as high as around 50% during the campaign period becausemore » graphite tiles were exposed to the air for about two months and the hydrogen contents at the tiles are not fully pumped out due to the lack of baking on the PFC in the 2009 campaign. The validation of the spectroscopy method was checked by using the Zeeman effects and the ratio of hydrogen to the deuterium is compared with results from the residual gas analysis. During the tokamak shot, the ratio is low below 10% initially and saturated after around 1 s. When there is a hydrogen injection to the vessel via ion cyclotron wall conditioning and the boronization process where the carbone is used, the ratio of the hydrogen to the deuterium is increased by up to 100% and it recovers to around 50% after one day of operation. However it does not decrease below 50% at the end of the experimental campaign. It was found that the full baking on the PFC (with a high temperature and sufficient vacuum pumping) is required for the ratio control which guarantees the efficient ICRF heating at the KSTAR 2010 experimental campaign.« less

Princeton Plasma Physics Laboratory: Annual report, October 1, 1986--September 30, 1987

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1987-01-01

This report contains papers on the following topics: Principle Parameters Achieved in Experimental Devices (FY87); Tokamak Fusion Test Reactor; Princeton Beta Experiment-Modification; S-1 Spheromak; Current-Drive Experiment; X-Ray Laser Studies; Theoretical Division; Tokamak Modeling; Compact Ignition Tokamak; Engineering Department; Project Planning and Safety Office; Quality Assurance and Reliability; Administrative Operations; and PPPL Patent Invention Disclosures (FY87).

Alternative approaches to plasma confinement

NASA Technical Reports Server (NTRS)

Roth, J. R.

1977-01-01

The potential applications of fusion reactors, the desirable properties of reactors intended for various applications, and the limitations of the Tokamak concept are discussed. The principles and characteristics of 20 distinct alternative confinement concepts are described, each of which may be an alternative to the Tokamak. The devices are classed as Tokamak-like, stellarator-like, mirror machines, bumpy tori, electrostatically assisted, migma concept, and wall-confined plasma.

Bootstrap current in a tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C.E.

1994-03-01

The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart Zweben; Samuel Cohen; Hantao Ji

Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bourham, Mohamed A.; Gilligan, John G.

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing componentsmore » safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Y.; Tobias, B.; Chang, Y. -T.

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

NASA Astrophysics Data System (ADS)

Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

2016-10-01

The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

Multichannel reconfigurable measurement system for hot plasma diagnostics based on GEM-2D detector

NASA Astrophysics Data System (ADS)

Wojenski, A. J.; Kasprowicz, G.; Pozniak, K. T.; Byszuk, A.; Chernyshova, M.; Czarski, T.; Jablonski, S.; Juszczyk, B.; Zienkiewicz, P.

2015-12-01

In the future magnetically confined fusion research reactors (e.g. ITER tokamak), precise determination of the level of the soft X-ray radiation of plasma with temperature above 30 keV (around 350 mln K) will be very important in plasma parameters optimization. This paper presents the first version of a designed spectrography measurement system. The system is already installed at JET tokamak. Based on the experience gained from the project, the new generation of hardware for spectrography measurements, was designed and also described in the paper. The GEM detector readout structure was changed to 2D in order to perform measurements of i.e. laser generated plasma. The hardware structure of the system was redesigned in order to provide large number of high speed input channels. Finally, this paper also covers the issue of new control software, necessary to set-up a complete system of certain complexity and perform data acquisition. The main goal of the project was to develop a new version of the system, which includes upgraded structure and data transmission infrastructure (i.e. handling large number of measurement channels, high sampling rate).

Plasma Radiofrequency Discharges as Cleaning Technique for the Removal of C-W Coatings

NASA Astrophysics Data System (ADS)

Cremona, A.; Vassallo, E.; Caniello, R.; Ghezzi, F.; Grosso, G.; Laguardia, L.

2013-06-01

Erosion of materials by chemical and physical sputtering is one of the most concern of plasma wall interaction in tokamaks. In divertor ITER-like tokamaks, where carbon and tungsten are planned to be used, hydrogenated C-W mixed compounds are expected to form by erosion, transport and re-deposition processes. The selection of these materials as divertor components involves lifetime and safety issues due to tritium retention in carbon co-deposits. In this paper a cleaning technique based on RF (13.56 MHz) capacitively coupled H2/Ar plasmas has been used to remove C-W mixed materials from test specimens. The dependence of the removal rate on the H2/Ar ratio and on the plasma pressure has been investigated by X-ray photoelectron spectroscopy, atomic force microscopy, profilometry as regards the solid phase and by Langmuir probe and optical emission spectroscopy as regards the plasma phase. The best result has been obtained with a H2/Ar ratio of 10/90 at a pressure of 1 Pa. An explanation based on a synergistic effect between physical sputtering due to energetic ions and chemical etching due to radicals, together with the pressure dependence of the ion energy distribution function, is given.

Regularization of soft-X-ray imaging in the DIII-D tokamak

DOE PAGES

Wingen, A.; Shafer, M. W.; Unterberg, E. A.; ...

2015-03-02

We developed an image inversion scheme for the soft X-ray imaging system (SXRIS) diagnostic at the DIII-D tokamak in order to obtain the local soft X-ray emission at a poloidal cross-section from the spatially line-integrated image taken by the SXRIS camera. The scheme uses the Tikhonov regularization method since the inversion problem is generally ill-posed. The regularization technique uses the generalized singular value decomposition to determine a solution that depends on a free regularization parameter. The latter has to be chosen carefully, and the so called {\\it L-curve} method to find the optimum regularization parameter is outlined. A representative testmore » image is used to study the properties of the inversion scheme with respect to inversion accuracy, amount/strength of regularization, image noise and image resolution. Moreover, the optimum inversion parameters are identified, while the L-curve method successfully computes the optimum regularization parameter. Noise is found to be the most limiting issue, but sufficient regularization is still possible at noise to signal ratios up to 10%-15%. Finally, the inversion scheme is applied to measured SXRIS data and the line-integrated SXRIS image is successfully inverted.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorelenkov, Nikolai N

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler [1]. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicityinduced AEsmore » (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).« less

Nonlinear 3D MHD verification study: SpeCyl and PIXIE3D codes for RFP and Tokamak plasmas

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Cappello, S.; Chacon, L.

2010-11-01

A strong emphasis is presently placed in the fusion community on reaching predictive capability of computational models. An essential requirement of such endeavor is the process of assessing the mathematical correctness of computational tools, termed verification [1]. We present here a successful nonlinear cross-benchmark verification study between the 3D nonlinear MHD codes SpeCyl [2] and PIXIE3D [3]. Excellent quantitative agreement is obtained in both 2D and 3D nonlinear visco-resistive dynamics for reversed-field pinch (RFP) and tokamak configurations [4]. RFP dynamics, in particular, lends itself as an ideal non trivial test-bed for 3D nonlinear verification. Perspectives for future application of the fully-implicit parallel code PIXIE3D to RFP physics, in particular to address open issues on RFP helical self-organization, will be provided. [4pt] [1] M. Greenwald, Phys. Plasmas 17, 058101 (2010) [0pt] [2] S. Cappello and D. Biskamp, Nucl. Fusion 36, 571 (1996) [0pt] [3] L. Chac'on, Phys. Plasmas 15, 056103 (2008) [0pt] [4] D. Bonfiglio, L. Chac'on and S. Cappello, Phys. Plasmas 17 (2010)

Source-to-incident-flux relation in a Tokamak blanket module

NASA Astrophysics Data System (ADS)

Imel, G. R.

The next-generation Tokamak experiments, including the Tokamak fusion test reactor (TFTR), will utilize small blanket modules to measure performance parameters such as tritium breeding profiles, power deposition profiles, and neutron flux profiles. Specifically, a neutron calorimeter (simply a neutron moderating blanket module) which permits inferring the incident 14 MeV flux based on measured temperature profiles was proposed for TFTR. The problem of how to relate this total scalar flux to the fusion neutron source is addressed. This relation is necessary since the calorimeter is proposed as a total fusion energy monitor. The methods and assumptions presented was valid for the TFTR Lithium Breeding Module (LBM), as well as other modules on larger Tokamak reactors.

Gamma ray imager on the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.

2016-04-15

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 electronsmore » 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.« less

Gamma ray imager on the DIII-D tokamak

DOE PAGES

Pace, D. C.; Cooper, C. M.; Taussig, D.; ...

2016-04-13

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 electronsmore » 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. In conclusion, 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.« less

The ITPA disruption database

DOE PAGES

Eidietis, N. W.; Gerhardt, S. P.; Granetz, R. S.; ...

2015-05-22

A multi-device database of disruption characteristics has been developed under the auspices of the International Tokamak Physics Activity magneto hydrodynamics topical group. The purpose of this ITPA Disruption Database (IDDB) is to find the commonalities between the disruption and disruption mitigation characteristics in a wide variety of tokamaks in order to elucidate the physics underlying tokamak disruptions and to extrapolate toward much larger devices, such as ITER and future burning plasma devices. Conversely, in order to previous smaller disruption data collation efforts, the IDDB aims to provide significant context for each shot provided, allowing exploration of a wide array ofmore » relationships between pre-disruption and disruption parameters. Furthermore, the IDDB presently includes contributions from nine tokamaks, including both conventional aspect ratio and spherical tokamaks. An initial parametric analysis of the available data is presented. Our analysis includes current quench rates, halo current fraction and peaking, and the effectiveness of massive impurity injection. The IDDB is publicly available, with instruction for access provided herein.« less

Radioactivity evaluation for the KSTAR tokamak.

PubMed

Kim, Hyunduk; Lee, Hee-Seock; Hong, Sukmo; Kim, Minho; Chung, Chinwha; Kim, Changsuk

2005-01-01

The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 10(16) s(-1) through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10(-4) mrem h(-1) in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 10(6) Bq kg(-1) in the carbon graphite of a plasma-facing wall.

Electron Cyclotron Radiation, Related Power Loss, and Passive Current Drive in Tokamaks: A Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fidone, Ignazio; Giruzzi, Gerardo; Granata, Giovanni

2001-01-15

A critical review on emission of weakly damped, high-harmonics electron cyclotron radiation, the related synchrotron power loss, and passive current drive in tokamaks with a fish-scale first wall is presented. First, the properties of overlapping harmonics are discussed using general analytical formulas and numerical applications. Next, the radiation power loss and efficiency of passive current drive in tokamak reactors are derived for the asymmetric fish-scale first wall. The radiation power loss is determined by the direction-averaged reflection coefficient {sigma}{sub 0} and the passive current drive by the differential reflectivity {delta}{sigma}/(1 - {sigma}{sub 0}). Finally, the problem of experimental investigations ofmore » the high harmonics radiation spectra, of {sigma}{sub 0} and {delta}{sigma}/(1 - {sigma}{sub 0}) in existing and next-step tokamaks, is discussed. Accurate measurements of the radiation spectra and the fish-scale reflectivity can be performed at arbitrary electron temperature using a partial fish-scale structure located near the tokamak equatorial plane.« less

Impedance of an intense plasma-cathode electron source for tokamak startup

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

2016-05-01

An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Realizing Steady State Tokamak Operation for Fusion Energy

NASA Astrophysics Data System (ADS)

Luce, T. C.

2009-11-01

Continuous operation of a tokamak for fusion energy has obvious engineering advantages, but also presents physics challenges beyond the achievement of conditions needed for a burning plasma. The power from 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 in the present generation of tokamaks, and significant progress has been made in the last decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are now operated routinely without disruptions close to the ideal MHD pressure limit, as needed for steady-state operation. Scenarios that project to high fusion gain have been demonstrated where more than half of the current is supplied by the ``bootstrap'' current generated by the pressure gradient in the plasma. Fully noninductive sustainment has been obtained for about a resistive time (the longest intrinsic time scale in the confined plasma) with normalized pressure and confinement approaching those needed for demonstration of steady-state conditions in ITER. One key challenge remaining to be addressed is how to handle the demanding heat and particle fluxes expected in a steady-state tokamak without compromising the high level of core plasma performance. Rather than attempt a comprehensive historical survey, this review will start from the plasma requirements of a steady-state tokamak powerplant, illustrate with examples the progress made in both experimental and theoretical understanding, and point to the remaining physics challenges.

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Modular coils and finite-β operation of a quasi-axially symmetric tokamak

NASA Astrophysics Data System (ADS)

Drevlak, M.

1998-09-01

Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nührenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (Merkel, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (Hirshman, S.P., Van Rij, W.I., Merkel, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak.

U. S. fusion programs: Struggling to stay in the game

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, M.

Funding for the US fusion energy program has suffered and will probably continue to suffer major cuts. A committee hand-picked by Energy Secretary James Watkins urged the Department of Energy to mount an aggressive program to develop fusion power, but congress cut funding from $323 million in 1990 to $275 million in 1991. This portends dire conditions for fusion research and development. Projects to receive top priority are concerned with the tokamaks and to keep the next big machine, the Burning Plasma Experiment, scheduled for beginning of construction in 1993 on schedule. Secretary Watkins is said to want to keepmore » the International Thermonuclear Energy Reactor (ITER) on schedule. ITER would follow the Burning Plasma Experiment.« less

Data processing and analysis for 2D imaging GEM detector system

NASA Astrophysics Data System (ADS)

Czarski, T.; Chernyshova, M.; Pozniak, K. T.; Kasprowicz, G.; Byszuk, A.; Juszczyk, B.; Kolasinski, P.; Linczuk, M.; Wojenski, A.; Zabolotny, W.; Zienkiewicz, P.

2014-11-01

The Triple Gas Electron Multiplier (T-GEM) is presented as soft X-ray (SXR) energy and position sensitive detector for high-resolution X-ray diagnostics of magnetic confinement fusion plasmas [1]. Multi-channel measurement system and essential data processing for X-ray energy and position recognition is consider. Several modes of data acquisition are introduced depending on processing division for hardware and software components. Typical measuring issues aredeliberated for enhancement of data quality. Fundamental output characteristics are presented for one and two dimensional detector structure. Representative results for reference X-ray source and tokamak plasma are demonstrated.

Gyrokinetic simulations and experiment

NASA Astrophysics Data System (ADS)

Ross, David W.; Bravenec, R. V.; Dorland, W.

2002-11-01

Nonlinear gyrokinetic simulations with the code GS2 have been carried out in an effort to predict transport fluxes and fluctuation levels in the tokamaks DIII-D and Alcator C-Mod.(W. Dorland et al. in Fusion Energy 2000 (International Atomic Energy Agency, Vienna, 2000).)^,( W. Ross and W. Dorland, submitted to Phys. Plasmas (2002).) These simulations account for full electron dynamics and, in some instances, electromagnetic waves.( D. W. Ross, W. Dorland, and B. N. Rogers, Bull. Am. Phys. Soc. 46, 115 (2001).) Here, some issues of the necessary resolution, precision and wave number range are examined in connection with the experimental comparisons and parameter scans.

Chaotic density fluctuations in L-mode plasmas of the DIII-D tokamak

DOE PAGES

Maggs, J. E.; Rhodes, Terry L.; Morales, G. J.

2015-03-05

Analysis of the time series obtained with the Doppler backscattering system (DBS) in the DIII-D tokamak shows that intermediate wave number plasma density fluctuations in low confinement (L-mode) tokamak plasmas are chaotic. Here, the supporting evidence is based on the shape of the power spectrum; the location of the signal in the complexity-entropy plane (C-H plane); and the population of the corresponding Bandt-Pompe probability distributions.

Method of sustaining a radial electric field and poloidal plasma rotation over most of the cross-section of a tokamak

DOEpatents

Darrow, Douglass S.; Ono, Masayuki

1990-03-06

A radial electric field of a desired magnitude and configuration is created throughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.

Method of sustaining a radial electric field and poloidal plasma rotation over most of the cross-section of a tokamak

DOEpatents

Darrow, Douglass S.; Ono, Masayuki

1990-01-01

A radial electric field of a desired magnitude and configuration is created hroughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

DOE PAGES

Mitarai, O.; Xiao, C.; McColl, D.; ...

2015-03-24

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

Preliminary measurements of neutrons from the D-D reaction in the COMPASS tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dankowski, J., E-mail: jan.dankowski@ifj.edu.pl; Kurowski, A.; Twarog, D.

Recent results of measured fast neutrons created in the D-D reaction on the COMPASS tokamak during ohmic discharges are presented in this paper. Two different type detectors were used during experiment. He-3 detectors and bubble detectors as a support. The measurements are an introduction for neutron diagnostic on tokamak COMPASS and monitoring neutrons during discharges with Neutral Beam Injection (NBI). The He-3 counters and bubble detectors were located in two positions near tokamak vacuum chamber at a distance less than 40 cm to the centre of plasma. The neutrons flux was observed in ohmic discharges. However, analysis of our resultsmore » does not indicate any clear source of neutrons production during ohmic discharges.« less

Plasma production and preliminary results from the ADITYA Upgrade tokamak

NASA Astrophysics Data System (ADS)

R, L. TANNA; J, GHOSH; Harshita, RAJ; Rohit, KUMAR; Suman, AICH; Vaibhav, RANJAN; K, A. JADEJA; K, M. PATEL; S, B. BHATT; K, SATHYANARAYANA; P, K. CHATTOPADHYAY; M, N. MAKWANA; K, S. SHAH; C, N. GUPTA; V, K. PANCHAL; Praveenlal, EDAPPALA; Bharat, ARAMBHADIYA; Minsha, SHAH; Vismay, RAULJI; M, B. CHOWDHURI; S, BANERJEE; R, MANCHANDA; D, RAJU; P, K. ATREY; Umesh, NAGORA; J, RAVAL; Y, S. JOISA; K, TAHILIANI; S, K. JHA; M, V. GOPALKRISHANA

2018-07-01

The Ohmically heated circular limiter tokamak ADITYA (R 0 = 75 cm, a = 25 cm) has been upgraded to a tokamak named the ADITYA Upgrade (ADITYA-U) with an open divertor configuration with divertor plates. The main goal of ADITYA-U is to carry out dedicated experiments relevant for bigger fusion machines including ITER, such as the generation and control of runaway electrons, disruption prediction, and mitigation studies, along with an improvement in confinement with shaped plasma. The ADITYA tokamak was dismantled and the assembly of ADITYA-U was completed in March 2016. Integration of subsystems like data acquisition and remote operation along with plasma production and preliminary plasma characterization of ADITYA-U plasmas are presented in this paper.

Newly developed double neural network concept for reliable fast plasma position control

NASA Astrophysics Data System (ADS)

Jeon, Young-Mu; Na, Yong-Su; Kim, Myung-Rak; Hwang, Y. S.

2001-01-01

Neural network is considered as a parameter estimation tool in plasma controls for next generation tokamak such as ITER. The neural network has been reported to be so accurate and fast for plasma equilibrium identification that it may be applied to the control of complex tokamak plasmas. For this application, the reliability of the conventional neural network needs to be improved. In this study, a new idea of double neural network is developed to achieve this. The new idea has been applied to simple plasma position identification of KSTAR tokamak for feasibility test. Characteristics of the concept show higher reliability and fault tolerance even in severe faulty conditions, which may make neural network applicable to plasma control reliably and widely in future tokamaks.

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mitarai, O.; Xiao, C.; McColl, D.

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

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.

Numerical modelling of geodesic acoustic mode relaxation in a tokamak edge

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Dorr, M.; ...

2013-05-08

Here, the edge of a tokamak in a high confinement (H mode) regime is characterized by steep density gradients and a large radial electric field. Recent analytical studies demonstrated that the presence of a strong radial electric field consistent with a subsonic pedestal equilibrium modifies the conventional results of the neoclassical formalism developed for the core region. In the present work we make use of the recently developed gyrokinetic code COGENT to numerically investigate neoclassical transport in a tokamak edge including the effects of a strong radial electric field. The results of numerical simulations are found to be in goodmore » qualitative agreement with the theoretical predictions and the quantitative discrepancy is discussed. In addition, the present work investigates the effects of a strong radial electric field on the relaxation of geodesic acoustic modes (GAMs) in a tokamak edge. Numerical simulations demonstrate that the presence of a strong radial electric field characteristic of a tokamak pedestal can enhance the GAM decay rate, and heuristic arguments elucidating this finding are provided.« less

Likelihood of Alfvénic instability bifurcation in experiments

NASA Astrophysics Data System (ADS)

Duarte, Vinicius; Gorelenkov, Nikolai; Schneller, Mirjam; Fredrickson, Eric; Berk, Herbert; Canal, Gustavo; Heidbrink, William; Kaye, Stanley; Podesta, Mario; van Zeeland, Michael; Wang, Weixing

2017-10-01

We apply a criterion for the likely nature of fast ion redistribution in tokamaks to be in the convective or diffusive nonlinear regimes. The criterion, which is shown to be rather sensitive to the relative strength of collisional or micro-turbulent scattering and drag processes, ultimately translates into a condition for the applicability of reduced quasilinear modeling for realistic tokamak eigenmodes scenarios. The criterion is tested and validated against different machines, where the chirping mode behavior is shown to be in accord with the model. It has been found that the anomalous fast ion transport is a likely mediator of the bifurcation between the fixed-frequency mode behavior and rapid chirping in tokamaks. In addition, micro-turbulence appears to resolve the disparity with respect to the ubiquitous chirping observation in spherical tokamaks and its rarer occurrence in conventional tokamaks. In NSTX, the tendency for chirping is further studied in terms of the beam beta and the plasma rotation shear. For more accurate quantitative assessment, numerical simulations of the effects of electrostatic ion temperature gradient turbulence on chirping are presently being pursued using the GTS code.

Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Mangiarotti, F.; Minervini, J.

2014-10-01

The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

Sub-Alfvénic reduced magnetohydrodynamic equations for tokamaks

NASA Astrophysics Data System (ADS)

Sengupta, W.; Hassam, A. B.; Antonsen, T. M.

2017-06-01

A reduced set of magnetohydrodynamic (MHD) equations is derived, applicable to large aspect ratio tokamaks and relevant for dynamics that is sub-Alfvénic with respect to ideal ballooning modes. This ordering optimally allows sound waves, Mercier modes, drift modes, geodesic-acoustic modes (GAM), zonal flows and shear Alfvén waves. Wavelengths long compared to the gyroradius but comparable to the minor radius of a typical tokamak are considered. With the inclusion of resistivity, tearing modes, resistive ballooning modes, Pfirsch-Schluter cells and the Stringer spin-up are also included. A major advantage is that the resulting system is two-dimensional in space, and the system incorporates self-consistent and dynamic Shafranov shifts. A limitation is that the system is valid only in radial domains where the tokamak safety factor, , is close to rational. In the tokamak core, the system is well suited to study the sawtooth discharge in the presence of Mercier modes. The systematic ordering scheme and methodology developed are versatile enough to reduce the more general collisional two-fluid equations or possibly the Vlasov-Maxwell system in the MHD ordering.

Long-distance delivery of multi-channel polarization signals in nuclear fusion research

NASA Astrophysics Data System (ADS)

Ko, Jinseok; Chung, Jinil; Lee, Kyuhang

2017-04-01

A polarization-preserving optical system that includes a dual photoelastic modulator (PEM) has been designed and fabricated for the motional Stark effect (MSE) diagnostic system which measures internal magnetic field structures inside the tokamak for the Korea Superconducting Tokamak Advanced Research. The collection optics located outside the vacuum window is composed of four lenses, a dielectric coated mirror, and a dichroic beam splitter in addition to the PEM and a polarizer. The fiber dissector is designed based on the focal plane that aligns 25 lines of sight, each of which constitutes a bundle of 19 600-μm fibers. The fibers run about 40 m from the front optics in the tokamak vacuum vessel to the detector in the diagnostic area remote from the tokamak hall. This takes the advantage of the fact that the polarization information is intensity-modulated once going through the PEM and the polarizer. The polarization signals measured by the MSE diagnostic successfully demonstrates its proof-of-principle physics that is critical in the stable and steady-state operation of the tokamak plasmas.

Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak.

PubMed

Li, G; Tan, Y; Liu, Y Q

2015-08-01

Eddy currents have an important effect on tokamak plasma equilibrium and control of magneto hydrodynamic activity. The vacuum vessel of the Sino-UNIted Spherical Tokamak is separated into two hemispherical sections by a toroidal insulating barrier. Consequently, the characteristics of eddy currents are more complex than those found in a standard tokamak. Thus, it is necessary to measure and analyze the eddy-current distribution. In this study, we propose an experimental method for measuring the eddy-current distribution in a vacuum vessel. By placing a flexible printed circuit board with magnetic probes onto the external surface of the vacuum vessel to measure the magnetic field parallel to the surface and then subtracting the magnetic field generated by the vertical-field coils, the magnetic field due to the eddy current can be obtained, and its distribution can be determined. We successfully applied this method to the Sino-UNIted Spherical Tokamak, and thus, we obtained the eddy-current distribution despite the presence of the magnetic field generated by the external coils.

Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

NASA Astrophysics Data System (ADS)

Ahmadi, S.; Salar Elahi, A.; Ghorannevis, M.

2018-03-01

In this paper we have studied an Ergodic Magnetic Limiter (EML) based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist) Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE PAGES

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.; ...

2018-03-14

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Electric Tokamak Results and Plans

NASA Astrophysics Data System (ADS)

Taylor, R. J.; Gauvreau, J.-L.; Gourdain, P.-A.; Kissick, M. W.; Leboeuf, J.-N.; Schmitz, L. W.

1999-11-01

Initial plasmas have been obtained in ET (R=5 m, a=1 m) at 100 G previously with plasma currents up to 100 kA. Firsts full field plasma at 2.5 kG with the possibility of electrode and RF driven H-modes are expected in August 1999. The initial goal of the Electric Tokamak, ET(See Web Site for more details.), is to "drift lock" the thermal ions to the magnetic surfaces by rapid poloidal rotation ω_pol > ω_bounce. The required radial electric field is Er = Vi × B/A. ET physics is an outcome of the TTF effort of the last decade with a "unity beta vision" added to it by Cowley(S. C. Cowley, P. K. Kaw, R. S. Kelly, R. M. Kulsrud, \\underlinePhys. fluids B) 3 (1991) 2066. Classical heat transport is expected at 2 keV temperatures if rotation and beta can be combined, leading to the realization of an omnigeneous magnetic confinement zone(D. Palumbo, \\underlineNuovo Cimento) 53B (1967) 507.. The plasma startup experience and the scientific program plan will be discussed.

Scaling Relationships for ELM Diverter Heat Flux on DIII D

NASA Astrophysics Data System (ADS)

Peters, E. A.; Makowski, M. A.; Leonard, A. W.

2015-11-01

Edge Localized Modes (ELMs) are periodic plasma instabilities that occur during H-mode operation in tokamaks. Left unmitigated, these instabilities result in concentrated particle and heat fluxes at the divertor and stand to cause serious damage to the plasma facing components of tokamaks. The purpose of this research is to find scaling relationships that predict divertor heat flux due to ELMs based on plasma parameters at the time of instability. This will be accomplished by correlating characteristic ELM parameters with corresponding plasma measurements and analyzing the data for trends. One early assessment is the effect of the heat transmission coefficient ? on the in/out asymmetry of the calculated ELM heat fluxes. Using IR camera data, further assessments in this study will continue to emphasize in/out asymmetry in ELMs, as this has important implications for ITER operation. Work supported in part by the US DOE, DE-AC52-07NA27344, DE-FC02-04ER54698, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

Fast, multi-channel real-time processing of signals with microsecond latency using graphics processing units.

PubMed

Rath, N; Kato, S; Levesque, J P; Mauel, M E; Navratil, G A; Peng, Q

2014-04-01

Fast, digital signal processing (DSP) has many applications. Typical hardware options for performing DSP are field-programmable gate arrays (FPGAs), application-specific integrated DSP chips, or general purpose personal computer systems. This paper presents a novel DSP platform that has been developed for feedback control on the HBT-EP tokamak device. The system runs all signal processing exclusively on a Graphics Processing Unit (GPU) to achieve real-time performance with latencies below 8 μs. Signals are transferred into and out of the GPU using PCI Express peer-to-peer direct-memory-access transfers without involvement of the central processing unit or host memory. Tests were performed on the feedback control system of the HBT-EP tokamak using forty 16-bit floating point inputs and outputs each and a sampling rate of up to 250 kHz. Signals were digitized by a D-TACQ ACQ196 module, processing done on an NVIDIA GTX 580 GPU programmed in CUDA, and analog output was generated by D-TACQ AO32CPCI modules.

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Stabilization of Tokamak Plasmas by the Addition of Nonaxisymmetric Coils

NASA Astrophysics Data System (ADS)

Reiman, Allan

2008-11-01

It has been recognized since the early days of the fusion program that stellarator coils can be used to stabilize current carrying, toroidal, magnetically confined plasmas.[1] More recently, it has been shown that the vertical mode in a tokamak can be stabilized by a relatively simple set of parallelogram-shaped, localized, nonaxisymmetric coils.[2] We show that by superposing sets of these parallelogram-shaped, nonaxisymmetric coils at different locations, it is possible to reproduce the coil current patterns for conventional stellarator coils as well as those for Furth-Hartman coils[3]. This allows us to gain insight into the physics of stabilization produced by various sets of nonaxisymmetric coils by analysis of the effect on stability of localized coils at differing locations. In particular, the relationship between the stabilization effect and the rotational transform generated by the nonaxisymmetric coils is clarified. [1] J. L. Johnson, C. R. Oberman, R. M. Kulsrud, and E. A. Frieman, Phys. Fluids 1, 281 (1958) [2] A. Reiman, Phys. Rev. Lett. 99, 135007, (2007). [3] H.P. Furth and C.W. Hartman, Phys. Fluids 11, 408 (1968).

Measurement of H/D ratio and ion temperature on a HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Wei, Lehan; Lin, Xiaodong

1997-01-01

By combining optical fibers with piezoelectric scanning Fabry-Perot interferometer, the profiles of Hα and Dα have been determined simultaneously in a single Tokamak discharge. Consequently, the ratio of hydrogen to deuterium and ion temperature are obtained. Not only is the uncertainty of shot-to-shot avoided, the results of the experiment indicate that this instrumentation has the advantage of rapid wavelength scanning, large dispersion, high resolution, and good adaptability of working in adverse circumstances such as at a Tokamak site.

Electron Injection by E-Field Drift and its Application in Starting-up Tokamaks at Low Loop Voltage

NASA Astrophysics Data System (ADS)

Pan, Yuan; Yan, Xiao-Lin; Liu, Bao-Hua

2003-05-01

We propose an innovative method of electron injection by E-field drift into a plasma device and discuss its application in starting-up tokamak plasmas at low loop voltage. The experimental results obtained from HT-6M Tokamak are also presented. The breakdown loop voltage is obviously reduced and the discharge performance is improved by using the electron injection method. It could be applied to some other types of plasma device.

Startup and stability of a small spherical tokamak

NASA Astrophysics Data System (ADS)

Garstka, Gregory Douglas

1997-09-01

The spherical tokamak (ST) is an evolutionary extension of the conventional tokamak concept where the aspect ratio is less than 2. These devices may possess significant advantages over standard tokamaks-they are capable of achieving higher values of /beta, seem to be more resilient to disruptions, and are significantly smaller than conventional tokamaks. Two important questions for the next generation of spherical tokamaks concern startup and internal reconnection events (IREs). Understanding startup is crucial due to the limited amount of ohmic flux in an ST. The IREs are disruption- like events observed on STs that do not result in termination of the current channel. Experiments have been conducted on the Madison EDUcational Small Aspect-ratio (MEDUSA) tokamak to answer some of the questions about startup and IREs in STs. MEDUSA is a small ohmic tokamak with an insulating vacuum vessel. Major parameters are R=12 cm, a=8 cm, Ip=10-40 kA, BT=0.2-0.45 T, /Delta tpulse=1-2 ms, /langle ne/rangle/approx5×1019/ m-3, and Te0/approx100 eV. The experiments in this work were aided by an internal magnetic probe array that constrained the reconstruction of MHD equilibria. It was found that startup efficiency, measured by the Ejima coefficient CE, improved with increasing loop voltage and toroidal field. Double tearing modes were found to be an important mechanism for current penetration in MEDUSA; their presence early in the discharge can improve the magnetic flux consumption. The lowest achieved value of the Ejima coefficient was 0.61 (0.13 for 'OH only') for a discharge with 0.375 T toroidal field and 9.4 V startup loop voltage. The study of internal reconnection events revealed the presence of a heretofore undiscovered precursor, which in MEDUSA was manifested as coherent oscillations in the internal poloidal field at 65-75 kHz for 100 μs prior to the IRE. These events were found to result in decreased /ell i and /beta, inward movement of the magnetic axis, dramatically increased q0 and slightly decreased q98. The magnetic helicity was found to change between -15% and +28% over an IRE.

Message from the Editor

NASA Astrophysics Data System (ADS)

Stambaugh, Ronald D.

2013-01-01

The journal Nuclear Fusion has played a key role in the development of the physics basis for fusion energy. That physics basis has been sufficiently advanced to enable construction of such major facilities as ITER along the tokamak line in magnetic fusion and the National Ignition Facility (NIF) in laser-driven fusion. In the coming decade, while ITER is being constructed and brought into deuterium-tritium (DT) operation, this physics basis will be significantly deepened and extended, with particular key remaining issues addressed. Indeed such a focus was already evident with about 19% of the papers submitted to the 24th IAEA Fusion Energy Conference in San Diego, USA appearing in the directly labelled ITER and IFE categories. Of course many of the papers in the other research categories were aimed at issues relevant to these major fusion directions. About 17% of the papers submitted in the 'Experiment and Theory' categories dealt with the highly ITER relevant and inter-related issues of edge-localized modes, non-axisymmetric fields and plasma rotation. It is gratifying indeed to see how the international community is able to make such a concerted effort, facilitated by the ITPA and the ITER-IO, around such a major issue for ITER. In addition to deepening and extending the physics bases for the mainline approaches to fusion energy, the coming decade should see significant progress in the physics basis for additional fusion concepts. The stellarator concept should reach a high level of maturity with such facilities as LHD operating in Japan and already producing significant results and the W7-X in the EU coming online soon. Physics issues that require pulses of hundreds of seconds to investigate can be confronted in the new superconducting tokamaks coming online in Asia and in the major stellarators. The basis for steady-state operation of a tokamak may be further developed in the upper half of the tokamak operating space—the wall stabilized regime. New divertor geometries are already being investigated. Progress should continue on additional driver approaches in inertial fusion. Nuclear Fusion will continue to play a major role in documenting the significant advances in fusion plasma science on the way to fusion energy. Successful outcomes in projects like ITER and NIF will bring sharply into focus the remaining significant issues in fusion materials science and fusion nuclear science and technology needed to move from the scientific feasibility of fusion to the actual realization of fusion power production. These issues are largely common to magnetic and inertial fusion. Progress in these areas has been limited by the lack of suitable major research facilities. Hopefully the coming decade will see progress along these lines. Nuclear Fusion will play its part with increased papers reporting significant advances in fusion materials and nuclear science and technology. The reputation and status of the journal remains high; paper submissions are increasing and the Impact Factor for the journal remains high at 4.09 for 2011. We look forward in the coming months to publishing expanded versions of many of the outstanding papers presented at the IAEA FEC in San Diego. We congratulate Dr Patrick Diamond of the University of California at San Diego for winning the 2012 Nuclear Fusion Prize for his paper [1] and Dr Hajime Urano of the Japan Atomic Energy Agency for winning the 2011 Nuclear Fusion Prize for his paper [2]. Papers of such quality by our many authors enable the high standard of the journal to be maintained. The Nuclear Fusion editorial office understands how much effort is required by our referees. The Editorial Board decided that an expression of thanks to our most loyal referees is appropriate and so, since January 2005, we have been offering ten of the most active referees over the past year a personal subscription to Nuclear Fusion with electronic access for one year, free of charge. This year, three of the top referees have reviewed five manuscripts in the period November 2011 to December 2012 and provided excellent advice to the authors. We have excluded our Board Members, Guest Editors of special editions and those referees who were already listed in recent years. The following people have been selected: Marina Becoulet, CEA-Cadarache, France Jiaqui Dong, Southwestern Institute of Physics, China Emiliano Fable, Max-Planck-Institut für Plasmaphysik, Germany Ambrogio Fasoli, Ecole Polytechnique Federale de Lausanne, Switzerland Eric Fredrickson, Princeton Plasma Physics Laboratory, USA Manuel Garcia-Munoz, Max-Planck-Institut fuer Plasmaphysik, Germany William Heidbrink, California University, USA Katsumi Ida, National Inst. For Fusion Science, Japan Peter Stangeby, Toronto University, Canada James Strachan, Princeton Plasma Physics Laboratory, USA Victor Yavorskij, Ukraine National Academy of Sciences, Ukraine In addition, there is a group of several hundred referees who have helped us in the past year to maintain the high scientific standard of Nuclear Fusion. At the end of this issue we give the full list of all referees for 2012. Our thanks to them!

Recent Progress on Spherical Torus Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 configurationmore » 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.« less

Compact Torus Injection Experiments on the H.I.T. teststand and the JFT-2M tokamak

NASA Astrophysics Data System (ADS)

Fukumoto, Naoyuki; Fujiwara, Makoto; Kuramoto, Keiji; Ageishi, Masaya; Nagata, Masayoshi; Uyama, Tadao; Ogawa, Hiroaki; Kasai, Satoshi; Hasegawa, Kouichi; Shibata, Takatoshi

1997-11-01

A spheromak-type compact torus (CT) acceleration and injection experiment has been carried out using the Himeji Institute of Technology Compact Torus Injector (HIT-CTI). We investigate the possibility of refueling, density control, current drive, and edge electric field control of tokamak plasmas by means of CT injection. The HIT-CTI produces a CT with a speed of 200 km/s and a density of 1× 10^21m-3. We have constructed new electrodes and power supplies, and will install the HIT-CTI on the JFT-2M tokamak at JAERI in Autumn 1997. The outer electrode serves as a common ground for both the formation bank (144μF, 20kV) and the acceleration bank (92.4μF, 40kV). If the external toroidal field of the tokamak is applied across the CT acceleration region, the CT kinetic energy might decrease during penetration into the field lines joining the inner and outer electrode. This could result in the CT not being able to reach the core of the tokamak plasma. Determining the optimum position of the inner electrode is one of the near term goals of this research. We will present magnetic probe, He-Ne interferometer and fast framing camera data from experiments at H.I.T., where a CT was accelerated into a transverse field. We will also present initial results from the operation of the HIT-CTI on the JFT-2M tokamak.

Impedance of an intense plasma-cathode electron source for tokamak startup

DOE PAGES

Hinson, Edward Thomas; Barr, Jayson L.; Bongard, Michael W.; ...

2016-05-31

In this study, an impedance model is formulated and tested for the ~1kV, ~1kA/cm 2, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (n arc ≈ 10 21 m -3) within the electron source, and the less dense external tokamak edge plasma (n edge ≈ 10 18 m -3) into which current is injected at the applied injector voltage, V inj. Experiments on the Pegasus spherical tokamak show the injected current, I inj, increases with V inj according to the standard double layer scaling I injmore » ~ V inj 3/2 at low current and transitions to I inj ~ V inj 1/2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density n b ~ I inj/V inj 1/2. For low tokamak edge density n edge and high I inj, the inferred beam density n b is consistent with the requirement n b ≤ n edge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, n b ~ n arc is observed, consistent with a limit to n b imposed by expansion of the double layer sheath. These results suggest that n arc is a viable control actuator for the source impedance.« less

Studies of Be migration in the JET tokamak using AMS with 10Be marker

NASA Astrophysics Data System (ADS)

Bykov, I.; Bergsåker, H.; Possnert, G.; Zhou, Y.; Heinola, K.; Pettersson, J.; Conroy, S.; Likonen, J.; Petersson, P.; Widdowson, A.

2016-03-01

The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a 10Be enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm2 sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The 10Be/9Be ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the 10B isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation.

Design of tangential multi-energy soft x-ray camera for NSTX-U

NASA Astrophysics Data System (ADS)

Delgado-Aparicio, Luis F.; Maddox, J.; Pablant, N.; Hill, K.; Bitter, M.; Stratton, B.; Efthimion, Phillip

2016-10-01

For tokamaks and future facilities to operate safely in a high-pressure long-pulse discharge, it is imperative to address key issues associated with impurity sources, core transport and high-Z impurity accumulation. Multi-energy SXR imaging provides a unique opportunity for measuring, simultaneously, a variety of important plasma properties (Te, nZ and ΔZeff). A new tangential multi-energy soft x-ray pin-hole camera is being design to sample the continuum- and line-emission from low-, medium- and high-Z impurities. This new x-ray diagnostic will be installed on an equatorial midplane port of NSTX-U tokamak and will measure the radial structure of the photon emissivity with a radial resolution below 1 cm at a 500 Hz frame rate and a photon-energy resolution of 500 eV. The layout and response expected of the new system will be shown for different plasma conditions and impurity concentrations. The effect of toroidal rotation driving poloidal asymmetries in the core radiation is also addressed. This effort is designed to contribute to the near- and long-term highest priority research goals for NSTX-U which will integrate a non-inductive operation at reduced collisionality, long energy-confinement-times and a transition to a divertor solution with metal walls.

Data processing for soft X-ray diagnostics based on GEM detector measurements for fusion plasma imaging

NASA Astrophysics Data System (ADS)

Czarski, T.; Chernyshova, M.; Pozniak, K. T.; Kasprowicz, G.; Byszuk, A.; Juszczyk, B.; Wojenski, A.; Zabolotny, W.; Zienkiewicz, P.

2015-12-01

The measurement system based on GEM - Gas Electron Multiplier detector is developed for X-ray diagnostics of magnetic confinement fusion plasmas. The Triple Gas Electron Multiplier (T-GEM) is presented as soft X-ray (SXR) energy and position sensitive detector. The paper is focused on the measurement subject and describes the fundamental data processing to obtain reliable characteristics (histograms) useful for physicists. So, it is the software part of the project between the electronic hardware and physics applications. The project is original and it was developed by the paper authors. Multi-channel measurement system and essential data processing for X-ray energy and position recognition are considered. Several modes of data acquisition determined by hardware and software processing are introduced. Typical measuring issues are deliberated for the enhancement of data quality. The primary version based on 1-D GEM detector was applied for the high-resolution X-ray crystal spectrometer KX1 in the JET tokamak. The current version considers 2-D detector structures initially for the investigation purpose. Two detector structures with single-pixel sensors and multi-pixel (directional) sensors are considered for two-dimensional X-ray imaging. Fundamental output characteristics are presented for one and two dimensional detector structure. Representative results for reference source and tokamak plasma are demonstrated.

Compact ECEI system with in-vessel reflective optics for WEST.

PubMed

Nam, Y B; Park, H K; Lee, W; Yun, G S; Kim, M; Sabot, R; Elbeze, D; Lotte, P; Shen, J

2016-11-01

An electron cyclotron emission imaging (ECEI) diagnostic system for WEST (W Environment for Steady state Tokamak) is under development to study the MHD instabilities affected by tungsten impurities. The system will provide 2-D T e fluctuation images (width × height = ∼18 cm × ∼ 34 cm at low field side and ∼13 cm × ∼ 39 cm at high field side) from a poloidal cross section with high spatial (≤1.7 cm) and temporal (≤2 μs) resolutions. While the key concept and electronic structure are similar to that of prior ECEI systems on other tokamak devices such as KSTAR, DIII-D, or ASDEX-U, part of the imaging optics have to be placed inside the vacuum vessel in order to resolve issues on limited installation space and longer beam path to the detector position. The in-vessel optics consisting of two large curvature-radius mirrors are expected to withstand the extreme heating on long-pulse operation scenario (∼1000 s). The out-vessel optical housing is constructed as compact as possible to remove easily from the installation site in case of necessity. Commissioning of the system is scheduled on the second experimental WEST campaign end of 2017.

Application of very high harmonic fast waves for off-axis current drive in the DIII-D and FNSF-AT tokamaks

DOE PAGES

Prater, Ronald; Moeller, Charles P.; Pinsker, Robert I.; ...

2014-06-26

Fast waves at frequencies far above the ion cyclotron frequency and approaching the lower hybrid frequency (also called “helicons” or “whistlers”) have application to off-axis current drive in tokamaks with high electron beta. The high frequency causes the whistler-like behavior of the wave power nearly following field lines, but with a small radial component, so the waves spiral slowly toward the plasma center. The high frequency also contributes to strong damping. Modeling predicts robust off-axis current drive with good efficiency compared to alternatives in high performance discharges in DIII-D and Fusion Nuclear Science Facility (FNSF) when the electron beta ismore » above about 1.8%. Detailed analysis of ray behavior shows that ray trajectories and damping are deterministic (that is, not strongly affected by plasma profiles or initial ray conditions), unlike the chaotic ray behavior in lower frequency fast wave experiments. Current drive was found to not be sensitive to the launched value of the parallel index of refraction n||, so wave accessibility issues can be reduced. Finally, use of a traveling wave antenna provides a very narrow n|| spectrum, which also helps avoid accessibility problems.« less

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

NASA Astrophysics Data System (ADS)

Wang, Y.; Tobias, B.; Chang, Y.-T.; Yu, J.-H.; Li, M.; Hu, F.; Chen, M.; Mamidanna, M.; Phan, T.; Pham, A.-V.; Gu, J.; Liu, X.; Zhu, Y.; Domier, C. W.; Shi, L.; Valeo, E.; Kramer, G. J.; Kuwahara, D.; Nagayama, Y.; Mase, A.; Luhmann, N. C., Jr.

2017-07-01

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. Microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These have the potential to greatly advance microwave fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfvén eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today’s most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

DOE PAGES

Wang, Y.; Tobias, B.; Chang, Y. -T.; ...

2017-03-14

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Amiya K.

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 difficultmore » 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 neoclassical values by combined mechanisms of ExB and diamagnetic flow shear suppression of the ion temperature gradient (ITG) instabilities. However, even when the ion transport is strongly suppressed, the electron transport remains highly anomalous. The most plausible physics scenario for the anomalous electron transport is based on electron temperature gradient (ETG) instabilities. This instability is an electron analog of and nearly isomorphic to the ITG instability, which we had studied before extensively. However, this isomorphism is broken nonlinearily. It is noted that as the typical ETG mode growth rates are larger (in contrast to ITG modes) than ExB shearing rates in usual tokamaks, the flow shear suppression of ETG modes is highly unlikely. This motivated a broader range of investigations of other physics scenarios of nonlinear saturation and transport scaling of ETG modes.« less

Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

NASA Astrophysics Data System (ADS)

Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

2017-10-01

The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

Feasibility of AN Ecrh System for Jet:. Plant Layout, Auxiliaries and Services

NASA Astrophysics Data System (ADS)

Lennholm, M.; Bouquey, F.; Braune, H.; Farthing, J.; Garavaglia, S.; Giruzzi, G.; Granucci, G.; Jennison, M.; Parkin, A.

2011-02-01

A study conducted over the last year to asses the desirability and feasibility of installing an ECRH system on the JET tokamak has concluded that such a system is indeed both desirable and feasible. Details of physics studies, launcher and transmission line design, and power supplies are presented elsewhere in these proceedings. This paper concentrates on the logistical implications of installing this system at JET. The paper addresses issues such as port allocation and plant location. The study has concluded that a new building will be needed to house the ECRH plant. Building layout proposals are presented together with considerations regarding the required auxiliary equipment.

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Superconducting magnet development for tokamaks and mirrors: a technical assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laverick, C.; Jacobs, R. B.; Boom, R. W.

1977-11-01

The role of superconducting magnets in Magnetic Fusion Energy Research and Development is assessed from a consideration of program plans and schedules, the present status of the programs and the research and development suggestions arising from recent studies and workshops. A principal conclusion is that the large superconducting magnet systems needed for commercial magnetic fusion reactors can be constructed. However such magnets working under severe conditions, with increasingly stringent reliability, safety and cost restrictions can never be built unless experience is first gained in a number of important installations designed to prove physics and technology steps on the way tomore » commercial power demonstration. The immediate problem is to design a technology program in the absence of definite device needs and specifications, giving a priority weighting to the multiplicity of good, high quality development program suggestions when all proposals cannot be supported.« less

What is the fate of runaway positrons in tokamaks?

DOE PAGES

Liu, Jian; Qin, Hong; Fisch, Nathaniel J.; ...

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.

Geodesic acoustic modes in noncircular cross section tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P.; Konovaltseva, L. V.

2017-03-15

The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.

What is the fate of runaway positrons in tokamaks?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Jian; Qin, Hong, E-mail: hongqin@ustc.edu.cn; Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543

2014-06-15

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.

Gyrokinetic-Vlasov simulations of the ion temperature gradient turbulence in tokamak and helical systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watanabe, T.-H.; Sugama, H.; Graduate University for Advanced Studies

2006-11-30

Recent progress of the gyrokinetic-Vlasov simulations on the ion temperature gradient (ITG) turbulence in tokamak and helical systems is reported, where the entropy balance is checked as a reference for the numerical accuracy. The tokamak ITG turbulence simulation carried out on the Earth Simulator clearly captures a nonlinear generation process of zonal flows. The tera-flops and tera-bytes scale simulation is also applied to a helical system with the same poloidal and toroidal periodicities of L = 2 and M = 10 as in the Large Helical Device.

Final technical report for DE-SC00012633 AToM (Advanced Tokamak Modeling)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, Christopher; Orlov, Dmitri; Izzo, Valerie

This final report for the AToM project documents contributions from University of California, San Diego researchers over the period of 9/1/2014 – 8/31/2017. The primary focus of these efforts was on performing validation studies of core tokamak transport models using the OMFIT framework, including development of OMFIT workflow scripts. Additional work was performed to develop tools for use of the nonlinear magnetohydrodynamics code NIMROD in OMFIT, and its use in the study of runaway electron dynamics in tokamak disruptions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Conversion of magnetic energy to runaway kinetic energy during the termination of runaway current on the J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Dai, A. J.; Chen, Z. Y.; Huang, D. W.; Tong, R. H.; Zhang, J.; Wei, Y. N.; Ma, T. K.; Wang, X. L.; Yang, H. Y.; Gao, H. L.; Pan, Y.; the J-TEXT Team

2018-05-01

A large number of runaway electrons (REs) with energies as high as several tens of mega-electron volt (MeV) may be generated during disruptions on a large-scale tokamak. The kinetic energy carried by REs is eventually deposited on the plasma-facing components, causing damage and posing a threat on the operation of the tokamak. The remaining magnetic energy following a thermal quench is significant on a large-scale tokamak. The conversion of magnetic energy to runaway kinetic energy will increase the threat of runaway electrons on the first wall. The magnetic energy dissipated inside the vacuum vessel (VV) equals the decrease of initial magnetic energy inside the VV plus the magnetic energy flowing into the VV during a disruption. Based on the estimated magnetic energy, the evolution of magnetic-kinetic energy conversion are analyzed through three periods in disruptions with a runaway current plateau.

Kinetic shear Alfvén instability in the presence of impurity ions in tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, Gaimin; Shen, Y.; Xie, T.

2013-10-15

The effects of impurity ions on the kinetic shear Alfvén (KSA) instability in tokamak plasmas are investigated by numerically solving the integral equations for the KSA eigenmode in the toroidal geometry. The kinetic effects of hydrogen and impurity ions, including transit motion, finite ion Larmor radius, and finite-orbit-width, are taken into account. Toroidicity induced linear mode coupling is included through the ballooning-mode representation. Here, the effects of carbon, oxygen, and tungsten ions on the KSA instability in toroidal plasmas are investigated. It is found that, depending on the concentration and density profile of the impurity ions, the latter can bemore » either stabilizing or destabilizing for the KSA modes. The results here confirm the importance of impurity ions in tokamak experiments and should be useful for analyzing experimental data as well as for understanding anomalous transport and control of tokamak plasmas.« less

ETF system code: composition and applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reid, R.L.; Wu, K.F.

1980-01-01

A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies,more » such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system.« less

Saturated internal instabilities in advanced-tokamak plasmas

NASA Astrophysics Data System (ADS)

Hua, M.-D.; Chapman, I. T.; Pinches, S. D.; Hastie, R. J.; MAST Team

2010-06-01

"Advanced tokamak" (AT) scenarios were developed with the aim of reaching steady-state operation in future potential tokamak fusion power plants. AT scenarios exhibit non-monotonic to flat safety factor profiles (q, a measure of the magnetic field line pitch), with the minimum q (qmin) slightly above an integer value (qs). However, it has been predicted that these q profiles are unstable to ideal magnetohydrodynamic instabilities as qmin approaches qs. These ideal instabilities, observed and diagnosed as such for the first time in MAST plasmas with AT-like q profiles, have far-reaching consequences like confinement degradation, flattening of the toroidal core rotation or enhanced fast ion losses. These observations motivate the stability analysis of advanced-tokamak plasmas, with a view to provide guidance for stability thresholds in AT scenarios. Additionally, the measured rotation damping is compared to the self-consistently calculated predictions from neoclassical toroidal viscosity theory.

Theoretical study on the laser-driven ion-beam trace probe in toroidal devices with large poloidal magnetic field

NASA Astrophysics Data System (ADS)

Yang, X.; Xiao, C.; Chen, Y.; Xu, T.; Yu, Y.; Xu, M.; Wang, L.; Wang, X.; Lin, C.

2018-03-01

Recently, a new diagnostic method, Laser-driven Ion-beam Trace Probe (LITP), has been proposed to reconstruct 2D profiles of the poloidal magnetic field (Bp) and radial electric field (Er) in the tokamak devices. A linear assumption and test particle model were used in those reconstructions. In some toroidal devices such as the spherical tokamak and the Reversal Field Pinch (RFP), Bp is not small enough to meet the linear assumption. In those cases, the error of reconstruction increases quickly when Bp is larger than 10% of the toroidal magnetic field (Bt), and the previous test particle model may cause large error in the tomography process. Here a nonlinear reconstruction method is proposed for those cases. Preliminary numerical results show that LITP could be applied not only in tokamak devices, but also in other toroidal devices, such as the spherical tokamak, RFP, etc.

Destruction of tungsten limiters in the T-10 Tokamak under high plasma heat loads

NASA Astrophysics Data System (ADS)

Grashin, S. A.; Arkhipov, I. I.; Budaev, V. P.; Giniyatulin, R. N.; Karpov, A. V.; Klyuchnikov, L. A.; Krupin, V. A.; Litunovskiy, N. V.; Masul, I. V.; Makhankov, F. N.; Martynenko, Yu V.; Sarytchev, D. V.; Solomatin, R. Yu; Khimchenko, L. N.

2017-10-01

Tungsten limiters were tested in the T-10 tokamak. The limiters were made from the ITER-grade WMP “POLEMA” tungsten. The influence of the edge tokamak plasma on tungsten limiters leads to significant cracking of tungsten. The heat load of up to 2 MW · m-2 leads to the micro-crack development at the grain boundaries accompanied by the loss of grains. The heat loads that exceed 5 MW · m-2 lead to the macro crack development. Under the present T-10 tokamak conditions, the heat and particle fluxes in the edge plasma lead to the significant destruction of tungsten limiters during the experimental campaign. During the disruption and runaway electron formation, extreme heat loads of more than 1 GW/m2 cause strong melting of tungsten on the inner and outer part of the ring limiter.

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

Macroscopic erosion of divertor and first wall armour in future tokamaks

NASA Astrophysics Data System (ADS)

Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-12-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

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.

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 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. PMID:24892099

Development of Numerical Methods to Estimate the Ohmic Breakdown Scenarios of a Tokamak

NASA Astrophysics Data System (ADS)

Yoo, Min-Gu; Kim, Jayhyun; An, Younghwa; Hwang, Yong-Seok; Shim, Seung Bo; Lee, Hae June; Na, Yong-Su

2011-10-01

The ohmic breakdown is a fundamental method to initiate the plasma in a tokamak. For the robust breakdown, ohmic breakdown scenarios have to be carefully designed by optimizing the magnetic field configurations to minimize the stray magnetic fields. This research focuses on development of numerical methods to estimate the ohmic breakdown scenarios by precise analysis of the magnetic field configurations. This is essential for the robust and optimal breakdown and start-up of fusion devices especially for ITER and its beyond equipped with low toroidal electric field (ET <= 0.3 V/m). A field-line-following analysis code based on the Townsend avalanche theory and a particle simulation code are developed to analyze the breakdown characteristics of actual complex magnetic field configurations including the stray magnetic fields in tokamaks. They are applied to the ohmic breakdown scenarios of tokamaks such as KSTAR and VEST and compared with experiments.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar, E-mail: awadheshkrai@rediffmail.com

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 asmore » “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.« less

Enabling co-simulation of tokamak plant models and plasma control systems

DOE PAGES

Walker, M. L.

2017-12-22

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

Enabling co-simulation of tokamak plant models and plasma control systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walker, M. L.

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, 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, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr 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, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr 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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C. E.; Poli, F. M.; Ghantous, K.

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 βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 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 βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 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 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.« less

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

DOE PAGES

Jaworski, M. A.; Brooks, A.; Kaita, R.; ...

2016-08-08

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physicsmore » and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.« less

On current drive by Ohkawa mechanism of electron cyclotron wave in large inverse aspect ratio tokamaks

NASA Astrophysics Data System (ADS)

Zheng, Pingwei; Gong, Xueyu; Lu, Xingqiang; He, Lihua; Cao, Jingjia; Huang, Qianhong; Deng, Sheng

2018-03-01

A localized and efficient current drive method in the outer-half region of the tokamak with a large inverse aspect ratio is proposed via the Ohkawa mechanism of electron cyclotron (EC) waves. Further off-axis Ohkawa current drive (OKCD) via EC waves was investigated in high electron beta β e HL-2M-like tokamaks with a large inverse aspect ratio, and in EAST-like tokamaks with a low inverse aspect ratio. OKCD can be driven efficiently, and the driven current profile is spatially localized in the radial region, ranging from 0.62 to 0.85, where the large fraction of trapped electrons provides an excellent advantage for OKCD. Furthermore, the current drive efficiency increases with an increase in minor radius, and then drops when the minor radius beyond a certain value. The effect of trapped electrons greatly enhances the current driving capability of the OKCD mechanism. The highest current drive efficiency can reach 0.183 by adjusting the steering mirror to change the toroidal and poloidal incident angle, and the total driven current by OKCD can reach 20-32 kA MW-1 in HL-2M-like tokamaks. The current drive is less efficient for the EAST-like scenario due to the lower inverse aspect ratio. The results show that OKCD may be a valuable alternative current drive method in large inverse aspect ratio tokamaks, and the potential capabilities of OKCD can be used to suppress some important magnetohydrodynamics instabilities in the far off-axis region.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Footprint of Field lines

NASA Astrophysics Data System (ADS)

Figgins, Montoya; Ali, Halima; Punjabi, Alkesh

2000-10-01

We are working with the Simple Map^1 to find the footprint of field lines on the diverter plate in a single-null tokamak. Footprint of a field line is the position of the line when it escapes across the divertor plate. The Simple Map represents the magnetic field in a single-null divertor tokamak. The path of a field line is given by the equations: X_n+1=X_n-kY_n(1-Y_n) and Y_n+1=Y_n+kX_n+1. In order to find the footprint, we must first find the last good surface which is Y=0.997135768 and X=0. The value of k is fixed at 0.6. The starting values X0 are fixed at X_0=0. We use 10,000 points between the last good surface and the X-point. The X-point is located at (0,1). We also use the Continuous Analog of the Simple Map given by the equations: X(φ)=X_0-kY0 (1-Y_0)φ and Y(φ)=Y_0+kX(φ)φ. This will tell us what the (φ,X) is which represents the field lines crossing the divertor plate. The divertor plate is located at Y=1. When graphed, the footprint of field lines looks like the rings of Saturn. This work is supported by US DOES OFES. Ms. Montoya Figgins is HU CFRT Summer Fusion High School Scholar from E. E. Smith High School in North Carolina. She is supported by NASA under its NASA SHARP Plus Program. 1. Punjabi A, Verma A, and Boozer A, Phys Rev Lett, 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994)

The prospects for magnetohydrodynamic stability in advanced tokamak regimes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manickam, J.; Chance, M.S.; Jardin, S.C.

1994-05-01

Stability analysis of advanced regime tokamaks is presented. Here advanced regimes are defined to include configurations where the ratio of the bootstrap current, [ital I][sub BS], to the total plasma current, [ital I][sub [ital p

Current status and future R&D for reduced-activation ferritic/martensitic steels

NASA Astrophysics Data System (ADS)

Hishinuma, A.; Kohyama, A.; Klueh, R. L.; Gelles, D. S.; Dietz, W.; Ehrlich, K.

1998-10-01

International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe-(7-9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

Prediction of nonlinear evolution character of energetic-particle-driven instabilities

DOE PAGES

Duarte, Vinicius N.; Berk, H. L.; Gorelenkov, N. N.; ...

2017-03-17

A general criterion is proposed and found to successfully predict the emergence of chirping oscillations of unstable Alfvénic eigenmodes in tokamak plasma experiments. The model includes realistic eigenfunction structure, detailed phase-space dependences of the instability drive, stochastic scattering and the Coulomb drag. The stochastic scattering combines the effects of collisional pitch angle scattering and micro-turbulence spatial diffusion. Furthermore, the latter mechanism is essential to accurately identify the transition between the fixed-frequency mode behavior and rapid chirping in tokamaks and to resolve the disparity with respect to chirping observation in spherical and conventional tokamaks.

Prediction of nonlinear evolution character of energetic-particle-driven instabilities

NASA Astrophysics Data System (ADS)

Duarte, V. N.; Berk, H. L.; Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; Nazikian, R.; Pace, D. C.; Podestà, M.; Tobias, B. J.; Van Zeeland, M. A.

2017-05-01

A general criterion is proposed and found to successfully predict the emergence of chirping oscillations of unstable Alfvénic eigenmodes in tokamak plasma experiments. The model includes realistic eigenfunction structure, detailed phase-space dependences of the instability drive, stochastic scattering and the Coulomb drag. The stochastic scattering combines the effects of collisional pitch angle scattering and micro-turbulence spatial diffusion. The latter mechanism is essential to accurately identify the transition between the fixed-frequency mode behavior and rapid chirping in tokamaks and to resolve the disparity with respect to chirping observation in spherical and conventional tokamaks.

Anomalous transport scaling in the DIII-D tokamak matched by supercomputer simulation.

PubMed

Candy, J; Waltz, R E

2003-07-25

Gyrokinetic simulation of tokamak transport has evolved sufficiently to allow direct comparison of numerical results with experimental data. It is to be emphasized that only with the simultaneous inclusion of many distinct and complex effects can this comparison realistically be made. Until now, numerical studies of tokamak microturbulence have been restricted to either (a) flux tubes or (b) electrostatic fluctuations. Using a newly developed global electromagnetic solver, we have been able to recover via direct simulation the Bohm-like scaling observed in DIII-D L-mode discharges. We also match, well within experimental uncertainty, the measured energy diffusivities.

Measurement of H/D ratio and ion temperature on a HT-6M Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wei, L.; Lin, X.

1997-01-01

By combining optical fibers with piezoelectric scanning Fabry{endash}Perot interferometer, the profiles of H{sub {alpha}} and D{sub {alpha}} have been determined simultaneously in a single Tokamak discharge. Consequently, the ratio of hydrogen to deuterium and ion temperature are obtained. Not only is the uncertainty of shot-to-shot avoided, the results of the experiment indicate that this instrumentation has the advantage of rapid wavelength scanning, large dispersion, high resolution, and good adaptability of working in adverse circumstances such as at a Tokamak site. {copyright} {ital 1997 American Institute of Physics.}

A fluid model for the edge pressure pedestal height and width in tokamaks based on the transport constraint of particle, energy, and momentum balance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stacey, W. M., E-mail: weston.stacey@nre.gatech.edu

2016-06-15

A fluid model for the tokamak edge pressure profile required by the conservation of particles, momentum and energy in the presence of specified heating and fueling sources and electromagnetic and geometric parameters has been developed. Kinetics effects of ion orbit loss are incorporated into the model. The use of this model as a “transport” constraint together with a “Peeling-Ballooning (P-B)” instability constraint to achieve a prediction of edge pressure pedestal heights and widths in future tokamaks is discussed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Dong-Hwan; Hong, Suk-Ho; National Fusion Research Institute

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 reliabilitymore » of the method.« less

Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration, operation, improvements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohm, P., E-mail: bohm@ipp.cas.cz; Bilkova, P.; Melich, R.

2014-11-15

The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to ∼1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles.

Li Experiments at the Tokamak T-11M Toward PFC Concept of Steady State Tokamak-Reactor

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2009-11-01

As practical method of using a liquid lithium as a renewable plasma-facing component (PCF) for steady state tokamak-reactor the concept of lithium emitter-collector is considered [1]. It is based on lithium filled capillary porous system proposed by V.A. Evtikhin et al. (1996). The lithium circulation process consists of four steps: (1) Li emission from the PFC emitter into the plasma; (2) plasma boundary cooling by non-coronal Li radiation; (3) Li ion capture by the collector (before they are lost to the tokamak chamber wall); (4) Li return from the collector to the emitter. T-11M tokamak experiments have used three local rail limiters made from lithium, molybdenum and graphite as lithium collectors. The lithium behavior was studied by analysis of the witness samples, and by a mobile graphite probe. The key findings are: (1) lithium collection on the ion side of the lithium limiter is 2-3 times larger than on the electron side; (2) total efficiency of Li collection integrated over all three rail limiters can reach 50-70% of the lithium emission during the discharge pulse, while the theoretical limit is about 90%. [1] S.V. Mirnov, J. Nucl. Mat., 390-391, 876 (2009).

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ko, J., E-mail: jinseok@nfri.re.kr

2016-11-15

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 ismore » 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, q{sub 0}, during the sawtooth events.« less

An enhanced tokamak startup model

NASA Astrophysics Data System (ADS)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

Hiro and Evans currents in Vertical Disruption Event

NASA Astrophysics Data System (ADS)

Zakharov, Leonid; Xujing Li Team; Sergei Galkin Team

2014-10-01

The notion of Tokamak Magneto-Hydrodynamics (TMHD), which explicitly reflects the anisotropy of a high temperature tokamak plasma is introduced. The set of TMHD equations is formulated for simulations of macroscopic plasma dynamics and disruptions in tokamaks. Free from the Courant restriction on the time step, this set of equations is appropriate for high performance plasmas and does not require any extension of the MHD plasma model. At the same time, TMHD requires the use of magnetic field aligned numerical grids. The TMHD model was used for creation of theory of the Wall Touching Kink and Vertical Modes (WTKM and WTVM), prediction of Hiro and Evans currents, design of an innovative diagnostics for Hiro current measurements, installed on EAST device. While Hiro currents have explained the toroidal asymmetry in the plasma current measurements in JET disruptions, the Evans currents explain the tile current measurements in tokamaks. The recently developed Vertical Disruption Code (VDE) have demonstrated 5 regimes of VDE and confirmed the generation of both Hiro and Evans currents. The results challenge the 24 years long misinterpretation of the tile currents in tokamaks as ``halo'' currents, which were a product of misuse of equilibrium reconstruction for VDE. This work is supported by US DoE Contract No. DE-AC02-09-CH1146.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 4.Technology and Material Research in Fusion Power Plant Development

NASA Astrophysics Data System (ADS)

Akiba, Masato; Matsui, Hideki; Takatsu, Hideyuki; Konishi, Satoshi

Technical issues regarding the fusion power plant that are required to be developed in the period of ITER construction and operation, both with ITER and with other facilities that complement ITER are described in this section. Three major fields are considered to be important in fusion technology. Section 4.1 summarizes blanket study, and ITER Test Blanket Module (TBM) development that focuses its effort on the first generation power blanket to be installed in DEMO. ITER will be equipped with 6 TBMs which are developed under each party's fusion program. In Japan, the solid breeder using water as a coolant is the primary candidate, and He-cooled pebble bed is the alternative. Other liquid options such as LiPb, Li or molten salt are developed by other parties' initiatives. The Test Blanket Working Group (TBWG) is coordinating these efforts. Japanese universities are investigating advanced concepts and fundamental crosscutting technologies. Section 4.2 introduces material development and particularly, the international irradiation facility, IFMIF. Reduced activation ferritic/martensitic steels are identified as promising candidates for the structural material of the first generation fusion blanket, while and vanadium alloy and SiC/SiC composite are pursued as advanced options. The IFMIF is currently planning the next phase of joint activity, EVEDA (Engineering Validation and Engineering Design Activity) that encompasses construction. Material studies together with the ITER TBM will provide essential technical information for development of the fusion power plant. Other technical issues to be addressed regarding the first generation fusion power plant are summarized in section 4.3. Development of components for ITER made remarkable progress for the major essential technology also necessary for future fusion plants, however many still need further improvements toward power plant. Such areas includes; the divertor, plasma heating/current drive, magnets, tritium, and remote handling. There remain many other technical issues for power plant which require integrated efforts.

Recent Progress on Spherical Torus Research and Implications for Fusion Energy Development Path

NASA Astrophysics Data System (ADS)

Ono, Masayuki

2014-10-01

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A =R0 / a) reduced to A near 1.5, well below the normal tokamak operating range of A equal to 2.5 or greater. As the aspect ratio is reduced, the ideal tokamak beta (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as 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 plasma elongation 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 the longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in the UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all areas of fusion research, including fundamental fusion energy science as well as technological innovation. These results suggest exciting future prospects for ST research in both the near and longer term. The talk will summarize the key physics results from worldwide ST experiments, and describe ST community plans to provide the database for FNSF design while improving predictive capabilities for ITER and beyond. This work supported by DoE Contract No. DE-AC02-09CH11466.

Physics objectives of PI3 spherical tokamak program

NASA Astrophysics Data System (ADS)

Howard, Stephen; Laberge, Michel; Reynolds, Meritt; O'Shea, Peter; Ivanov, Russ; Young, William; Carle, Patrick; Froese, Aaron; Epp, Kelly

2017-10-01

Achieving net energy gain with a Magnetized Target Fusion (MTF) system requires the initial plasma state to satisfy a set of performance goals, such as particle inventory (1021 ions), sufficient magnetic flux (0.3 Wb) to confine the plasma without MHD instability, and initial energy confinement time several times longer than the compression time. General Fusion (GF) is now constructing Plasma Injector 3 (PI3) to explore the physics of reactor-scale plasmas. Energy considerations lead us to design around an initial state of Rvessel = 1 m. PI3 will use fast coaxial helicity injection via a Marshall gun to create a spherical tokamak plasma, with no additional heating. MTF requires solenoid-free startup with no vertical field coils, and will rely on flux conservation by a metal wall. PI3 is 5x larger than SPECTOR so is expected to yield magnetic lifetime increase of 25x, while peak temperature of PI3 is expected to be similar (400-500 eV) Physics investigations will study MHD activity and the resistive and convective evolution of current, temperature and density profiles. We seek to understand the confinement physics, radiative loss, thermal and particle transport, recycling and edge physics of PI3.

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Brooks; A.H. Reiman; G.H. Neilson

High-beta, low-aspect-ratio (compact) stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady-state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has amore » substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.« less

Initial Simulations of RF Waves in Hot Plasmas Using the FullWave Code

NASA Astrophysics Data System (ADS)

Zhao, Liangji; Svidzinski, Vladimir; Spencer, Andrew; Kim, Jin-Soo

2017-10-01

FullWave is a simulation tool that models RF fields in hot inhomogeneous magnetized plasmas. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. In an rf field, the hot plasma dielectric response is limited to the distance of a few particles' Larmor radii, near the magnetic field line passing through the test point. The localization of the hot plasma dielectric response results in a sparse matrix of the problem thus significantly reduces the size of the problem and makes the simulations faster. We will present the initial results of modeling of rf waves using the Fullwave code, including calculation of nonlocal conductivity kernel in 2D Tokamak geometry; the interpolation of conductivity kernel from test points to adaptive cloud of computational points; and the results of self-consistent simulations of 2D rf fields using calculated hot plasma conductivity kernel in a tokamak plasma with reduced parameters. Work supported by the US DOE ``SBIR program.

High heat flux testing of CFC composites for the tokamak physics experiment

NASA Astrophysics Data System (ADS)

Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

1996-10-01

High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

Synchrotron emission diagnostic of full-orbit kinetic simulations of runaway electrons in tokamaks plasmas

NASA Astrophysics Data System (ADS)

Carbajal Gomez, Leopoldo; Del-Castillo-Negrete, Diego

2017-10-01

Developing avoidance or mitigation strategies of runaway electrons (RE) for the safe operation of ITER is imperative. Synchrotron radiation (SR) of RE is routinely used in current tokamak experiments to diagnose RE. We present the results of a newly developed camera diagnostic of SR for full-orbit kinetic simulations of RE in DIII-D-like plasmas that simultaneously includes: full-orbit effects, information of the spectral and angular distribution of SR of each electron, and basic geometric optics of a camera. We observe a strong dependence of the SR measured by the camera on the pitch angle distribution of RE, namely we find that crescent shapes of the SR on the camera pictures relate to RE distributions with small pitch angles, while ellipse shapes relate to distributions of RE with larger pitch angles. A weak dependence of the SR measured by the camera with the RE energy, value of the q-profile at the edge, and the chosen range of wavelengths is found. Furthermore, we observe that oversimplifying the angular distribution of the SR changes the synchrotron spectra and overestimates its amplitude. Research sponsored by the LDRD Program of ORNL, managed by UT-Battelle, LLC, for the U. S. DoE.

The Effects of Aperiodic Perturbation on the Last Good Surface of a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Jordan, Joseph; Punjabi, Alkesh; Ali, Halima

2003-10-01

An area-preserving discrete mapping, called the Simple Map is used to analyze the effects of aperiodic perturbation on the last good magnetic surface of a single-null divertor tokamak. The SM was developed by Punjabi and Boozer /1/. The SM equations are modified to reflect the effect of aperiodic perturbation: Xn + 1 = Xn - (k+δ_aperiodics(n))Y_n(1-Y_n), Y_n+1 = Yn + kX_n+1 where δ_aperiodic is the amplitude of the aperiodic perturbation. The equations are coded into FORTRAN and used in conjunction with a graphing program to show the trajectories for different amplitudes of aperiodic perturbation. As the aperiodic perturbation is increased, the trajectories break from a clean curve and develop strange behavior as they approach chaos. The progression of the trajectories from a clean curve to a chaotic one is observed, and the results are presented in this paper. This project is supported by NASA SHARP and US DOE Grant Number DE-F02-02ER54673. The research was done under the mentorship of Profs. Punjabi and Ali. 1. Punjabi A., Verma A., and Boozer A., Phys. Rev. Lett, 69, 3322 (1992)

Origin of Non-Gaussian Spectra Observed via the Charge Exchange Recombination Spectroscopy Diagnostic in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Sulyman, Alex; Chrystal, Colin; Haskey, Shaun; Burrell, Keith; Grierson, Brian

2017-10-01

The possible observation of non-Maxwellian ion distribution functions in the pedestal of DIII-D will be investigated with a synthetic diagnostic that accounts for the effect of finite neutral beam size. Ion distribution functions in tokamak plasmas are typically assumed to be Maxwellian, however non-Gaussian features observed in impurity charge exchange spectra have challenged this concept. Two possible explanations for these observations are spatial averaging over a finite beam size and a local ion distribution that is non-Maxwellian. Non-Maxwellian ion distribution functions could be driven by orbit loss effects in the edge of the plasma, and this has implications for momentum transport and intrinsic rotation. To investigate the potential effect of finite beam size on the observed spectra, a synthetic diagnostic has been created that uses FIDAsim to model beam and halo neutral density. Finite beam size effects are investigated for vertical and tangential views in the core and pedestal region with varying gradient scale lengths. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program, DE-FC02-04ER54698, and DE-AC02-09CH11466.

Analysis of the Zeeman effect on D α spectra on the EAST tokamak

NASA Astrophysics Data System (ADS)

Gao, Wei; Huang, Juan; Wu, Chengrui; Xu, Zong; Hou, Yumei; Jin, Zhao; Chen, Yingjie; Zhang, Pengfei; Zhang, Ling; Wu, Zhenwei; EAST Team

2017-04-01

Based on the passive spectroscopy, the {{{D}}}α atomic emission spectra in the boundary region of the plasma have been measured by a high resolution optical spectroscopic multichannel analysis (OSMA) system in EAST tokamak. The Zeeman splitting of the {{{D}}}α spectral lines has been observed. A fitting procedure by using a nonlinear least squares method was applied to fit and analyze all polarization π and +/- σ components of the {{{D}}}α atomic spectra to acquire the information of the local plasma. The spectral line shape was investigated according to emission spectra from different regions (e.g., low-field side and high-field side) along the viewing chords. Each polarization component was fitted and classified into three energy categories (the cold, warm, and hot components) based on different atomic production processes, in consistent with the transition energy distribution by calculating the gradient of the {{{D}}}α spectral profile. The emission position, magnetic field intensity, and flow velocity of a deuterium atom were also discussed in the context. Project supported by the National Natural Science Foundation of China (Grant Nos. 11275231 and 11575249) and the National Magnetic Confinement Fusion Energy Research Program of China (Grant No. 2015GB110005).

Overview of MST Research

NASA Astrophysics Data System (ADS)

Sarff, J. S.

2016-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with ohmic heating and minimal external magnetization, (2) predictive capability in toroidal confinement physics, and (3) basic plasma physics is summarized. Validation of key plasma models is a program priority. Programmable power supplies (PPS) are being developed to maximize inductive capability. Well-controlled flattops with current as low as 0.02 MA are produced with an existing PPS, and Ip <= 0.8 MA is anticipated with a second PPS under construction. The Lundquist number spans S =10(4 - 9) for 0.02-0.8 MA, allowing nonlinear MHD validation using NIMROD and DEBS at low S to be connected to highest S experiments. The PPS also enables MST tokamak operation for studying transients and runaway electron suppression with RMPs. Gyrokinetic modeling with GENE predicts unstable TEM in improved-confinement plasmas. Fluctuations are measured with TEM properties including a density-gradient threshold larger than for tokamak plasmas. Probe measurements hint that drift waves are also excited via the turbulent cascade in standard RFP plasmas. Turbulent energization of an electron tail occurs during sawtooth reconnection. New diagnostics are being developed to measure the energetic ion profile and transport from EP instabilities with NBI. Supported by US DoE and NSF.

Alpha-channeling simulation experiment in the DIII-D tokamak.

PubMed

Wong, K L; Budny, R; Nazikian, R; Petty, C C; Greenfield, C M; Heidbrink, W W; Ruskov, E

2004-08-20

Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

ECRH Studies on Tokamak Plasmas.

DTIC Science & Technology

1980-10-10

r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

Stabilization of the Vertical Mode in Tokamaks by Localized Nonaxisymmetric Fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reiman, A.

Vertical instability of a tokamak plasma can be controlled by nonaxisymmetric magnetic fields localized near the plasma edge at the bottom and top of the torus. The required magnetic fields can be produced by a relatively simple set of parallelogram-shaped coils.

Dust-Particle Transport in Tokamak Edge Plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K

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 comprehensivemore » 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.« less

Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma

NASA Astrophysics Data System (ADS)

Alipour, Ramin; Ghanbari, Mohamad R.

2018-04-01

Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less than them at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level.

Plasma density injection and flow during coaxial helicity injection in a tokamak

NASA Astrophysics Data System (ADS)

Hooper, E. B.

2018-02-01

Whole device, resistive MHD simulations of spheromaks and tokamaks have used a large diffusion coefficient that maintains a nearly constant density throughout the device. In the present work, helicity and plasma are coinjected into a low-density plasma in a tokamak with a small diffusion coefficient. As in previous simulations [Hooper et al., Phys. Plasmas 20, 092510 (2013)], a flux bubble is formed, which expands to fill the tokamak volume. The injected plasma is non-uniform inside the bubble. The flow pattern is analyzed; when the simulation is not axisymmetric, an n = 1 mode on the surface of the bubble generates leakage of plasma into the low-density volume. Closed flux is generated following injection, as in experiments and previous simulations. The result provides a more detailed physics analysis of the injection, including density non-uniformities in the plasma that may affect its use as a startup plasma [Raman et al., Phys. Rev. Lett. 97, 175002 (2006)].

Time dependent 14 MeV neutrons measurement using a polycrystalline chemical vapor deposited diamond detector at the JET tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Angelone, M.; Pillon, M.; Bertalot, L.

A polycrystalline chemical vapor deposited (CVD) diamond detector was installed on a JET tokamak in order to monitor the time dependent 14 MeV neutron emission produced by D-T plasma pulses during the Trace Tritium Experiment (TTE) performed in October 2003. This was the first tentative ever attempted to use a CVD diamond detector as neutron monitor in a tokamak environment. Despite its small active volume, the detector was able to detect the 14 MeV neutron emission (>1.0x10{sup 15} n/shot) with good reliability and stability during the experimental campaign that lasted five weeks. The comparison with standard silicon detectors presently usedmore » at JET as 14 MeV neutron monitors is reported, showing excellent correlation between the measurements. The results prove that CVD diamond detectors can be reliably used in a tokamak environment and therefore confirm the potential of this technology for next step machines like ITER.« less

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

DOE PAGES

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; ...

2015-05-15

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10 x compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid,more » exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Finally, Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.« less

Simple Map with Low MN Perturbation for a Single-Null Divertor Tokamak with Constant Width of Stochastic Layer

NASA Astrophysics Data System (ADS)

Verma, Arun; Smith, Terry; Punjabi, Alkesh; Boozer, Allen

1996-11-01

In this work, we investigate the effects of low MN perturbations in a single-null divertor tokamak with stochastic scrape-off layer. The unperturbed magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). We choose the combinations of the map parameter k, and the strength of the low MN perturbation such that the width of stochastic layer remains unchanged. We give detailed results on the effects of low MN perturbation on the magnetic topology of the stochastic layer and on the footprint of field lines on the divertor plate given the constraint of constant width of the stochastic layer. The low MN perturbations occur naturally and therefore their effects are of considerable importance in tokamak divertor physics. This work is supported by US DOE OFES. Use of CRAY at HU and at NERSC is gratefully acknowledged.

Electron Cyclotron Current Drive Efficiency in General Tokamak Geometry and Its Application to Advanced Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Lin-Liu, Y. R.; Chan, V. S.; Luce, T. C.; Prater, R.

1998-11-01

Owing to relativistic mass shift in the cyclotron resonance condition, a simple and accurate interpolation formula for estimating the current drive efficiency, such as those(S.C. Chiu et al.), Nucl. Fusion 29, 2175 (1989).^,(D.A. Ehst and C.F.F. Karney, Nucl. Fusion 31), 1933 (1991). commonly used in FWCD, is not available in the case of ECCD. In this work, we model ECCD using the adjoint techniques. A semi-analytic adjoint function appropriate for general tokamak geometry is obtained using Fisch's relativistic collision model. Predictions of off-axis ECCD qualitatively and semi-quantitatively agrees with those of Cohen,(R.H. Cohen, Phys. Fluids 30), 2442 (1987). currently implemented in the raytracing code TORAY. The dependences of the current drive efficiency on the wave launch configuration and the plasma parameters will be presented. Strong absorption of the wave away from the resonance layer is shown to be an important factor in optimizing the off-axis ECCD for application to advanced tokamak operations.

Algorithms development for the GEM-based detection system

NASA Astrophysics Data System (ADS)

Czarski, T.; Chernyshova, M.; Malinowski, K.; Pozniak, K. T.; Kasprowicz, G.; Kolasinski, P.; Krawczyk, R.; Wojenski, A.; Zabolotny, W.

2016-09-01

The measurement system based on GEM - Gas Electron Multiplier detector - is developed for soft X-ray diagnostics of tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an Xray source. The focal measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the serial data acquisition is applied for the dynamic plasma diagnostics. The charge clusters are counted in the space determined by 2D position, charge value and time intervals. Radiation source characteristics are presented by histograms for a selected range of position, time intervals and cluster charge values corresponding to the energy spectra.

A self-consistent model of an isothermal tokamak

NASA Astrophysics Data System (ADS)

McNamara, Steven; Lilley, Matthew

2014-10-01

Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.

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.

Coherent nonlinear coupling between a long-wavelength mode and small-scale turbulence in the TEXT tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsui, H.Y.W.; Rypdal, K.; Ritz, C.P.

1993-04-26

Bispectral analysis of Langmuir probe data indicates that coherent nonlinear coupling, in addition to the noncoherent turbulent interactions, exists in the edge plasma of the tokamak TEXT. Not all the modes involved reside within the spectral region of the usual broadband turbulence. At a major resonant surface the small-scale turbulent activity interacts [ital coherently] with a localized long-wavelength mode; a signature of regular or coherent structure. By the observed coupling to the transport related turbulence, the long-wavelength mode can influence plasma confinement indirectly. These observations signify the influence of low-order resonant surfaces on the edge turbulence in tokamaks.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rasouli, C.; Abbasi Davani, F., E-mail: fabbasidavani@gmail.com

A series of experiments and numerical calculations have been done on the Damavand tokamak for accurate determination of equilibrium parameters, such as the plasma boundary position and shape. For this work, the pickup coils of the Damavand tokamak were recalibrated and after that a plasma boundary shape identification code was developed for analyzing the experimental data, such as magnetic probes and coils currents data. The plasma boundary position, shape and other parameters are determined by the plasma shape identification code. A free-boundary equilibrium code was also generated for comparison with the plasma boundary shape identification results and determination of requiredmore » fields to obtain elongated plasma in the Damavand tokamak.« less

Study of runaway electrons with Hard X-ray spectrometry of tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shevelev, A.; Chugunov, I.; Khilkevitch, E.

2014-08-21

Hard-X-ray spectrometry is a tool widely used for diagnostic of runaway electrons in existing tokamaks. In future machines, ITER and DEMO, HXR spectrometry will be useful providing information on runaway electron energy, runaway beam current and its profile during disruption.

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).

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

MHD Instabilities and Toroidal Field Effects on Plasma Column Behavior in Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khorshid, Pejman; Plasma Physics Research Center, Islamic Azad University, 14665-678, Tehran; Wang, L.

2006-12-04

In the edge plasma of the CT-6B and IRAN-T1 tokamaks the shape of plasma column based on MHD behavior has been studied. The bulk of plasma behavior during plasma column rotation as non-rigid body plasma has been investigated. We found that mode number and rotation frequency of plasma column are different in angle position, so that the mode number detected from Mirnov coils array located in poloidal angle on the inner side of chamber is more than outer side which it can be because of toroidal magnetic field effects. The results of IR-T1 and CT-6B tokamaks compared with each other,more » so that in the CT-6B because of its coils number must be less, but because of its Iron core the effect of toroidal magnetic field became more effective with respect to IR-T1. In addition, it is shown that the plasma column behaves as non-Rigid body plasma so that the poloidal rotation velocity variation in CT-6B is more than IR-T1. A relative correction for island rotation frequency has been suggested in connection with IRAN-T1 and CT-6B tokamak results, which can be considered for optical measurement purposes and also for future advanced tokamak control design.« less

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.

Scoping study for compact high-field superconducting net energy tokamaks

NASA Astrophysics Data System (ADS)

Mumgaard, R. T.; Greenwald, M.; Freidberg, J. P.; Wolfe, S. M.; Hartwig, Z. S.; Brunner, D.; Sorbom, B. N.; Whyte, D. G.

2016-10-01

The continued development and commercialization of high temperature superconductors (HTS) may enable the construction of compact, net-energy tokamaks. HTS, in contrast to present generation low temperature superconductors, offers improved performance in high magnetic fields, higher current density, stronger materials, higher temperature operation, and simplified assembly. Using HTS along with community-consensus confinement physics (H98 =1) may make it possible to achieve net-energy (Q>1) or burning plasma conditions (Q>5) in DIII-D or ASDEX-U sized, conventional aspect ratio tokamaks. It is shown that, by operating at high plasma current and density enabled by the high magnetic field (B>10T), the required triple products may be achieved at plasma volumes under 20m3, major radii under 2m, with external heating powers under 40MW. This is at the scale of existing devices operated by laboratories, universities and companies. The trade-offs in the core heating, divertor heat exhaust, sustainment, stability, and proximity to known plasma physics limits are discussed in the context of the present tokamak experience base and the requirements for future devices. The resulting HTS-based design space is compared and contrasted to previous studies on high-field copper experiments with similar missions. The physics exploration conducted with such HTS devices could decrease the real and perceived risks of ITER exploitation, and aid in quickly developing commercially-applicable tokamak pilot plants and reactors.

Observation of finite-. beta. MHD phenomena in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 = 1more » 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.« less

OMFIT Tokamak Profile Data Fitting and Physics Analysis

DOE PAGES

Logan, N. C.; Grierson, B. A.; Haskey, S. R.; ...

2018-01-22

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

OMFIT Tokamak Profile Data Fitting and Physics Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Logan, N. C.; Grierson, B. A.; Haskey, S. R.

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

Public Data Set: A Power-Balance Model for Local Helicity Injection Startup in a Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus G.

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.L. Barr et. al, 'A Power-Balance Model for Local Helicity Injection Startup in a Spherical Tokamak,' Nuclear Fusion 58, 076011 (2018).

[The sawtooth oscillation phenomenon of visible spectral signal in HT-6M Tokamak].

PubMed

Xu, W; Fang, Z; Wan, B; Li, J; Luo, J; Yin, F

1997-02-01

The sawtooth oscillation phenomenon of visible spectral signal in HT-6M Tokamak is presented. The influences of electron temperature, electron density and atomic ground density on the spectral signal discussed. This phenomenon results mainly from the change of electron temperature at the edge.

Final Report Advanced Quasioptical Launcher System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeffrey Neilson

2010-04-30

This program developed an analytical design tool for designing antenna and mirror systems to convert whispering gallery RF modes to Gaussian or HE11 modes. Whispering gallery modes are generated by gyrotrons used for electron cyclotron heating of fusion plasmas in tokamaks. These modes cannot be easily transmitted and must be converted to free space or waveguide modes compatible with transmission line systems.This program improved the capability of SURF3D/LOT, which was initially developed in a previous SBIR program. This suite of codes revolutionized quasi-optical launcher design, and this code, or equivalent codes, are now used worldwide. This program added functionality tomore » SURF3D/LOT to allow creating of more compact launcher and mirror systems and provide direct coupling to corrugated waveguide within the vacuum envelope of the gyrotron. Analysis was also extended to include full-wave analysis of mirror transmission line systems. The code includes a graphical user interface and is available for advanced design of launcher systems.« less

Analysis of electric current flow through the HTc multilayered superconductors

NASA Astrophysics Data System (ADS)

Sosnowski, J.

2016-02-01

Issue of the flow of the transport current through multilayered high-temperature superconductors is considered, depending on the direction of the electric current towards the surface of the superconducting CuO2 layers. For configuration of the current flow inside of the layers and for perpendicular magnetic field, it will be considered the current limitations connected with interaction of pancake type vortices with nano-sized defects, created among other during fast neutrons irradiation. So it makes this issue associated with work of nuclear energy devices, like tokamak ITER, LHC and actually developed accelerator Nuclotron-NICA, as well as cryocables. Phenomenological analysis of the pinning potential barrier formation will be in the paper given, which determines critical current flow inside the plane. Comparison of theoretical model with experimental data will be presented too as well as influence of fast neutrons irradiation dose on critical current calculated. For current direction perpendicular to superconducting planes the current-voltage characteristics are calculated basing on model assuming formation of long intrinsic Josephson's junctions in layered HTc superconductors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Virginia Finley

The results of the 1999 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 1999. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program ismore » to create innovations to make fusion power a practical reality--an alternative energy source. 1999 marked the first year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion reactors. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. The 1999 performance of the Princeton Plasma Physics Laboratory was rated ''outstanding'' by the U.S. Department of Energy in the Laboratory Appraisal report issued early in 2000. The report cited the Laboratory's consistently excellent scientific and technological achievements, its successful management practices, and included high marks in a host of other areas including environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey Department of Environmental Protection. PPPL monitored for the presence of non-radiological contaminants, mainly volatile organic compounds (components of degreasing solvents). Monitoring revealed the presence of low levels of volatile organic compounds in an area adjacent to PPPL. Also, PPPL's radiological monitoring program characterized the ambient, background levels of tritium in the environment and from the TFTR stack; the data are presented in this report.« less

Use of soft x-ray diagnostic on the COMPASS tokamak for investigations of sawteeth crash neighborhood and of plasma position using fast inversion methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Imrisek, M.; Faculty of Mathematics and Physics, Charles University in Prague, Prague; Weinzettl, V.

2014-11-15

The soft x-ray diagnostic is suitable for monitoring plasma activity in the tokamak core, e.g., sawtooth instability. Moreover, spatially resolved measurements can provide information about plasma position and shape, which can supplement magnetic measurements. In this contribution, fast algorithms with the potential for a real-time use are tested on the data from the COMPASS tokamak. In addition, the soft x-ray data are compared with data from other diagnostics in order to discuss possible connection between sawtooth instability on one side and the transition to higher confinement mode, edge localized modes and productions of runaway electrons on the other side.

Plasma rotation measurement in small tokamaks using an optical spectrometer and a single photomultiplier as detector.

PubMed

Severo, J H F; Nascimento, I C; Kuznetov, Yu K; Tsypin, V S; Galvão, R M O; Tendler, M

2007-04-01

The method for plasma rotation measurement in the tokamak TCABR is reported in this article. During a discharge, an optical spectrometer is used to scan sequentially spectral lines of plasma impurities and spectral lines of a calibration lamp. Knowing the scanning velocity of the diffraction grating of the spectrometer with adequate precision, the Doppler shifts of impurity lines are determined. The photomultiplier output voltage signals are recorded with adequate sampling rate. With this method the residual poloidal and toroidal plasma rotation velocities were determined, assuming that they are the same as those of the impurity ions. The results show reasonable agreement with the neoclassical theory and with results from similar tokamaks.

Features of self-organized plasma physics in tokamaks

NASA Astrophysics Data System (ADS)

Razumova, K. A.

2018-01-01

The history of investigations the role of self-organization processes in tokamak plasma confinement is presented. It was experimentally shown that the normalized pressure profile is the same for different tokamaks. Instead of the conventional Fick equation, where the thermal flux is proportional to a pressure gradient, processes in the plasma are well described by the Dyabilanin’s energy balance equation, in which the heat flux is proportional to the difference of normalized gradients for self-consistent and real pressure profiles. The transport coefficient depends on the values of heat flux, which compensates distortion of the pressure profile with external impacts. Radiative cooling of the plasma edge decreases the heat flux and improves the confinement.

Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma.

PubMed

Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H

2012-10-01

To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

Far-infrared laser diagnostics on the HT-6M tokamak

NASA Astrophysics Data System (ADS)

Gao, X.; Lu, H. J.; Guo, Q. L.; Wan, Y. X.; Tong, X. D.

1995-01-01

A multichannel far-infrared (FIR) hydrogen cyanide (HCN) laser interferometer was developed to measure plasma electron density profile on the HT-6M tokamak. The structure of the seven-channel FIR laser interferometer is described. The laser source used in the interferometer was a continuous-wave glow discharge HCN laser with a cavity length of 3.4 m and power output of about 100 mW at 337 μm. The detection sensitivity was 1/15 fringe with a temporal resolution of 0.1 ms. Experimental results were measured by the seven-channel FIR HCN laser interferometer with edge Ohmic heating, a pumping limiter, and ion cyclotron resonant heating on the HT-6M tokamak are reported.

A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

PubMed

Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

2011-09-23

New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

Can a Penning ionization discharge simulate the tokamak scrape-off plasma conditions?

NASA Technical Reports Server (NTRS)

Finkenthal, M.; Littman, A.; Stutman, D.; Kovnovich, S.; Mandelbaum, P.; Schwob, J. L.; Bhatia, A. K.

1990-01-01

The tokamak scrape-off (the region between the vacuum vessel wall and the magnetically confined fusion plasma edge), represents a source/sink for the hot fusion plasma. The electron densities and temperatures are in the ranges 10 to the 11th - 10 to the 13th/cu cm and 1-40 eV, respectively (depending on the size, magnetic field intensity and configuration, plasma current, etc). In the work reported, the electron temperature and density have been estimated in a Penning ionization discharge by comparing its spectroscopic emission in the VUV with that predicted by a collisional radiative model. An attempt to directly compare this emission with that of the tokamak edge is briefly described.

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

NASA Astrophysics Data System (ADS)

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

2018-02-01

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

DTT: a divertor tokamak test facility for the study of the power exhaust issues in view of DEMO

NASA Astrophysics Data System (ADS)

Albanese, R.; WPDTT2 Team; DTT Project Proposal Contributors, the

2017-01-01

In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€.

Development of a new gas puff imaging diagnostic on the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Yuan, B.; Xu, M.; Yu, Y.; Zang, L.; Hong, R.; Chen, C.; Wang, Z.; Nie, L.; Ke, R.; Guo, D.; Wu, Y.; Long, T.; Gong, S.; Liu, H.; Ye, M.; Duan, X.; HL-2A team

2018-03-01

A new gas puff imaging (GPI) diagnostic has been developed on the HL-2A tokamak to study two-dimensional plasma edge turbulence in poloidal vs. radial plane. During a discharge, neutral helium or deuterium gas is puffed at the edge of the plasma through a rectangular multi\

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).

Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

ScienceCinema

None

2018-01-16

From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

Public Data Set: Initiation and Sustainment of Tokamak Plasmas with Local Helicity Injection as the Majority Current Drive

DOE Data Explorer

Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609); Bodner, Grant M. [University of Wisconsin-Madison] (ORCID:0000000324979172); 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); Pachicano, Jessica L. [University of Wisconsin-Madison] (ORCID:0000000207255693); Pierren, Christopher [University of Wisconsin-Madison] (ORCID:0000000228289825); Reusch, Joshua A. [University of Wisconsin-Madison] (ORCID:0000000284249422); Rhodes, Alexander T. [University of Wisconsin-Madison] (ORCID:0000000280735714); Richner, Nathan J. [University of Wisconsin-Madison] (ORCID:0000000155443915); Rodriguez Sanchez, Cuauhtemoc [University of Wisconsin-Madison] (ORCID:0000000334712586); Schaefer, Carolyn E. [University of Wisconsin-Madison] (ORCID:0000000248848727); Weberski, Justin D. [University of Wisconsin-Madison] (ORCID:0000000256267914)

2018-05-22

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.M. Perry et al., 'Initiation and Sustainment of Tokamak Plasmas with Local Helicity Injection as the Majority Current Drive,' accepted for publication in Nuclear Fusion.

Public Data Set: Non-inductively Driven Tokamak Plasmas at Near-Unity βt in the Pegasus Toroidal Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.A. Reusch et al., 'Non-inductively Driven Tokamak Plasmas at Near-Unity βt in the Pegasus Toroidal Experiment,' Phys. Plasmas 25, 056101 (2018).

Edge-localized-modes in tokamaks

DOE PAGES

Leonard, Anthony W.

2014-09-11

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heatmore » flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.« less

Investigation of neutral particle dynamics in Aditya tokamak plasma with DEGAS2 code

NASA Astrophysics Data System (ADS)

Dey, Ritu; Ghosh, Joydeep; Chowdhuri, M. B.; Manchanda, R.; Banerjee, S.; Ramaiya, N.; Sharma, Deepti; Srinivasan, R.; Stotler, D. P.; Aditya Team

2017-08-01

Neutral particle behavior in Aditya tokamak, which has a circular poloidal ring limiter at one particular toroidal location, has been investigated using DEGAS2 code. The code is based on the calculation using Monte Carlo algorithms and is mainly used in tokamaks with divertor configuration. This code has been successfully implemented in Aditya tokamak with limiter configuration. The penetration of neutral hydrogen atom is studied with various atomic and molecular contributions and it is found that the maximum contribution comes from the dissociation processes. For the same, H α spectrum is also simulated and matched with the experimental one. The dominant contribution around 64% comes from molecular dissociation processes and neutral particle is generated by those processes have energy of ~2.0 eV. Furthermore, the variation of neutral hydrogen density and H α emissivity profile are analysed for various edge temperature profiles and found that there is not much changes in H α emission at the plasma edge with the variation of edge temperature (7-40 eV).

Avoiding Tokamak Disruptions by Applying Static Magnetic Fields That Align Locked Modes with Stabilizing Wave-Driven Currents [Avoiding Tokamak Disruptions by Magnetically Aligning Locked Modes with Stabilizing Wave-Driven Currents

DOE PAGES

Volpe, F. A.; Hyatt, Alan; La Haye, Robert J.; ...

2015-10-19

The international ITER tokamak has the objective of demonstrating the scientific feasibility of magnetic confinement fusion as a source of energy. A concern towards the achievement of this goal is represented by major disruptions: complete losses of confinement often initiated by a non-rotating ('locked') magnetic island created by magnetic reconnection. During disruptions, energy and particles accumulated in the plasma volume over many seconds are lost in a few milliseconds and released on the plasma-facing materials. In addition, multi-MA level currents flowing in the tokamak plasma for its sustainment and confinement are lost, also in milliseconds, thus terminating the plasma dischargemore » and causing electromagnetic stresses that, if unmitigated, could lead to excessive device wear. Moreover it is shown that magnetic perturbations can be used to avoid disruptions by "guiding" the magnetic island to lock in a position where it is accessible to millimetre wave beams that fully stabilize it.« less

Neutral helium beam probe

NASA Astrophysics Data System (ADS)

Karim, Rezwanul

1999-10-01

This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

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.

Investigation of neutral particle dynamics in Aditya tokamak plasma with DEGAS2 code

DOE PAGES

Dey, Ritu; Ghosh, Joydeep; Chowdhuri, M. B.; ...

2017-06-09

Neutral particle behavior in Aditya tokamak, which has a circular poloidal ring limiter at one particular toroidal location, has been investigated using DEGAS2 code. The code is based on the calculation using Monte Carlo algorithms and is mainly used in tokamaks with divertor configuration. This code has been successfully implemented in Aditya tokamak with limiter configuration. The penetration of neutral hydrogen atom is studied with various atomic and molecular contributions and it is found that the maximum contribution comes from the dissociation processes. For the same, H α spectrum is also simulated which was matched with the experimental one. Themore » dominant contribution around 64% comes from molecular dissociation processes and neutral particle is generated by those processes have energy of ~ 2.0 eV. Furthermore, the variation of neutral hydrogen density and H α emissivity profile are analysed for various edge temperature profiles and found that there is not much changes in H α emission at the plasma edge with the variation of edge temperature (7 to 40 eV).« less

Integrated Tokamak modeling: When physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poli, Francesca Maria

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Inductive flux usage and its optimization in tokamak operation

DOE PAGES

Luce, Timothy C.; Humphreys, David A.; Jackson, Gary L.; ...

2014-07-30

The energy flow from the poloidal field coils of a tokamak to the electromagnetic and kinetic stored energy of the plasma are considered in the context of optimizing the operation of ITER. The goal is to optimize the flux usage in order to allow the longest possible burn in ITER at the desired conditions to meet the physics objectives (500 MW fusion power with energy gain of 10). A mathematical formulation of the energy flow is derived and applied to experiments in the DIII-D tokamak that simulate the ITER design shape and relevant normalized current and pressure. The rate ofmore » rise of the plasma current was varied, and the fastest stable current rise is found to be the optimum for flux usage in DIII-D. A method to project the results to ITER is formulated. The constraints of the ITER poloidal field coil set yield an optimum at ramp rates slower than the maximum stable rate for plasmas similar to the DIII-D plasmas. Finally, experiments in present-day tokamaks for further optimization of the current rise and validation of the projections are suggested.« less

Paleoclassical transport explains electron transport barriers in RTP and TEXTOR

NASA Astrophysics Data System (ADS)

Hogeweij, G. M. D.; Callen, J. D.; RTP Team; TEXTOR Team

2008-06-01

The recently developed paleoclassical transport model sets the minimum level of electron thermal transport in a tokamak. This transport level has proven to be in good agreement with experimental observations in many cases when fluctuation-induced anomalous transport is small, i.e. in (near-)ohmic plasmas in small to medium size tokamaks, inside internal transport barriers (ITBs) or edge transport barriers (H-mode pedestal). In this paper predictions of the paleoclassical transport model are compared in detail with data from such kinds of discharges: ohmic discharges from the RTP tokamak, EC heated RTP discharges featuring both dynamic and shot-to-shot scans of the ECH power deposition radius and off-axis EC heated discharges from the TEXTOR tokamak. For ohmically heated RTP discharges the Te profiles predicted by the paleoclassical model are in reasonable agreement with the experimental observations, and various parametric dependences are captured satisfactorily. The electron thermal ITBs observed in steady state EC heated RTP discharges and transiently after switch-off of off-axis ECH in TEXTOR are predicted very well by the paleoclassical model.

Fast island phase identification for tearing mode feedback control on J-TEXT tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rao, B., E-mail: borao@hust.edu.cn; Li, D.; Hu, F. R.

A new method to control the tearing mode (TM) in tokamaks has been proposed [Q. Hu and Q. Yu, Nucl. Fusion 56, 034001 (5pp.) (2016)], according to which, the external resonant magnetic perturbation needs to be applied in certain magnetic island phase regions. Therefore, it is very important to identify the helical phase of magnetic islands in real time. The TM in tokamak plasmas is normally rotating and carries magnetic oscillations, which are known as Mirnov oscillations and can be detected by Mirnov probes. When the O-point or X-point of the magnetic island passes through the probe, the signal willmore » experience a zero-crossing. A poloidal Mirnov probe array and a corresponding island phase identification method are presented. A field-programmable gate array is used to provide the magnetic island helical phase in real time by using multichannel zero crossing detection. This system has been developed on the J-TEXT tokamak and works well. This paper introduces the establishment of the fast magnetic island phase identifying system.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE PAGES

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Control system of neoclassical tearing modes in real time on HL-2A tokamak.

PubMed

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

NASA Astrophysics Data System (ADS)

Adamek, J.; Müller, H. W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horacek, J.; Kurzan, B.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R.; Fernandes, H.; Figueiredo, H.

2016-04-01

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (ΦBPP) and the floating potential (Vfl) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula Te = (ΦBPP - Vfl)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

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

DOE PAGES

Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; ...

2016-06-20

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 β p and β N despite strong ITBs. Good confinement has been achieved with reduced toroidal rotation. These high β p plasmas challenge the energy transport understanding, especiallymore » in the electron energy channel. A new turbulent transport model, named 2 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. Finally, more investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.« less

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes.

PubMed

Adamek, J; Müller, H W; Silva, C; Schrittwieser, R; Ionita, C; Mehlmann, F; Costea, S; Horacek, J; Kurzan, B; Bilkova, P; Böhm, P; Aftanas, M; Vondracek, P; Stöckel, J; Panek, R; Fernandes, H; Figueiredo, H

2016-04-01

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (ΦBPP) and the floating potential (Vfl) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula Te = (ΦBPP - Vfl)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

Control system of neoclassical tearing modes in real time on HL-2A tokamak

NASA Astrophysics Data System (ADS)

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Desgranges, C.; Hernandez, C.; Mateus, C.; Maier, H.; Böswirth, B.; Greuner, H.; Samaille, F.; Bucalossi, J.; Missirlian, M.

2017-12-01

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

A novel flexible field-aligned coordinate system for tokamak edge plasma simulation

NASA Astrophysics Data System (ADS)

Leddy, J.; Dudson, B.; Romanelli, M.; Shanahan, B.; Walkden, N.

2017-03-01

Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (i.e. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature.

Physics of GAM-initiated L-H transition in a tokamak

NASA Astrophysics Data System (ADS)

Askinazi, L. G.; Belokurov, A. A.; Bulanin, V. V.; Gurchenko, A. D.; Gusakov, E. Z.; Kiviniemi, T. P.; Lebedev, S. V.; Kornev, V. A.; Korpilo, T.; Krikunov, S. V.; Leerink, S.; Machielsen, M.; Niskala, P.; Petrov, A. V.; Tukachinsky, A. S.; Yashin, A. Yu; Zhubr, N. A.

2017-01-01

Based on experimental observations using the TUMAN-3M and FT-2 tokamaks, and the results of gyrokinetic modeling of the interplay between turbulence and the geodesic acoustic mode (GAM) in these installations, a simple model is proposed for the analysis of the conditions required for L-H transition triggering by a burst of radial electric field oscillations in a tokamak. In the framework of this model, one-dimensional density evolution is considered to be governed by an anomalous diffusion coefficient dependent on radial electric field shear. The radial electric field is taken as the sum of the oscillating term and the quasi-stationary one determined by density and ion temperature gradients through a neoclassical formula. If the oscillating field parameters (amplitude, frequency, etc) are properly adjusted, a transport barrier forms at the plasma periphery and sustains after the oscillations are switched off, manifesting a transition into the high confinement mode with a strong inhomogeneous radial electric field and suppressed transport at the plasma edge. The electric field oscillation parameters required for L-H transition triggering are compared with the GAM parameters observed at the TUMAN-3M (in the discharges with ohmic L-H transition) and FT-2 tokamaks (where no clear L-H transition was observed). It is concluded based on this comparison that the GAM may act as a trigger for the L-H transition, provided that certain conditions for GAM oscillation and tokamak discharge are met.

High-beta spherical tokamak startup in TS-4 merging experiment by use of toroidal field ramp-up

NASA Astrophysics Data System (ADS)

Kaminou, Yasuhiro; , Toru, II; Kato, Joji; Inomoto, Michiaki; Ono, Yasushi; TS Group Team; National InstituteFusion Science Collaboration

2014-10-01

We demonstrated the formation method of an ultrahigh-beta spherical tokamak by use of a field-reversed configuration and a spheromak in TS-4 device (R ~ 0.5 m, A ~ 1.5, Ip ~ 30-100 kA, B ~ 100 mT). This method is composed of the following steps: 1. Two spheromaks are merged together and a high-beta spheromak or FRC is formed by reconnection heating. 2. External toroidal magnetic field is added (current rising time ~50 μs), and spherical tokamak-like configuration is formed. In this way, the ultrahigh-beta ST is formed. The ultrahigh-beta ST formed by FRC has a diamagnetic toroidal field, and it presumed to be in a second-stable state for ballooning stability, and the one formed by spheromak has a weak paramagnetic toroidal magnetic field, while a spheormak has a strong paramagnetic toroidal magnetic field. This diamagnetic current derives from inductive electric field by ramping up the external toroidal magnetic field, and the diamagnetic current sustains high thermal pressure of the ultrahigh-beta spherical tokamak. And the beta of the ultrahigh-beta ST formed by FRC reaches about 50%. To sustain the high-beta state, 0.6 MW neutral beam injection and center solenoid coils are installed to the TS-4 device. In the poster, we report the experimental results of ultrahigh-beta spherical tokamak startup and sustainment by NBI and CS current driving experiment.

Isotope effects on L-H threshold and confinement in tokamak plasmas

NASA Astrophysics Data System (ADS)

Maggi, C. F.; Weisen, H.; Hillesheim, J. C.; Chankin, A.; Delabie, E.; Horvath, L.; Auriemma, F.; Carvalho, I. S.; Corrigan, G.; Flanagan, J.; Garzotti, L.; Keeling, D.; King, D.; Lerche, E.; Lorenzini, R.; Maslov, M.; Menmuir, S.; Saarelma, S.; Sips, A. C. C.; Solano, E. R.; Belonohy, E.; Casson, F. J.; Challis, C.; Giroud, C.; Parail, V.; Silva, C.; Valisa, M.; Contributors, JET

2018-01-01

The dependence of plasma transport and confinement on the main hydrogenic ion isotope mass is of fundamental importance for understanding turbulent transport and, therefore, for accurate extrapolations of confinement from present tokamak experiments, which typically use a single hydrogen isotope, to burning plasmas such as ITER, which will operate in deuterium-tritium mixtures. Knowledge of the dependence of plasma properties and edge transport barrier formation on main ion species is critical in view of the initial, low-activation phase of ITER operations in hydrogen or helium and of its implications on the subsequent operation in deuterium-tritium. The favourable scaling of global energy confinement time with isotope mass, which has been observed in many tokamak experiments, remains largely unexplained theoretically. Moreover, the mass scaling observed in experiments varies depending on the plasma edge conditions. In preparation for upcoming deuterium-tritium experiments in the JET tokamak with the ITER-like Be/W Wall (JET-ILW), a thorough experimental investigation of isotope effects in hydrogen, deuterium and tritium plasmas is being carried out, in order to provide stringent tests of plasma energy, particle and momentum transport models. Recent hydrogen and deuterium isotope experiments in JET-ILW on L-H power threshold, L-mode and H-mode confinement are reviewed and discussed in the context of past and more recent isotope experiments in tokamak plasmas, highlighting common elements as well as contrasting observations that have been reported. The experimental findings are discussed in the context of fundamental aspects of plasma transport models.

Full orbit computations of ripple-induced fusion {alpha}-particle losses from burning tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClements, K.G.

A full orbit code is used to compute collisionless losses of fusion {alpha} particles from three proposed burning plasma tokamaks: the International Tokamak Experimental Reactor (ITER); a spherical tokamak power plant (STPP) [T. C. Hender, A. Bond, J. Edwards, P. J. Karditsas, K. G. McClements, J. Mustoe, D. V. Sherwood, G. M. Voss, and H. R. Wilson, Fusion Eng. Des. 48, 255 (2000)]; and a spherical tokamak components test facility (CTF) [H. R. Wilson, G. M. Voss, R. J. Akers, L. Appel, A. Dnestrovskij, O. Keating, T. C. Hender, M. J. Hole, G. Huysmans, A. Kirk, P. J. Knight, M.more » Loughlin, K. G. McClements, M. R. O'Brien, and D. Yu. Sychugov, Proceedings of the 20th IAEA Fusion Energy Conference, Invited Paper FT/3-1Ra]. It has been suggested that {alpha} particle transport could be enhanced due to cyclotron resonance with the toroidal magnetic field ripple. However, calculations for inductive operation in ITER yield a loss rate that appears to be broadly consistent with the predictions of guiding center theory, falling monotonically as the number of toroidal field coils N is increased (and hence the ripple amplitude is decreased). For STPP and CTF the loss rate does not decrease monotonically with N, but collisionless losses are generally low in absolute terms. As in the case of ITER, there is no evidence that finite Larmor radius effects would seriously degrade fusion {alpha}-particle confinement.« less

Design, simulation and construction of the Taban tokamak

NASA Astrophysics Data System (ADS)

H, R. MIRZAEI; R, AMROLLAHI

2018-04-01

This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

Impact of helical boundary conditions in MHD modeling of RFP and tokamak plasmas

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Cappello, S.; Escande, D. F.; Piovesan, P.; Veranda, M.; Chacón, L.

2012-10-01

Helical boundary conditions imposed by the active control system of the RFX-mod device provide a handle to govern the plasma dynamics in both RFP and Ohmic tokamak discharges [1]. By applying an edge radial magnetic field with proper helicity, it is possible to increase the persistence of the spontaneous helical RFP states at high current,and to stimulate them also at low current or high density. Helical BCs even allow to access helical states with different helicity than the spontaneous one [2]. In Ohmic tokamak operation at q(a)<2, the presence of the 2/1 RWM reduces the sawtoothing activity of the 1/1 internal kink, which takes a stationary snake-like character instead. Many of these features are qualitatively reproduced in 3D nonlinear MHD modeling. We study the impact of helical BCs on the MHD dynamics in both RFP and tokamak with two successfully benchmarked numerical tools, SpeCyl and PIXIE3D [3]. We recover the bifurcation from a sawtooth to a snake solution when imposing a 2/1 BC in the tokamak case and we interpret this as a toroidal/nonlinear coupling effect. We show that the bifurcation is more easily stimulated with a 1/1 BC.[4pt] [1] P. Piovesan, invited talk this meeting[0pt] [2] M. Veranda et al EPS-ICPP Conference (2012) P4.004[0pt] [3] D. Bonfiglio et al Phys. Plasmas (2010)

Multi-frequency ICRF diagnostic of Tokamak plasmas

NASA Astrophysics Data System (ADS)

Lafonteese, David James

This thesis explores the diagnostic possibilities of a fast wave-based method for measuring the ion density and temperature profiles of tokamak plasmas. In these studies fast waves are coupled to the plasma at frequencies at the second harmonic of the ion gyrofrequency, at which wave energy is absorbed by the finite-temperature ions. As the ion gyrofrequency is dependent upon the local magnetic field, which varies as l/R in a tokamak, this power absorption is radially localized. The simultaneous launching of multiple frequencies, all resonating at different plasma positions, allows local measurements of the ion density and temperature. To investigate the profile applications of wave damping measurements in a simulated tokamak, an inhouse slab-model ICRF code is developed. A variety of analysis methods are presented, and ion density and temperature profiles are reconstructed for hydrogen plasmas for the Electric Tokamak (ET) and ITER parameter spaces. These methods achieve promising results in simulated plasmas featuring bulk ion heating, off-axis RF heating, and density ramps. The experimental results of similar studies on the Electric Tokamak, a high aspect ratio (R/a = 5), low toroidal field (2.2 kG) device are then presented. In these studies, six fast wave frequencies were coupled using a single-strap, low-field-side antenna to ET plasmas. The frequencies were variable, and could be tuned to resonate at different radii for different experiments. Four magnetic pickup loops were used to measure of the toroidal component of the wave magnetic field. The expected greater eigenmode damping of center-resonant frequencies versus edge-resonant frequencies is consistently observed. Comparison of measured aspects of fast wave behavior in ET is made with the slab code predictions, which validate the code simulations under weakly-damped conditions. A density profile is measured for an ET discharge through analysis of the fast wave measurements, and is compared to an electron density profile derived from Thomson scattering data. The methodology behind a similar measurement of the ion temperature profile is also presented.

Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks—A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

NASA Astrophysics Data System (ADS)

Plyusnin, V. V.; Reux, C.; Kiptily, V. G.; Pautasso, G.; Decker, J.; Papp, G.; Kallenbach, A.; Weinzettl, V.; Mlynar, J.; Coda, S.; Riccardo, V.; Lomas, P.; Jachmich, S.; Shevelev, A. E.; Alper, B.; Khilkevitch, E.; Martin, Y.; Dux, R.; Fuchs, C.; Duval, B.; Brix, M.; Tardini, G.; Maraschek, M.; Treutterer, W.; Giannone, L.; Mlynek, A.; Ficker, O.; Martin, P.; Gerasimov, S.; Potzel, S.; Paprok, R.; McCarthy, P. J.; Imrisek, M.; Boboc, A.; Lackner, K.; Fernandes, A.; Havlicek, J.; Giacomelli, L.; Vlainic, M.; Nocente, M.; Kruezi, U.; COMPASS Team; TCV Team; ASDEX-Upgrade Team; EUROFusion MST1 Team; contributors, JET

2018-01-01

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70-75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations.

High power heating of magnetic reconnection in merging tokamak experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Y.; Tanabe, H.; Gi, K.

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 magneticmore » 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.« less

Measurement of H/H+D Ratio and Recycling in Ion Cyclotron Resonance Heating HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Ding, Liancheng; Jiang, Guangkuan; Wei, Lehan

1994-12-01

A scanning Fabry-Perot interferometer has been used to measure the Hα and Dα lines obtain the H/H+D ratio in ion cyclotron resonance heating HT-6M tokamak for determing the energy absorption mechanism. The recycling is observed by changing the working gas from deuterium to hydrogen.

Public Data Set: Continuous, Edge Localized Ion Heating During Non-Solenoidal Plasma Startup and Sustainment in a Low Aspect Ratio Tokamak

DOE Data Explorer

Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Hinson, Edward T. [University of Wisconsin-Madison] (ORCID:000000019713140X); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609); Reusch, Joshua A. [University of Wisconsin-Madison] (ORCID:0000000284249422); Schlossberg, David J. [University of Wisconsin-Madison] (ORCID:0000000287139448)

2017-05-16

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in M.G. Burke et. al., 'Continuous, Edge Localized Ion Heating During Non-Solenoidal Plasma Startup and Sustainment in a Low Aspect Ratio Tokamak,' Nucl. Fusion 57, 076010 (2017).

AE3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spong, Donald A

AE3D solves for the shear Alfven eigenmodes and eigenfrequencies in a torodal magnetic fusion confinement device. The configuration can be either 2D (e.g. tokamak, reversed field pinch) or 3D (e.g. stellarator, helical reversed field pinch, tokamak with ripple). The equations solved are based on a reduced MHD model and sound wave coupling effects are not currently included.

Ignition in tokamaks with modulated source of auxiliary heating

NASA Astrophysics Data System (ADS)

Morozov, D. Kh

2017-12-01

It is shown that the ignition may be achieved in tokamaks with the modulated power source. The time-averaged source power may be smaller than the steady-state source power, which is sufficient for the ignition. Nevertheless, the maximal power must be large enough, because the ignition must be achieved within a finite time interval.

Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

NASA Astrophysics Data System (ADS)

Fukuda, Takeshi

The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

DOE PAGES

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; ...

2018-02-16

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. Here, this Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and timemore » scales of cold-pulse experiments in tokamak plasmas.« less

Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

NASA Astrophysics Data System (ADS)

Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

2017-12-01

The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

Runaway Electrons Modeling and Nanoparticle Plasma Jet Penetration into Tokamak Plasma

NASA Astrophysics Data System (ADS)

Galkin, S. A.; Bogatu, I. N.

2017-10-01

A novel idea to probe runaway electrons (REs) by superfast injection of high velocity nanoparticle plasma jet (NPPJ) from a plasma accelerator needs to be sustained by both RE dynamics modeling and simulation of NPPJ penetration through increasing tokamak magnetic field. We present our recent progress in both areas. RE simulation is based on the model, including Dreicer and ``avalanche'' mechanisms of RE generation, with emphasis on high Zeff effects. The high-density hyper-velocity C60 and BN NPPJ penetration through transversal B-field is conducted with the Hybrid Electro-Magnetic code (HEM-2D) in cylindrical coordinates, with 1/R B-field dependence for both DIII-D and ITER tokamaks. Work is supported in part by US DOE SBIR Grant.

Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cui, Z. Q.; Chen, Z. J.; Xie, X. F.

2014-11-15

The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic fieldmore » of 200 G.« less

Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas

NASA Technical Reports Server (NTRS)

Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.

1980-01-01

The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.

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

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. Here, this Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and timemore » scales of cold-pulse experiments in tokamak plasmas.« less

Plasma shaping effects on tokamak scrape-off layer turbulence

NASA Astrophysics Data System (ADS)

Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo

2017-03-01

The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\

Setup for potential bias experiments on the Saha Institute of Nuclear Physics tokamak

NASA Astrophysics Data System (ADS)

Ghosh, J.; Pal, R.; Chattopadhyay, P. K.

1999-12-01

An experimental setup for studying the influence of the radial electric field on very low qa plasma on the Saha Institute of Nuclear Physics tokamak is presented. A high current, high voltage pulsed power supply, using a semiconductor controlled rectifier (SCR) as a dc switch is developed and used to bias a tungsten electrode inserted inside the plasma. The electrode's exposed length and its position inside the plasma are controlled by a double bellows assembly to optimize the electrode-exposed length. We show that using the force commutation method to turn the SCR off to get the power pulse desired has good potential for carrying out similar kinds of studies, especially in a low budget small tokamak.

Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experimenta)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmitt, J. C.; Bialek, J.; Lazerson, S.

2014-11-01

The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnositc signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and themore » equilibrium reconstruction are presented.« less

Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmitt, J. C., E-mail: jschmitt@pppl.gov; Lazerson, S.; Majeski, R.

2014-11-15

The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnostic signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and themore » equilibrium reconstruction are presented.« less

Fusion pumped laser

DOEpatents

Pappas, Daniel S.

1989-01-01

Apparatus is provided for generating energy in the form of laser radiation. A tokamak fusion reactor is provided for generating a long, or continuous, pulse of high-energy neutrons. The tokamak design provides a temperature and a magnetic field which is effective to generate a neutron flux of at least 10.sup.15 neutrons/cm.sup.2.s. A conversion medium receives neutrons from the tokamak and converts the high-energy neutrons to an energy source with an intensity and an energy effective to excite a preselected lasing medium. The energy source typically comprises fission fragments, alpha particles, and radiation from a fission event. A lasing medium is provided which is responsive to the energy source to generate a population inversion which is effective to support laser oscillations for generating output radiation.

The cluster charge identification in the GEM detector for fusion plasma imaging by soft X-ray diagnostics

NASA Astrophysics Data System (ADS)

Czarski, T.; Chernyshova, M.; Malinowski, K.; Pozniak, K. T.; Kasprowicz, G.; Kolasinski, P.; Krawczyk, R.; Wojenski, A.; Zabolotny, W.

2016-11-01

The measurement system based on gas electron multiplier detector is developed for soft X-ray diagnostics of tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an X-ray source. The focal measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the serial data acquisition is applied for the dynamic plasma diagnostics. The charge clusters are counted in the space determined by 2D position, charge value, and time intervals. Radiation source characteristics are presented by histograms for a selected range of position, time intervals, and cluster charge values corresponding to the energy spectra.

The cluster charge identification in the GEM detector for fusion plasma imaging by soft X-ray diagnostics.

PubMed

Czarski, T; Chernyshova, M; Malinowski, K; Pozniak, K T; Kasprowicz, G; Kolasinski, P; Krawczyk, R; Wojenski, A; Zabolotny, W

2016-11-01

The measurement system based on gas electron multiplier detector is developed for soft X-ray diagnostics of tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an X-ray source. The focal measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the serial data acquisition is applied for the dynamic plasma diagnostics. The charge clusters are counted in the space determined by 2D position, charge value, and time intervals. Radiation source characteristics are presented by histograms for a selected range of position, time intervals, and cluster charge values corresponding to the energy spectra.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oughton, S.; Matthaeus, W. H.; Dmitruk, P.

Originally proposed as an efficient approach to computation of nonlinear dynamics in tokamak fusion research devices, reduced magnetohydrodynamics (RMHD) has subsequently found application in studies of coronal heating, flux tube dynamics, charged particle transport, and, in general, as an approximation to describe plasma turbulence in space physics and astrophysics. Given the diverse set of derivations available in the literature, there has emerged some level of discussion and a lack of consensus regarding the completeness of RMHD as a turbulence model, and its applicability in contexts such as the solar wind. Some of the key issues in this discussion are examinedmore » here, emphasizing that RMHD is properly neither 2D nor fully 3D, being rather an incomplete representation that enforces at least one family of extraneous conservation laws.« less

Three-dimensional drift kinetic response of high- β plasmas in the DIII-D tokamak

DOE PAGES

Wang, Zhirui R.; Lanctot, Matthew J.; Liu, Y. Q.; ...

2015-04-07

A quantitative interpretation of the experimentally measured high pressure plasma response to externally applied three-dimensional (3D) magnetic field perturbations, across the no-wall Troyon limit, is achieved. The key to success is the self-consistent inclusion of the drift kinetic resonance effects in numerical modeling using the MARS-K code. This resolves an outstanding issue of ideal magneto-hydrodynamic model, which signi cantly over-predicts the plasma induced field ampli fication near the no-wall limit, as compared to experiments. The self-consistent drift kinetic model leads to quantitative agreement not only for the measured 3D field amplitude and toroidal phase, but also for the measured internalmore » 3D displacement of the plasma.« less

Kinetic simulation of edge instability in fusion plasmas

NASA Astrophysics Data System (ADS)

Fulton, Daniel Patrick

In this work, gyrokinetic simulations in edge plasmas of both tokamaks and field reversed. configurations (FRC) have been carried out using the Gyrokinetic Toroidal Code (GTC) and A New Code (ANC) has been formulated for cross-separatrix FRC simulation. In the tokamak edge, turbulent transport in the pedestal of an H-mode DIII-D plasma is. studied via simulations of electrostatic driftwaves. Annulus geometry is used and simulations focus on two radial locations corresponding to the pedestal top with mild pressure gradient and steep pressure gradient. A reactive trapped electron instability with typical ballooning mode structure is excited in the pedestal top. At the steep gradient, the electrostatic instability exhibits unusual mode structure, peaking at poloidal angles theta=+- pi/2. Simulations find this unusual mode structure is due to steep pressure gradients in the pedestal but not due to the particular DIII-D magnetic geometry. Realistic DIII-D geometry has a stabilizing effect compared to a simple circular tokamak geometry. Driftwave instability in FRC is studied for the first time using gyrokinetic simulation. GTC. is upgraded to treat realistic equilibrium calculated by an MHD equilibrium code. Electrostatic local simulations in outer closed flux surfaces find ion-scale modes are stable due to the large ion gyroradius and that electron drift-interchange modes are excited by electron temperature gradient and bad magnetic curvature. In the scrape-off layer (SOL) ion-scale modes are excited by density gradient and bad curvature. Collisions have weak effects on instabilities both in the core and SOL. Simulation results are consistent with density fluctuation measurements in the C-2 experiment using Doppler backscattering (DBS). The critical density gradients measured by the DBS qualitatively agree with the linear instability threshold calculated by GTC simulations. One outstanding critical issue in the FRC is the interplay between turbulence in the FRC. core and SOL regions. While the magnetic flux coordinates used by GTC provide a number of computational advantages, they present unique challenges at the magnetic field separatrix. To address this limitation, a new code, capable of coupled core-SOL simulations, is formulated, implemented, and successfully verified.

RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrada, Juan J; Reiersen, Wayne T

U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C andmore » 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.« less

Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 1996

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.D. Levine; V.L. Finley

1998-03-01

The results of the 1996 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the US Department of Energy and the public with information on the level of radioactive and nonradioactive pollutants, if any, that are added to the environment as a result of PPPL's operations. During Calendar Year 1996, PPPL's Tokamak Fusion Test Reactor (TFTR) continued to conduct fusion experiments. Having set a world record on November 2, 1994, by achieving approximately 10.7 million watts of controlled fusion power during the deuterium-tritium (D-T) plasmamore » experiments, researchers turned their attention to studying plasma science experiments, which included ''enhanced reverse shear techniques.'' Since November 1993, more than 700 tritium-fueled experiments were conducted, which generated more than 4 x 10(superscript 20) neutrons and 1.4 gigajoules of fusion energy. In 1996, the overall performance of Princeton Plasma Physics Laboratory was rated ''excellent'' by the US Department of Energy in the Laboratory Appraisal report issued in early 1997. The report cited the Laboratory's consistently excellent scientific and technological achievements and its successful management practices, which included high marks for environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey Department of Environmental Protection. PPPL monitored for the presence of nonradiological contaminants, mainly volatile organic compounds (components of degreasing solvents) and petroleum hydrocarbons (past leaks of releases of diesel fuel from underground storage tanks). Also, PPPL's radiological monitoring program characterized the ambient, background levels of tritium in the environment and from the TFTR stack; the data are presented in this report. During 1996, PPPL completed the removal of contaminated soil from two locations that were identified through the monitoring program: petroleum hydrocarbons along a drainage swale and chromium adjacent to the cooling tower.« less

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosea, J.; Adler, J.H.; Alling, P.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning;more » possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.« less

Modeling RF Fields in Hot Plasmas with Parallel Full Wave Code

NASA Astrophysics Data System (ADS)

Spencer, Andrew; Svidzinski, Vladimir; Zhao, Liangji; Galkin, Sergei; Kim, Jin-Soo

2016-10-01

FAR-TECH, Inc. is developing a suite of full wave RF plasma codes. It is based on a meshless formulation in configuration space with adapted cloud of computational points (CCP) capability and using the hot plasma conductivity kernel to model the nonlocal plasma dielectric response. The conductivity kernel is calculated by numerically integrating the linearized Vlasov equation along unperturbed particle trajectories. Work has been done on the following calculations: 1) the conductivity kernel in hot plasmas, 2) a monitor function based on analytic solutions of the cold-plasma dispersion relation, 3) an adaptive CCP based on the monitor function, 4) stencils to approximate the wave equations on the CCP, 5) the solution to the full wave equations in the cold-plasma model in tokamak geometry for ECRH and ICRH range of frequencies, and 6) the solution to the wave equations using the calculated hot plasma conductivity kernel. We will present results on using a meshless formulation on adaptive CCP to solve the wave equations and on implementing the non-local hot plasma dielectric response to the wave equations. The presentation will include numerical results of wave propagation and absorption in the cold and hot tokamak plasma RF models, using DIII-D geometry and plasma parameters. Work is supported by the U.S. DOE SBIR program.

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

DOE PAGES

Pace, D. C.; Collins, C. S.; Crowley, B.; ...

2016-09-28

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak. Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significantmore » changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. As a result, developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.« less

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, D. C.; Collins, C. S.; Crowley, B.

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak. Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significantmore » changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. As a result, developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.« less

Effect of Dipole Perturbation on a Good Confining Surface Near X Point

NASA Astrophysics Data System (ADS)

Prakash, Sushant; Witmer, Lisa; Punjabi, Alkesh; Ali, Halima

2002-11-01

We analyze effects of dipole perturbation on a good magnetic surface of a single null divertor tokamak using the method of maps developed by Punjabi and Boozer /1/. Unperturbed fields are represented by the Symmetric Simple Map /2/. Effects of high MN perturbation are represented by the Dipole Map (DM) /2/ given by: x_n+1 = x_n+2δ s^3x_n[(y_n-y_s+s) / [x_n+1^2+(y_n-y_s+s)^2]^2], y_n+1 = y_n+δ s^3x_n[[(y_n-y_s+s)^2-x_n+1^2] / [x_n+1^2+(y_n-y_s+s)^2]^2] The good surface is x_0=0 and y_0=0.952873. We start with the strength of perturbation, δ=0, increase it slowly, and observe how the surface changes. The results will be presented. This project is supported by the QEM network NASA Sharp Plus program and by DE-FG02-02ER54673. The work is done on the FUSION1 server under Profs. Ali and Punjabi 1. Punjabi A., Verma A. and Boozer A., Phys. Rev. Lett., 69, 3322 (1992) 2. Ali H., Punjabi A. and Boozer A., Dipole Map for Single-null Divertor Tokamak, 2002 International Congress on Plasma Physics, Manly, Australia.

Calculation of the Area of Stochastic Layer for a Single-Null Divertor Tokamak with the Effects of Dipole Coil Using Method of Maps

NASA Astrophysics Data System (ADS)

Basemore, Alphonso; Ali, Halima; Watson, Michael; Punjabi, Alkesh

1996-11-01

We calculate the variation in area of the stochastic scrape-off layer of a single-null divertor tokamak resulting from the effects of an externally placed dipole coil using the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effects of the dipole coil are repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The strength of the dipole perturbation and the distance of the coil from last good surface are varied. The area of the stochastic layer is calculated using the method of fractal dimension. This work is supported by US DOE OFES. Alphonso Basemore is a HU CFRT Summer Fusion High School Workshop scholar from Mount Tabor High School in North Carolina. He is supported by NASA under its NASA SharpPlus Program.

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

NASA Astrophysics Data System (ADS)

Pace, D. C.; Collins, C. S.; Crowley, B.; Grierson, B. A.; Heidbrink, W. W.; Pawley, C.; Rauch, J.; Scoville, J. T.; Van Zeeland, M. A.; Zhu, Y. B.; The DIII-D Team

2017-01-01

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614). Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significant changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. Developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Ion Bernstein wave heating research

NASA Astrophysics Data System (ADS)

Ono, Masayuki

1993-02-01

Ion Bernstein wave heating (IBWH) utilizes the ion Bernstein wave (IBW), a hot plasma wave, to carry the radio frequency (rf) power to heat the tokamak reactor core. Earlier wave accessibility studies have shown that this finite-Larmor-radius (FLR) mode should penetrate into a hot dense reactor plasma core without significant attenuation. Moreover, the IBW's low perpendicular phase velocity (ω/k⊥≊VTi≪Vα) greatly reduces the otherwise serious wave absorption by the 3.5 MeV fusion α particles. In addition, the property of IBW's that k⊥ρi≊1 makes localized bulk ion heating possible at the ion cyclotron harmonic layers. Such bulk ion heating can prove useful in optimizing fusion reactivity. In another vein, with proper selection of parameters, IBW's can be made subject to strong localized electron Landau damping near the major ion cyclotron harmonic resonance layers. This property can be useful, for example, for rf current drive in the reactor plasma core. IBW's can be excited with loop antennas or with a lower-hybrid-like waveguide launcher at the plasma edge, the latter structure being one that is especially compatible with reactor application. In either case, the mode at the plasma edge is an electron plasma wave (EPW). Deeper in the plasma, the EPW is mode transformed into an IBW. Such launching and mode transformation of IBW's were first demonstrated in experiments in the Advanced Concepts Torus-1 (ACT-1) [Phys. Rev. Lett. 45, 1105 (1980)] plasma torus and in particle simulation calculations. These and other aspects of IBW heating physics have been investigated through a number of experiments performed on ACT-1, the Japanese Institute of Plasma Physics Tokamak II-Upgrade (JIPPTII-U) [Phys. Rev. Lett. 54, 2339 (1985)], the Tokyo University Non-Circular Tokamak (TNT) [Nucl. Fusion 26, 1097 (1986)], the Princeton Large Tokamak (PLT) [Phys. Rev. Lett. 60, 294 (1988)], and Alcator-C [Phys. Rev. Lett. 60, 298 (1988)]. In these experiments both linear and nonlinear heating processes have been observed. Interestingly, improvement of plasma confinement was also observed in the PLT and Alcator-C experiments, opening up the possible use of IBW's for the active control of plasma transport. Two theoretical explanations have been proposed: one based on four-wave mixing of IBW with low-frequency turbulence, the other on the nonlinear generation of a velocity-shear layer. Both models are consistent with the observed threshold power level of a few hundred kW in the experiments. Experiments on lower field plasmas on JFTII-M [Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Physics, New York, 1989), p. 350] and DIII-D [Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Physics, New York, 1989), p. 314] have raised some concern with the IBW wave-launching process. The experiments showed serious impurity release from the walls but little or no core heating, a combination of circumstances strongly suggestive of edge heating. Possible parasitic channels could include the excitation of short wavelength modes by the Faraday shield's fringing fields, antenna-sheath-wave excitation, an axial-convective loss channel, and nonlinear processes such as parametric instability and ponderomotive effects. Suggested remedies include changes in the antenna phasing, the use of low-Z insulators, operating at higher frequencies, positioning the plasma differently with respect to the antenna, eliminating the Faraday shields, and using a waveguide launcher. The recent JIPPTII-U experiment, employing a 0-π phased antenna array with a higher frequency 130 MHz source, demonstrated that those remedies can indeed work. Looking to the future, one seeks additional ways in which IBWH can improve tokamak performance. The strong ponderomotive potential of the IBWH antenna may be used to stabilize external kinks and, acting as an rf limiter, to control the plasma edge. Control of the plasma pressure profile with local IBWH heating is already an important part of the Princeton Beta Experiment-Modified (PBX-M) [Ninth Topical Conference on Radio-Frequency Power in Plasmas, Charleston, SC, 1991 (American Institute of Physics, New York, 1991), p. 129] program in its exploration of the second-stability regime. Application of IBWH may also improve the performance of neutral beam heating and the efficiency and localization of lower-hybrid current drive for current profile control. Used with pellet injection, IBWH may also prolong the period of good confinement. The three planned high-power IBWH experiments covering vastly different parameters: f=40-80 MHz for PBX-M; f=130 MHz for JIPPT-II-U; and f=430 MHz for the Frascati Tokamak-Upgrade (FT-U) [16th European Physical Society Conference on Controlled Fusion and Plasma Physics, Venice, Italy, 1989 (European Physical Society, Amsterdam, 1989), Vol. III, p. 1069] appear to be well positioned to explore these possibilities and to clarify other issues including the physics of wave launching and associated nonlinear processes.

Kinetic description of rotating Tokamak plasmas with anisotropic temperatures in the collisionless regime

NASA Astrophysics Data System (ADS)

Cremaschini, Claudio; Tessarotto, Massimo

2011-11-01

A largely unsolved theoretical issue in controlled fusion research is the consistent kinetic treatment of slowly-time varying plasma states occurring in collisionless and magnetized axisymmetric plasmas. The phenomenology may include finite pressure anisotropies as well as strong toroidal and poloidal differential rotation, characteristic of Tokamak plasmas. Despite the fact that physical phenomena occurring in fusion plasmas depend fundamentally on the microscopic particle phase-space dynamics, their consistent kinetic treatment remains still essentially unchallenged to date. The goal of this paper is to address the problem within the framework of Vlasov-Maxwell description. The gyrokinetic treatment of charged particles dynamics is adopted for the construction of asymptotic solutions for the quasi-stationary species kinetic distribution functions. These are expressed in terms of the particle exact and adiabatic invariants. The theory relies on a perturbative approach, which permits to construct asymptotic analytical solutions of the Vlasov-Maxwell system. In this way, both diamagnetic and energy corrections are included consistently into the theory. In particular, by imposing suitable kinetic constraints, the existence of generalized bi-Maxwellian asymptotic kinetic equilibria is pointed out. The theory applies for toroidal rotation velocity of the order of the ion thermal speed. These solutions satisfy identically also the constraints imposed by the Maxwell equations, i.e., quasi-neutrality and Ampere's law. As a result, it is shown that, in the presence of nonuniform fluid and EM fields, these kinetic equilibria can sustain simultaneously toroidal differential rotation, quasi-stationary finite poloidal flows and temperature anisotropy.

Analysis of wall-embedded Langmuir probe signals in different conditions on the Tokamak à Configuration Variable

NASA Astrophysics Data System (ADS)

Février, O.; Theiler, C.; De Oliveira, H.; Labit, B.; Fedorczak, N.; Baillod, A.

2018-05-01

This paper presents the current wall-embedded Langmuir probe system installed on the Tokamak à Configuration Variable (TCV), as well as the analysis tool chain used to interpret the current-voltage characteristic obtained when the probes are operated in swept-bias conditions. The analysis is based on a four-parameter fit combined with a minimum temperature approach. In order to reduce the effect of plasma fluctuations and measurement noise, several current-voltage characteristics are usually averaged before proceeding to the fitting. The impact of this procedure on the results is investigated, as well as the possible role of finite resistances in the circuitry, which could lead to an overestimation of the temperature. We study the application of the procedure in a specific regime, the plasma detachment, where results from other diagnostics indicate that the electron temperature derived from the Langmuir probes might be overestimated. To address this issue, we explore other fitting models and, in particular, an extension of the asymmetric double probe fit, which features effects of sheath expansion. We show that these models yield lower temperatures (up to approximately 60%) than the standard analysis in detached conditions, particularly for a temperature peak observed near the plasma strike point, but a discrepancy with other measurements remains. We explore a possible explanation for this observation, the presence of a fast electron population, and assess how robust the different methods are in such conditions.

Particle fueling experiments with a series of pellets in LHD

NASA Astrophysics Data System (ADS)

Baldzuhn, J.; Damm, H.; Dinklage, A.; Sakamoto, R.; Motojima, G.; Yasuhara, R.; Ida, K.; Yamada, H.; LHD Experiment Group; Wendelstein 7-X Team

2018-03-01

Ice pellet injection is performed in the heliotron Large Helical Device (LHD). The pellets are injected in short series, with up to eight individual pellets. Parameter variations are performed for the pellet ice isotopes, the LHD magnetic configurations, the heating scenario, and some others. These experiments are performed in order to find out whether deeper fueling can be achieved with a series of pellets compared to single pellets. An increase of the fueling efficiency is expected since pre-cooling of the plasma by the first pellets within a series could aid deeper penetration of later pellets in the same series. In addition, these experiments show which boundary conditions must be fulfilled to optimize the technique. The high-field side injection of pellets, as proposed for deep fueling in a tokamak, will not be feasible with the same efficiency in a stellarator or heliotron because there the magnetic field gradient is smaller than in a tokamak of comparable size. Hence, too shallow pellet fueling, in particular in a large device or a fusion reactor, will be an issue that can be overcome only by extremely high pellet velocities, or other techniques that will have to be developed in the future. It turned out by our investigations that the fueling efficiency can be enhanced by the injection of a series of pellets to some extent. However, further investigations will be needed in order to optimize this approach for deep particle fueling.

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.

H-mode achievement and edge features in RFX-mod tokamak operation

NASA Astrophysics Data System (ADS)

Spolaore, M.; Cavazzana, R.; Marrelli, L.; Carraro, L.; Franz, P.; Spagnolo, S.; Zaniol, B.; Zuin, M.; Cordaro, L.; Dal Bello, S.; De Masi, G.; Ferro, A.; Finotti, C.; Grando, L.; Grenfell, G.; Innocente, P.; Kudlacek, O.; Marchiori, G.; Martines, E.; Momo, B.; Paccagnella, R.; Piovesan, P.; Piron, C.; Puiatti, M. E.; Recchia, M.; Scarin, P.; Taliercio, C.; Vianello, N.; Zanotto, L.

2017-11-01

The RFX-mod experiment is a fusion device designed to operate as a reversed field pinch (RFP), with a major radius R = 2 m and a minor radius a = 0.459 m. Its high versatility recently allowed operating it also as an ohmic tokamak, allowing comparative studies between the two configurations in the same device. The device is equipped with a state of the art MHD mode feedback control system providing a magnetic boundary effective control, by applying resonant or non-resonant magnetic perturbations (MP), both in RFP and in tokamak configurations. In the fusion community the application of MPs is widely studied as a promising tool to limit the impact of plasma filaments and ELMs (edge localized modes) on plasma facing components. An important new research line is the exploitation of the RFX-mod active control system for ELM mitigation studies. As a first step in this direction, this paper presents the most recent achievements in term of RFX-mod tokamak explored scenarios, which allowed the first investigation of the ohmic and edge biasing induced H-mode. The production of D-shaped tokamak discharges and the design and deployment of an insertable polarized electrode were accomplished. Reproducible H-mode phases were obtained with insertable electrode negative biasing in single null discharges, representing an unexplored scenario with this technique. Important modifications of the edge plasma density and flow properties are observed. During the achieved H-mode ELM-like electromagnetic composite filamentary structures are observed. They are characterized by clear vorticity and parallel current density patterns.

Metal droplet erosion and shielding plasma layer under plasma flows typical of transient processes in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru

It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.

Doppler-broadening measurements of x-ray lines for determination of the ion temperature in tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bitter, M.; von Goeler, S.; Horton, R.

1979-01-29

Ion-temperature results are deduced from Doppler-broadening measurements of the K..cap alpha.. (1s-2p) resonance line emitted from heliumlike iron impurity ions in the hot central core of PLT (Princeton Large Torus) tokamak discharges. The measurements were performed using a high-resolution Bragg-crystal spectrometer with a multiwire proportional counter.

Measurement of electron density profiles on HT-6M tokamak by 7-channel FIR HCN laser interferometer

NASA Astrophysics Data System (ADS)

Xiang, Gao; Qiliang, Guo

1990-12-01

Electron density measurements are periormed on HT-6M tokamak using a 7 channel Far-Infrared HCN laser interferometer. From the measured line integrals--7 channel phase shifts the electron density profile is reconstructed by a fit procedure. Results were tested by comparison to Abel inverted. Some recent interesting experimental results were reported.

Tokamak startup using point-source dc helicity injection.

PubMed

Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C

2009-06-05

Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

DOE PAGES

Chrystal, Colin; Burrell, Keith H.; Grierson, Brian A.; ...

2015-10-20

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in-situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination diagnostic (CER) at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain informationmore » about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. Lastly, the methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.« less

A Riccati solution for the ideal MHD plasma response with applications to real-time stability control

NASA Astrophysics Data System (ADS)

Glasser, Alexander; Kolemen, Egemen; Glasser, A. H.

2018-03-01

Active feedback control of ideal MHD stability in a tokamak requires rapid plasma stability analysis. Toward this end, we reformulate the δW stability method with a Hamilton-Jacobi theory, elucidating analytical and numerical features of the generic tokamak ideal MHD stability problem. The plasma response matrix is demonstrated to be the solution of an ideal MHD matrix Riccati differential equation. Since Riccati equations are prevalent in the control theory literature, such a shift in perspective brings to bear a range of numerical methods that are well-suited to the robust, fast solution of control problems. We discuss the usefulness of Riccati techniques in solving the stiff ordinary differential equations often encountered in ideal MHD stability analyses—for example, in tokamak edge and stellarator physics. We demonstrate the applicability of such methods to an existing 2D ideal MHD stability code—DCON [A. H. Glasser, Phys. Plasmas 23, 072505 (2016)]—enabling its parallel operation in near real-time, with wall-clock time ≪1 s . Such speed may help enable active feedback ideal MHD stability control, especially in tokamak plasmas whose ideal MHD equilibria evolve with inductive timescale τ≳ 1s—as in ITER.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seo, Seong-Heon; Wi, H. M.; Lee, W. R.

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 ofmore » 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.« less

Optimization study of normal conductor tokamak for commercial neutron source

NASA Astrophysics Data System (ADS)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Advances in stellarator gyrokinetics

NASA Astrophysics Data System (ADS)

Helander, P.; Bird, T.; Jenko, F.; Kleiber, R.; Plunk, G. G.; Proll, J. H. E.; Riemann, J.; Xanthopoulos, P.

2015-05-01

Recent progress in the gyrokinetic theory of stellarator microinstabilities and turbulence simulations is summarized. The simulations have been carried out using two different gyrokinetic codes, the global particle-in-cell code EUTERPE and the continuum code GENE, which operates in the geometry of a flux tube or a flux surface but is local in the radial direction. Ion-temperature-gradient (ITG) and trapped-electron modes are studied and compared with their counterparts in axisymmetric tokamak geometry. Several interesting differences emerge. Because of the more complicated structure of the magnetic field, the fluctuations are much less evenly distributed over each flux surface in stellarators than in tokamaks. Instead of covering the entire outboard side of the torus, ITG turbulence is localized to narrow bands along the magnetic field in regions of unfavourable curvature, and the resulting transport depends on the normalized gyroradius ρ* even in radially local simulations. Trapped-electron modes can be significantly more stable than in typical tokamaks, because of the spatial separation of regions with trapped particles from those with bad magnetic curvature. Preliminary non-linear simulations in flux-tube geometry suggest differences in the turbulence levels in Wendelstein 7-X and a typical tokamak.

The reconstruction and research progress of the TEXT-U tokamak in China

NASA Astrophysics Data System (ADS)

Zhuang, G.; Pan, Y.; Hu, X. W.; Wang, Z. J.; Ding, Y. H.; Zhang, M.; Gao, L.; Zhang, X. Q.; Yang, Z. J.; Yu, K. X.; Gentle, K. W.; Huang, H.; J-TEXT Team

2011-09-01

The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 1019 m-3, and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

Radial force on the vacuum chamber wall during thermal quench in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru

The radial force balance during a thermal quench in tokamaks is analyzed. As a rule, the duration τ{sub tp} of such events is much shorter than the resistive time τ{sub w} of the vacuum chamber wall. Therefore, the perturbations of the magnetic field B produced by the evolving plasma cannot penetrate the wall, which makes different the magnetic pressures on its inner and outer sides. The goal of this work is the analytical estimation of the resulting integral radial force on the wall. The plasma is considered axially symmetric; for the description of radial forces on the wall, the resultsmore » of V.D. Shafranov’s classical work [J. Nucl. Energy C 5, 251 (1963)] are used. Developed for tokamaks, the standard equilibrium theory considers three interacting systems: plasma, poloidal field coils, and toroidal field coils. Here, the wall is additionally incorporated with currents driven by ∂B/∂t≠0 accompanying the fast loss of the plasma thermal energy. It is shown that they essentially affect the force redistribution, thereby leading to large loads on the wall. The estimates prove that these loads have to be accounted for in the disruptive scenarios in large tokamaks.« less

Effect of ECRH and resonant magnetic fields on formation of magnetic islands in the T-10 tokamak plasma

NASA Astrophysics Data System (ADS)

Shestakov, E. A.; Savrukhin, P. V.

2017-10-01

Experiments in the T-10 tokamak demonstrated possibility of controlling the plasma current during disruption instability using the electron cyclotron resonance heating (ECRH) and the controlled operation of the ohmic current-holding system. Quasistable plasma discharge with repeating sawtooth oscillations can be restored after energy quench using auxiliary ECRH power when PEC / POH > 2-5. The external magnetic field generation system consisted of eight saddle coils that were arranged symmetrically relative to the equatorial plane of the torus outside of the vacuum vessel of the T-10 tokamak to study the possible resonant magnetic field effects on the rotation frequency of magnetic islands. The saddle coils power supply system is based on four thyristor converters with a total power of 300 kW. The power supply control system is based on Siemens S7 controllers. As shown by preliminary experiments, the interaction efficiency of external magnetic fields with plasma depends on the plasma magnetic configuration. Optimal conditions for slowing the rotation of magnetic islands were determined. Additionally, the direction of the error magnetic field in the T-10 tokamak was determined, and the threshold value of the external magnetic field was determined.

Compact Torus Fueling of the STOR-M Tokamak

NASA Astrophysics Data System (ADS)

Xiao, C.; Hirose, A.; Zawalski, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.

1996-11-01

Tangential injection of accelerated compact torus (CT) has been performed on the STOR-M tokamak (R/a=46/12 cm, B_t<1 T, I_p<= 50 kA, barn_e=(0.5 - 1)×10^13 cm-3) using the University of Saskatchewan Compact Torus Injector (USCTI). The CT parameters are: m~=1 μg, v=120 km/sec, B=0.1 T and n=(2 - 4)×10^15 cm-3. After CT injection, the electron density in tokamak doubles and the poloidal β-value increases. Indications of reduction in the loop voltage and H_α emission level have also been observed. Currently, following efforts are being made: (a) to coat chromium on the electrode surface, (b) to increase the on-line baking temperature, and (c) to reduce the neutral gas load which follows the CT plasma. In addition, numerical calculation of CT motion in a tokamak magnetic field has been carried out. For horizontal injection, the initial CT magnetic dipole direction should be aligned with the CT velocity for deeper penetration. In the case of vertical injection, the CT trajectory is independent of the initial magnetic dipole direction and central penetration is facilitated by off-axis injection.

3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

NASA Astrophysics Data System (ADS)

Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

2017-10-01

The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ < 0.5) fueling from neutral ionization sources is decreased by 40-60% with RMPs. This work was funded by the US Department of Energy under Grant DE-SC0012315.

On the applicability of the standard approaches for evaluating a neoclassical radial electric field in a tokamak edge region

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Simakov, A. N.; ...

2013-08-27

The use of the standard approaches for evaluating a neoclassical radial electric field E r, i.e., the Ampere (or gyro-Poisson) equation, requires accurate calculation of the difference between the gyroaveraged electron and ion particle fluxes (or densities). In the core of a tokamak, the nontrivial difference appears only in high-order corrections to a local Maxwellian distribution due to the intrinsic ambipolarity of particle transport. The evaluation of such high-order corrections may be inconsistent with the accuracy of the standard long wavelength gyrokinetic equation (GKE), thus imposing limitations on the applicability of the standard approaches. However, in the edge of amore » tokamak, charge-exchange collisions with neutrals and prompt ion orbit losses can drive non-intrinsically ambipolar particle fluxes for which a nontrivial (E r-dependent) difference between the electron and ion fluxes appears already in a low order and can be accurately predicted by the long wavelength GKE. As a result, the parameter regimes where the radial electric field dynamics in the tokamak edge region is dominated by the non-intrinsically ambipolar processes, thus allowing for the use of the standard approaches, are discussed.« less

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

NASA Astrophysics Data System (ADS)

Chrystal, C.; Burrell, K. H.; Grierson, B. A.; Pace, D. C.

2015-10-01

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination (CER) diagnostic at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain information about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. The methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamek, J., E-mail: adamek@ipp.cas.cz; Horacek, J.; Bilkova, P.

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (Φ{sub BPP}) and the floating potential (V{sub fl}) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula T{submore » e} = (Φ{sub BPP} − V{sub fl})/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.« less

Simulation of MST tokamak discharges with resonant magnetic perturbations

NASA Astrophysics Data System (ADS)

Cornille, B. S.; Sovinec, C. R.; Chapman, B. E.; Dubois, A.; McCollam, K. J.; Munaretto, S.

2016-10-01

Nonlinear MHD modeling of MST tokamak plasmas with an applied resonant magnetic perturbation (RMP) reveals degradation of flux surfaces that may account for the experimentally observed suppression of runaway electrons with the RMP. Runaway electrons are routinely generated in MST tokamak discharges with low plasma density. When an m = 3 RMP is applied these electrons are strongly suppressed, while an m = 1 RMP of comparable amplitude has little effect. The computations are performed using the NIMROD code and use reconstructed equilibrium states of MST tokamak plasmas with q (0) < 1 and q (a) = 2.2 . Linear computations show that the (1 , 1) -kink and (2 , 2) -tearing modes are unstable, and nonlinear simulations produce sawtoothing with a period of approximately 0.5 ms, which is comparable to the period of MHD activity observed experimentally. Adding an m = 3 RMP in the computation degrades flux surfaces in the outer region of the plasma, while no degradation occurs with an m = 1 RMP. The outer flux surface degradation with the m = 3 RMP, combined with the sawtooth-induced distortion of flux surfaces in the core, may account for the observed suppression of runaway electrons. Work supported by DOE Grant DE-FC02-08ER54975.

Book Review:

NASA Astrophysics Data System (ADS)

Rogister, A. L.

2004-03-01

John Wesson’s well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author’s task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2 9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11 12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of tokamak physics, from basic principles to interpretation of experimental data, and to a wider readership an elegant and authoritative introduction to the challenges that are associated with the development of the tokamak reactor, a source of limitless and clean thermonuclear power. This reference book should be on the shelf of every fusion scientist and graduate student.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

Status of fusion research and implications for D/He-3 systems

NASA Technical Reports Server (NTRS)

Miley, George H.

1988-01-01

World wide programs in both magnetic confinement and inertial confinement fusion research have made steady progress towards the experimental demonstration of energy breakeven. However, after breakeven is achieved, considerable time and effort must still be expended to develop a usable power plant. The main program described is focused on Deuterium-Tritium devices. In magnetic confinement, three of the most promising high beta approaches with a reasonable experimental data base are the Field Reversed Configuration, the high field tokamak, and the dense Z-pinch. The situation is less clear in inertial confinement where the first step requires an experimental demonstration of D/T spark ignition. It appears that fusion research has reached a point in time where an R and D plan to develop a D/He-3 fusion reactor can be laid out with some confidence of success.

Investigation of the plasma radiation power in the Globus-M tokamak by means of SPD silicon photodiodes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iblyaminova, A. D., E-mail: a.iblyaminova@mail.ioffe.ru; Avdeeva, G. F.; Aruev, P. N.

2016-10-15

Radiation losses from the plasma of the Globus-M tokamak are studied by means of SPD silicon photodiodes developed at the Ioffe Institute, Russian Academy of Sciences. The results from measurements of radiation losses in regimes with ohmic and neutral beam injection heating of plasmas with different isotope compositions are presented. The dependence of the radiation loss power on the plasma current and plasma–wall distance is investigated. The radiation power in different spectral ranges is analyzed by means of an SPD spectrometric module. Results of measurements of radiation losses before and after tokamak vessel boronization are presented. The time evolution ofmore » the sensitivity of the SPD photodiode during its two-year exploitation in Globus-M is analyzed.« less

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

The computation in diagnostics for tokamaks: systems, designs, approaches

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał; Linczuk, Paweł; Czarski, Tomasz; Wojeński, Andrzej; Chernyshova, Maryna; Poźniak, Krzysztof; Kolasiński, Piotr; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol; Gaska, Michał

2017-08-01

The requirements given for GEM (Gaseous Electron Multiplier) detector based acquisition system for plasma impurities diagnostics triggered a need for the development of a specialized software and hardware architecture. The amount of computations with latency and throughput restrictions cause that an advanced solution is sought for. In order to provide a mechanism fitting the designated tokamaks, an insight into existing solutions was necessary. In the article there is discussed architecture of systems used for plasma diagnostics and in related scientific fields. The developed solution is compared and contrasted with other diagnostic and control systems. Particular attention is payed to specific requirements for plasma impurities diagnostics in tokamak thermal fusion reactor. Subsequently, the details are presented that justified the choice of the system architecture and the discussion on various approaches is given.

Isotopic effect in experiments on lower hybrid current drive in the FT-2 tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lashkul, S. I., E-mail: Serguey.lashkul@mail.ioffe.ru; Altukhov, A. B.; Gurchenko, A. D., E-mail: aleksey.gurchenko@mail.ioffe.ru

To analyze factors influencing the limiting value of the plasma density at which lower hybrid (LH) current drive terminates, the isotopic factor (the difference in the LH resonance densities in hydrogen and deuterium plasmas) was used for the first time in experiments carried out at the FT-2 tokamak. It is experimentally found that the efficiency of LH current drive in deuterium plasma is appreciably higher than that in hydrogen plasma. The significant role of the parametric decay of the LH pumping wave, which hampers the use of the LH range of RF waves for current drive at high plasma densities,more » is confirmed. It is demonstrated that the parameters characterizing LH current drive agree well with the earlier results obtained at large tokamaks.« less

Adaptive optimal stochastic state feedback control of resistive wall modes in tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2006-01-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least-square method with exponential forgetting factor and covariance resetting is used to identify (experimentally determine) the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time-dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

Adaptive Optimal Stochastic State Feedback Control of Resistive Wall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2007-06-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least square method with exponential forgetting factor and covariance resetting is used to identify the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal.

PubMed

Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Haskey, S R; Kaplan, D H

2016-11-01

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. The unique combination of experimentally measured main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.

Investigation of the validity of quasilinear theory for electron Landau damping in a tokamak using a broad-band wave effect

DOE PAGES

Lee, Jungpyo; Bonoli, Paul; Wright, John

2011-01-01

The quasilinear diffusion coefficient assuming a constant magnetic field along the electron orbit is widely used to describe electron Landau damping of waves in a tokamak where the magnitude of the magnetic field varies on a flux surface. To understand the impact of violating the constant magnetic field assumption, we introduce the effect of a broad-bandwidth wave spectrum which has been used in the past to validate quasilinear theory for the fast decorrelation process between resonances. By the reevaluation of the diffusion coefficient through the level of the phase integral for the tokamak geometry with the broad-band wave effect included,more » we identify the three acceptable errors for the use of the quasilinear diffusion coefficient.« less

The effect of the inductive electric field on ion poloidal rotation in all collisionality regimes for the primary ions in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pan Chengkang; Wang Shaojie; Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031

2007-11-15

The expression for the poloidal rotation velocity of the primary ions that is caused by the parallel inductive electric field in tokamaks and valid in all collisionality regimes is derived via the Hirshman-Sigmar moment approach. Also the expression of the collisional impurity ions poloidal rotation velocity that is caused by the parallel inductive electric field in tokamaks is derived. The poloidal rotation velocities of the primary ions and the impurity ions are sensitive to the primary ion collisionality parameter and the impurity strength parameter. The poloidal rotation velocities of the primary ions and the impurity ions decrease with the primarymore » ion collisionality parameter and decrease with the impurity strength parameter.« less

Energetic electrons, hard x-ray emission and MHD activity studies in the IR-T1 tokamak.

PubMed

Agah, K Mikaili; Ghoranneviss, M; Elahi, A Salar

2015-01-01

Determinations of plasma parameters as well as the Magnetohydrodynamics (MHD) activity, energetic electrons energy and energy confinement time are essential for future fusion reactors experiments and optimized operation. Also some of the plasma information can be deduced from these parameters, such as plasma equilibrium, stability, and MHD instabilities. In this contribution we investigated the relation between energetic electrons, hard x-ray emission and MHD activity in the IR-T1 Tokamak. For this purpose we used the magnetic diagnostics and a hard x-ray spectroscopy in IR-T1 tokamak. A hard x-ray emission is produced by collision of the runaway electrons with the plasma particles or limiters. The mean energy was calculated from the slope of the energy spectrum of hard x-ray photons.

A spatially resolving x-ray crystal spectrometer for measurement of ion-temperature and rotation-velocity profiles on the Alcator C-Mod tokamak.

PubMed

Hill, K W; Bitter, M L; Scott, S D; Ince-Cushman, A; Reinke, M; Rice, J E; Beiersdorfer, P; Gu, M-F; Lee, S G; Broennimann, Ch; Eikenberry, E F

2008-10-01

A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra (lambda/d lambda>6000) of He-like and H-like Ar K alpha lines with good spatial (approximately 1 cm) and temporal (approximately 10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (T(i)), and toroidal plasma rotation velocity (upsilon(phi)) from the line Doppler widths and shifts. The data analysis techniques, T(i) and upsilon(phi) profiles, analysis of fusion-neutron background, and predictions of performance on other tokamaks, including ITER, will be presented.

Axisymmetric electrostatic magnetohydrodynamic oscillations in tokamaks with general cross-sections and toroidal flow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, M. S.; Guo, Wenfeng

2016-06-15

The frequency spectrum and mode structure of axisymmetric electrostatic oscillations [the zonal flow (ZF), sound waves (SW), geodesic acoustic modes (GAM), and electrostatic mean flows (EMF)] in tokamaks with general cross-sections and toroidal flows are studied analytically using the electrostatic approximation for magnetohydrodynamic modes. These modes constitute the “electrostatic continua.” Starting from the energy principle for a tokamak plasma with toroidal rotation, we showed that these modes are completely stable. The ZF, the SW, and the EMF could all be viewed as special cases of the general GAM. The Euler equations for the general GAM are obtained and are solvedmore » analytically for both the low and high range of Mach numbers. The solution consists of the usual countable infinite set of eigen-modes with discrete eigen-frequencies, and two modes with lower frequencies. The countable infinite set is identified with the regular GAM. The lower frequency mode, which is also divergence free as the plasma rotation tends to zero, is identified as the ZF. The other lower (zero) frequency mode is a pure geodesic E×B flow and not divergence free is identified as the EMF. The frequency of the EMF is shown to be exactly 0 independent of plasma cross-section or its flow Mach number. We also show that in general, sound waves with no geodesic components are (almost) completely lost in tokamaks with a general cross-sectional shape. The exception is the special case of strict up-down symmetry. In this case, half of the GAMs would have no geodesic displacements. They are identified as the SW. Present day tokamaks, although not strictly up-down symmetric, usually are only slightly up-down asymmetric. They are expected to share the property with the up-down symmetric tokamak in that half of the GAMs would be more sound wave-like, i.e., have much weaker coupling to the geodesic components than the other half of non-sound-wave-like modes with stronger coupling to the geodesic displacements. Based on the general notion that the geodesic component of the GAM is more effective in tearing up the eddies in the electrostatic turbulence, it is important to preferentially excite the GAMs that are non-sound-wave like to maximize the efficiency on turbulence suppression through external means. Finally, approximate formulae for the frequencies of the EMF, ZF, SW, and the GAM for a large aspect ratio circular tokamak rotating at low Mach numbers are also provided.« less

Effects of radial envelope modulations on the collisionless trapped-electron mode in tokamak plasmas

NASA Astrophysics Data System (ADS)

Chen, Hao-Tian; Chen, Liu

2018-05-01

Adopting the ballooning-mode representation and including the effects of radial envelope modulations, we have derived the corresponding linear eigenmode equation for the collisionless trapped-electron mode in tokamak plasmas. Numerical solutions of the eigenmode equation indicate that finite radial envelope modulations can affect the linear stability properties both quantitatively and qualitatively via the significant modifications in the corresponding eigenmode structures.

Deuterium velocity and temperature measurements on the DIII-D tokamak.

PubMed

Grierson, B A; Burrell, K H; Solomon, W M; Pablant, N A

2010-10-01

Newly installed diagnostic capabilities on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 46, 6114 (2002)] enable the measurement of main ion (deuterium) velocity and temperature by charge exchange recombination spectroscopy. The uncertainty in atomic physics corrections for determining the velocity is overcome by exploiting the geometrical dependence of the apparent velocity on the viewing angle with respect to the neutral beam.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Presentation Stations of the General Atomics Fusion Educational Program

NASA Astrophysics Data System (ADS)

Lee, R. L.; Fusion Group Education Outreach Team

1996-11-01

The General Atomics Fusion Group's Educational Program has been actively promoting fusion science and applications throughout San Diego County's secondary school systems for over three years. The educational program allows many students to learn more about nuclear fusion science, its applications, and what it takes to become an active participant in an important field of study. It also helps educators to better understand how to teach fusion science in their classroom. Tours of the DIII--D facility are a centerpiece of the program. Over 1000 students visited the DIII--D research facility during the 1995--1996 school year for a half-day of presentations, discussions, and hands-on learning. Interactive presentations are provided at six different stations by GA scientists and engineers to small groups of students during the tours. Stations include topics on energy, plasma science, the electromagnetic spectrum, radiation and risk assessment, and data acquisition. Included also is a tour of the DIII--D machine hall and model where students can see and discuss many aspects of the tokamak. Portions of each station will be presented and discussed.

Data Acquisition Systems

NASA Technical Reports Server (NTRS)

1994-01-01

In the mid-1980s, Kinetic Systems and Langley Research Center determined that high speed CAMAC (Computer Automated Measurement and Control) data acquisition systems could significantly improve Langley's ARTS (Advanced Real Time Simulation) system. The ARTS system supports flight simulation R&D, and the CAMAC equipment allowed 32 high performance simulators to be controlled by centrally located host computers. This technology broadened Kinetic Systems' capabilities and led to several commercial applications. One of them is General Atomics' fusion research program. Kinetic Systems equipment allows tokamak data to be acquired four to 15 times more rapidly. Ford Motor company uses the same technology to control and monitor transmission testing facilities.

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

NASA Astrophysics Data System (ADS)

Wang, Y. F.; Xu, G. S.; Wan, B. N.; Li, G. Q.; Yan, N.; Li, Y. L.; Wang, H. Q.; Peng, Y.-K. Martin; Xia, T. Y.; Ding, S. Y.; Chen, R.; Yang, Q. Q.; Liu, H. Q.; Zang, Q.; Zhang, T.; Lyu, B.; Xu, J. C.; Feng, W.; Wang, L.; Chen, Y. J.; Luo, Z. P.; Hu, G. H.; Zhang, W.; Shao, L. M.; Ye, Y.; Lan, H.; Chen, L.; Li, J.; Zhao, N.; Wang, Q.; Snyder, P. B.; Liang, Y.; Qian, J. P.; Gong, X. Z.; EAST team

2018-05-01

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (I p = 600 kA), pedestal in the long-pulse H-mode discharge (I p = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge {Z}{{e}{{f}}{{f}}} value.

Study of Globus-M Tokamak Poloidal System and Plasma Position Control

NASA Astrophysics Data System (ADS)

Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

2017-12-01

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

The possible role of Reynolds stress in the creation of a transport barrier in tokamak edge plasmas

NASA Astrophysics Data System (ADS)

Vergote, M.; van Schoor, M.; Xu, Y.; Jachmich, S.; Weynants, R.; Hron, M.; Stöckel, J.

2005-03-01

To obtain a good confinement, mandatory in a fusion reactor, the understanding of the formation of transport barriers in the edge plasma of a tokamak is essential. Turbulence, the major candidate to explain anomalous transport, can be quenched by sheared flows in the edge which rip the convective cells apart, thus forming a barrier. Experimental evidence from the Chinese HT-6M tokamak [Y.H. Xu et al.: Phys. Rev. Lett. 84 (2000) 3867], points to the fact that momentum transfer from the turbulence can create these sheared flows via the Reynolds stresses. A new 1-D fluid model for the generation of the poloidal flow, has been developed taking into account the driving force of the Reynolds stress and the friction forces due to neutrals and parallel viscosity. Special attention has been dedicated to the computation of the flux-surface-averaging for the various terms. This model has been confronted with the experimental results obtained in the HT-6M tokamak, where Reynolds stresses were generated by application of a turbulent heating pulse. If the model is applied in cylindrical geometry, the calculated Reynolds stress-induced flow agrees well with the measured poloidal velocity in the plasma edge. However, when the full toroidal geometry is taken into account, it seems that the Reynolds stresses are too small to explain the observed rotation. This indicates that the role of the Reynolds stresses in inducing macroscopic flow in the torus is weakened. A combined system of probes allowing to measure the Reynolds stress and the rotation velocity simultaneously, has been developed and installed on the CASTOR tokamak (Prague). We report here on the first results obtained.

Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, T. E.

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Punjabi, Alkesh; Ali, Halima

A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates ({psi},{theta}) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. {psi} is the toroidal magnetic flux and {theta} is the poloidal angle. Natural canonical coordinates ({psi},{theta},{phi}) can be transformed to physical position (R,Z,{phi}) using a canonical transformation. (R,Z,{phi}) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonicalmore » coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.« less

The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2016-10-01

The replacement of the metal centre-post in spherical tokamaks (STs) by a plasma centre-post (PCP, the TF current carrier) is the ideal scenario for a ST reactor. A simple rodless ultra low aspect-ratio tokamak (RULART) using a screw-pinch PCP ECR-assisted with an external solenoid has been proposed in the most compact RULART [Ribeiro C, SOFE-15]. There the solenoid provided the stabilizing field for the PCP and the toroidal electrical field for the tokamak start-up, which will stabilize further the PCP, acting as stabilizing closed conducting surface. Relative low TF will be required. The compactness (high ratio of plasma-spherical vessel volume) may provide passive stabilization and easier access to L-H mode transition. It is presented here: 1) stability analysis of the PCP (initially MHD stable due to the hollow J profile); 2) tokamak equilibrium simulations, and 3) potential use for aneutronic reactions studies via pairs of proton p and boron 11B ion beams in He plasmas. The beams' line-of-sights sufficiently miss the sources of each other, thus allowing a near maximum relative velocities and reactivity. The reactions should occur close to the PCP mid-plane. Some born alphas should cross the PCP and be dragged by the ion flow (higher momentum exchange) towards the anode but escape directly to a direct electricity converter. Others will reach evenly the vessel directly or via thermal diffusion (favourable heating by the large excursion 2a), leading to the lowest power wall load possible. This might be a potential hybrid direct-steam cycle conversion reactor scheme, nearly aneutronic, and with no ash or particle retention problems, as opposed to the D-T thermal reaction proposals.

Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices

DOE PAGES

Evans, T. E.

2015-11-13

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima

2008-12-01

A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates (ψ,θ) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. ψ is the toroidal magnetic flux and θ is the poloidal angle. Natural canonical coordinates (ψ,θ,φ) can be transformed to physical position (R,Z,φ) using a canonical transformation. (R,Z,φ) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonical coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotationmore » measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.« less

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal

DOE PAGES

Grierson, B. A.; Burrell, K. H.; Chrystal, C.; ...

2016-09-12

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. Furthermore, the unique combination of experimentally measuredmore » main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.« less

Residual zonal flows in tokamaks and stellarators at arbitrary wavelengths

NASA Astrophysics Data System (ADS)

Monreal, Pedro; Calvo, Iván; Sánchez, Edilberto; Parra, Félix I.; Bustos, Andrés; Könies, Axel; Kleiber, Ralf; Görler, Tobias

2016-04-01

In the linear collisionless limit, a zonal potential perturbation in a toroidal plasma relaxes, in general, to a non-zero residual value. Expressions for the residual value in tokamak and stellarator geometries, and for arbitrary wavelengths, are derived. These expressions involve averages over the lowest order particle trajectories, that typically cannot be evaluated analytically. In this work, an efficient numerical method for the evaluation of such expressions is reported. It is shown that this method is faster than direct gyrokinetic simulations performed with the Gene and EUTERPE codes. Calculations of the residual value in stellarators are provided for much shorter wavelengths than previously available in the literature. Electrons must be treated kinetically in stellarators because, unlike in tokamaks, kinetic electrons modify the residual value even at long wavelengths. This effect, that had already been predicted theoretically, is confirmed by gyrokinetic simulations.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high-field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and density regimemore » consistent with a hot-ion-mode fusion reactor. These simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

H-Mode Behavior Induced by Modulated Toroidal Current on HT-7 and HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Mao, J. S.; Luo, J. R.; Xu, Y. H.; Zhao, J. Y.; Zhang, X. M.; Li, J. G.; Zhang, X. M.; Gao, X.; Li, Y. D.; Jie, Y. X.; Wu, Z. W.; Hu, L. Q.; Liu, S. X.; Zhang, X. D.; Bao, Y.; Yang, K.; Wang, G. X.; Chen, L.; Shi, Y. J.; Qin, P. J.; Gu, X. M.; Cui, N. Z.; Fan, H. Y.; Chen, Y. F.; Xia, C. Y.; Ruan, H. L.; Tong, X. D.; Phillips, P. E.

2001-10-01

An improved Ohmic confinement phase (similar to H-mode) has been observed during Modulating Toroidal Current on the Hefei Tokamak-6M (HT-6M) and Hefei super-conducting Tokamak-7 (HT-7). This improved plasma confinement phase is characterized by: (a) an increase in ne and T_e(0); (b) reduced H_α radiation from the edge; (c) steeper density and temperature profiles at the edge; (d) a more negative radial electric field inside the limiter; (e) a deeper electrostatic potential well at the edge; (f) reduced magnetic fluctuations at the edge; (g) MHD suppressing; (h) and by an increase in global energy confinement time, τ _e, by 27%-45%. The well-like structure of the radial electric field E_r, appears at an L-H like transition.

Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

NASA Astrophysics Data System (ADS)

Pathan, F. S.; Khan, Z.; Semwal, P.; Raval, D. C.; Joshi, K. S.; Thankey, P. L.; Dhanani, K. R.

2008-05-01

Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN2 panels during sniffer test and pressure drop test using helium gas.

Turbulent equipartition pinch of toroidal momentum in spherical torus

NASA Astrophysics Data System (ADS)

Hahm, T. S.; Lee, J.; Wang, W. X.; Diamond, P. H.; Choi, G. J.; Na, D. H.; Na, Y. S.; Chung, K. J.; Hwang, Y. S.

2014-12-01

We present a new analytic expression for turbulent equipartition (TEP) pinch of toroidal angular momentum originating from magnetic field inhomogeneity of spherical torus (ST) plasmas. Starting from a conservative modern nonlinear gyrokinetic equation (Hahm et al 1988 Phys. Fluids 31 2670), we derive an expression for pinch to momentum diffusivity ratio without using a usual tokamak approximation of B ∝ 1/R which has been previously employed for TEP momentum pinch derivation in tokamaks (Hahm et al 2007 Phys. Plasmas 14 072302). Our new formula is evaluated for model equilibria of National Spherical Torus eXperiment (NSTX) (Ono et al 2001 Nucl. Fusion 41 1435) and Versatile Experiment Spherical Torus (VEST) (Chung et al 2013 Plasma Sci. Technol. 15 244) plasmas. Our result predicts stronger inward pinch for both cases, as compared to the prediction based on the tokamak formula.

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

NASA Astrophysics Data System (ADS)

Azizov, E. A.; Gladush, G. G.; Dokuka, V. N.; Khayrutdinov, R. R.

2015-12-01

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

Electron Bernstein waves in spherical torus plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saveliev, A. N.

2006-11-30

Propagation and absorption of the electron Bernstein waves (EBWs) in spherical tokamaks (STs) have been intensively discussed in recent years because the EBWs coupled with an externally launched electromagnetic beam seem to be the only opportunity for microwave plasma heating and current drive in the electron cyclotron (EC) frequency range in the STs. The whole problem of the electron Bernstein heating and current drive (EBWHCD) in spherical plasmas is naturally divided into three major parts: coupling of incident electromagnetic waves (EMWs) to the EBWs near the upper hybrid resonance (UHR) surface, propagation and absorption of the EBWs in the plasmamore » interior and generation of noninductive current driven by the EBWs. The present paper is a brief survey of the most important theoretical and numerical results on the issue of EBWs.« less

Interpretation of the molecular fluxes measured at the periphery of a magnetically confined plasma

NASA Astrophysics Data System (ADS)

Liu-Hinz, C.; Terreault, B.; Martin, F.

1995-04-01

A new instrument ("Variable Geometry Sniffer Probe" or VGSP), allowing one to sample and mass analyse atoms, ions or molecules moving in different directions and at different locations at the periphery of a plasma, has been built and used in plasma edge studies in the TdeV tokamak. Three different regimes of particle sampling have been identified. First, the VGSP can measure the fluxes of hydrogen and impurity molecules issuing from the walls. Second, it has the capability of detecting low energy charge-exchange and Franck-Condon neutrals. Finally, there is a parallel ion flux sampling regime, for which it is shown that both the connection lengths to the divertor plates and the X × B plasma flows induced by edge electric fields play major roles.

The cluster charge identification in the GEM detector for fusion plasma imaging by soft X-ray diagnostics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Czarski, T., E-mail: tomasz.czarski@ifpilm.pl; Chernyshova, M.; Malinowski, K.

2016-11-15

The measurement system based on gas electron multiplier detector is developed for soft X-ray diagnostics of tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an X-ray source. The focal measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the serial data acquisition is applied for the dynamic plasma diagnostics. The charge clusters are counted in the space determined by 2D position, charge value, and time intervals. Radiation source characteristics are presented by histograms for a selected range of position, time intervals,more » and cluster charge values corresponding to the energy spectra.« less

Mechanisms of Stochastic Diffusion of Energetic Ions in Spherical Tori

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ya.I. Kolesnichenko; R.B. White; Yu.V. Yakovenko

Stochastic diffusion of the energetic ions in spherical tori is considered. The following issues are addressed: (I) Goldston-White-Boozer diffusion in a rippled field; (ii) cyclotron-resonance-induced diffusion caused by the ripple; (iii) effects of non-conservation of the magnetic moment in an axisymmetric field. It is found that the stochastic diffusion in spherical tori with a weak magnetic field has a number of peculiarities in comparison with conventional tokamaks; in particular, it is characterized by an increased role of mechanisms associated with non-conservation of the particle magnetic moment. It is concluded that in current experiments on National Spherical Torus eXperiment (NSTX) themore » stochastic diffusion does not have a considerable influence on the confinement of energetic ions.« less

Overview of Edge Simulation Laboratory (ESL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.; Dorr, M.; Hittinger, J.; Rognlien, T.; Umansky, M.; Xiong, A.; Xu, X.; Belli, E.; Candy, J.; Snyder, P.; Colella, P.; Martin, D.; Sternberg, T.; van Straalen, B.; Bodi, K.; Krasheninnikov, S.

2006-10-01

The ESL is a new collaboration to build a full-f electromagnetic gyrokinetic code for tokamak edge plasmas using continuum methods. Target applications are edge turbulence and transport (neoclassical and anomalous), and edge-localized modes. Initially the project has three major threads: (i) verification and validation of TEMPEST, the project's initial (electrostatic) edge code which can be run in 4D (neoclassical and transport-timescale applications) or 5D (turbulence); (ii) design of the next generation code, which will include more complete physics (electromagnetics, fluid equation option, improved collisions) and advanced numerics (fully conservative, high-order discretization, mapped multiblock grids, adaptivity), and (iii) rapid-prototype codes to explore the issues attached to solving fully nonlinear gyrokinetics with steep radial gradiens. We present a brief summary of the status of each of these activities.

The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

NASA Astrophysics Data System (ADS)

Rubel, Marek; Petersson, Per; Alves, Eduardo; Brezinsek, Sebastijan; Coad, Joseph Paul; Heinola, Kalle; Mayer, Matej; Widdowson, Anna

2016-03-01

First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1-10), high sensitivity and combination of several methods in a single run. The role of 3He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. 15N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

Experiment to investigate current drive by fast Alfven waves in a small tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gahl, J.; Ishihara, O.; Wong, K.

1985-07-01

An experiment has been carried out to study current generation by Doppler shifted cyclotron resonance heating of minority ions with a unidirectional wave in the small tokamak at Texas Tech University. One of the objectives of the experiment is to understand in detail the wave-particle interactions through which fast (compressional) Alfven waves in the ion cyclotron range of frequencies drive currents in toroidal devices.

SUNIST Microwave Power System

NASA Astrophysics Data System (ADS)

Feng, Songlin; Yang, Xuanzong; Feng, Chunhua; Wang, Long; Rao, Jun; Feng, Kecheng

2005-06-01

Experiments on the start-up and formation of spherical tokamak plasmas by electron cyclotron heating alone without ohmic heating and electrode discharge assisted electron cyclotron wave current start-up will be carried out on the SUNIST (Sino United Spherical Tokamak) device. The 2.45 GHz/100kW/30 ms microwave power system and 1000 V/50 A power supply for electrode discharge are ready for experiments with non-inductive current drive.

Design of geometric phase measurement in EAST Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lan, T.; Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031; Liu, H. Q., E-mail: hqliu@ipp.ac.cn

2016-07-15

The optimum scheme for geometric phase measurement in EAST Tokamak is proposed in this paper. The theoretical values of geometric phase for the probe beams of EAST Polarimeter-Interferometer (POINT) system are calculated by path integration in parameter space. Meanwhile, the influences of some controllable parameters on geometric phase are evaluated. The feasibility and challenge of distinguishing geometric effect in the POINT signal are also assessed in detail.

Transition to subcritical turbulence in a tokamak plasma

NASA Astrophysics Data System (ADS)

van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.

2016-12-01

Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.

New receiving line for the remote-steering antenna of the 140 GHz CTS diagnostics in the FTU Tokamak

NASA Astrophysics Data System (ADS)

D'Arcangelo, O.; Bin, W.; Bruschi, A.; Cappelli, M.; Fanale, F.; Gittini, G.; Pallotta, F.; Rocchi, G.; Tudisco, O.; Garavaglia, S.; Granucci, G.; Moro, A.; Tuccillo, A. A.

2018-01-01

A new receiving antenna for collecting signals of the Collective Thomson Scattering (CTS) diagnostics in FTU Tokamak has been recently installed. The squared corrugated section and the precisely defined length make it possible to receive from different directions by remotely steering the receiving mirrors. This type of Remote-Steering (RS) antennas, being studied on FTU for the DEMO Electron Cyclotron Heating (ECH) system launch, is already installed on the W7- X stellarator and will be tested in the next campaign. The transmission of the signal from the antenna in the tokamak hall to the CTS diagnostics hall will be mainly realized by means of oversized circular corrugated waveguides carrying the hybrid HE11 (quasi-gaussian) waveguide mode, with inclusion of a special smooth-waveguide section and a short run of reduced-size square-corrugated waveguide through the tokamak bio-shield. The coupling between different waveguide types is made with ellipsoidal focusing mirrors, using quasi-optical matching formulas between the gaussian-shaped beams in input and output to the waveguides. In this work, after a complete study of feasibility of the overall line, a design for the receiving line will be proposed, in order to realize an executive layout to be used as a guideline for the commissioning phase.

Fusion Breeding for Sustainable, Mid Century, Carbon Free Power

NASA Astrophysics Data System (ADS)

Manheimer, Wallace

2015-11-01

If ITER achieves Q ~10, it is still very far from useful fusion. The fusion power, and the driver power will allow only a small amount of power to be delivered, <~50MW for an ITER scale tokamak. It is unlikely, considering ``conservative design rules'' that tokamaks can ever be economical pure fusion power producers. Considering the status of other magnetic fusion concepts, it is also very unlikely that any alternate concept will either. Laser fusion does not seem to be constrained by any conservative design rules, but considering the failure of NIF to achhieve ignition, at this point it has many more obstacles to overcome than magnetic fusion. One way out of this dilemma is to use an ITER size tokamak, or a NIF size laser, as a fuel breeder for searate nuclear reactors. Hence ITER and NIF become ends in themselves, instead of steps to who knows what DEMO decades later. Such a tokamak can easily live within the consrtaints of conservative design rules. This has led the author to propose ``The Energy Park'' a sustainable, carbon free, economical, and environmently viable power source without prolifertion risk. It is one fusion breeder fuels 5 conventional nuclear reactors, and one fast neutron reactor burns the actinide wastes.

On the breakdown modes and parameter space of Ohmic Tokamak startup

NASA Astrophysics Data System (ADS)

Peng, Yanli; Jiang, Wei; Zhang, Ya; Hu, Xiwei; Zhuang, Ge; Innocenti, Maria; Lapenta, Giovanni

2017-10-01

Tokamak plasma has to be hot. The process of turning the initial dilute neutral hydrogen gas at room temperature into fully ionized plasma is called tokamak startup. Even with over 40 years of research, the parameter ranges for the successful startup still aren't determined by numerical simulations but by trial and errors. However, in recent years it has drawn much attention due to one of the challenges faced by ITER: the maximum electric field for startup can't exceed 0.3 V/m, which makes the parameter range for successful startup narrower. Besides, this physical mechanism is far from being understood either theoretically or numerically. In this work, we have simulated the plasma breakdown phase driven by pure Ohmic heating using a particle-in-cell/Monte Carlo code, with the aim of giving a predictive parameter range for most tokamaks, even for ITER. We have found three situations during the discharge, as a function of the initial parameters: no breakdown, breakdown and runaway. Moreover, breakdown delay and volt-second consumption under different initial conditions are evaluated. In addition, we have simulated breakdown on ITER and confirmed that when the electric field is 0.3 V/m, the optimal pre-filling pressure is 0.001 Pa, which is in good agreement with ITER's design.

Polarized fusion, its implications, and plans for a proof-of-principle experiment at the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Sandorfi, A. M.; Deur, A.; Lowry, M. M.; Wei, X.; Pace, D.; Eidietis, N.; Hyatt, A.; Jackson, G. L.; Lanctot, M.; Smith, S.; St-John, H.; Miller, G. W.; Zheng, X.; Baylor, L. R.

2015-10-01

The cross section for the primary fusion reaction in a tokamak, D+t --> α +n, would increase by a factor of 1.5 if the fuels were spin polarized parallel to the local field, rather than randomly oriented. Simulations show further gains in reaction rate would accompany this increase in large-scale machines such as ITER, due to increased alpha heating. The potential realization of such benefits rests on the crucial question of the survival of spin polarization for periods comparable to the energy containment time. Despite encouraging calculations, technical challenges in preparing and handling polarized materials have prevented any direct tests. Advances in three areas - polarized material technologies developed for nuclear and particle physics as well as medical imaging, polymer pellets developed for Inertial Confinement, and cryogenic injection guns developed for fueling tokamaks - have matured to the point where a direct in situ measurement is possible using the mirror reaction, D+3He --> α +p. Designs and simulations of a proof-of-principle experiment at the DIII-D tokamak in San Diego will be discussed. Work carried out under US DOE Contract DE-AC05-06OR23177 supporting Jefferson Lab and General Atomics Internal R&D funding.

Dependence of intrinsic torque and momentum confinement on normalized gyroradius and collisionality in the DIII-D tokamak

DOE PAGES

Chrystal, C.; Grierson, B. A.; Solomon, W. M.; ...

2017-03-29

We measured the dependence of intrinsic torque and momentum confinement time on normalized gyroradius (ρ *) and collisionality (v *) in the DIII-D tokamak. The intrinsic torque normalized to temperature is found to have ρ * and v * dependencies of ρ * -1.5 ± 0.8 and v * -0.26 ± 0.04. This dependence on ρ * is unexpectedly favorable (increasing as ρ * decreases). The choice of normalization is important, and the implications are discussed. The unexpected dependence on ρ * is found to be robust, despite some uncertainty in the choice of normalization. Furthermore, the dependence of momentummore » confinement on ρ * does not clearly demonstrate Bohm or gyro-Bohm like scaling, and a weaker dependence on v * is found. The calculations required to use these dependencies to determine the intrinsic torque in future tokamaks such as ITER are presented, and the importance of the normalization is explained. Based on the currently available information, the intrinsic torque predicted for ITER is 33 N m, comparable to the expected torque available from neutral beam injection. The expected average intrinsic rotation associated with this intrinsic torque is small compared to current tokamaks, but it may still aid stability and performance in ITER. Published by AIP Publishing.« less

Extracting 3D Information from 1D and 2D Diagnostic Systems on the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Brookman, Michael

2017-10-01

The interpretation of tokamak data often hinges on assumptions of axisymetry and flux surface equilibria, neglecting 3D effects. This work discusses examples on the DIII-D tokamak where this assumption is an insufficient approximation, and explores the diagnostic information available to resolve 3D effects while preserving 1D profiles. Methods for extracting 3D data from the electron cyclotron emission radiometers, density profile reflectometer, and Thomson scattering system are discussed. Coordinating diagnostics around the tokamak shows the significance of 3D features, such as sawteeth[1] and resonant magnetic perturbations. A consequence of imposed 3D perturbations is a shift in major radius of measured profiles between diagnostics at different toroidal locations. Integrating different diagnostics requires a database containing information about their toroidal, poloidal, and radial locations. Through the data analysis framework OMFIT, it is possible to measure the magnitude of the apparent shifts from 3D effects and enforce consistency between diagnostics. Using the existing 1D and 2D diagnostic systems on DIII-D, this process allows the effects of the 3D perturbations on 1D profiles to be addressed. Supported by US DOE contracts DE-FC02-04ER54698, DE-FG03-97ER54415.

Largescale Long-term particle Simulations of Runaway electrons in Tokamaks

NASA Astrophysics Data System (ADS)

Liu, Jian; Qin, Hong; Wang, Yulei

2016-10-01

To understand runaway dynamical behavior is crucial to assess the safety of tokamaks. Though many important analytical and numerical results have been achieved, the overall dynamic behaviors of runaway electrons in a realistic tokamak configuration is still rather vague. In this work, the secular full-orbit simulations of runaway electrons are carried out based on a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different timescales spanning 11 orders. A detailed analysis of the collisionless neoclassical scattering is provided when considering the coupling between the rotation of momentum vector and the background field. In large timescale, the initial condition of runaway electrons in phase space globally influences the runaway distribution. It is discovered that parameters and field configuration of tokamaks can modify the runaway electron dynamics significantly. Simulations on 10 million cores of supercomputer using the APT code have been completed. A resolution of 107 in phase space is used, and simulations are performed for 1011 time steps. Largescale simulations show that in a realistic fusion reactor, the concern of runaway electrons is not as serious as previously thought. This research was supported by National Magnetic Connement Fusion Energy Research Project (2015GB111003, 2014GB124005), the National Natural Science Foundation of China (NSFC-11575185, 11575186) and the GeoAlgorithmic Plasma Simulator (GAPS) Project.

SCTE: An open-source Perl framework for testing equipment control and data acquisition

NASA Astrophysics Data System (ADS)

Mostaço-Guidolin, Luiz C.; Frigori, Rafael B.; Ruchko, Leonid; Galvão, Ricardo M. O.

2012-07-01

SCTE intends to provide a simple, yet powerful, framework for building data acquisition and equipment control systems for experimental Physics, and correlated areas. Via its SCTE::Instrument module, RS-232, USB, and LAN buses are supported, and the intricacies of hardware communication are encapsulated underneath an object oriented abstraction layer. Written in Perl, and using the SCPI protocol, enabled instruments can be easily programmed to perform a wide variety of tasks. While this work presents general aspects of the development of data acquisition systems using the SCTE framework, it is illustrated by particular applications designed for the calibration of several in-house developed devices for power measurement in the tokamak TCABR Alfvén Waves Excitement System. Catalogue identifier: AELZ_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AELZ_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: GNU General Public License Version 3 No. of lines in distributed program, including test data, etc.: 13 811 No. of bytes in distributed program, including test data, etc.: 743 709 Distribution format: tar.gz Programming language: Perl version 5.10.0 or higher. Computer: PC. SCPI capable digital oscilloscope, with RS-232, USB, or LAN communication ports, null modem, USB, or Ethernet cables Operating system: GNU/Linux (2.6.28-11), should also work on any Unix-based operational system Classification: 4.14 External routines: Perl modules: Device::SerialPort, Term::ANSIColor, Math::GSL, Net::HTTP. Gnuplot 4.0 or higher Nature of problem: Automation of experiments and data acquisition often requires expensive equipment and in-house development of software applications. Nowadays personal computers and test equipment come with fast and easy-to-use communication ports. Instrument vendors often supply application programs capable of controlling such devices, but are very restricted in terms of functionalities. For instance, they are not capable of controlling more than one test equipment at a same time or to automate repetitive tasks. SCTE provides a way of using auxiliary equipment in order to automate experiment procedures at low cost using only free, and open-source operational system and libraries. Solution method: SCTE provides a Perl module that implements RS-232, USB, and LAN communication allowing the use of SCPI capable instruments [1]. Therefore providing a straightforward way of creating automation and data acquisition applications using personal computers and testing instruments [2]. SCPI Consortium, Standard Commands for Programmable Instruments, 1999, http://www.scpiconsortium.org. L.C.B. Mostaço-Guidolin, Determinação da configuração de ondas de Alfvén excitadas no tokamak TCABR, Master's thesis, Universidade de São Paulo (2007), http://www.teses.usp.br/teses/disponiveis/43/43134/tde-23042009-230419/.

A Real-Time Data Acquisition and Processing Framework Based on FlexRIO FPGA and ITER Fast Plant System Controller

NASA Astrophysics Data System (ADS)

Yang, C.; Zheng, W.; Zhang, M.; Yuan, T.; Zhuang, G.; Pan, Y.

2016-06-01

Measurement and control of the plasma in real-time are critical for advanced Tokamak operation. It requires high speed real-time data acquisition and processing. ITER has designed the Fast Plant System Controllers (FPSC) for these purposes. At J-TEXT Tokamak, a real-time data acquisition and processing framework has been designed and implemented using standard ITER FPSC technologies. The main hardware components of this framework are an Industrial Personal Computer (IPC) with a real-time system and FlexRIO devices based on FPGA. With FlexRIO devices, data can be processed by FPGA in real-time before they are passed to the CPU. The software elements are based on a real-time framework which runs under Red Hat Enterprise Linux MRG-R and uses Experimental Physics and Industrial Control System (EPICS) for monitoring and configuring. That makes the framework accord with ITER FPSC standard technology. With this framework, any kind of data acquisition and processing FlexRIO FPGA program can be configured with a FPSC. An application using the framework has been implemented for the polarimeter-interferometer diagnostic system on J-TEXT. The application is able to extract phase-shift information from the intermediate frequency signal produced by the polarimeter-interferometer diagnostic system and calculate plasma density profile in real-time. Different algorithms implementations on the FlexRIO FPGA are compared in the paper.

Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Medley, S.S.

The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10/sup -3/ Torr is required to maintain the analyzer chamber at a pressure of < or approx. = 10/sup -6/ Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of < or approx.more » = 10/sup -8/ Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus.« less

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

3D toroidal physics: Testing the boundaries of symmetry breakinga)

NASA Astrophysics Data System (ADS)

Spong, Donald A.

2015-05-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to provide the plasma control needed for a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D edge localized mode suppression fields to stellarators with more dominant 3D field structures. This motivates the development of physics models that are applicable across the full range of 3D devices. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with the requirements of future fusion reactors.

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

NASA Astrophysics Data System (ADS)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Physics through the 1990s: Plasmas and fluids

NASA Technical Reports Server (NTRS)

1986-01-01

The volume contains recommendations for programs in, and government support of, plasma and fluid physics. Four broad areas are covered: the physics of fluids, general plasma physics, fusion, and space and astrophysical plasmas. In the first section, the accomplishments of fluid physics and a detailed review of its sub-fields, such as combustion, non-Newtonian fluids, turbulence, aerodynamics, and geophysical fluid dynamics, are described. The general plasma physics section deals with the wide scope of the theoretical concepts involved in plasma research, and with the machines; intense beam systems, collective and laser-driven accelerators, and the associated diagnostics. The section on the fusion plasma research program examines confinement and heating systems, such as Tokamaks, magnetic mirrors, and inertial-confinement systems, and several others. Finally, theory and experiment in space and astrophysical plasma research is detailed, ranging from the laboratory to the solar system and beyond. A glossary is included.

EDITORIAL: Special issue containing papers presented at the 11th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems Special issue containing papers presented at the 11th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems

NASA Astrophysics Data System (ADS)

Kolesnichenko, Ya.

2010-08-01

The history of fusion research resembles the way in which one builds skyscrapers: laying the first foundation stone, one thinks about the top of the skyscraper. At the early stages of fusion, when it became clear that the thermonuclear reactor would operate with DT plasma confined by the magnetic field, the study of the `top item'—the physics of 3.5 MeV alpha particles produced by the DT fusion reaction—was initiated. The first publications on this topic appeared as long ago as the 1960s. At that time, because the physics of alpha particles was far from the experimental demand, investigations were carried out by small groups of theoreticians who hoped to discover important and interesting phenomena in this new research area. Soon after the beginning of the work, theoreticians discovered that alpha particles could excite various instabilities in fusion plasmas. In particular, at the end of the 1960s an Alfvén instability driven by alpha particles was predicted. Later it turned out that a variety of Alfvén instabilities with very different features does exist. Instabilities with perturbations of the Alfvénic type play an important role in current experiments; it is likely that they will affect plasma performance in ITER and future reactors. The first experimental manifestation of instabilities excited by superthermal particles in fusion devices was observed in the PDX tokamak in 1983. In this device a large-scale instability—the so called `fishbone instability'—associated with ions produced by the neutral beam injection resulted in a loss of a large fraction of the injected energy. Since then, the study of energetic-ion-driven instabilities and the effects produced by energetic ions in fusion plasmas has attracted the growing attention of both experimentalists and theorists. Recognizing the importance of this topic, the first conference on fusion alpha particles was held in 1989 in Kyiv under the auspices of the IAEA. The meeting in Kyiv and several subsequent meetings (Aspenäs (1991), Trieste (1993), Princeton (1995), and JET/Abingdon (1997)) were entitled `Alpha Particles in Fusion Research'. During the JET/Abingdon meeting in 1997 it was decided to extend the topic by including other suprathermal particles, in particular accelerated electrons, and rename the meetings accordingly. The subsequent meetings with the current name `Energetic Particles in Magnetic Confinement Systems' were held in Naka (1999), Gothenburg (2001), San Diego (2003), Takayama (2005) and Kloster Seeon (2007). The most recent meeting in this series was held in Kyiv, Ukraine, in September 2009. This was an anniversary meeting, 20 years after the first meeting. Like the first meeting, it was hosted by the Institute for Nuclear Research, National Academy of Sciences of Ukraine. It was attended by about 80 researchers from 18 countries, ITER, and EC. The program of the meeting consisted of 78 presentations, including 12 invited talks, 16 oral contributed talks, and 50 posters, which were selected by the International Advisory Committee (IAC). The IAC consisted of 11 people representing EC (L.-G. Eriksson), Germany (S. Günter), Italy (F. Zonca), Japan (K. Shinohara and K. Toi), Switzerland (A. Fasoli), UK (S. Sharapov), Ukraine (Ya. Kolesnichenko—IAC Chair), USA (H. Berk, W. Heidbrink, and R. Nazikian). The meeting program covered a wide range of physics issues concerning energetic ions in toroidal fusion facilities—tokamaks, stellarators, and spherical tori. Many new interesting and practically important results of both experimental and theoretical studies were reported. The research presented covered topics such as instabilities driven by energetic ions, transport of energetic ions caused by plasma microturbulence and destabilized eigenmodes, non-linear phenomena induced by the instabilities, classical transport processes, effects of runaway electrons, diagnostics of energetic ions and plasmas, and aspects of ITER physics. In addition to these topics, which were also covered at previous conferences in this series and have become conventional, experimental and theoretical results on the influence of energetic ions on bulk plasma transport properties were also reported. Some materials from the meeting are available on the web page http://www.kinr.kiev.ua/TCM/index.html. 24 of the works presented at the meeting are published in this special issue. These works were reviewed to the usual high standard of Nuclear Fusion. The guest editor of this special issue is grateful to the publishers for their cooperation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaing, K. C.

It is shown that potato orbits in the near-axis region of a high beta tokamak are squeezed in a magnetic well. The squeezing factor is the same as that for the banana orbits derived in an earlier work [Phys. Plasmas 3, 2843 (1996)]. It depends on the energy of the particle. For high-energy particles, the size of the squeezed orbits is independent of their energy. This implies improved confinement for high-energy particles and for high beta tokamaks with advanced fuels.

Tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wesson, J

1987-01-01

The word tokamak derives from the Russian term, toroidalnaya kamera magnitaya (toroidal chamber magnetic). The device was invented in the Soviet Union in 1950 and has since developed into one of the chief ways in which it is hoped to obtain usable power from plasmas through thermonuclear fusion. The present is meant to be an introduction to those entering the field, to those already engaged in research, and to those who want to gain some understanding of what it's all about.

Role of bremsstrahlung radiation in limiting the energy of runaway electrons in tokamaks.

PubMed

Bakhtiari, M; Kramer, G J; Takechi, M; Tamai, H; Miura, Y; Kusama, Y; Kamada, Y

2005-06-03

Bremsstrahlung radiation of runaway electrons is found to be an energy limit for runaway electrons in tokamaks. The minimum and maximum energy of runaway electron beams is shown to be limited by collisions and bremsstrahlung radiation, respectively. It is also found that a massive injection of a high-Z gas such as xenon can terminate a disruption-generated runaway current before the runaway electrons hit the walls.

Coupling Effect between Equilibrium Field and Heating Field and Modification of the Power Supply System on SUNIST Spherical Tokamak

NASA Astrophysics Data System (ADS)

He, Yexi; Li, Xiaoyan; Gao, Zhe

2005-02-01

Strong inductive coupling between the heating field and equilibrium field is confirmed to be responsible for the poor plasma equilibrium in initial discharges on the SUNIST spherical tokamak. A modification project for the power supply system of equilibrium field coils is successfully performed to increase the duration time of plasma current flattop from much less than 1ms to about 2 ms.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

Analytic study on low- external ideal infernal modes in tokamaks with large edge pressure gradients

NASA Astrophysics Data System (ADS)

Brunetti, Daniele; Graves, J. P.; Lazzaro, E.; Mariani, A.; Nowak, S.; Cooper, W. A.; Wahlberg, C.

2018-04-01

The problem of pressure driven infernal type perturbations near the plasma edge is addressed analytically for a circular limited tokamak configuration which presents an edge flattened safety factor. The plasma is separated from a metallic wall, either ideally conducting or resistive, by a vacuum region. The dispersion relation for such types of instabilities is derived and discussed for two classes of equilibrium profiles for pressure and mass density.

Determination of 3D Equilibria from Flux Surface Knowledge Only

DOE Office of Scientific and Technical Information (OSTI.GOV)

H.E. Mynick; N. Pomphrey

We show that the method of Christiansen and Taylor, from which complete tokamak equilibria can be determined given only knowledge of the shape of the flux surfaces, can be extended to 3-dimensional equilibria, such as those of stellarators. As for the tokamak case, the given geometric knowledge has a high degree of redundancy, so that the full equilibrium can be obtained using only a small portion of that information.

Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D.A.; Gibson, G.

1979-03-01

The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based onmore » two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma.« less

First Direct Observation of Runaway-Electron-Driven Whistler Waves in Tokamaks

NASA Astrophysics Data System (ADS)

Spong, D. A.; Heidbrink, W. W.; Paz-Soldan, C.; Du, X. D.; Thome, K. E.; Van Zeeland, M. A.; Collins, C.; Lvovskiy, A.; Moyer, R. A.; Austin, M. E.; Brennan, D. P.; Liu, C.; Jaeger, E. F.; Lau, C.

2018-04-01

DIII-D experiments at low density (ne˜1019 m-3 ) have directly measured whistler waves in the 100-200 MHz range excited by multi-MeV runaway electrons. Whistler activity is correlated with runaway intensity (hard x-ray emission level), occurs in novel discrete frequency bands, and exhibits nonlinear limit-cycle-like behavior. The measured frequencies scale with the magnetic field strength and electron density as expected from the whistler dispersion relation. The modes are stabilized with increasing magnetic field, which is consistent with wave-particle resonance mechanisms. The mode amplitudes show intermittent time variations correlated with changes in the electron cyclotron emission that follow predator-prey cycles. These can be interpreted as wave-induced pitch angle scattering of moderate energy runaways. The tokamak runaway-whistler mechanisms have parallels to whistler phenomena in ionospheric plasmas. The observations also open new directions for the modeling and active control of runaway electrons in tokamaks.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru

The radial force balance in a tokamak during fast transient events with a duration much shorter than the resistive time of the vacuum vessel wall is analyzed. The aim of the work is to analytically estimate the resulting integral radial force on the wall. In contrast to the preceding study [Plasma Phys. Rep. 41, 952 (2015)], where a similar problem was considered for thermal quench, simultaneous changes in the profiles and values of the pressure and plasma current are allowed here. Thereby, the current quench and various methods of disruption mitigation used in the existing tokamaks and considered for futuremore » applications are also covered. General formulas for the force at an arbitrary sequence or combination of events are derived, and estimates for the standard tokamak model are made. The earlier results and conclusions are confirmed, and it is shown that, in the disruption mitigation scenarios accepted for ITER, the radial forces can be as high as in uncontrolled disruptions.« less

Breakdown assisted by a novel electron drift injection in the J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Wang, Nengchao; Jin, Hai; Zhuang, Ge; Ding, Yonghua; Pan, Yuan; Cen, Yishun; Chen, Zhipeng; Huang, Hai; Liu, Dequan; Rao, Bo; Zhang, Ming; Zou, Bichen

2014-07-01

A novel electron drift injection (EDI) system aiming to improve breakdown behavior has been designed and constructed on the Joint Texas EXperiment Tokamak Tokamak. Electrons emitted by the system undergo the E×B drift, ∇B drift and curvature drift in sequence in order to traverse the confining magnetic field. A local electrostatic well, generated by a concave-shaped plate biased more negative than the cathode, is introduced to interrupt the emitted electrons moving along the magnetic field line (in the parallel direction) in an attempt to bring an enhancement of the injection efficiency and depth. A series of experiments have demonstrated the feasibility of this method, and a penetration distance deeper than 9.5 cm is achieved. Notable breakdown improvements, including the reduction of breakdown delay and average loop voltage, are observed for discharges assisted by EDI. The lower limit of successfully ionized pressure is expanded.

Analysis of higher harmonics on bidirectional heat pulse propagation experiment in helical and tokamak plasmas

NASA Astrophysics Data System (ADS)

Kobayashi, T.; Ida, K.; Inagaki, S.; Tsuchiya, H.; Tamura, N.; Choe, G. H.; Yun, G. S.; Park, H. K.; Ko, W. H.; Evans, T. E.; Austin, M. E.; Shafer, M. W.; Ono, M.; López-bruna, D.; Ochando, M. A.; Estrada, T.; Hidalgo, C.; Moon, C.; Igami, H.; Yoshimura, Y.; Tsujimura, T. Ii.; Itoh, S.-I.; Itoh, K.

2017-07-01

In this contribution we analyze modulation electron cyclotron resonance heating (MECH) experiment and discuss higher harmonic frequency dependence of transport coefficients. We use the bidirectional heat pulse propagation method, in which both inward propagating heat pulse and outward propagating heat pulse are analyzed at a radial range, in order to distinguish frequency dependence of transport coefficients due to hysteresis from that due to other reasons, such as radially dependent transport coefficients, a finite damping term, or boundary effects. The method is applied to MECH experiments performed in various helical and tokamak devices, i.e. Large Helical Device (LHD), TJ-II, Korea Superconducting Tokamak Advanced Research (KSTAR), and Doublet III-D (DIII-D) with different plasma conditions. The frequency dependence of transport coefficients are clearly observed, showing a possibility of existence of transport hysteresis in flux-gradient relation.

Alpha channeling with high-field launch of lower hybrid waves

DOE PAGES

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

2015-11-04

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

Electronic system for Langmuir probe measurements

NASA Astrophysics Data System (ADS)

Mitov, M.; Bankova, A.; Dimitrova, M.; Ivanova, P.; Tutulkov, K.; Djermanova, N.; Dejarnac, R.; Stöckel, J.; Popov, Tsv K.

2012-03-01

A newly developed Langmuir probe system for measurements of current-voltage (IV) characteristics in the tokamak divertor area is presented and discussed. The system is partially controlled by a computer allowing simultaneous and independent feeding and registration of signals. The system is mounted in the COMPASS tokamak, Institute of Plasma Physics, Academy of Sciences of the Czech Republic. The new electronic circuit boards include also active low-pass filters which smooth the signal before recording by the data acquisition system (DAQ). The signal is thus less noisy and the data processing is much easier. We also designed and built a microcontroller-driven waveform generator with resolution of 1 Ms/s. The power supply is linear and uses a transformer. We avoided the use of a switching power supply because of the noise that it could generate. Examples of measurements of the IV characteristics by divertor probes in the COMPASS tokamak and evaluation of the EEDF are presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

H.E. Mynick, N. Pomphrey and P. Xanthopoulos

Recent progress in reducing turbulent transport in stellarators and tokamaks by 3D shaping using a stellarator optimization code in conjunction with a gyrokinetic code is presented. The original applications of the method focussed on ion temperature gradient transport in a quasi-axisymmetric stellarator design. Here, an examination of both other turbulence channels and other starting configurations is initiated. It is found that the designs evolved for transport from ion temperature gradient turbulence also display reduced transport from other transport channels whose modes are also stabilized by improved curvature, such as electron temperature gradient and ballooning modes. The optimizer is also appliedmore » to evolving from a tokamak, finding appreciable turbulence reduction for these devices as well. From these studies, improved understanding is obtained of why the deformations found by the optimizer are beneficial, and these deformations are related to earlier theoretical work in both stellarators and tokamaks.« less

Application of automatic gain control for radiometer diagnostic in SST-1 tokamak.

PubMed

Makwana, Foram R; Siju, Varsha; Edappala, Praveenlal; Pathak, S K

2017-12-01

This paper describes the characterisation of a negative feedback type of automatic gain control (AGC) circuit that will be an integral part of the heterodyne radiometer system operating at a frequency range of 75-86 GHz at SST-1 tokamak. The developed AGC circuit is a combination of variable gain amplifier and log amplifier which provides both gain and attenuation typically up to 15 dB and 45 dB, respectively, at a fixed set point voltage and it has been explored for the first time in tokamak radiometry application. The other important characteristics are that it exhibits a very fast response time of 390 ns to understand the fast dynamics of electron cyclotron emission and can operate at very wide input RF power dynamic range of around 60 dB that ensures signal level within the dynamic range of the detection system.

RF assisted Glow Discharge Condition experiment for SST-1 Tokamak

NASA Astrophysics Data System (ADS)

Raval, Dilip; Khan, Ziauddin; George, Siju; Dhanani, Kalpeshkumar R.; Paravastu, Yuvakiran; Semwal, Pratibha; Thankey, Prashant; Shoaib Khan, Mohammad; Kakati, Bharat; Pradhan, Subrata

2017-04-01

Impurity control reduces the radiation loss from plasma and hence enhances the plasma operation. Oxygen and water vapors are the most common impurities in tokamak devices. Water vapour can be reduced with extensive baking while in order to have a significant reduction in oxygen it is necessary to use glow discharge condition (GDC). RF assisted glow discharge cleaning system will be implemented to remove low z impurities at PFC installed SST-1 vacuum vessel. A RF assisted Glow discharge conditioning is studied at laboratory to find the optimum operating parameters in a view to implement at SST-1 tokamak. Helium is used as a fuel gas in the present experiment. It is observed that the ultimate impurity level is reduced significantly below to the accepted level for plasma operation after RF assisted GDC. The experimental findings of RF assisted Glow discharge conditioning is discussed in details in this paper.

Phase Contrast Imaging on the HL-2A Tokamak

NASA Astrophysics Data System (ADS)

Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team

2016-10-01

In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).

Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

PubMed

Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

2012-01-10

Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

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.

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henline, P.A.

1995-12-31

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DIII-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henline, P.A.

1995-10-01

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.

2015-12-15

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket basedmore » on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.« less

Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieland, R.M.; Houlberg, W.A.

1982-02-01

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Depositionmore » profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10/sup 13/ - 2 x 10/sup 14/ cm/sup -3/, and beam orientation ranging from perpendicular to tangential to the inside wall.« less

Poloidal flux profile reconstruction from pointwise measurements via extended Kalman filtering in the DIII-D Tokamak

DOE PAGES

Wang, Hexiang; Barton, Justin E.; Schuster, Eugenio

2015-09-01

The accuracy of the internal states of a tokamak, which usually cannot be measured directly, is of crucial importance for feedback control of the plasma dynamics. A first-principles-driven plasma response model could provide an estimation of the internal states given the boundary conditions on the magnetic axis and at plasma boundary. However, the estimation would highly depend on initial conditions, which may not always be known, disturbances, and non-modeled dynamics. Here in this work, a closed-loop state observer for the poloidal magnetic flux is proposed based on a very limited set of real-time measurements by following an Extended Kalman Filteringmore » (EKF) approach. Comparisons between estimated and measured magnetic flux profiles are carried out for several discharges in the DIII-D tokamak. The experimental results illustrate the capability of the proposed observer in dealing with incorrect initial conditions and measurement noise.« less

Ideal MHD Stability Prediction and Required Power for EAST Advanced Scenario

NASA Astrophysics Data System (ADS)

Chen, Junjie; Li, Guoqiang; Qian, Jinping; Liu, Zixi

2012-11-01

The Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. The ideal magnetohydrodynamic (MHD) stability and required power for the EAST advanced tokamak (AT) scenario with negative central shear and double transport barrier (DTB) are investigated. With the equilibrium code TOQ and stability code GATO, the ideal MHD stability is analyzed. It is shown that a moderate ratio of edge transport barriers' (ETB) height to internal transport barriers' (ITBs) height is beneficial to ideal MHD stability. The normalized beta βN limit is about 2.20 (without wall) and 3.70 (with ideal wall). With the scaling law of energy confinement time, the required heating power for EAST AT scenario is calculated. The total heating power Pt increases as the toroidal magnetic field BT or the normalized beta βN is increased.

Tomographic reconstruction of tokamak plasma light emission using wavelet-vaguelette decomposition

NASA Astrophysics Data System (ADS)

Schneider, Kai; Nguyen van Yen, Romain; Fedorczak, Nicolas; Brochard, Frederic; Bonhomme, Gerard; Farge, Marie; Monier-Garbet, Pascale

2012-10-01

Images acquired by cameras installed in tokamaks are difficult to interpret because the three-dimensional structure of the plasma is flattened in a non-trivial way. Nevertheless, taking advantage of the slow variation of the fluctuations along magnetic field lines, the optical transformation may be approximated by a generalized Abel transform, for which we proposed in Nguyen van yen et al., Nucl. Fus., 52 (2012) 013005, an inversion technique based on the wavelet-vaguelette decomposition. After validation of the new method using an academic test case and numerical data obtained with the Tokam 2D code, we present an application to an experimental movie obtained in the tokamak Tore Supra. A comparison with a classical regularization technique for ill-posed inverse problems, the singular value decomposition, allows us to assess the efficiency. The superiority of the wavelet-vaguelette technique is reflected in preserving local features, such as blobs and fronts, in the denoised emissivity map.

Tomographic reconstruction of tokamak plasma light emission from single image using wavelet-vaguelette decomposition

NASA Astrophysics Data System (ADS)

Nguyen van yen, R.; Fedorczak, N.; Brochard, F.; Bonhomme, G.; Schneider, K.; Farge, M.; Monier-Garbet, P.

2012-01-01

Images acquired by cameras installed in tokamaks are difficult to interpret because the three-dimensional structure of the plasma is flattened in a non-trivial way. Nevertheless, taking advantage of the slow variation of the fluctuations along magnetic field lines, the optical transformation may be approximated by a generalized Abel transform, for which we propose an inversion technique based on the wavelet-vaguelette decomposition. After validation of the new method using an academic test case and numerical data obtained with the Tokam 2D code, we present an application to an experimental movie obtained in the tokamak Tore Supra. A comparison with a classical regularization technique for ill-posed inverse problems, the singular value decomposition, allows us to assess the efficiency. The superiority of the wavelet-vaguelette technique is reflected in preserving local features, such as blobs and fronts, in the denoised emissivity map.

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.

Currents in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Azari, A.; Eidietis, N. W.

2012-10-01

Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.