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U.S. Burning Plasma Organization e-News
September 15, 2009 (Issue 36)
CONTENTS
Director's Corner Reports
Feature Article
Upcoming Burning Plasma-related Events 2009 Events
2010 Events
2011 Events
Dear Burning Plasma Aficionados:
This newsletter provides a short update on U.S. Burning Plasma Organization activities. Comments on articles in the newsletter may be sent to the editor (R. Nazikian rnazikian at pppl.gov) or assistant editor (Rita Wilkinson ritaw at mail.utexas.edu).
Thank you for your interest in Burning Plasma research in the U.S.!
Director's Corner by J. W. Van Dam
New Topical Group Leaders
Each year, the leadership for five of the ten USBPO Topical Groups rotates. We’d like to express our sincere thanks to the following outgoing leaders, whose respective topical groups are indicated in parentheses: Dennis Whyte (Boundary), Cynthia Phillips (Plasma-Wave Interactions), Nermin Uckan (Fusion Engineering Science), Dave Humphreys (Operations and Control), and Raffi Nazikian (Energetic Particles). These five persons were the very first leaders for their topical groups, having assumed these responsibilities when the USBPO was established three years ago.
Now we welcome new leaders and deputy leaders for these five topical groups. For each of the groups, the USBPO Council has approved promoting the existing deputy leaders to become the topical group leaders. Also, we are pleased to introduce new deputy leaders.
| Topical Group | New Leader | New Deputy Leader |
| Boundary | Tom Rognlien | Tony Leonard |
| Plasma-Wave Interactions | Steve Wukitch | Gary Taylor |
| Fusion Engineering Science | Richard Nygren | Larry Baylor |
| Operations and Control | David Gates | Mike Walker |
| Energetic Particles | Don Spong | Eric Fredrickson |
We express appreciation to our outgoing leaders, and also to the new leaders and deputy leaders of these Topical Groups. A complete list of the Topical Group leadership is posted on the USBPO web site at http://burningplasma.org/groups.html.
ITER Plasma Control Workshop
The ITER Organization recently announced that it plans to hold an ITER Plasma Control Workshop at Cadarache during December 8-11, 2009. The workshop will serve to launch conceptual design activities for the plasma control system of the ITER facility. Dr. Joe Snipes (joseph.snipes at iter.org) will be responsible for the technical coordination of the workshop.
In the announcement for the workshop, the proposed plasma control system for ITER is described as needing to encompassing these functionalities: (1) ion and/or electron cyclotron wall conditioning and tritium removal, (2) plasma axisymmetric magnetic control, (3) power and particle flux control, (4) plasma kinetic control, (5) non-axisymmetric stability control, and (6) exception handling. This control system must be operational for ITER first plasma in 2018 and must have much of its foreseen functionality available by 2021.
A major emphasis of the workshop will be a consideration of how best to take advantage of existing plasma control expertise for the design and development of the ITER plasma control system. The seven Domestic Agencies have been asked to nominate experts in the design and use of control systems as participants at the workshop. For further information, please contact Tonia McPeters (865-574-5955, mcpeterstl at ornl.gov) at the US ITER Project Office.
Update about TBM Simulation Experiments on DIII-D
At the request of ITER, the DIII-D program plans to carry out experiments that simulate the impact of Test Blanket Modules on plasma performance in ITER. An insertable module has been designed that simulates the magnetic field perturbations produced by ferromagnetic materials used in the ITER Test Blanket Modules. Key physics issues to be addressed will include the impact of the perturbations on H-mode performance and transport, the H-mode pedestal, the L-H transition, plasma rotation, and fast-ion transport, as well as MHD stability and operational limits. Mike Schaffer (DIII-D) and Joe Snipes (ITER) will lead this task force, with Chuck Greenfield serving as deputy. An international team of 13 experts from the seven ITER Members, which includes three US scientists from PPPL and ORNL, will participate as part of this TBM task force.
The DIII-D TBM simulation module is now under construction, with installation planned for September and October 2009. Experiments with the module are planned for a two-week period during November 5-20, 2009, immediately following the APS-DPP Annual Meeting.
If you are interested in staying informed about the planning for the TBM task force experiments, please sign up on the mailing list which is available on the task force web page at:
http://fusion.gat.com/global/TBM2010.
ITER Project Baseline
The ITER Organization is currently busy completing the documentation that describes the details of the technical scope, schedule, and value estimate of the ITER project. This documentation consists of dozens of individual documents, amounting to hundreds of pages, which collectively are referred to as the Project Baseline. According to an article by Principal Deputy Director-General Norbert Holtkamp in the ITER Newsline, the Project Baseline
“…includes information and documentation on processes that were implemented as a response to the Briscoe Panel review such as new tools for cost assessment and a full risk analysis down to the last screw and connection. It describes the project all the way from the beginning of construction through commissioning, major ‘need dates’ for equipment delivery and required inter-project links, which describe the interconnection of deliveries between DAs or to the phased assembly on site. At the end, we know that all the activity timelines meet the 2018 project completion date and on to Deuterium-Tritium operation in 2026, having integrated the overall schedule with all the detailed schedules from the Domestic Agencies. This by itself is a tremendous achievement of the integrated Scheduling Group, which incorporates people from the project offices of the ITER Organization and the Domestic Agencies.”
The highest-level of these documents will be considered by the ITER Council for approval at its meeting in November. Prior to that, the Science and Technology Advisory Committee and the Management Advisory Committee of the ITER Council will review many of the documents that are part of the Project Baseline at their respective meetings, to be held in October.
Vacuum Vessel Review
At the March 18 meeting of the ITER Organization with the Domestic Agencies, several outstanding technical issues were identified with the design of the ITER vacuum vessel and the in-vessel components. The five primary issues concerned electromagnetic loads on blanket supports, nuclear heating of the toroidal field coils, field joint design, design of the in-vessel coils, and blanket manifold. Other technical issues concerned magnetic field ripple and ferromagnetic insert forces. In addition, manufacturability studies by industries in Korea and Europe, which are the two Members responsible for constructing the vacuum vessel sectors, indicated challenges to the cost and schedule.
The ITER Organization immediately launched a three-month study of an alternative design as a backup option and also of possible solutions for the baseline design of the vacuum vessel and blanket module. Guenter Janeschitz (ITER) reported about this effort at the STAC-6 meeting in May.
The results of the study were assessed during a comprehensive design review held in Cadarache during July 7-10. Rich Hawryluk (US) chaired the design review. Scientists and engineers from all ITER Members participated in the meeting. The review committee concluded that the baseline reference design is more mature at this time and, with some modifications, is closer to meeting the technical requirements of ITER. The committee, therefore, recommended proceeding with the modified reference design, with the understanding that the design and integration of the in-vessel coils can be satisfactorily resolved and an analysis of the nuclear heat load can be completed.
ITER Contributed Oral Session at APS-DPP Meeting
Please mark Thursday afternoon, November 5, on your calendars to attend the special session of contributed orals about “Research in Support of ITER.” The session chair will be all-around ITER expert Rich Hawryluk. The program for the APS-DPP Meeting
has recently been posted. You can find the list of talks in the ITER session here:
http://meetings.aps.org/Meeting/DPP09/SessionIndex2/?SessionEventID=110105
ReNeW Report Published
The final report of the Magnetic Fusion Energy Sciences Research Needs Workshop (ReNeW for short) has been released. It may be downloaded from the ReNeW web site: http://burningplasma.org/web/renew.html. Printed copies will be available in a few weeks. Although ReNeW was not a USBPO effort per se, the USBPO contributed mightily to this activity by hosting the ReNeW web site and providing lots of people-power. In fact, of the 22 members of the USBPO Research Committee, 16 were involved in ReNeW (two as theme leaders, three as panel leaders, and 11 as panel members). Moreover, of the 12 members of the USBPO Council, 10 were involved (four as theme leaders, three as panel leaders, and three as panel members). When you read the 425 pages of this report, I think you will agree that it is extremely well written, thanks to the assiduous efforts of many participants. The Office of Fusion Energy Sciences plans to use this document for charting a strategic course for the US fusion program during the next decade and beyond.
Reports
ITPA Topical Group on Diagnostics—Report on Activities in the period July 2008 – June 2009, Extract from the annual report
Written by R. L. Boivin (General Atomics, USA)
The coordinated activities of the Topical Group on Diagnostics were continued over the period of July 2008 to June 2009. There were two meetings of the ITPA Topical Group (TG) on Diagnostics during this period. Good progress has been made in the tasks designated as high priority: (i) development of methods of measuring the energy and density distribution of escaping alpha particles, (ii) assessment of the calibration strategy and calibration source strength needed for the neutron detector, (iii) determination of the lifetime of plasma-facing mirrors used in optical systems, (iv) development of requirements and assessment of techniques for measurement of hot dust, (v) assessment of the impact of in-vessel wall reflections on diagnostics, and (vi) assessment of the measurement requirements for plasma initiation and identification of potential gaps in planned measurement techniques.
Development of methods of measuring the energy and density distribution of escaping a-particles
Escaping alpha particles
One key task was to identify alpha particle orbits that could reach a possible detector outside the plasma. Orbits were calculated in two scenarios. Analysis showed that a smooth/flat outer wall would prevent direct detection of alpha particles. Options, including wall modification, are being evaluated.
It was agreed that a significant push on orbit calculations, along with detection efficiency analysis, is required in order to assess the possibility of direct alpha loss detection. An evaluation of activation techniques is continuing, but is likely to lack time resolution.
Confined alpha particles
The ITER fast ion collective Thomson scattering system is designed to measure the confined alpha particles. The work on confined a-particle measurements has been moved to intermediate priority, and the in-port components of this system are now included in the revised ITER diagnostic system.
Assessment of the calibration strategy and required calibration source strength
The Neutron Working Group has further developed the calibration strategy and begun to specify the required strength of the calibration source. Also, the need and location of a neutron test area have been further evaluated.
Calibrations with a neutron source mounted in the vacuum vessel will be required, and a determination of the optimum number of calibration points is being determined. This determination requires numerical simulations that are time consuming and subject to changes in the ITER overall configuration. Even with the best calibration strategy that can be devised and implemented, extensive modelling with neutronics codes will be needed to correct for the heavy support structure of the neutron generator and other components within the vacuum vessel. In addition, these simulations will need to account for any changes to the machine structure that occur after the calibration, e.g., additional port plugs.
Determination of the lifetime of plasma-facing mirrors used in optical systems
The report of the Specialist Working Group on First Mirrors gave an overview of all activities in the field of first mirrors. Further progress was reported in the field of deposition mitigation (e.g., by flowing gas in front of the mirror) and mirror cleaning, coated mirrors, mirror manufacturing, and irradiation testing of mirrors. With the progress accomplished so far, it was agreed that the development of mitigation methods for metallic (beryllium, tungsten) deposition is rapidly becoming urgent. Candidate mitigation methods against deposition need to be reviewed to identify the most promising ones for further development.
Promising results were presented on the active control of carbon deposition in diagnostic ducts and remote areas. For example, the complete suppression of carbon deposition was attained by use of deuterium gas fed in the duct interior. Encouraging results were achieved on the cleaning of mirrors. Softer carbon films that formed on the surfaces of mirrors exposed in the divertor of DIII-D were cleaned completely and the reflectivity was restored. Harder films originating from TEXTOR were mostly removed, leading to a significant increase in the mirror reflectivity.
Work on laser irradiation cleaning is starting at various Russian and EU institutions. Promising results from laser cleaning were reported from the HL-2A tokamak, where an Nd:YAG laser was used to remove carbon deposits. Applicability of these techniques for ITER conditions should be assessed.
A new task has been undertaken to assess the present risk associated with First Mirror failures (erosion/deposition) and their impacts of diagnostic performance. Preliminary findings were based on three main criteria, wavelength of interest, location, and solid angle sustained by the mirror. By assigning a risk level (high, medium, low) for each criterion, one could then identify the high-risk areas (systems) and direct resources to address the most urgent tasks.
Development of requirements and assessment of techniques for measurement of hot dust
Recent studies and discussions within the ITER Organization reached the conclusion that the inventories for dust and tritium are expected to reach their maximum limits on a timescale comparable to the target erosion lifetime. Based on this information, a control strategy for dust and tritium has been formulated. Dust will be removed during the scheduled divertor replacements (approximately every 4 years). Additionally, the dust will be monitored during and before shutdown. Local measurements will be benchmarked versus the tritium and dust recovered during the replacement of the divertor cassettes. The first benchmarking will be done in the hydrogen phase.
Over the last year, a few additional diagnostics were enabled in ITER for measuring dust and erosion. They are the divertor erosion monitor, removable samples (dust generation), micro-balance (dust), and laser-induced desorption (tritium).
An outstanding measurement issue is the detection of hot dust, for which a finalization of the requirements is still underway. Techniques to address this need have not been identified.
Assessment of the impact of in-vessel wall reflections on diagnostics
Many optical diagnostics will have to work against a background of stray reflected light coming from the plasma. Because the ITER plasma is much larger than existing tokamak plasmas, this problem will be more severe than that experienced thus far. The problem needs to be evaluated through a process of modeling and measurements on existing machines and through measurements of the reflectivity of relevant materials.
Recently, measurements and modelling of reflected radiation has been performed for infra-red measurements on Tore Supra. Without this analysis, false readings of temperatures can be made on internal surfaces. Further work is needed to standardize this approach for ITER and to develop a data bank of reflectivity coefficients for a variety of materials.
Assessment of the measurement requirements for plasma initiation and identification of potential gaps in planned measurement techniques
The early phase of plasma formation and control may require additional or special measurements that are different from those used during the flat-top phase. This task aims at assessing such needs and identifying any gaps in the associated proposed techniques.
Of particular interest is the need for measurements between discharges (e.g., wall conditioning, gas composition, erosion, etc.), at breakdown (e.g., null structure, impurity levels, etc.), in the ramp-up phase (density, current profile, etc.), and ultimately in the ramp-down phase (e.g., density, radiation levels, etc.).
Note of Appreciation
After many years of loyal and dedicated service in support of the ITPA and ITER expert groups, Dr. Alan Costley has stepped down as the co-chair of the ITPA Diagnostic Topical Group, in connection with his planned retirement from ITER. We are very much indebted to him and sincerely thank him for all his tireless efforts.
ITPA Pedestal Topical Group Response to ITER Urgent Tasks
Written by H. R. Wilson (University of York, UK), A. Loarte (ITER Organization), N. Oyama (JAEA, Japan), P. Lang (Nagoya University, Japan), M. Fenstermacher (LLNL, USA), R. Satori (JAEA, Japan), and P. B. Snyder (General Atomics, USA)
The ITER Organization has identified a number of urgent R&D issues regarding the pedestal. In the present report, we highlight those issues that the ITPA Pedestal Topical Group is well positioned to address, describe an updated work plan, and report progress over the last year.
The focus here is on urgent ITER issues. Less urgent, but nevertheless important, issues will be addressed in parallel with this program (aiming to avoid their becoming urgent). The interested reader should consult the summaries of the Pedestal Topical Group meetings for progress in these areas. There are five main areas (urgent issues) that the Topical Group will address:
1. Conditions for ELM suppression using resonant magnetic perturbations
2. Conditions for ELM pacing using pellets
3. Impact of the TF ripple on the pedestal characteristics
4. The impact of heating source on pedestal structure and ELM size
5. L-H transition physics
The ITPA Pedestal Topical Group has set up five working groups to address each of these areas.
1. ELM suppression with RMP coils
It is extremely likely that control of Type I ELMs will be necessary for ITER to meet its objectives fully. Coils to provide Resonant Magnetic Perturbations (RMPs) are presently the only tool available that is known to suppress ELMs in high-performance regimes. However, the technique has only been proven on DIII-D and the physics remains uncertain. The work plan presented here aims to improve our understanding and so reduce uncertainties in ELM control scenarios for ITER.
Objectives:
1. Reproduce ELM suppression with RMPs on at least one tokamak other than DIII-D.
2. Identify the criteria for ELM suppression from experimental data and theoretical models.
3. Quantify the impact of ELM suppression by RMPs on the pedestal pressure and core confinement, and develop and validate theoretical models.
4. Quantify the power loading on the walls and the divertor with RMP-suppressed ELMs; make recommendations on any requirement for rotating RMPs.
5. Explore the capability to suppress or mitigate ELMs during the current ramp phase (i.e., close to the L-H transition threshold and with q95 varying in time).
6. Demonstrate ELM control with ITER-like pellet fuelling.
7. Model the performance of the ITER ELM control coil set, and propose changes to the design as appropriate. This is likely to require further developments in modelling the plasma response, which is very challenging.Progress:
Progress is being made in understanding parameters that appear to influence ELM suppression. These include q95 resonance conditions, collisionality, density pump-out, beta, and island overlap width. DIII-D experience initially indicated that the island overlap condition is a good ordering parameter and that ELM suppression is correlated with achieving a minimum island overlap width. Recent results from MAST now demonstrate that, while this may be a necessary condition, it is not sufficient. A major task is to quantify the importance of key parameters and relate these to theoretical models.
2. ELM pace-making with pellets
If pellets can be used to trigger ELMs and increase their frequency by a factor of ten or more, and if the ELM size falls by a similar factor, then the ELMs on ITER will be tolerable. This programme of work will explore whether an increase in ELM frequency is possible through pellet pace-making, and then whether this results in a corresponding drop in ELM size. The compatibility with fueling will also be explored. The facilities that can contribute to this R&D in the next two years are AUG, DIII-D, and JET. JET provides a unique opportunity for pellet pace-making since the fueling from the pellets will not be the dominant effect (otherwise, the rise in density would itself modify the ELM frequency and mask the effect of the pellets). This package of work will also strive to develop a physics understanding of the mechanism by which pellets trigger ELMs, thus helping to minimize uncertainty in extrapolating the technique to ITER. A high-frequency pellet injector is probably not required for this physics study.
Objectives:
1. Achieve a frequency of pellet-triggered ELMs that is greater than ten times the natural ELM frequency with minimal density rise in an ITER-relevant scenario (e.g., shape, q, etc.).
2. Explore the dependence of the ELM size on frequency for pellet-triggered ELMs.
3. Quantify the minimum pellet size for triggering ELMs.
4. Develop a model for the ELM-triggering mechanism.
5. Optimise the injection angle.
6. Study the compatibility of ELM pacing by pellets with the required bulk plasma fueling (by pellets) in ITER reference scenarios.
7. Explore alternative options for pellet material.
8. Recommend pellet pace-making options for ITER.Progress:
Delays in the new pellet pace-making system on JET have meant that experiments have been confined to using the fueling pellets. Results confirm those found on AUG. On DIII-D, the pellet dropper has yet to successfully trigger ELMs. The high-frequency pellet injector is expected to be available on JET toward the end of the summer to enable a campaign directed towards the objectives of this topical group before JET shuts down for an extended period. There is information on the minimum pellet depth required to trigger an ELM: the pellet triggers the ELM before it is halfway into the pedestal region. On DIII-D, when RMPs are applied, the pellets must penetrate a little deeper into the pedestal to trigger the ELM.
3. Impact of toroidal field ripple
Toroidal field ripple can affect the confinement of energetic particles (alpha particles and beam ions) and result in local heat deposition on plasma-facing components. In addition, recent experimental results from JT-60U and JET show that the large amplitude of the toroidal field ripple can degrade the pedestal performance. However, the acceptable level of the toroidal field ripple required to achieve the ITER mission of Q=10 has not yet been confirmed. Since the toroidal field ripple increases the loss of fast and thermal ions, which can produce counter-current plasma rotation, the influence of the ripple on the pedestal through toroidal rotation modification should also be examined.
Objectives:
1. Survey the relationship between toroidal field ripple amplitude and pedestal performance in existing devices.
2. Survey the relationship between toroidal rotation and pedestal performance in existing devices.
3. Identify the experimental conditions for the degradation of pedestal and H-mode performance with the toroidal field ripple amplitude expected in ITER.
4. Develop and validate the model of ripple-induced losses of energetic and thermal particles, and ripple-induced toroidal rotation.
5. Assess the effects of toroidal field ripple on the lower hybrid power threshold.
6. Recommend the maximum acceptable ripple amplitude for ITER.Progress:
From the dedicated ripple experiments on JET, it has been found that the effect of ripple on H-mode properties (stored energy and density) varies depending on plasma background parameters. The effect of ripple was seen in plasmas with lower density (or collisionality?) at 1.7 MA and 2.5 MA, whereas, no significant difference was seen in plasmas with strong gas fueling. Therefore, there is no simple correlation between fast ion losses/torque/rotation and confinement that explains the JET results. The inter-machine experiment between JET and DIII-D demonstrated that it is possible to achieve the matched pedestal structure and the same H-mode quality (HH~1), although the ripple amplitude of these devices was different, 0.08% in JET and 0.35% in DIII-D. Therefore, it is suggested that the ripple amplitude of 0.35% might not affect the H-mode quality.
4. Pedestal structure
The overall performance of ITER will depend to a large extent on the pressure at the top of the pedestal. Two effects influence this: the pedestal width and the pressure gradient in the pedestal region. Ideal MHD is likely to set the maximum achievable gradient (though it is possible that two-fluid effects like diamagnetism may permit higher gradients in certain regimes). Recent experimental data suggests that the pedestal width scales as the square root of bp, and weakly with r*. Together, these results provide a prediction for the pedestal height on ITER. One issue where uncertainty remains is whether the pedestal height depends on the heating source. It is important to test this before a final decision on the mix of heating power for ITER is taken, so this is an urgent issue. A related issue is how ELM type and size depend on the heating power mix. Other, less urgent (but, nevertheless, important) issues to address include characterizing the transport processes in the pedestal and developing an understanding of the density and temperature pedestal heights.
Objectives:
1. Explore whether the pedestal pressure height and width depend on the heating source; quantify any differences, and interpret the results in terms of emerging models.
2. Explore whether the density pedestal properties depend on heating source—for example, through modified fueling sources (i.e., enhanced core fueling with NBI compared to that with ICRF).
3. Assess the impact of heating source on ELM size and explore prospects for interpretation in terms of peeling-ballooning theory.
4. Quantify the impact of torque on the pedestal structure and on ELMs.
5. Assess the potential viability of the Quiescent H-Mode (QH) as a high pedestal, ELM-free regime for ITER.
6. Develop theoretical models for the observed scaling of the pedestal width with plasma parameters.Progress:
Stability calculations have continued to provide a useful way of exploring pedestal structure and ELM characteristics in terms of the peeling-ballooning theory. Experimental observations continue to suggest weak or no dependence of the pedestal width on gyroradius. Plans have been discussed with leaders of a range of tokamaks to initiate experiments on exploring how (or whether) pedestal structure depends on heating source. Some data is available about the impact of torque on pedestal structure and ELMs, but more work is planned. One result from JT-60U is slightly lower pedestal height in counter injection, whereas DIII-D results indicate a weak, or no, effect at fixed beta. (JT-60U experiments were at fixed power, not fixed beta.). In light of recent progress in extending the parameter range of the Quiescent H-Mode (QH) to a broad range of input torque, rotation, and density values, we have established a new objective that is focused on determining the potential viability of QH mode as a high pedestal, ELM-free regime for ITER.
5. L-H Transition
The aim of the Pedestal Topical Group, in collaboration with the Transport and Confinement Topical Group, is to reduce the level of uncertainty in achieving and maintaining H-modes on ITER.
The mechanism(s) responsible for the L-to-H transition remains poorly understood. While it is thought that sheared flows play an important role, the mechanism for the spontaneous generation of flows remains unclear. There is, therefore, considerable uncertainty related to the trigger mechanism for the L-H transition and, as a consequence, the power threshold for ITER. This is important because the heating power available to ITER may be marginal for accessing the H-mode, according to some scaling laws. There are three key issues that this Topical Group will address: (1) Does ITER have sufficient power to access the H-mode, and how can this be optimized? (2) Can ITER stay in a high performance H-mode regime as the density and current are increased to achieve the fusion performance? (3) With the heating power available on ITER, is the quality of the H-mode sufficient to access Q=10 regimes?
Objectives:
1. Develop an understanding of the impact of radiated power on the L-H transition power threshold.
2. Identify any possible dependence of the L-H transition power threshold on the plasma heating mechanism and the effect of momentum injection.
3. Determine the characteristics of the H-mode when the power is marginally above threshold.
4. Characterise the conditions under which a high performance H-mode plasma makes a back transition to a regime of reduced performance (e.g., Type III or dithering H-mode, or an L-mode) for fixed global plasma parameters (power, fueling, etc).
5. Determine whether, and how, the L-H power threshold is modified by current ramps.
6 Determine the dependence of the L-H transition and pedestal characteristics on the plasma ion species.
7. Provide a first-principles model of the L-H transition.
Summary of the 2008-2009 Annual Report of the Transport and Confinement ITPA Task Group
S.M. Kaye (PPPL, USA)
Two topical group meetings have been held this past year. The first was held on Oct. 20-22, 2008, in Milan, Italy, following the IAEA Fusion Energy Conference. This meeting was held jointly with the Pedestal Topical Group, and it covered L-H threshold physics (the joint topic of interest), particle and impurity transport (most notably density peaking), rotation and momentum confinement, core transport, and modeling.
The second meeting was held jointly with the Integrated Operating Scenarios Topical Group in Naka, Japan, from March 31 to April 2, 2009. The joint topic of interest was transport modeling and physics model validation, but other topics addressed at the meeting were the effect of rotation on performance, momentum transport, and electron transport.
In addition to the joint experiments between facilities and the joint analysis activities, database work is still ongoing, although to a lesser extent than in previous years.
The activity of the Transport and Confinement Topical Group over the past year was broad, but it did address parts of all the high-priority issues identified in the ITER Urgent R&D Needs document. The transport-related high-priority issues include the following:
1. Transport and confinement in transient phases
2. Access to high-confinement regimes during steady-state and ramp-up/down H, D, and DT phases (including L-H threshold physics)
3. Characterization of proposed schemes for ELM control, and compatibility with scenario requirements (to be addressed by the Pedestal Topical Group)
4. Determination of ripple effects on ITER plasma performance and on fast particle confinement (although some of this was addressed by the Transport and Confinement Topical Group, mostly it will be addressed by the Energetic Particle Topical Group)
5. Particle transport and fueling in ITER reference scenarios.
Below are summaries of the work by the group in selected high-priority areas.
Particle and impurity transport: density peaking
The flattening of the density profile with increasing collisionality has been well documented and is a persistent feature among most conventional as well as low-aspect-ratio tokamaks. The empirical extrapolations to ITER collisionalities suggest that density peaking in that device would be ne(r=0.2)/<ne> ~ 1.5.
There has been progress in understanding the physics of the density peaking. Measurements from JT-60U indicate longer density profile scale lengths when the turbulence in the ITG-range exhibits smaller correlation lengths, supporting the conjecture of turbulence-driven pinches. LHD results indicate greater density pump-out with increased turbulence.
There has been also a great deal of theoretical progress towards understanding the source for density peaking, which could put predictions for ITER on a firmer physics basis. Gyrokinetic calculations that include both Trapped Electron Mode and Ion Temperature Gradient Modes have shown that the particle fluxes are a complex combination of inward and outward contributions at different wavenumbers and energies of trapped particles in phase space, and that a dependence on collisionality is exhibited. The slower trapped electrons cause most of the inward transport, while the faster ones give rise to the outward transport. Ion Temperature Gradient Mode simulations with the GS2 code were used to parameterize the normalized density gradient scale length as functions of collisionality, Te/Ti, and neutral beam particle flux. These calculations showed that the density peaking is primarily a function of collisionality. Experimental data was used to compute the expected R/Ln values, which were then compared to theoretical predictions. Good agreement between the two was found. Given this agreement, theory was used to predict the density peaking for ITER, and it was found to be ~1.5, which is consistent with empirical estimates.
L-H transition physics
Initial results from density ramp and power step-up and step-down experiments in JET indicated little difference in power threshold for L-H or H-L transitions, with the power threshold being ~1.2Pscaling, where Pscaling = 0.3neBR2.5. This result indicates no hysteresis. However, recent experiments show mixed results, calling into question the usefulness of ne as a fundamental parameter controlling the forward and back transitions.
Results from both ASDEX-U and JET indicate that a Type I ELMy regime is required for achieving H-factors of ~1 at powers just above the L-H threshold power. For ASDEX-U, Type I ELMs can occur when P~PLH; however, on JET, P~1.5PLH is required. At lower power, smaller, Type III ELMs are observed, and these degrade confinement by ~20%. This is true as well on ASDEX-U, where H<1 when Type III ELMs occur.
Experiments on ASDEX-U indicate that the species dependence of PLH is favorable for an ITER helium phase of operation, with power thresholds for helium the same as those for deuterium plasmas. Experiments on DIII-D indicated much higher thresholds for hydrogen than for deuterium plasmas.
NSTX has shown that there is a strong reduction in PLH with application of lithium wall coatings. A very important consideration has emerged with respect to applied external fields for ELM suppression. Preliminary results from JET, NSTX, and MAST were presented, and this topic is likely to be the subject of a new ITPA joint experiment in 2010.
Model validation during ramp-up/ramp-down phases
The objective of this work is to identify physics-based models that can be used for ITER scenario development and to understand the plasma evolution during the early and late discharge phases. This has particular application to determining whether the planned hardware provides sufficient flexibility for plasma control and achievement of performance objectives. The approach is to validate models at a high level (e.g., Te, li agreement) in ITER “DEMO” discharges on various devices such as ASDEX-U, C-Mod, DIII-D, and JET. This work is cross cutting with the Integrated Operating Scenarios Topical Group.
Many simulations have been performed. The obvious conclusion from the work so far is that the models and simulations do not provide a robust prediction for ITER. The Coppi-Tang-Redi Model has been used as a basis for modeling the ramp-up phase, and this model, as published, leads to overestimates of Te in the core and underestimates near the edge, leading to more rapid current penetration and higher li than desired. A reduction in the model diffusivity in the outer regions gives somewhat better agreement locally, but even worse agreement farther in. This work has been carried out in JET, DIII-D, and C-Mod plasmas. To date, no other physics-based models have been tested for the current ramp-up phase.
The Integrated Simulations and Modeling group of EFDA has focused on an empirical model and, more recently, on GLF23 simulations. It was shown from work on JET, however, that the results of the simulations are extremely sensitive to assumptions made about unmeasured quantities such as Zeff profiles and specific impurity content.
The results of these efforts so far do not give a robust and confident picture for ITER predictions, and the two topical groups held a discussion about whether too much is being expected from the modeling at this stage. In particular, an actual prediction for the L-H transition is not imminent, and there is a need for a physics-based prediction of pedestal temperature. A prediction of the edge Te based on peeling-ballooning mode theory is presently being developed. However, it was felt that, although we should not stop these benchmarking efforts, we do need to redefine and refocus them, perhaps taking an alternative approach. This approach would be to adopt a set of Te profiles from the ramp-up phases of existing experiments and adjust the magnitude of the profiles in response to changes in heating power. Then, without needing to predict the Te, one could still assess whether an acceptable li can be obtained with the available heating power.
Feature Article
Shattered Pellet Disruption Mitigation Tool Now Under Test for ITER
Written by L. R. Baylor (ORNL, USA)
A disruption is a sudden unexpected termination of a plasma discharge, which can potentially damage the plasma chamber from sudden thermal loads, high magnetic forces, and runaway electrons formed during the current quench. A new technique has been developed to mitigate these effects. It involves injecting very large quantities of solid deuterium (with each pellet nearly the size of a wine cork) into the disrupting plasma. The solid pellet is formed in situ in a liquid helium-cooled barrel. It is fired by a pneumatic gun and shattered against a V-shaped plate just before entering the plasma, to produce a spray of material that will quickly dissipate the plasma energy in a few milliseconds before any significant plasma chamber effects can take place. A new system using this technique has been installed on the DIII-D tokamak and was first tested this summer. This technique can potentially be employed on ITER to prevent any unexpected disruptions from reducing machine availability for physics experiments.
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| Fig. 1 Installation of the shattered pellet disruption mitigation system on DIII-D. |
The “shotgun” style pellet injector was developed for disruption mitigation with support and hardware from the Virtual Laboratory for Technology fueling program. The wine cork size D2 pellets (15.3 mm diameter × 22 mm long ~ 4000 torr-L) are formed in about 10 minutes and shot into the V-shaped shatter plate installed inside DIII-D, as shown in Fig. 1. When the pellet hits the plate at a typical speed of ~ 500 m/s, it shatters into a spray of solid particles, as well as liquid and gas, which are aimed toward the plasma magnetic axis. Measurements from a foil witness plate indicate that hundreds of particles less than 0.5 mm in size are produced and more than 50 particles are produced that are greater than 1 mm in size. Further details on the design and testing of the hardware are available from Steve Combs ( CombsSK at ornl.gov ).
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Fig. 2 Pipe gun originally developed |
This shotgun pellet injector system (see Fig. 2) has now been used in an initial plasma disruption mitigation experiment. Preliminary results indicate very rapid shutdowns with an apparently strong central deposition of the particles. The disruptions with the shattered D2 pellets were observed to have minimal divertor heat loads and halo currents. The deposition of injected material occurred throughout the core plasma, unlike massive gas injection (MGI), which deposits at the edge and is then mixed in during the thermal quench. At the APS-DPP meeting in November, Nicolas Commaux and Tom Jernigan will present full results of the experiments and implications for use on ITER.
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| Fig. 3 Possible ITER shattered pellet disruption mitigation configuration. Injector from DIII-D is shown (to scale) below the upper port plug. |
The application of this technique for disruption mitigation in ITER is promising. The amount of material needed for ITER collisional runaway electron suppression (~ 2 x 1025 atoms of D2) is significantly more than that used in the DIII-D pellets. The equivalent required single-pellet size would be nearly the size of a softball. A more practical design would employ many tens of pellets of a similar size to the DIII-D wine cork pellet, all fired in parallel from different ports, as shown schematically in Fig. 3. This would spread out the injected mass and provide significant redundancy in case of individual injector failure. While initial results look promising, significantly more research is needed to understand how best to mitigate the runaway electron formation in ITER, which is the most challenging disruption threat. In addition to an effective mitigation system, it is clear that early and accurate detection of precursors to disruptions and fast actuation of the mitigation system will be critical to provide safe and reliable disruption mitigation in ITER.
Announcments
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Upcoming Burning Plasma Events
2009 Events
| Sept 13-17 9th European Conference on Applied Superconductivity (EUCAS) Dresden, Germany |
| Sept 21-23 11th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems Kiev, Ukraine |
| Sept 21-24 14th International Symposium on Laser-Aided Plasma Diagnostics (LAPD-14) Castelbrando, Treviso, Italy |
| Sept 24-25 ITPA Energetic Particle Topical Group Meeting Kiev, Ukraine |
Sept 28-Oct 2 6th International Conference on Plasma Physics & Plasma Technology Minsk, Belarus |
| Sept 30-Oct 2 12th International Workshop on "H-mode Physics and Transport Barriers" Princeton, New Jersey USA |
| Oct 5-7 ITPA Transport & Confinement Topical Group Meeting Princeton, New Jersey USA |
| Oct 5-7 ITPA Pedestal & Edge Physics Topical Group Meeting Princeton, New Jersey USA |
| Oct 6-8 Meeting of the ITPA Topical Group on MHD Abingdon, UK |
| Oct 6-9 21st International Conference on Numerical Simulation of Plasmas Lisboa, Portugal |
| Oct 11-16 9th International Symposium on Fusion Nuclear Technology (ISFNT-9) Dalian, China |
| Oct 12-16 ITPA Diagnostics Topical Group Meeting Pohang, Korea |
| Oct 18-23 21st International Conference on Magnet Technology (MT-21) Hefei, China |
| Nov 2-6 51st APS-DPP Meeting Atlanta, Georgia USA |
| Nov 9-11 14th Workshop on MHD Stability and Control Princeton, New Jersey USA |
| Nov 15-19 ANS Winter Meeting Washington, DC USA |
| Dec 2-3 Fusion Power Associates Annual Meeting Washington, DC USA |
| Dec 8-11 The 19th International Toki Conference (ITC19) Ceratopia Toki-City, Gifu, Japan |
| Dec 14-17 ITPA SOL & Divertor Topical Group Meeting San Diego, California USA |
| 2010 Events |
| Week of 22 March ITPA Transport & Confinement Topical Group Meeting Oxfordshire, UK |
| Spring ITPA IOC Topical Group Meeting Princeton, New Jersey USA |
| April 12-15 16th EC Meeting China |
| April 13-16 Transport Task Force Annapolis, Maryland USA |
| April 13-16 International Conference on Plasma Diagnostics Pont-à-Mousson, France |
| April 19-21 Sherwood Fusion Theory Conference Seattle, Washington USA |
| May 16-20 18th ITPA Diagnostics & HTPD Topical Group Meetings Wildwood, New Jersey USA |
| May 19-21 STAC-8 Cadarache, France |
| May 24-28 Plasma Surface Interactions San Diego, California, USA |
| Fall ITPA Transport and Confinement Topical Group Meeting (following IAEA) South Korea |
| Fall ITPA IOC Topical Group Meeting (following IAEA) South Korea |
| Fall ITPA Diagnostics Topical Group Meeting (following IAEA) Japan |
| Sept 27-Oct 1 26th Symposium on Fusion Technology (SOFT2010) Porto, Portugal |
| Oct 24-29 9th International Conference on Tritium Science and Technology Nara, Japan |
| 2011 Events |
| Spring ITPA Transport & Confinement Topical Group Meeting (following US/EU TIF) San Diego, California USA |
Fusion Research-related events can also be seen on the USBPO website at http://burningplasma.org/events.html