Princeton Plasma Physics Laboratory And Alcator C-Mod ... · Princeton Plasma Physics Laboratory...

Princeton Plasma Physics LaboratoryAnd Alcator C-Mod Collaboration

Five Year Plan 2003-2008

PPPL StaffMay, 2003

Plasma Physics LaboratoryPrinceton UniversityPrinceton, NJ, USA

- 2 -

Contents

1. Introduction .......................................................................................................................................42. Advanced Tokamak...........................................................................................................................5

2.1. Highlights of Recent Research ..................................................................................................52.1.1. Lower Hybrid Launcher.....................................................................................................52.1.2. MSE Diagnostic Startup and Improvements .....................................................................5

2.2. Proposed Research.....................................................................................................................62.2.1 Lower Hybrid Current Drive ..............................................................................................62.2.2. MSE Current Distribution Measurements .........................................................................8

3. Burning Plasma Experiments.........................................................................................................103.1. Highlights of Recent Research ................................................................................................10

3.1.1. Initial Single-null/Double-null Studies............................................................................103.2. Proposed Research...................................................................................................................11

3.2.1. Compare Single-null/Double-null Diverted Discharges.................................................113.2.2. ICRF Heating at High Power Levels...............................................................................113.2.3. D(3He) Minority Heating .................................................................................................113.2.4. Transport...........................................................................................................................113.2.5. Divertor and Plasma Boundary........................................................................................123.2.6. RF (ICRF and LHCD)......................................................................................................133.2.7. Global Stability ................................................................................................................133.2.8. Advanced Tokamak .........................................................................................................143.2.9. Integrated Scenarios .........................................................................................................14

4. Transport.........................................................................................................................................154.1. Highlights of Recent Research ................................................................................................15

4.1.1. Marginal Stability and Turbulence ..................................................................................154.1.2. ITB Modeling ...................................................................................................................16

4.2. Proposed Research...................................................................................................................164.2.1. Marginal Stability and Turbulence ..................................................................................164.2.2. Electron Transport............................................................................................................174.2.3. ITB Modeling ...................................................................................................................17

5. Divertor and Plasma Boundary ......................................................................................................175.1. Highlights of Recent Research ................................................................................................17

5.1.1. GPI Edge Turbulence Visualization ................................................................................175.1.2. Edge Neutrals Modeling ..................................................................................................215.1.3. Reflectometer Upgrade ....................................................................................................22

5.2. Proposed Research...................................................................................................................225.2.1. Edge Turbulence Visualization Extended .......................................................................225.2.2. Comparison with Theoretical Simulations ......................................................................245.2.3. Edge Turbulence Control Experiments ...........................................................................265.2.4. Edge Modeling Extended.................................................................................................28

6. RF Physics and Wave-particle Interactions...................................................................................296.1. Highlights of Recent Research ................................................................................................29

6.1.1. D(H) Minority Heating Experiments...............................................................................296.1.2. Mode Conversion Experiments .......................................................................................296.1.3. ICRF Modeling ................................................................................................................296.1.4. LH Modeling ....................................................................................................................30

6.2. Proposed Research...................................................................................................................306.2.1. D(3He) Minority Heating Experiments ...........................................................................30

- 3 -

6.2.2. FWCD experiments..........................................................................................................316.2.3. MCCD experiments .........................................................................................................316.2.4. Flow drive studies ............................................................................................................316.2.5. LH wave physics ..............................................................................................................316.2.6. LHCD physics ..................................................................................................................316.2.7. RF Modeling.....................................................................................................................31

7. Theory and Computation of Macroscopic Stability .......................................................................337.1. Overview ..................................................................................................................................337.2. Model Development ................................................................................................................33

7.2.1 Ideal MHD.........................................................................................................................337.2.2 Two-fluid Extended-MHD................................................................................................347.2.3 Kinetic Extended-MHD ....................................................................................................34

7.3 Beta limiting MHD modes .......................................................................................................357.3.1 Sawtooth Phenomena ........................................................................................................357.3.2 Neoclassical Tearing Modes .............................................................................................357.3.3 Energetic Particle Modes ..................................................................................................36

7.4 Edge MHD Stability and the behavior of ELMs .....................................................................367.4.1 Linear Analysis .................................................................................................................377.4.2 Nonlinear Physics of ELMs & Evolution of Free Boundary Modes...............................37

7.5 Prediction of the Cause and Effect of Disruptions ..................................................................377.5.1 Physics of the Disruption ..................................................................................................377.5.2 Disruption Forces ..............................................................................................................38

7.6 Control Issues............................................................................................................................387.6.1 Profile and Shape Control .................................................................................................387.6.2 The Physics of Pellet Fueling ...........................................................................................387.6.3 Internal Mode Control.......................................................................................................39

7.7 Appendix: Macrostability Codes Supported by PPPL Theory Department .........................408. Facility .............................................................................................................................................42

8.1. Highlights of Recent Research ................................................................................................438.1.1. Rework of 4 ICRF transmitters........................................................................................438.1.2. Fabrication, installation, upgrading of 4-strap ICRF antenna ........................................438.1.3. Ongoing ICRF engineering support ................................................................................448.1.4. Completion of LHCD launcher .......................................................................................44

8.2. Proposed Research...................................................................................................................458.2.1. LHCD launcher #1 installation and commissioning .......................................................458.2.2 Fabrication of LHCD launcher #2 ....................................................................................468.2.3 Participate in 4-strap ICRF antenna #2 ............................................................................468.2.4 Participate in tunable cavities for ICRF transmitters 1 and 2 ..........................................46

9. Budget ..............................................................................................................................................4710. Manpower ......................................................................................................................................4811. Contributors ...................................................................................................................................49

- 4 -

1. Introduction

The purpose of the PPPL C-Mod collaboration is to conduct and enable forefront scientificresearch on the Alcator C-Mod tokamak and to perform engineering/technical support for thejoint MIT/PPPL team.

Research aims include:

• Research on the effectiveness of off-axis current drive via Lower Hybrid current drive andits effect on plasma performance. This program includes design and fabrication of thelauncher and participation in the associated research;

• Experimental study of basic ICRF plasma-wave interaction processes and their comparisonwith theory in order to gain predictive capability for heating and current drive in reactor-grade experiments;

• Creation and understanding of internal transport barriers, off-axis current drive forPEP/ITB mode studies, and low frequency (ω<Ωci) current drive for reactor application viaICRF heating and mode conversion current drive; and

• Core confinement, and H-mode behavior including pedestal characteristics andfluctuations.

Recent and proposed hardware upgrades include both plasma control and diagnostic components:

• ICRF antennas for plasma heating and current drive;

• LHCD launchers and coupling hardware for control of the plasma current profile;

• Current profile diagnostics to increase understanding of current drive and plasma behavior(in conjunction with a C-Mod-provided diagnostic neutral beam);

• Edge diagnostics (edge fluctuation measurements at the plasma periphery withreflectometry and 2-D imaging of edge turbulence) to increase understanding of turbulenceand transport.

Engineering and technical support for RF power systems include:

• Engineering assistance in tuning and maintaining the ICRF transmitters;

• Technical assistance in modifying the new 4-strap ICRF antenna following its initialoperation in FY2000 through FY2002; and

• Engineering participation in the design, fabrication, and installation of the Lower Hybridcurrent drive system as part of the Lower Hybrid project.

In all these areas PPPL provides assistance in areas where PPPL has competence and capabilitiesneeded by the C-Mod program while enhancing research opportunities for PPPL scientists.

- 5 -

2. Advanced Tokamak

2.1. Highlights of Recent Research

2.1.1. Lower Hybrid Launcher

Based on our experience with Lower Hybrid current drive on both the PLT and PBX tokamaks,PPPL proposed the design and fabrication of a Lower Hybrid launcher for C-Mod. This will beused to launch directed waves and drive an off-axis current, which together with ICRF heatingwill result in current and pressure distribution profile modifications producing the AdvanceTokamak discharge configuration.

2.1.2. MSE Diagnostic Startup and Improvements

A key objective of the Alcator C-Mod experimental program for next five years is improvedunderstanding of Advanced Tokamak (AT) physics in plasma regimes relevant to long-pulseburning plasma experiments and fusion reactors. Broadly speaking, AT plasma regimes arecharacterized by a degree of control of temperature, density, transport, current and flow profiles

- 6 -

that yield significant improvements in energy confinement and beta limits. Historically, ATregimes were developed on a number of tokamaks using modified plasma growth and earlyplasma heating to create hollow q(r) profiles, coupled with beam-driven current and plasmaflows. Concurrently, development of the Motional Stark Effect (MSE) diagnostic throughout the1990's provided local measurements of q(r) that were crucial to understanding the improvedconfinement and beta and to empirical optimization of AT performance.

The inductive plasma startup and beam-driven flows give the experimenter considerableflexibility to alter the magnetic shear and flow-shear profiles which has proven very useful inunraveling their role in controlling confinement and stability. Unfortunately, these techniqueshave limited applicability to high density, long-pulse tokamaks including fusion reactors. Tocomplement the ongoing AT research at NSTX, DIII-D and elsewhere, the Alcator program willexploit its unique capabilities -- especially LHCD and high density -- to study AT physics issuescrucial to establishing the feasibility of AT plasma scenarios in reactors and long-pulse burningplasma experiments.

As elsewhere, measurements of the q-profile on C-Mod will be essential to study the physics ofnoninductive current drive (LHCD, MCCD) and the influence of magnetic shear on transport.Working closely with C-Mod staff, PPPL has developed and installed a 10-channel MSEdiagnostic which will be the cornerstone diagnostic for measuring the q-profile in AT physicsstudies.

The C-Mod environment is hostile to MSE diagnostics: limited access mandates the use of in-vessel optical components which are subject to large accelerations (hundreds of g's) duringdisruptions. Prior to the 2003 run period, the mechanical support structure for two large in-vesselglass mirrors was redesigned to reduce the disruptive forces and to provide resilience againstaccelerations up to 500g. In parallel, PPPL has designed silver-coated replacement Inconelmirrors which will be available for installation, if needed, in FY2003.

2.2. Proposed Research

2.2.1 Lower Hybrid Current Drive

The Advanced Tokamak scenario requires all the plasma current to be driven by non inductivemethods. Since it is not practical to drive the all current with exterior methods such as LHCD andNBI the bootstrap current has to be optimized. This is obtained by actively controlling transportbarriers and current profiles.

The LHCD system is equipped with the maximum flexibility possible for power deposition andincludes the possibility of changing the spectrum (therefore the driven current) in real time.

Figure 2.1 shows the extent of the power spectrum vs. the refractive index in the toroidaldirection.

- 7 -

-10 -5 0 5 100

2

4

6

1 2 3 4 5n||

0

2

4

6

POWER SPECTRUM

24 waveguides

0.55 cm opening 0.15 cm septum

90º120º

150º

180º

60º

P (a.u.)

P (a.u.)

Fig 2.1. Power spectrum of the PPPL/C-Mod Lower-Hybrid launcher.

As seen in the figure, the maximum directivity is obtained for very fast waves which damp on fastelectrons: the addition of the second coupler will allow us to use one coupler (one spectrum) togenerate a fast electron tail in the desired radial position. This might require high n||, therefore theefficiency might be low. The other coupler will then launch waves with faster phase velocities,thereby increasing the efficiency.

0

1

2

3

4

5

6

1 2 3 4 5n||

P (a.u.)

Figure 2 n|| spectrum for various progressive phase differences(line), compared to the spectra obtained by keeping the the high power phase shifter at ∆Φ=90° while changing the phase in the low power phase shifter (gray).

∆Φ=60° 90°

120°

150°

180°

- 8 -

The phase relation between the two adjacent waveguide columns fed by the same Klystron isset by a mechanical phase shifter that can be changed only between pulses. It is possible toobtain a slightly less directional power spectrum – but in real time - using the low power phaseshifter before the Klystron. An example of this spectrum is shown as the shaded areas in Figure2.2.

2.2.2. MSE Current Distribution Measurements

Improved Spatial Resolution - Additional Spatial Channels

The number of MSE radial measurement locations will be doubled by installing 10additional detector systems (PMT's, filters, electronics) and rearranging the present fiberbundles (FY04).

Improved Spatial Resolution - Beam Collimation

The present short-pulse Diagnostic Neutral Beam (DNB) which is on loan to C-Mod untilmid-FY04 has a circular footprint with a 1/e diameter of approximately 8 cm that limitsthe radial resolution to about 4 cm (∆ r/a = 0.18) at r = a/2. The permanent, long-pulsereplacement DNB to be installed in FY05 has improved divergence corresponding to a 6-cm diameter. Further improvements in the radial resolution will require beam collimation;initial design calculations indicate that a beam as small as 2-3 cm could be realized at thecost of a factor 2-3 reduction in signal strength. PPPL will design a variable aperturesystem capable of varying the beam diameter from the 2-3 cm range desired for someMSE studies up to the present beam size to accommodate the needs of other beam-baseddiagnostics (BES, CXRS) that prefer maximum signal strength (FY05).

Edge Electric Field Measurement

A second optical system will be added to view the plasma from a different angle todiscriminate between the usual MSE electric field (v × B) and radial electric fields at theplasma edge (FY05).

Measurement and Analysis Improvements

Hardware upgrades in combination with ongoing maturation of the MSE analysis and itsintegration into equilibrium codes (EFIT) will provide increasingly detailed and accurateq-profile and shear-profile measurements over the 2003-8 campaign. While various stagesof the MSE development will certainly overlap, a representative progression of steps ispresented below. The steps are characterized by an ever-increasing requirement foraccuracy of the measured q-profile, e.g. merely demonstrating that LH drives currentsrequires only modest accuracy, while detailed comparisons of transport barrier behaviorwith microturbulence codes involves the radial derivative of the q-profile and hence needshighly accurate MSE measurements.

These improvements will allow us to:

• Confirm existence of LH-driven currents (FY04).

- 9 -

• Parameterize the LH current drive efficiency and radial localization of LH-drivencurrents. Compare with ACCOME code calculations and TRANSP using the LSCcode as the LH package (FY04-05).

• Characterize the effect of modified magnetic shear by LHCD on transport barriers,e.g. correlate radial position of transport barriers with ρmin, qmin (FY05-06).

• Extend magnetic shear studies to plasmas having high bootstrap current fraction inaddition to LHCD (FY06-07).

• Compare transport barrier behavior (location, magnitude, parametric scaling) andmicroturbulence model calculations that include the measured magnetic-shear andflow-shear profiles (FY05-7).

Extensive code calculations by the C-Mod staff have identified plasma regimes in whichLHCD can drive sufficient current to reverse the q-profile and thereby access improvedconfinement. LHCD is most effective at high electron temperature and with densityrelatively low for C-Mod, though still high compared to most other LH experiments.Central temperatures of over 5 keV have been demonstrated in experiments with ne(0) =1.5 x 1020 m-3 with ICRH applied during current ramp-up. Modeling based on such a targetscenario with increased ICRH predicts that LH will drive a current of 390 kA and generatereversed shear over a broad region, with (r/a)qmin ~ 0.50. We expect adequate signal-to-noise ratios for the MSE measurement over this density range with the present DNB andits proposed long-pulse replacement.

Truly reactor-relevant AT regimes require a high bootstrap fraction in addition to non-inductive magnetic-shear profile control. To achieve a high bootstrap fraction, it isnecessary to operate at higher plasma density and confinement. Modeling based on targetplasmas with edge and/or core transport barriers have identified regimes with bootstrapfractions of 65-75%, reduced LH driven current (270 kA), but maintaining a stronglyreversed current profile. These plasmas have central densities in the range 3 x 1020 m-3

which, based on prior experience, is expected to produce inadequate signal-to-noise forcore q-profile measurements. In these high-density regimes MSE may be supplementedby a proposed 20-channel polarimeter that views the across the plasma from a horizontalport.

The MSE diagnostic will contribute important q-profile measurements to a number of otherresearch topics in the Alcator C-mod program, including:

Stabilizing NTM

Both LHCD and Mode Conversion Current Drive have been proposed as a tool forstabilizing Neoclassical Tearing Modes in burning plasma experiments. Under the RFphysics program, MSE will measure the current drive efficiency and radial localization ofboth techniques for comparison with theory and code predictions. Then under theauspices of the C-Mod MHD Stability program, MSE will provide measurements todocument the magnitude and radial localization of driven currents that are needed tostabilize NTM's in open-loop. The understanding gained from these programs will formthe basis for development of a feedback algorithm and control scheme to demonstratefeedback stabilization in a high-performance H-mode plasma.

- 10 -

Sawtooth Stabilization

C-Mod's Burning Plasma program proposes to include a study of the feasibility of usingfast wave current drive to control the amplitude, extent and frequency of sawtooth activity.A purely empirical evaluation of FWCD on sawtooth activity can be carried out without adetailed knowledge of the q-profile, but MSE will contribute information pertaining to theefficiency and radial localization of FWCD which may prove useful in guiding theexperiments.

Low Frequency FWCD

Fast wave current drive with ω < ωci for all species is potentially attractive for burningplasmas because it avoids the possibility of parasitic absorption on the α-particles. MSEwill document the current drive efficiency and radial localization as part of the BurningPlasma program.

3. Burning Plasma Experiments

Alcator C-Mod is well suited to addressing a broad range of physics R&D issues of interest tonext step burning plasma experiments such as ITER or FIRE. Alcator C-Mod is roughly a 1/3scale model of FIRE with all metal plasma facing components and the flexibility to operate withdouble null poloidal divertors. Plasma control capabilities for plasma heating (ICRF), currentdrive (LHCD) and pellet fueling are very similar to the capabilities envisioned for FIRE. AlcatorC-Mod has the capability of operating in the H-Mode; in the Enhanced D-Alpha mode; and in theRF controlled Internal Transport Barrier modes and has just completed the LHCD Project whichwill allow the exploration of reversed shear advanced tokamak regimes. Therefore, C-Mod is avaluable facility for addressing specific physics issues as well as integrated tests of severalpotential operating scenarios for burning plasma experiments. If fully implemented, the burningplasma support program and capabilities proposed (C-Mod 5 Year Proposal Section 4) by theAlcator C-Mod group would address many critical issues for ITER and FIRE. This sectionhighlights the areas of particular interest to ITER and FIRE within PPPL.

The PPPL burning plasma study group has begun active participation in the C-Modcollaboration only recently, and so our specific proposed collaborative efforts in this area are lesswell defined than for the Advanced Tokamak, RF, and Transport research thrusts. Broadlyspeaking, PPPL proposes to develop and carry out Miniproposals that will exploit C-Mod’sunique parametric and diagnostic capabilities to address critical burning-plasmas issues.


3.1.1. Initial Single-null/Double-null Studies

A number of C-Mod research results described in Section 4 of the C-Mod 5 Year proposal areimportant for ITER and FIRE experiments. Of particular interest are the studies just beginning oncomparison of single-null diverted discharges with double-null diverted discharges. The type ofdivertor configuration – single null or double null and pumping capability – single pump or

- 11 -

double pump – has a significant effect on plasma performance and engineering for burningplasma experiments. These studies involving H-modes will seek to quantify the effect of SN/DNand plasma cross-section triangularity on confinement (pedestal), ELMs and divertor detachment.Also of interest is the determination of disruption characteristics for DN plasmas with a neutralstability region.


3.2.1. Compare Single-null/Double-null Diverted Discharges

The studies described in section 3.1.1 will not be completed in the present 5-year plan, and willbe carried over into the proposed 5-year program. The scope of the experiments will be extendedto include the effect of divertor pumping as that capability is added to C-Mod.

3.2.2. ICRF Heating at High Power Levels

C- Mod has the capability for high power density ICRF and this is of interest to both ITER andFIRE. The practical experience gained on C-Mod will be invaluable in finalizing launcher designfor ITER and for optimizing the choice of launcher and coupling physics for FIRE.

3.2.3. D(3He) Minority Heating

Minority heating using 3He will be used as a primary heating method on ITER and FIRE forheating in H, He, D and T plasmas. Extending these studies to power levels and parametersapproaching those in ITER and FIRE will be an important contribution to the burning plasmaprogram. See section 6.2 for specific RF physics issues.

3.2.4. Transport

Confinement Scaling of H-Mode with SN/DN and High Triangularity

Results from ASDEX Upgrade, JET and JT-60U all show an increase in the ITER H-modescaling multiplier H98(y,2) as the triangularity is increased for densities in the range 0.7 <n/nG < 1. Data from C-Mod in this area would help fill out the ITPA Confinement andModeling Data Base and the ITPA Pedestal Data Base and would strengthen the physicsbasis for ITER and FIRE.

Similarity Discharge Studies

Similarity H-Mode discharges with the same ν* and β as ITER and FIRE will beinvestigated as another technique for understanding and projecting ITER and FIRE H-mode performance. Of particular interest is to also extend the range of similar parametersto the pedestal region, and to the SOL/divertor region.

Transport Scaling in ITBs

- 12 -

The internal transport barriers (ITB) that are induced by off-axis ICRF and then controlledby on-axis ICRF heating are of particular importance to an ICRF heated burning plasma.An important question is whether strong central alpha heating will deplete the transportbarrier in the same manner as on-axis ICRF heating appears to do on Alcator C-Mod.

Rotation with RF Only and Impact on Confinement

Flow shear associated with beam-induced rotation has been exploited in several tokamaksto study the physics of transport barriers and to substantially improve both local transportand global performance. It may become progressively more difficult to use beams toprovide such control as plasmas become larger and denser in burning plasma experiments.Therefore, understanding the mechanism for plasma rotation without externally injectedmomentum that has been observed on C-Mod is very important for future reactor plasmas.Continued experiments and modeling in this area will be carried out in close cooperationwith the C-Mod scientific staff.

Particle Transport (Peaked Density Profiles-Fuel Mix Optimization)

Increased attention is needed on particle transport for burning plasmas since controllingthe fueling mixture and profile has very high leverage for a burning plasma experimentlike ITER or FIRE. Pellet injection from the high field side is a potentially powerfultechnique for fueling the core of a burning plasma. A key question is whetherconfinement modes with improved energy confinement will also have improved particleconfinement that leads to unacceptable impurity accumulation in long pulse discharges.Studies of particle (fuel and impurities) transport in conjunction with H-Mode, EDA, ITBand AT plasma regimes will be carried out to address this issue.

3.2.5. Divertor and Plasma Boundary

The divertor and plasma boundary issues are very important for a burning plasma experimentand the follow-on reactor plasma. In the ARIES advanced tokamak power plant plasmas, aradiative divertor with metal plasma facing components was proposed to handle the large plasmapower exhausts of Pheat/R ~ 80 MW/m while maintaining a low tritium inventory. This willrequire significant radiation in the SOL and in the divertor to achieve reasonable powerdeposition densities on the first wall and divertor. In addition, the capability to exhaust thehelium ash must be maintained. C-Mod has all metal PFCs and has shown that the tritiumretention is in the acceptable range of ~0.2%. THE C-Mod exhaust power densities are Pheat/R ~10 MW/m are significant and will provide data relevant to both ITER (Pheat/R ~ 20 MW/m) andmore compact burning experiments such as FIRE (also Pheat/R ~ 20 MW/m).

Tests of a Tungsten Bush Module

The studies of power deposition, tritium retention and effects of ELMs will be extended totests of a FIRE-like tungsten brush module as part of the proposed C-Mod program. Thiswill be of interest to ITER as well.

Impact of SN/DN Configuration on ELMs

- 13 -

Type I ELMs are projected to severely damage the divertor target plates in burning plasmaexperiments and even more so in an ARIES scale reactor. ASDEX-U, JT-60U and DIII-Dhave observed a transition to smaller Type II Elms which would be acceptable in FIREand possibly in ARIES-RS. A set of experiments is proposed to investigate ELMbehavior as the plasma configuration is changed from SN to DN and as the triangularity isincreased. This will also be done for radiative divertor conditions. This is of great interestto both ITER and FIRE.

Optimum Pumping Divertor Configuration

A major issue for burning plasma and reactor design is to determine the optimum particlepumping configuration. In the present C-Mod configuration that divertor pumping is notsufficient to prevent significant particle recycling from the first wall. The baseline C-Modplan is to investigate a configuration where the lower divertor takes the plasma exhaustheat load and the upper divertor chamber takes the particle exhaust load. Another set ofexperiments is proposed to investigate the effectiveness of a full double null configurationwith both divertor chambers pumping that would be compared to partial pumpingconfigurations.

Integrated Detached Divertor Operation with All Metal Walls and AdvancedTokamak

This is a crucial need for the future of an advanced burning plasma experiment and wouldbe one of the major accomplishments of the US Base program. The proposed C-Modprogram would have the capability to carryout this task.

3.2.6. RF (ICRF and LHCD)

Coupling Physics of ICRF and LHCD

The proposed experimental program will be of great importance in providing experience on thecoupling of ICRF in various RF heating scenarios of interest to FIRE. It will be important toresolve the LHCD efficiency issue that was raised at Snowmass so that uncertainties in the ITERand FIRE LHCD current drive calculations can be reduced.

ICRF and LHCD Launcher Development

The experience gained on the operational characteristics of the high power launchers will bevaluable for the design of launchers for both FIRE and ITER-FEAT.

3.2.7. Global Stability

NTM Stabilization by LHCD

FIRE proposes to use LHCD as one of the techniques for stabilizing the neoclassicaltearing mode in 10 Tesla H-mode operation. Initial experiments on Compass-D haveshown promise for stabilizing the NTM mainly by ∆’ modification. Experiments on C-Mod would extend these results toward FIRE-like parameters.

- 14 -

Disruption mitigation with a neutral stability point due to DN

Disruption mitigation and prevention are very important for burning plasma experimentsand are essential for a tokamak reactor. A neutral stability point has been observed to slowthe onset of vertical disruptions in JT-60U. A balanced DN configuration has a neutralstability point that may allow implementation of a fast vertical feedback system to controland possibly eliminate vertical disruptions. If the C-Mod configuration evolves to abalanced DN configuration then this area of research could be expanded to include ultrafast feedback coils imbedded in the vacuum vessel wall to control disruptions.

3.2.8. Advanced Tokamak

Simulation of AT Regimes

The C-mod advanced tokamak regime using LHCD with the capability to producestrongly reversed shear resulting in fbs~ 65% and βN ≈ 3 without a conducting wall are ofgreat interest to ITER and FIRE. PPPL has significant modeling capability in theTokamak Simulation Code (TSC) and current drive (LSX) that would be useful in refiningthe AT scenarios and in analyzing the experimental results. PPPL and FIRE wouldpropose to have significant collaboration in this area.

3.2.9. Integrated Scenarios

Conventional Edge Barriers (H-Mode, EDA) with Radiative All-metal Divertor

The integration of confinement, MHD stability, divertor and plasma boundary for SN/DNplasmas with radiative divertors would be an important contribution to the physics basicsfor both FIRE and ITER. C-Mod offers the unique features of all metal PFCs and reactorlevel magnetic fields in these experiments.

Advanced Tokamak with Radiative, All-Metal Divertor

This is a critical item for burning plasma R&D leading directly along the path to anattractive tokamak reactor concept, and C-Mod is uniquely positioned to carryout anextensive R&D program in this area. This is the mode of operation for ARIES-RS/AT andfor advanced tokamak modes in FIRE and ITER. ITER would be more interested in theSN configuration while FIRE would be most interested in the full integration of a DNdivertor with double pumping. The flexibility to investigate a range of scenarios withfrom weakly reversed central shear to strongly reversed central shear in essentially steady-state discharges with ~ 5 current redistribution times will provide a strong basis for ATmodes with parameters up to fbs~ 65% and βN ≈ 3. Of particular interest will be the energyand particle confinement within the qmin surface and its scaling with respect to dimensionalparameters (Ip, P, B, etc.) as well as dimensionless scaling. If the confinement barrier iseffective then perhaps higher βN could be explored with resistive wall mode stabilizationas a future upgrade. TRANSP analysis of transport properties and TSC/LSX simulation ofthe overall discharge are capabilities that PPPL can contribute to this collaboration.

- 15 -

4. Transport


4.1.1. Marginal Stability and Turbulence

The measured temperature gradients exceeded - by 50% or more - the best estimates availablein the late 90s of the theoretical critical gradient for ion-temperature-gradient (ITG) modes. Thetheoretical effort was strengthened with more complete simulations undertaken by PPPLresearchers, and linear micro-stability theory continued to be at odds with the availablemeasurements.

This motivated nonlinear turbulence simulations using the gyrokinetic code GS2 [1,2], whichshowed that the discrepancy could be understood as a substantial nonlinear upshift (the so-called‘Dimits shift’) in the ‘effective’ critical gradient due to stabilization of ITG modes by zonalflows. This upshift had been previously predicted theoretically, but no experimental evidence forit had been put forward. These results are illustrated in Fig. 5.1 where the power conducted by thecomputed turbulence is plotted against the normalized ion temperature gradient. The experimentalgradient likely to be 50% higher than the linear critical gradient at R/LTi=4.5.

0

5

10

15

20

25

4 5 6 7

Con

duct

ed p

ower

(M

W)

R/LT

IFS-PPPLmodel

Nonlinear GS2 simulations

Measured R/LTe

Pheat

Lower νe & ν

i

Lower νi

Standardcollisionality

C-Mod EDA H-modermid

=0.56a

Kinetic electrons and ions

Fig. 4.1. Conducted power as a function of R/LT as computed by the IFS-PPPL model and theGS2 code.

- 16 -

This work led to a further discovery with potentially great importance because it highlightedthe importance of a kinetic treatment of the electron dynamics [3], which had been ignored inprevious theoretical predictions of the Dimits shift. It appears from the current work that theDimits shift may be much smaller than previously thought in lower-collisionality tokamaks suchas JET, DIII-D, and ITER (see the lower νe & νi curve in the Figure). The shift appears to belarge in C-Mod because its highly collisional plasma largely erases the non-adiabatic features ofthe electron dynamics that can weaken the Dimits shift.

References

1. M. Kotschenreuther, G. Rewoldt, W.M. Tang, Computer Phys. Comm, 88, 128, 1995.2. W. Dorland, F. Jenko, M. Kotschenreuther, B.N., Phys. Rev. Lett. 85, 5579, 2000.3. D. Mikkelsen, et al., "Nonlinear Simulations of Drift-Wave Turbulence in Alcator C-

Mod", in Proceedings of the 19th IAEA Fusion Energy Conference", Lyon, IAEA-CN-94/EX/P5-03, 2002.

4.1.2. ITB Modeling

Linear and nonlinear calculations of gyrokinetic microturbulence have been carried out withthe GS2 flux tube code at the trigger time for formation of an internal transport barrier in the off-axis RF H-mode. It is found that ITG modes are unstable outside the core plasma, that only weakinstabilities are present in the plasma core and that in the barrier region the plasma instabilitiesare quiescent at the trigger time. Sensitivity studies indicate that the normalized electrontemperature gradient drives the suppression of instability most strongly in the region where theITB will form, consistent with experimental measurements.

References

1. “Alcator C-Mod H-modes", APS conference, April 5-9, 2003, Philadelphia, PA.2. "Gyrokinetic Microturbulence and Transport Calculations for NSTX and Alcator C-

Mod", Transport Task Force Conference, April 2-5, 2003, Madison, WI.3. "Gyrokinetic Calculations of Microturbulence and Transport for NSTX and Alcator-

CMOD H-modes", to be presented at the 30th EPS Conference on Plasma Physics andControlled Fusion, July 7-11, 2003, St. Petersburg, RU.


4.2.1. Marginal Stability and Turbulence

It has become clear that nonlinear processes are crucial in determining both the saturated levelof turbulence and its associated transport and the ‘effective’ critical gradient (which waspreviously thought to be more easily calculable via simple linear stability). In order to trulyunderstand transport, one must measure the turbulence directly and understand the nonlinearsaturation processes. C-Mod provides plasmas that are complementary to those of othertokamaks, and thus helps to complete our understanding.

- 17 -

PPPL will contribute to this understanding by providing and operating a reflectometer whichwill measure turbulent fluctuations. We will carry out detailed comparisons of data from all theC-Mod fluctuation diagnostics with nonlinear gyrokinetic turbulence simulations using the GS2and GYRO codes.

As data from these diagnostics becomes available we will help to plan experiments that shouldoptimize the overlap between diagnostic capability and theoretically expected fluctuationcharacteristics. Key parameters will be identified in simulations, and experiments will bedesigned to test the importance of those parameters in real plasmas. This work will also benefitfrom the experience gained by PPPL researchers active in complementary studies at other sites.

4.2.2. Electron Transport

As the fusion program progresses toward reactor-grade plasmas, with predominantly electronheating and strongly coupled electron and ion temperatures, the importance of electron transportnecessarily grows. Experimental opportunities for electron fluctuation measurements aredescribed in the C-Mod 5-year plan, and relevant turbulence simulations will be carried out atPPPL.

An experiment proposed by PPPL personnel and carried out in C-Mod with MIT hasdemonstrated that it is possible to measure the electron temperature gradient scale length to highprecision (5-10%) when sawtooth heat pulses can be minimized. We plan to search for a way toexploit this capability by designing experiments that will be in a regime with a ‘critical gradient’dependence on electron temperature, but with a weak dependence on the ion temperature gradient(which is more difficult to determine). We would then compare the experimental and theoreticaldependences of the electron temperature gradient as an important parameter such as the q profileis varied.

4.2.3. ITB Modeling

The nonlinear ITG calculations will be processed with the Nevin’s GKV post-processor andheat and particle fluxes will be compared with transport analysis for the off-axis RF experiment.The existence of a Geodesic Acoustic Mode in the plasma core at the trigger time will beexplored. Initial GYRO code simulations of CMOD will be extended and compared with thosefrom GS2. The Ohmic H-mode ITB (Fiore, Phys. Plas. 2001) has many similarities to the off-axis RF H-mode ITB, but does not appear to be triggered by reduced electron temperaturegradient, so a parallel study of gyrokinetic microstability will be interesting and can be used toplan for future experimental tests. Nonlinear calculations of ETG turbulence for CMOD will givesome insight into the role of ETG on electron thermal transport on this ITER-relevant tokamak.

5. Divertor and Plasma Boundary


5.1.1. GPI Edge Turbulence Visualization

- 18 -

The PPPL/MIT gas puff imaging (GPI) diagnostic of edge turbulence was proposed in 1998,designed in 1999, and installed on Alcator C-Mod in early 2000. The first 2-D images werereported at the APS meeting in 2000, and initial results were presented in invited talks at the APSmeetings in 2001 [1] and 2002 [2]. This project is a close collaboration between PPPLresearchers and Jim Terry, Brian LaBombard, and others at MIT, in addition to people fromseveral other institutions (LANL, Garching, Dartmouth, and Princeton Scientific Instruments,Inc.).

The first GPI results were obtained with a 60 Hz Xybion intensified camera gated at 2µsec/frame, as shown in Fig. 5.1 [1]. The strongly turbulent structure was evident, but the motionof the turbulence could not be seen due to the low framing rate. The frequency spectrum andfluctuation level were checked to be similar to those seen by a Langmuir probe at the same radialposition. Initial comparisons were made with the 3-D fluid simulations of Hallatschek andRogers based on a local model, i.e. using the gradients at a representative point in the edge.

The GPI diagnostic was significantly improved in 2001 by the addition of an ultra-fast CCDcamera obtained through an SBIR with Princeton Scientific Instruments, Inc. With this PSI-3camera the motion of the turbulent blobs could be seen for the first time, as illustrated in Fig. 5.2[2-3]. This allowed us to visualize the rapid outward motion of the turbulent “blobs”, which hadbeen predicted theoretically [4]. An upgraded 28 frame PSI-4 camera was operated on C-Mod in2002 and many additional movies of the turbulence have been created under various conditions.

- 19 -

Fig. 5.1. 2-D images of edge turbulence in C-Mod taken with the GPI diagnostic near the outermidplane at 60 frames/sec with an exposure time of 2 µsec per frame. The diamond-shapedregion is 6 cm x 6 cm, with the radially outward direction toward the left, and the separatrix is theblack line through the center of the images.

Fig. 5.2. 2-D images of edge turbulence in C-Mod taken with the Princeton ScientificInstruments PSI-3 camera at 250,000 frames/sec with the same field of view as in Fig. 1.The rapid radial motion of the turbulent “blobs” can be seen with reference to the fixed arrows.

- 20 -

Theoretical simulations of C-Mod edge turbulence were improved by Hallatschek in 2002 toinclude nonlocal effects, i.e. the radial profile of the edge parameters. A direct comparison of thek-poloidal spectrum between GPI and simulation is shown in Fig. 5.3 [2]. The spectral range andshape agree fairly well in the SOL, but the simulation underestimates the turbulence in the limitershadow (right hand side). Nevertheless, this is among the most successful of the few directcomparisons to date between tokamak turbulence measurements and numerical simulations ofturbulence.

Fig. 5.3. Direct comparison between the GPI results and the NLET simulations of edgeturbulence in C-Mod. On the left is the measured k-poloidal spectrum of turbulence vs. minorradius, and on the right is the simulated k-poloidal spectrum from NLET. The NLET simulationshave been post-processed to calculate the actual spectrum of Dα light as measured by GPI. Theagreement is fairly good in the SOL, but the simulation underestimates the turbulence in thelimiter shadow (right hand side).

Also in 2002 the GPI diagnostic was further improved by the addition of a second Xybioncamera, which allowed two-time and two-color imaging of the edge turbulence. Initial resultswere reported at the APS 2002 meeting [5]. This work will be continued in 2003 with the aim ofmeasuring the motion of the turbulence during the entire discharge and the ratio of electrondensity to temperature fluctuations.

Modeling of the atomic and neutral physics of the GPI diagnostic was also done using theDEGAS-2 code. Simulations of GPI experiments on Alcator C-Mod demonstrated that theconnection between the emission patterns and plasma parameters is sufficiently complicated toprohibit a direct inversion of the camera images into separate density and temperature fluctuationimages [6]. However, it was also shown that the same spatial structure was apparent in thedensity and/or temperature patterns and the GPI line emission patterns, suggesting that awavenumber analysis would yield the same spectrum in both cases.

References

1 . S.J. Zweben, D.P. Stotler, J.L. Terry, B. LaBombard, M. Greenwald et al, Phys.Plasmas 9, p. 1981 (2002).

- 21 -

2. J.L. Terry, S.J. Zweben et al, “Observations of the Turbulence in the SOL of Alcator C-Mod and NSTX and Comparison with Simulation”, to be published in Phys. Plasmas(2003).

3. J.L. Terry, S.J. Zweben et al, Fusion Energy Conference IAEA 20014. S. Krasheninnikov, Phys. Lett. A 283, 368 (2001); D.A. D’Ippolito et al, Phys. Plasmas

9, 222 (2002).5. B. Bai, J.L. Terry et al, Bull. Am. Phys. Soc. 47, 9 p. 236 (APS DPP meeting ’02).6. D.P. Stotler, B. LaBombard, J. Terry, S. Zweben, J. Nucl. Mat. 313-316, 1066 (2003)

5.1.2. Edge Neutrals Modeling

The behavior of neutral gas in the periphery of magnetic fusion devices plays a crucial role incontrolling the core density and in establishing the boundary conditions for the core. The plasmadensity at the last closed flux surface determines the relationship between limits on the edge andscrape-off layer density and the core density. Furthermore, transport theories based on criticalgradients suggest that the plasma temperature just inside the last closed flux surface will stronglyaffect the core temperature. The core parameters in burning plasma experiments then in turn setthe fusion power.

The hypothesis of and growing evidence for the dominance of main chamber recycling overthat of the divertor must be examined and understood in order for scenarios for power and particlecontrol to be developed in future devices. That understanding can be deduced from anexamination of detailed computer models of the plasma and neutral transport that reproduce theavailable measurements in the edge and divertor of present machines.

Neutral pressure measurements and imaging of visible light emissions (e.g. the Balmer- α or D α

line) are the most useful diagnostics signals for comparison with neutral gas transport models.However, these models require a specification of the plasma parameters everywhere in themachine as well as the location and strength of neutral gas sources. The extensive diagnosticsavailable on Alcator C-Mod have permitted the development of simple models of the plasmavariation in the scrape-off layer and divertor that are largely determined by diagnostic data. As aresult, the neutral gas behavior can be examined by a stand-alone code such as the DEGAS-2Monte Carlo neutral transport code.

A first attempt along these lines yielded neutral pressures and D α signals much smaller than themeasured values [1]; the discrepancy was blamed on the omission of ion-neutral and neutral-neutral elastic scattering from the model. The subsequent addition of these processes to DEGAS-2 coincided with the execution of experiments dedicated to the study of neutral gas transport inthe Alcator C-Mod divertor and sub-divertor [2]. In spite of the model improvements, thediscrepancies persisted. Potential explanations for the discrepancy were then evaluated bydetermining the sensitivity of the simulated values to changes in the assumptions underlying thesimulations [3]. Models of the plasma-surface interaction processes (such as reflection andabsorption) and of the interpolation of plasma parameters between probe locations werehighlighted as potentially viable explanations.

References

1. “DEGAS 2 Neutral Transport Modeling of High Density, Low Temperature Plasmas,”D. P. Stotler, et al., PPPL Report PPPL-3221 (January 1997). Proceedings of the

- 22 -

Sixteenth International Conference on Plasma Physics and Controlled Nuclear FusionResearch, (Montreal, Canada, September 1996), Vol. 2, p. 633–639.

2 . “Modeling of Alcator C-Mod Divertor Baffling Experiments,” D. P. Stotler, et al.,Proceedings of the Fourteenth International Conference on Plasma Surface Interactionsin Controlled Fusion Devices, (Rosenheim, Germany, May 22–26, 2000). PPPL Report,PPPL-3523 (November 2000). J. Nucl. Mater. 290–293, 967–971 (April 2001).

3. “Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor,” D.P. Stotler, et al., PPPL Report PPPL-3690 (May 2002). Atomic Processes in Plasmas,13th APS Topical Conference on Atomic Processes in Plasmas, (Gatlinburg, Tennessee,April 22–25, 2002), p. 251–260.

5.1.3. Reflectometer Upgrade

Two new, higher frequency, channels have been added to the C-Mod microwavereflectometer, which will allow us to measure plasma fluctuation farther up the edge pedestal, andpossibly to the ITB region at low density. The hardware for these new 130 and 140 GHzchannels has been installed and checked out. Analysis codes have been ported from JT-60U,where they were developed and first used by G. Kramer, to the PPPL Petrel cluster for use withthe C-Mod measurements.


5.2.1. Edge Turbulence Visualization Extended

In FY’03 and FY’04 the GPI diagnostic on C-Mod will be used to explore the edge turbulencebehavior under a wide variety of edge conditions. This will be greatly helped by the anticipateduse of a 312 frame PSI-5 camera instead of the present 28 frame PSI-4 camera, since over 10times more data can be obtained per shot.

The main goal will be to make detailed quantitative comparisons between the turbulencemeasurements and theoretical simulations. A secondary goal is to determine the empiricalcorrelations between the edge turbulence and edge transport; for example, during the EDA andELM-free H-modes vs. L-mode, and near the Greenwald density limit. In each of these casesthere is reason to believe that the turbulence plays a major role in determining the edge transportstate, but so far no clear causal relationship has been established.

In FY ’05-’06 we would like to significantly increase the capabilities of the edge turbulenceimaging system on C-Mod. Several options listed below, one or more of which could beimplemented during this period. In FY ’07 and ’08 we will use the upgraded GPI system tofurther explore the physics of edge turbulence, and compare the results with other C-Moddiagnostics and with theoretical simulations. Note that each of these potential upgrades could bedone independently of the existing GPI system and each other, so that more than one can be doneat the same time or over several years.

1. 2-D imaging of edge turbulence near the inner SOL.2. 2-D imaging of edge turbulence near the X-point.3. 2-D imaging with two wavelengths for ne/Te measurements.

- 23 -

4. Supersonic gas injector for increased radial coverage.5. High resolution 2-D imaging of small-scale edge turbulence.6. Wide angle 2-D imaging for large-scale edge turbulence.

Some details about these diagnostic upgrades are discussed below:

1. The added capability of 2-D imaging at the inner SOL would be important for testingtheory, which generally predicts that inner wall turbulence is stabilized by favorablemagnetic curvature. A first step has already been taken in this direction on C-Mod bymeasuring the gas puff emission on discrete chords near the inner limiter [1], whichshowed that the fluctuation level there was about ten times lower than on the same fluxsurface at the outer limiter, in qualitative agreement with simulations. Three-dimensionalsimulations such as BOUT have already been used to calculate the poloidal distributionof edge turbulence in tokamaks [2], and strong changes in the poloidal distribution havebeen predicted near the density limit and during L-H transitions, but little or noconnection with the measured poloidal distribution of turbulence has yet been made.

2. The X-point of a diverted tokamak is important for edge studies since the local magneticshear and small poloidal field can strongly change the edge stability and transport. Anew class of modes called resistive x-point modes (RX) was recently discovered and aconnection between these modes and the H-mode transition and density limit has beenproposed [2]. However, very few measurements have been made of edge turbulence nearthe X-point, and so a 2-D imaging system focused on the X-point would be quiteinteresting for comparing with theory. For example, BOUT simulations predict that asthe density limit is approached, the dominant modes shift from the resistive X-point modelocalized near the X-point to the resistive ballooning mode localized near the outermidplane.

3. The present GPI system uses one spectral line at a time to image the turbulence, so bothelectron density and electron temperature fluctuations can contribute to the observedturbulent structure. In theory, these fluctuations are closely linked; nevertheless, it wouldbe interesting to measure them separately. A technique to do this involving the use oftwo cameras viewing the same GPI image in two different spectral lines has already beenproposed and tested on C-Mod [3], but further development is needed to increase ourconfidence in the results. This would mainly involve data taking and analysis, rather thannew hardware.

4. The present GPI system is limited to the region where neutrals can penetrate into theedge, which extends to just barely inside the separatrix at the outer midplane. Theaddition of a supersonic neutral gas injector inside the vessel could increase the radialviewing range of the GPI images, thus making it possible to view the H-mode barrier andpedestal region using GPI. This nozzle might also be useful to change the fuelinglocation and increase the fueling efficiency in C-Mod. A nozzle design compatible withthe required fueling rate and in-vessel constraints would need to be designed and testedbefore installation on C-Mod.

5. The spatial resolution of the present 2-D GPI imaging system is limited to ≈ 0.3 cm bythe in-vessel optics, but might be significantly improved in order to look for the small-scale turbulence with kpolρs > 0.5, i.e. λ ≤ 0.4 cm. This may be in the wavelength rangeof the nonlinear inverse-cascade spectrum of ETG modes (which have linear growth nearkpolρs ≈ 50, which is too small to see with GPI). This upgrade would probably require a

- 24 -

new magnifying telescope inside the vessel, and perhaps a smaller neutral gas injectorand a higher resolution coherent fiber bundle. A detailed study of the limiting spatialresolution of the entire GPI system would have to be completed before this wasimplemented.

6. At the other end of the size-scale spectrum, the present GPI system is limited to kpolρs >0.03, i.e. λ ≤ 4 cm, due to the limited size of the injected gas cloud, the limited field ofview of the in-vessel optics, and (to some extent) the curvature of the magnetic field.Since the amplitude of the observed turbulence is still high at the low-k end of thespectrum (Fig. 5.3), it would be very interesting to image a larger poloidal length todetermine where the k-spectrum begins to fall, and whether there is any coherentstructure at these larger scales. This could potentially be done by using multiple andoverlapping GPI systems similar to the present one, or by designing an in-vessel“fisheye” view, or by imaging a large gas cloud in the toroidal vs. poloidal plane from theside. A detailed study of the limiting spatial resolution of the entire GPI system wouldhave to be completed before this was implemented.

References

1 . S.J. Zweben, D.P. Stotler, J.L. Terry, B. LaBombard, M. Greenwald et al, Phys.Plasmas 9, p. 1981 (2002)

2 . S.J. Zweben, D.P. Stotler, J.L. Terry, B. LaBombard, M. Greenwald et al, Phys.Plasmas 9, p. 1981 (2002)

3. B. Bai, J.L. Terry et al, Bull. Am. Phys. Soc. 47, 9 p. 236 (APS DPP meeting ’02).

5.2.2. Comparison with Theoretical Simulations

It is crucial to compare the results of these measurements with the best available models andtheoretical simulations in order to understand what we are seeing. In FY’03 and FY’04 we willcontinue to analyze the existing GPI data and work with theorists to compare the GPI results withedge turbulence simulations and models. This process will continue throughout the 5-year planperiod, or until a good understanding of the physics of edge turbulence has been obtained.

In order to compare experiment and theory it will be important to calculate the expected GPIlight emission patterns from the theoretical simulations of the spatial structure of the plasmaturbulence. This modeling was begun using DEGAS 2 in Ref. [1] and will be extended so thatthe full atomic physics and response of the neutral density can be examined. For example, a moredetailed atomic physics model allowing the metastable states of HeI to be treated accurately willbe added to DEGAS 2, and their impact on the effective density and temperature dependence ofthe emission rates will be examined (FY-2003). In addition, a toroidally resolved (i.e., 3-D)simulation of the GPI experiment will be developed, including a 3-D representation of the gaspuff and the camera field of view. DEGAS 2 modeling of the various proposed hardwareupgrades will also be quire useful, e.g. for the two-color imaging, the supersonic gas nozzle, andthe extension of the k-spectral resolution of GPI.

There are several distinct areas for comparison with theory:

1. 2-D spatial structure and k-spectrum.2. Presence of zonal flow or streamers.3. Dynamics of coherent structures (blobs).

- 25 -

4. Scaling with dimensionless parameters.5. Influence of neutral density and/or atomic physics.6. Effect of boundary conditions, e.g. separatrix and/or limiters.

Some details about these theoretical connections are below:

1. The 2-D spatial structure of edge turbulence has already been measured with GPI in C-Mod. The measured kpol wavenumber spectrum agreed fairly well with the 3-D nonlocalfluid simulations of Hallatschek’s NLET edge turbulence simulation code, as shown inFig. 5.3. Further cases with different edge conditions will be examined using theexisting system, and the radial wavenumber spectra will also be compared whereverpossible. In particular, the behavior near the density limit and will be studied in detailusing He puffing into D discharges (or vice versa) to reduce the background fromnatural plasma light emission, and the results will be compared with the NLET andBOUT codes.

2. Recent turbulence simulations have pointed out the importance of poloidally directedzonal flows and radially directed streamers for the regulation of turbulence and possiblythe origin of the H-mode [2]. Although a related coherent poloidal zonal flow called ageodesic acoustic mode (GAM) has been measured in DIII-D using BES edgeturbulence data and in TEXT with the HIBP [3], the more interesting broadband, near-zero frequency zonal flows and streamers discussed in most theoretical papers have notyet been identified in experiment. The high-speed 2-D images from GPI in C-Modshould be an ideal way to identify these structures, although it is likely that the largerdata sets such as those from the 312 frame PSI-5 camera will be needed. Directcomparison with theoretical simulation of C-Mod will be made where possible.

3. The presence of coherent structures called “blobs” is obvious from the existing C-ModGPI data (Figs. 5.1 and 5.2), and simple models for blob motion have been discussedtheoretically [4]. A first attempt at comparison of the GPI data with the blob model willbe started in FY ’03-’04, but most likely the model and database will have to beextended in order to obtain a good quantitative comparison. For example, it is not yetclear from either the model or the data where and how blobs are formed or why theobserved blobs move poloidally in addition to the usual radially outward blob motion.Hopefully a clear connection between the simplified blob model and the detailedturbulence simulations will also be established over time.

4. Probably the best check of a theoretical model comes from establishing the scaling ofthe data with the main parameters in the theory. The main local scaling variables ofinterest are the collisionality, ion gyroradius, beta, and magnetic shear. At present thereis relatively little information about the scaling of edge turbulence in general, althoughthere are experimental and theoretical indications that the density limit [5] and the L-Htransition [6] are associated with higher and lower levels of turbulence, respectively. AB-field scaling study using GPI data was started on C-Mod in FY ’02, and similarstudies will be continued in FY ’03-’04 and beyond.

5 . The potentially important effect of neutrals and atomic physics on edge plasmaturbulence and transport has been noted frequently but seldom quantified eitherexperimentally or theoretically. For example, it is clear that edge temperatures andglobal confinement can be higher with clean and low recycling walls, but it is notunderstood why. The present or upgraded GPI system can be used to explore these

- 26 -

dependences in C-Mod, e.g. by measuring the observed turbulence with various types ofdeuterium or impurity gas puffs. This type of study is difficult due to incompleteknowledge and control of the edge neutral and impurity profiles, so it will extend overseveral years.

6. The plasma magnetic and/or limiter boundary conditions are also well known to affectboth the edge and global plasma confinement, but few systematic measurements of edgeturbulence have been made to clarify these dependences and compare them with theory.For example, the electrostatic potential and spatial distribution of the limiters in the SOLshould affect the radial electric field and parallel flow, thus influencing the edgeturbulence and transport. Thus it might be possible to introduce biased limiters orelectrodes in C-Mod to study to SOL, similar to experiments done on limited machines[7] and planned for MAST. Alternately, a study of the effect of varying “gaps” betweenthe diverted plasma separatrix and the wall would be interesting and could be comparedwith simulations.

References

1. D.P. Stotler, B. LaBombard, J. Terry, S. Zweben, J. Nucl. Mat. 313-316, 1066 (2003)2. P.H. Diamond et al., Nucl. Fusion 41, 0167 (2001); X. Garbet, Plas. Phys. Cont.

Fusion F43, A251 (2001), Z. Lin et al, Phys. Rev. Lett. 88, 195004 (2002).3. M. Jakubowski et al, Phys. Rev. Lett 89, 265003 (2002); P.M. Schoch et al., Rev. Sci.

Inst.74, 1846 (2003).4. S. Krasheninnikov, Phys. Lett. A 283, 368 (2001); D.A. D’Ippolito et al., Phys.

Plasmas 9, 222 (2002).5. M. Greenwald, Plasma Phys. Cont. Fusion F44, R27 (2002); B. LaBombard et al., Phys.

Plasmas 8, 2107 (2001).6. K.H. Burrell et al., Phys. Plasmas 4, 1499 (1997).7. J. Boedo et al., Nucl. Fusion 40, 1397 (2000).

5.2.3. Edge Turbulence Control Experiments

Experiments to control edge turbulence have both a practical and an intellectual value. On thepractical side, it would be quite useful to be able to change the edge conditions of an advancedtokamak in an externally controllable way; for example, to increase the width of the edge pedestalduring an H-mode. On the intellectual side, a well designed control experiment can be anexcellent test of the underlying physics; for example, when the control is based on a specificmodel of edge turbulence.

There are several possible edge turbulence control experiments in C-Mod:

1. Edge minority heating for H-mode control.2. RF sheath effects for edge turbulence control.3. Edge fueling control (pellets and strong puffs).4. Lower hybrid current drive for edge control.5. Localized limiter or electrode biasing.

Some details about these edge turbulence control experiments are discussed below:

- 27 -

1. In FY ’03 and FY ’04 we plan to start edge turbulence control experiments with an edgeminority heating experiment, which aim to change the edge radial electric field throughcontrolled generation and loss of minority tail ions. An initial calculation by C.S. Changhas indicated that the edge radial electric field can be significantly changed in C-Modwith 1 MW of RF coupling to the minority tail ions. If this method works, it may bepossible to externally trigger an H-mode or to increase the pedestal width of an existingH-mode. Even if this technique does not succeed in C-Mod, it would be interesting tostudy the local effect of RF tail ions on the edge turbulence through GPI measurements.

2. It is well known, and previously seen on C-Mod [1], that DC electric field sheaths can begenerated by rectification of the RF-driven currents to the RF antenna structures. Usuallythese sheaths are considered a nuisance, since they can lead to edge ion acceleration andimpurity generation at the antenna. However, some theoretical work [2] has suggestedthat these sheaths can be used to modify edge turbulence through the flows generated bythe resulting convective cells in the SOL. If so, the RF system might be used to controlheat and particle transport across the SOL. Initial experiments might be done using theexisting RF antennas, and future control experiments could potentially be designedthrough modeling and simulation.

3. It was recently shown on DIII-D that edge pellet injection can be used to trigger an H-mode below the conventional power threshold [3]. This discovery has already inspired anew model of edge turbulence which emphasizes the role of the local density gradient onthe H-mode threshold [4]. We should be able to check and extend these results using thelithium pellet injector on C-Mod, which can inject Li pellets into the edge. If successful,this can lead to an external control for the onset on H-modes in advanced tokamaks.Alternatively, a strong localized deuterium or low-Z impurity gas puff may be used,perhaps with the same supersonic nozzle mentioned in Sec. 5.2.1.

4. Lower hybrid current drive might be a powerful tool for edge turbulence and transportcontrol, since the LH waves can generate localized edge current and possibly fastelectrons at the edge. A first step would be to investigate the theoretical dependence ofedge turbulence on the local current density. A second step would be to evaluate whetherLH waves can create fast electron loss to the wall, which could change the edge electricfield similarly to fast ion loss. After the LH system is installed on C-Mod it would beinteresting to observe the effect on edge turbulence, and vice versa.

5. Limiter or electrode biasing has been used previously on tokamaks to change the radialelectric field profile and, for example, to induce an H-mode [5]. Although this techniqueis unlikely to be relevant to a burning plasma experiment, it might be used to control theedge on C-Mod and explore the effects of Er on the edge transport, the L-H transition,and the density limit. Initial experiments could be started with a biased probe or smallplate inserted into the SOL, and its effect can be monitored using GPI and other edgediagnostics.

References:

1. B. LaBombard, private communication.2. D.A. D’Ippolito et al, Nucl. Fusion 42, 1357 (2002).3. P. Gohil et al, Phys. Rev. Lett 86, 644 (2001).4. P.N. Guzdar et al, Phys. Rev. Lett. 89, 265004 (2002).5. J. Boedo et al, Nucl. Fusion 40, 1397 (2000).

- 28 -

5.2.4. Edge Modeling Extended

The toroidally axisymmetric simulations of Ref. [3] of Section 5.1.2 incorporate values for theneutral gas conductances between the various parts of the C-Mod divertor and sub-divertor thatare necessarily approximate. By definition, these conductances determine the relationshipbetween the neutral pressure near the target plate (set mostly by the plasma parameters) and thatat the location of the experimental diagnostics. In reality, the sub-divertor hardware in C-Mod isthree-dimensional. The only way to calibrate the assumed axisymmetric conductances is toanalyze that three-dimensional structure.

A series of neutral pressure measurements with calibrated gas puffs has provided experimentalvalues for these conductances. Because the impact of the plasma on these measurements iseither nonexistent (some experiments were done without a plasma present) or modest, they areideal for use in calibrating a neutral gas transport model.

The Alcator C-Mod vacuum vessel and sub-divertor can be represented by an axisymmetricobject with localized asymmetries. DEGAS 2 currently utilizes a very flexible code for setting upaxisymmetric geometries. Modest extensions of this code allow the required asymmetric objectsto be straightforwardly specified. This is all that is required for the simulation of theseexperiments since the core of the DEGAS 2 code is naturally three-dimensional.

A successful benchmarking of DEGAS 2 against these conductance measurements would laythe groundwork for subsequent investigations of the interaction of neutral atoms and moleculeswith the plasma. For example, simulations of the impact of the divertor bypass on plasmaconditions could be revisited. Or, by combining this neutral transport model with moresophisticated plasma transport models such as UEDGE or OSM, the issue of main chamberrecycling could be addressed in a quantitative way.

Monte Carlo neutral transport calculations are essential to the design of new particle controlhardware. In the event that the Alcator C-Mod team opts to install a cryopump on the tokamak,the calibrated DEGAS 2 model could be used to evaluate and optimize the design.

The neutral densities routinely achieved in Alcator C-Mod are high enough that reabsorptionof photons emitted by excited atoms can alter the neutral atom excited state distribution. M.Adams has previously used the CRETIN code to simulate radiation transport in the C-Moddivertor. However, CRETIN is not currently capable of treating details of the transport of neutralspecies. Although a coupling between CRETIN and DEGAS 2 is potentially complex andcomputationally demanding, the prospect of obtaining a self-consistent solution is sufficientlyattractive to make the effort worthwhile.

5.2.5 Reflectometer Extensions

The possibility exists for adding additional reflectometer channels at frequencies above140 GHz. These would allow the measurement of plasma turbulence even closer to the core, andhelp to gain insight into the behavior and possibly influence of turbulence on internal transportbarriers.

- 29 -

6. RF Physics and Wave-particle Interactions


6.1.1. D(H) Minority Heating Experiments

The D(H) minority regime has been used extensively for plasma heating in a wide variety of C-Mod experiments for many years. It will continue to provide heating for many of the physicsinitiatives even after the installation of LH heating hardware in FY03. PPPL scientists haveparticipated in many C-Mod experiments that rely on D(H) minority heating, and we expect tocontinue active participation throughout the upcoming five year program.

6.1.2. Mode Conversion Experiments

As the minority ion concentration is raised, or as a third ion species is introduced, the launchedFast Wave may mode convert to either an Ion Bernstein Wave or an Ion Cyclotron Wave (firstproposed by R. Perkins, PPPL). C-Mod’s high RF power level and its new and unique wavepropagation diagnostic have allowed a careful investigation of these processes to take place.

6.1.3. ICRF Modeling

A significant part of the PPPL collaborations program on C-Mod has been focused on thedevelopment of quantitatively accurate RF simulations packages to support the experimentalprogram objectives and to advance the understanding of the dynamics of electromagnetic waveinteractions in inhomogeneous magnetized plasmas. The METS 1D all-orders kinetic wave solverhas been used to explore minority heating and mode conversion scenarios in C-Mod, while theTRANSP time-dependent transport analysis code has been used to analyze ICRF heatingexperiments and to simulate lower-hybrid driven AT scenarios. These simulations studies havebeen conducted in close collaboration with Dr. Paul Bonoli (PSFC).

The transition from the minority heating regime to the mode conversion regime was analyzedwith the 1D METS code and compared to the 2D simulations provided by the TORIC 2D FLRkinetic wave code. Though the simulations from both codes agreed moderately well in the lowconcentration minority heating regime, there was a pronounced disagreement as the minorityconcentration was raised and mode conversion processes became significant [1,2]. The studiesindicated that TORIC was over-predicting the amount of minority ion absorption, probably due toinaccurate solutions for the RF wave fields. Subsequent work by Dr. Bonoli indicated that aninsufficient number of poloidal modes were retained in the simulations, leading to poorlyconverged solutions. By parallelizing the TORIC code and increasing the number of poloidalmodes, the first 2D simulations of mode conversion in C-Mod were obtained [3].

Time-dependent simulations of RF heated discharges provide a critical tool for understandingtransport and stability of C-Mod plasmas. Towards this end, the TORIC code was implemented inthe TRANSP code to provide the capability to model mode conversion as well as minorityheating experiments for the C-Mod program [4]. Because the previous ICRF package inTRANSP, SPRUCE, was based on a reduced order algorithm, it was not capable of analyzingexperiments in which mode conversion processes are significant. The TORIC code retains boththe fast wave and the mode converted waves in the field and power deposition solutions, though

- 30 -

its range of validity is restricted to plasmas in which the Larmor radii of the various plasmaspecies is small compared to the perpendicular wavelength of the modes. The upgradedTRANSP/TORIC package was benchmarked successfully against the TRANSP/SPRUCEpackage in the minority heating regimes. It was subsequently used to study the RF-induced ITBdischarges in C-Mod [3].

References

1. C.K. Phillips et al, “Modeling of ICRF Experiments in C-Mod”, Bull. Am. Phys. Soc.Vol. 44, No. 7 (1999) pg 206, abstract KP1-45.

2. P.T. Bonoli et al, Phys. Plasmas 7 (2000) 1886.3. P.T. Bonoli et al., “Numerical Modeling of ICRF Physics Experiments in the Alcator C-

Mod Tokamak”, Proc. 18th IAEA Fusion Energy Conf. (Sorrento) EXP4/01(2000).4. C.K. Phillips et al., US-Japan Workshop on Radio Frequency Physics, March 14-16,

2000 (Plasma Physics Laboratory, Princeton, NJ).

6.1.4. LH Modeling

The TRANSP package mentioned above also includes a lower hybrid package, LSC, which hasbeen used to explore the potential for maintaining AT discharges in C-Mod using lower hybridcurrent drive.


The C-Mod ICRF program allows PPPL to participate in a variety of experiments that followon from our previous work on PLT and TFTR and that complement on-going experiments onNSTX. The advances in wave diagnostics such as the PCI allow more detailed measurements ofwave propagation and absorption. This coupled with our improved modeling capability can leadto a much fully understanding of the multi-ion species ICRF application. Mode ConversionCurrent Drive (MCCD) and flow shear drive were topics just demonstrated but not fully exploredor exploited on TFTR. The C-Mod program presents an opportunity to evaluate these regimes fortheir Advanced Tokamak application. Fast Wave Current Drive is being utilized on the NSTXexperiment at high harmonics of the cyclotron frequency. The C-Mod experiment will allow us toexplore the same physics in a different parameter regime and to explore the technical issuesinvolved in controlling an effective antenna spectrum for current drive with different plasma edgeand antenna geometry.

Both NSTX and C-Mod have observed transport barriers in the presence of RF heating.Exploiting these barriers, as C-Mod has already begun to do with their density controlexperiments, will be important for achieving the potential of the Advanced Tokamak.

6.2.1. D(3He) Minority Heating Experiments

These experiments were initiated on PPPL’s PLT tokamak. C-Mod allows a valuableextension into its unique parameter space, with additional and new diagnostics for increasedexperimental information. These measurements will then allow detailed comparisons with theD(H) minority heating scenario to be performed.

- 31 -

6.2.2. FWCD experiments

FWCD is useful in AT discharges to provide on-axis current drive and q(0) control. TheFWCD experiments on C-Mod will be complimentary to those on NSTX since on-axis currentdrive minimizes the reduction in current drive efficiency due to electron trapping and the requiredantenna spectral control is different since combined radial position control and high efficiency arenot simultaneous constraints. On the other hand, the issues of edge plasma interaction withasymmetric antenna phasing will be important on both machines while other edge parasiticeffects may or may not be similar.

6.2.3. MCCD experiments

MCCD can be utilized for on or off-axis supplemental current drive. On TFTR efficientcurrent drive was seen under both conditions. On C-Mod, the PCI diagnostic will allow a bettermeasurement of the correlation between the wave damping and the driven current. The modeconversion scenario also naturally leads into flow drive studies.

6.2.4. Flow drive studies

RF-driven poloidal flow offers the possibility of stabilizing turbulence locally and generatingInternal Transport Barriers. These experiments were initially attempted on TFTR in its lastmonths of operation with the just-installed poloidal rotation diagnostic, with incomplete success.In fact, data analysis took place after the end of TFTR experiments. Subsequent theory andmodeling work has increased interest in this type of rotation control. C-Mod’s higher RF powerdensity and the added and new diagnostics will increase our information yield considerably.

6.2.5. LH wave physics

These experiments will study Lower Hybrid wave launching, propagation, and damping andpower deposition. Launcher-plasma separation and phasing will be varied systematically untilthe optimum directed wave is achieved.

6.2.6. LHCD physics

The purpose of the installation of the LH system is to provide a capability to drive the rightamount of current at a selected radial position to modify the current profile as desired. This phaseof the experiment will require a considerable experimental time as many parameters will have tobe adjusted to obtain the control of the driven current.

Throughout the LHCD physics experiments, an X-ray diagnostic will provide some informationabout the power deposition profile and possible radial diffusion of energetic electrons, but themost essential diagnostic is Motional Stark Effect to measure the radial location of drivencurrents and changes in the q-profile.

6.2.7. RF Modeling

- 32 -

The C-Mod program will be embarking on a campaign to study the dynamics of long pulse ATplasmas, supported by high bootstrap current fractions, ICRF heating and current drive, and lowerhybrid current profile control. The TRANSP / TORIC / LSC package, which provides the abilityto analyze the transport properties of these long pulse AT discharges, will be utilized to supportthe AT experimental program on C-Mod. A new version of the TORIC code, which correctsdifficulties encountered in previous version with spurious edge modes, is being tested andinstalled in TRANSP collaboratively by PPPL [Douglas McCune, C.K. Phillips, B. LeBlanc andthe PPPL-CPPG group], the IPP-Garching group [M. Brambilla] and the C-Mod team [PaulBonoli and John Wright]. Depending on guidance provided by more sophisticated but time-independent modeling studies with the ACCOME, TORIC and CQL3D packages as well as withcomparison to experimental observations, the LSC package may be upgraded or replaced with amore complete model. These simulation studies will be conducted in close collaboration with theC-Mod team.

A potentially important aspect of both the lower hybrid and the ICRF studies will be theinclusion of non-Maxwellian electron and ion particle velocity distributions in the wave field andabsorption calculations. As part of the RF SciDAC collaborative project [1] both the METS 1Dand the TORIC 2D full wave modeling codes are being modified to incorporate the effects ofnon-Maxwellian species on both the wave propagation and absorption. This work will continueafter the completion of the SciDAC project in July 2004. In the lower hybrid regime, previousnumerical and theoretical studies have indicated the lower hybrid wave-induced non-Maxwellianelectron distribution function can play a key role in filling in the “spectral gap” and in improvingthe overall current drive efficiency [2].

On C-Mod, fast electron distributions will be measured in the core plasma using imaging hardx-ray spectrometry. These measurements, along with current density measurements from MSE,will be used to benchmark and verify the physics contained in the generalized METS1D andTORIC-LH simulation codes. In minority ICRF heating regimes, nearly 30 years of experimentaland theoretical studies have demonstrated that an energetic, non-Maxwellian ion tail distributioncan be driven by the RF waves. Similarly, fusion-born alpha particles, which are highly energeticand non-Maxwellian, will be a significant component in burning plasmas. These energetic ionpopulations may alter the wave absorption, propagation and mode conversion properties of thesedischarges. Detailed experimental studies of the dependence of ICRF mode conversion processeson the plasma composition, the plasma density, and the spatial location of the resonance, modeconversion and cutoff surfaces will be compared to simulations from the METS and TORICcodes. These studies will provide the basis for understanding the relative importance of theseprocesses in AT scenarios and in future burning plasmas.

References

1. SciDAC research program “Numerical Computation of Wave-Plasma Interactions inMulti-dimensional Systems” D.B. Batchelor, L.A. Berry, M.D. Carter, E. F. Jaeger, E.D'Azevedo (ORNL); C.K. Phillips, A. Pletzer (PPPL); P.T. Bonoli, J. C. Wright (MIT);D.N. Smithe (Mission Research Corporation); R.W. Harvey (CompX Corporation); andD.A. D'Ippolito, J.R. Myra (Lodestar Research Corporation).

2. S. Bernabei et al., Phys. Plasmas 4 (1997) 125.

- 33 -

7. Theory and Computation of Macroscopic Stability

7.1. Overview

Macroscopic stability is central to the theory of magnetically confined plasmas. Aconfiguration that violated the ideal-MHD stability criteria would rearrange itself in a timecomparable to the Alfven-wave transit time, which is sub-microsecond in modern devices.Configurations that are stable to ideal-MHD must also be examined for stability to dissipative andkinetically-driven modes that can destroy magnetic surfaces, shed energetic particles, and/or leadto discharge termination.

The dominant theme in the near term tokamak research is to strengthen the physics basis forthe International Thermonuclear Engineering Reactor (ITER). The critical topics as listed by theInternational Toroidal Physics Activity (ITPA) topical physics group on MHD, Disruption andControl are listed here and described more in the following sections:

• Beta limiting MHD modes and their active control [including neoclassical tearingmodes (NTM), kink modes, and resistive wall modes (RWM)]

• Edge MHD stability and the behavior of Edge Localized Modes (ELMs)• The prediction, avoidance, and mitigation of disruptions in tokamak plasmas with

edge safety factor q95 ~ 3. This includes calculating forces due to halo currentsduring disruptions, and the prediction, avoidance, and mitigation of large runawaypopulations during disruptions

• Control issues in standard and advanced scenarios: shape and position, performanceincluding profile control with emphasis on integrated scenarios for burning plasmaexperiments.

Formally, the PPPL / Alcator C-Mod collaboration does not include direct financial support fortheory and computation of macroscopic stability physics issues. However, Alcator C-Mod canaccess regions of parameter space that are not obtainable in other tokamak facilities, and as suchit provides a very useful benchmark for comparison against macroscopic stability theory andvarious codes. In some cases the Alcator C-Mod parameters are closer to those expected inburning plasma experiments. There is continuing interest within the PPPL macroscopic theoryand computation groups to work closely with the Alcator C-Mod scientific staff over the 2003-8program period for theory/experiment tests.

7.2. Model Development

7.2.1 Ideal MHD

Ideal MHD theory and computations are now quite mature. The PEST I/II codes have been thecommunity standard for evaluating linear ideal MHD stability in axisymmetric devices for someyears. We continue to support these codes, to improve them, to support the use of these codes tobenchmark and validate newer codes, and to apply them to new applications. Applicationsinclude continuing efforts to understand and optimize the current and pressure limits of toroidaldevices, and their sensitivity to plasma shaping and profiles.

- 34 -

An active emphasis area for tokamaks is to better understand the role of ideal MHD in thetransport barrier in reversed shear discharges, and the role of current on the open field lines (halocurrents) in providing stability to diverted discharges, and the role of moderate-n ideal MHDModes in ELMs.

7.2.2 Two-fluid Extended-MHD

Two-fluid Extended-MHD models utilize fluid-like equations for the electrons and the ionswhere the difference in their relative velocity and the anisotropy of the pressure tensor areaccounted for.

There is not presently a universally agreed-upon unique form for the two-fluid equations in afusion plasma. The theoretical challenge is to define the ion and electron stress tensors and heatfluxes in a way to include finite (ion) Larmor radius (FLR) effects (which lead to gyroviscosity)and the effect of long collision lengths parallel to the magnetic field, including magnetic particletrapping (important in neoclassical effects). Defining appropriate closure relations and testingtheir regions of validity is an important and long-term activity.

The M3D code now has the capability for including the ion-gyroviscous contributions to theion stress tensor, and for taking into account the difference in the ion and electron fluid velocities.These extensions account for several important physical effects, including the “omega-star”stabilization of ideal and resistive MHD modes, and the Hall term in Ohm’s law, which cangreatly speed-up magnetic reconnection. This capability has enabled a whole new class ofsimulations where we should be able to explain many new experimental phenomena that cannotbe explained by resistive MHD alone. For example, without these terms, resistive ballooningmodes are seen to go unstable in the simulations where they are not observed in the experiments.These terms are also essential to get the trigger and the crash time for the sawtooth. We note herethat an in-depth study of these important phenomena will take many years because of the need toperform many cross-checks, including analytic, cross-code, and experimental comparisons.Numerical resolution and stability requirements also need to be clarified.

• (2004/7) Simulate the effect of the two-fluid terms on tokamak sawteeth andaxisymmetric reconnection events (current hole). Benchmark with experimental resultsas available.

7.2.3 Kinetic Extended-MHD

To model the nonlinear interaction of ions with MHD waves and to include large gyro-orbitand neoclassical effects more self-consistently, we have developed several hybrid methods, whereeither an energetic ion component or the whole ion population are modeled using the gyrokineticor drift kinetic equations. In the methods implemented so far, the ion fluid velocity is calculatedby solving the momentum equation, and the calculated ion fluid velocity is used in the Ohm’slaw, assuming quasineutrality. The ion pressure tensor is taken to be in the CLG form and thegyroviscous part of the stress tensor is calculated from the particles. Hot particle populations, forexample fusion-produced α -particles or injected neutral beam ions, can be simulated bycombining a gyrokinetic hot particle population with a fluid model for the background

We have linear codes that calculate the self-consistent interaction of energetic-ion populationswith MHD modes: NOVA-K (and NOVA-KN is the nonperturbative version) for modes withlow to medium toroidal mode numbers (n), and HINST for high-n modes. We have now

- 35 -

completed the project of incorporating a high-energy particle component in the massively parallelunstructured mesh nonlinear MHD code M3D. This represents a unique capability for studyingthe nonlinear development of energetic particle modes in shaped geometry, of arbitrary aspectratio fusion devices. It is thus applicable to interpret results from NSTX and present day tokamakand stellarator experiments, and is also a unique tool for assessing the effects of energetic particlemodes in a burning plasma. New exploratory applications of this code are being carried out evenas it is being benchmarked against theory, the linear-codes, and other non-linear codes.

• (2005/6) Use NOVA codes to predict the nonlinear mode saturation as well as the alphaparticle induced transport in burning plasmas.

• (2004/6) Model experiments in CMOD with NOVA and HINST codes for betterunderstanding the TAE and energetic particle resonant mode in conditions of the multiplemode excitation.

• (2004/5) Extend the M3D capability to allow kinetic representation of majority ions in anonlinear calculation. Benchmark with experimental results as available

• (2004/8) Simulate the effect of the kinetic terms on tokamak sawteeth with M3D.Benchmark with experimental results as available.

7.3 Beta limiting MHD modes

7.3.1 Sawtooth Phenomena

The sawtooth instability, present when the current in the tokamak peaks sufficiently so that thecentral safety-factor is below the stability criteria, is a concern for the next generation ofinductively driven burning plasma experiments: including ITER, FIRE, and IGNITOR. Thesemodes introduce an intrinsic non-axisymmetric nonlinearity into the device that can interact withother modes to degrade confinement, shed energetic particles, or induce a disruption. Previouswork on the onset conditions for the sawtooth instability needs to be revived and extended to theburning plasma regime. Excellent agreement has been obtained between a theoreticallymotivated criteria and TFTR data for the occurrence of sawtoothing. The relevant physicaleffects need to be incorporated into the 3D Extended-MHD code to develop a fully predictivemodel for when sawteeth occur, what their period is, and how they interact with other modes.There is also interest in modifying the linear stability codes to more effectively deal with the q0<1problem.

• (2004/5) Perform a systematic study of the sawtooth in the CDX-U tokamak usingrealistic values of parameters.

• (2006/8) Compute sawtooth behavior in CMOD plasmas with strong RF heating andcompare with experimental results as available.

• (2008) Study n=1 sawtooth in a burning plasma

7.3.2 Neoclassical Tearing Modes

The Neoclassical Tearing Mode (NTM) is observed to set the pressure limits in many longpulse discharges. One of the major thrusts of the non-linear MHD effort is to include sufficientphysics to simulate the NTM and to differentiate between the competing mechanisms, for whichan anisotropic thermal conduction model and a neoclassical closure for the ion and electronviscous stress tensors are essential. The goal of this work is to extend the scaling of the NTM

- 36 -

threshold to large devices and investigate relative changes in the importance of the differentthreshold mechanisms.

Another issue is whether the plasma can generate a seed island of sufficient width to exceed theNTM threshold. A wide range of mechanisms have been observed as precursors to NTMs,including the sudden onset of the internal kink, coupling to magnetic field errors, coupling to themagnetic perturbation of an Edge Localized Mode (ELM), and transition from resistive tearing toneoclassical tearing. Another issue is to better explain the cause and effects of island rotation.

The M3D two-fluid model should contain the relevant physics needed to realistically simulatethe growth of this mode, but the problem is challenging because of the very slow growth rates,necessitating long running times. The PIES code provides a complementary approach to NTMs.Because the mode grows on a resistive time scale, radial force balance is maintained, and thedynamical nonlinear evolution is described by Faraday’s law, a three-dimensional equilibriumcode with the correct constraints can be used to calculate the evolution. In collaboration with thestellarator group at NIFS in Japan, a three-dimensional bootstrap code has been coupled to PIES,allowing the calculation of the perturbed bootstrap current that drives the NTM. The predictionsof the bootstrap code have been benchmarked against Monte Carlo calculations with the ORBITcode, and this has led to an improvement in the bootstrap model near resonant surfaces. Atpresent, PIES maintains a flat current profile in the islands, appropriate to saturated modes. Afuture upgrade of the code will introduce gridding in the island interiors, allowing the code tofollow the island evolution. The M3D and PIES efforts are complimented by several otheractivities in this area. One is to evaluate the semi-analytic formulas of Hegna, et al, using thePEST-III code to calculate delta-prime.

• (2006/8) Develop a predictive model of neoclassical tearing modes in tokamaks.

7.3.3 Energetic Particle Modes

We plan to further investigate resonant destabilization of discrete and continuum Alfvenmodes by energetic particles such as fusion alphas or beam-injected ions. This effort includestoroidal Alfven eigenmodes (TAE), "fishbones", high beta modes (BAE) and energetic particlemodes. Current emphasis is on including more realistic models that take into account realisticgeometry, non-Maxwellian background distribution functions, and self-consistent mode structureso that the energetic particles are included non-perturbatively. Our interest is in bothconventional and reversed-shear plasmas. Long term goal is to provide quantitative reliableprediction of the amplitude saturation of collective modes and consequences for ITER, such asalpha particle transport and its effect on plasma burning.

• (2004/5) Carry out a systematic comparison of energetic particle effects in M3D withother linear and non-linear codes, as available.

• (2004/5) Perform a study of nonlinear effects of alpha particle modes in burning plasmaincluding excitation of the TAE and fishbone modes and stabilization of the sawtooth

7.4 Edge MHD Stability and the behavior of ELMs

- 37 -

The long term goal of this work is to understand the origin and dynamics of the edge localizedmodes, to make contact with the theory to the experimental characterization of (Type I, II, III andquiescent), and to understand under what conditions a burning plasma can access small ELMregimes.

7.4.1 Linear Analysis

Detailed comparison of the results of linear ideal MHD analysis and experimental results showa good correlation between the stability criteria for intermediate-n ideal peeling/ballooning MHDmode stability and Edge Localized Mode (ELM) activity in experiments. These calculations canprovide indications of what the critical gradients are, as a function of edge current density, foredge-localized MHD modes, and what the height of the H-mode pedestal is. What remains to bedetermined is the width of the region over which the plasma pressure profile can assume thecritical value, and the nature of the trigger mechanism for the ELM.

7.4.2 Nonlinear Physics of ELMs & Evolution of Free Boundary Modes

The M3D code is being extended to allow for real free-boundary modes, where a vacuumregion surrounds the plasma, which is in turn surrounded by a conducting wall. This requiressubstantial code development and optimization. The primary application of this extension will beto ELMs, although it will also be useful in the calculation of resistive wall modes (RWMs), aswell as for a more realistic description of primarily internal modes, and the onset of plasmadisruptions. The goal is to make closer contact with the ELM cycles observed in experiments,and to identify under what conditions the different ELM types will occur: i.e., Type-I, Type-II,etc.

• (2006/7) Begin the nonlinear study of free boundary modes in tokamaks and stellarators.

7.5 Prediction of the Cause and Effect of Disruptions

For tokamaks, the highest performance discharges are often terminated by a major disruption.This event causes rapid loss of thermal energy during the thermal quench phase and then loss ofcurrent during the current quench phase. The current quench phase also produces large electricfields that can accelerate substantial populations of electrons to the runaway regime by way of theavalanche process. In reactor scale devices, structural, divertor tile, and first wall damage causedby disruptions is a major concern.

7.5.1 Physics of the Disruption

For the tokamak to succeed as the embodiment of a Fusion Power Plant, we will need a muchbetter understanding of the mechanisms that lead to major disruptions. MHD simulations havebeen used to identify the mechanism of a particular type of high-ß disruption in the TokamakFusion Test Reactor (TFTR): a localized moderate-n ballooning mode nonlinearly destabilized byan internal kink. Further studies with more resolution and improved physics models will be doneto produce accurate criterion for such instabilities, which may shed light on ways to control oravoid them. Other disruption mechanisms need to be studied, such as the overlap of islands, andthe coupling of sawteeth modes with NTMs and ELMs. These studies will necessitate manydifferent nonlinear code runs with a credible physics model to build up a better understanding of

- 38 -

the different sequences of events that lead to disruptions, their statistics, and their dependence ontokamak operating parameters.

Another, related, thrust, is to understand the difference in disruption mechanisms betweentokamaks, current-carrying stellarators, and spherical tori. Of particular interest is to understandhow externally generated transform provides some level of disruption protection.

• (2004/8) Identify mechanisms for disruptions in tokamaks and operational regimes thatare free from these mechanisms.

7.5.2 Disruption Forces

We have developed a detailed axisymmetric model of disrupting tokamaks with the TSC code.This included a model of the thermal quench, current quench, halo currents, runaway electrongeneration, and the surrounding coils and structure. We have used this to study disruptionprevention and mitigation techniques in tokamaks. There is continuing need to apply these toolsto ITER in order to evaluate specific mitigation techniques, such as massive impurity injectionand massive gas injection by way of supersonic gas jets, extending previous “killer pellet”injection simulations.

There is also a need to calculate the vessel forces due to the non-axisymmetric nature of thedisrupting plasma. These non-axisymmetric halo currents have been identified by the ITPA as acritical item affecting the design of ITER. The vacuum and resistive wall additions to thenonlinear code M3D have enabled it to address these issues.

• (2004/2005) Calculate the range of expected non-axisymmetric halo currents in ITER.Benchmark these calculations with CMOD data as available.

7.6 Control Issues

7.6.1 Profile and Shape Control

The Tokamak Simulation Code (TSC) is a widely used tool for predicting the axisymmetricevolution of tokamaks and spherical torii. It solves for the transport-timescale evolution of theplasma parameters as well as the poloidal field coil currents and their associated control systemsand vessel currents. It was chosen by the ITER project as the standard code for projecting Volt-second requirements, plasma evolution, plasma shape control and several other functions. Weplan to keep developing TSC as the premier code of its type, and to continue to perform newcalibrations as available, and to use it to design new experiments and to help optimize existingexperiments as required. Recent applications of TSC include the study of tokamak startup bybootstrap overdrive, and a study of methods to prevent the production of runaway electrons indisrupting tokamaks. Future development will concentrate on incorporating better current-drivemodules, a better neutral beam package, and incorporation of a more modern graphics package.

7.6.2 The Physics of Pellet Fueling

- 39 -

Injecting small pellets of frozen hydrogen into a tokamak is known to be a viable method offueling. Early work by Parks, et al. gives a fairly accurate expression for the ablation rate forsuch pellets once they contact the high temperature plasma in the tokamak. However, it is knownfrom many experiments that the resulting density profile measured after the pellet has ablated isnot consistent with what one would infer by assuming the ablated material remained on the fluxsurfaces where the ablation occurred. The subsequent “anomalous” redistribution of mass isbelieved to be due to MHD processes.

This mass redistribution is most dramatic in experiments that compare density profilesresulting from “outside launch” and “inside launch”, referring to if the pellet is injected from theexterior, low field side of the torus or the interior, high field side. It has been clearlydemonstrated that pellets injected from the inside are more effective in fueling the center of theplasma. Near vertical launch is also an attractive option for next-generation experiments.

In an initial attempt to model this mass redistribution, the M3D code has been used to modelthe evolution of a localized “density blob” in a tokamak, and has verified that MHD effects causethe localized density perturbation to displace towards the low-field side of the plasma. However,this simulation did not have a pellet ablation model, did not follow the pellet trajectory, usedsingle-fluid resistive MHD, and the resolution was fairly coarse.

We are extending this modeling in several ways with a new Adaptive Mesh Refinement(AMR) MHD code AMRMHD, developed in conjunction with an applied math group atLawrence Berkeley Laboratory (LBL). The model being developed has two-fluid physics, theParks pellet ablation model, anisotropic heat conduction, and adaptive zoning to allow accuratemodeling of the expected range of space scales.

• (2004/2005) Develop a realistic 3D model of pellet injection into a CDX sized plasma• (2006/2007) Extend this model, benchmarking as available on CMOD

7.6.3 Internal Mode Control

It has been demonstrated experimentally that both the sawtooth and the NTM can be controlledby the application of RF heating and current drive. Our goal is to develop theoretical models ofthis stabilization and to use these models and simulations to interpret and extend the region ofapplicability of the experiments.

It has been demonstrated on JET and other tokamaks that application of ICRH near the q=1resonant surface can delay or completely stabilize the sawtooth. We propose to study thisphenomena from a fundamental level. There are now RF absorption codes that can accuratelycalculate the absorption of RF waves in fully 3D geometry and the resulting modification of thedistribution function. We plan to couple these with the nonlinear 3D Hybrid/MHD code to assessthe relative importance of current drive, plasma pressure modification, and energetic particles incontrolling this mode. This modeling should be of direct benefit to ITER, as they would like tohave the flexibility to control the sawtooth if it becomes necessary.

Where NTMs cannot be avoided, active stabilization is required. Both lower hybrid wave andelectron-cyclotron emission current drive have been proposed and tested experimentally asmechanisms for stabilizing tearing modes through the generation of current and heat in thevicinity of an unstable island. We propose implementing existing fluid-like models of RF drivencurrent in M3D to investigate the feedback stabilization of NTMs.

- 40 -

• (2006/7) Simulate RF stabilization of a MHD mode through integrated modeling using acoupled RF and 3D MHD code

7.7 Appendix: Macrostability Codes Supported by PPPLTheory Department

Free Boundary Evolving Axisymmetric Equilibrium

TSC is the Tokamak Simulation Code developed at PPPL and used extensively at Princetonand throughout the world. It can model the evolution of a free-boundary axisymmetrictokamak plasma on several different time scales. The plasma equilibrium and fieldevolution equations are solved on a two-dimensional Cartesian grid, while the surface-averaged transport equations for the pressures and densities are solved in magnetic fluxcoordinates. An arbitrary transport model can be used, but the Coppi-Tang model is usedmost frequently. Neoclassical-resistivity, bootstrap-current, auxiliary-heating, current-drive, alpha-heating, radiation, pellet-injection, sawtooth, and ballooning-mode transportmodels are all included. As an option, circuit equations are solved for all the poloidal fieldcoil systems with the effects of induced currents in passive conductors included. Realisticfeedback systems can be defined to control the time evolution of the plasma current,position, and shape. Required voltages for each coil system can be output as part of acalculation. Vertical stability and control can be studied, and a disrupting plasma can bemodeled. TSC can also be run in a "data comparison" mode, in which it reads speciallyprepared data files for the PBX-M, TFTR, or DIII-D experiments. For each of these, aspecial postprocessor is available to directly compare TSC predictions with both magneticsand kinetics data for particular shots from these experiments. In all modes, TSC calculatesthe ballooning-mode stability criteria internally, and it also writes files that are read by thePEST code to calculate ideal and resistive stability for low-n mode.

Inverse Equilibrium

The JSOLVER code uses the iterative metric method to simultaneously solve for plasmaequilibrium and the magnetic flux coordinates. During each iteration, it solves an ordinarydifferential equation for the toroidal field function g so that the surface averaged parallelcurrent density take on a prescribed form. It has an automatic zone doubling feature toallow efficient generation of high accuracy equilibrium.

The ESC Code uses a second-order generalized Newton’s method to solve the nonlinearGrad-Shafranov equation. For solving the intermediate linearized Grad-Shafranov equation,it uses either the gridless (sweeping) technique (which guarantees a prescribed accuracy)or the Runge-Kutta (faster) method.

3D Nonlinear MHD Equilibrium

The PIES code solves for 3D MHD equilibria without making any assumptions about theform of the magnetic field. It is therefore capable of handling equilibria with islands andstochastic regions, and with a divertor. For tearing unstable plasmas, the code can solvedirectly for the nonlinearly saturated state of the island, and can be used to track the island

- 41 -

as plasma profiles are varied. The code has been used to study error field effects, lockedmodes, and tearing modes in tokamaks, stellarator equilibria, and 3D tokamak equilibriawith ripple.

3D Nonlinear Time-Dependent MHD

The code M3D is a nonlinear resistive compressible 3D MHD code. It is a fully toroidalcode with no expansion approximation being used. It is an initial value code, and is runanalogous to running an actual experiment. As in the experiment, the initial conditions,such as initial density, temperature, and q profiles are first determined. Then, again as in theexperiment, the boundary conditions, such as the voltage at the wall, determine the timeevolution of the plasma discharge. The code can be run either using a finite differencestructured mesh or a finite element unstructured mesh. The code uses a streamfunction/potential representation for the magnetic vector potential and velocity that hasbeen designed to minimize spectral pollution. The basic solution algorithm is quasi-implicitin that only certain terms in the fluid part that are the most time-step limiting are solved forimplicitly, with explicit differencing being used for the remaining terms. M3D, the two-fluid version M3D-T, and the particle/MHD hybrid version M3D-K are components ofM3D project.

The AMRMHD code was developed in conjunction with the APDEC center at LBL,making use of their CHAMBO general adaptive mesh refinement software package. Thecode solves the MHD equation using a second-order in time generalized 8-wave upwindscheme for solving the MHD equations. The code is fully parallel, and has the feature thatzones use timesteps proportional to their linear dimension, resulting in a very efficientmethod for resolving problems with multiple space scales.

Linear Ideal low-n

Linearized stability analysis code based on a minimization of the Lagrangian. Usesa finite element technique, PEST-I solves for all three vector components of thedisplacement and returns a physical growth-time. PEST-II is a linearized stabilityanalysis code. Variational code which minimizes a scalar form of the Euler-Lagrange Eqns. obtained, using a model kinetic energy.

Linear Ideal high-n

The BALLOON code integrates the 2nd order ODE to find ballooning stability limits andthe critical-n from a WKB approximation. The Mercier criterion is calculated and a choiceof K.E. norm is available. Output includes contours of local magnetic shear, curvature, anddw contributions from the various driving terms.

The CAMINO code constructs the s-a stability curves generalized to arbitrary 2-D tokamakequilibria. The curves from all the flux surfaces generate a 3D stability ballooningboundary in (s, a, y). This can be used to tailor the plasma profiles to achieve the 2ndregion of stability or optimize the b for the 1st region.

Linear Non-Ideal low-n

- 42 -

PEST-III is an outgrowth of the PEST-II code that is used to calculate resistive instabilities.It uses singular finite element techniques to compute the jump in the logarithmic derivativeof global mode eigenfunctions across singular surfaces. This information gives the quantityknown as delta-prime, which determines the stability with respect to resistive instabilities.This delta-prime also enters into the theory for the evolution of the neo-classical tearingmode.

Linear low-n MHD + Particles

The NOVA-K code (and its non-perturbative version NOVA-KN) computes stability ofglobal MHD and non-MHD modes in the presence of energetic particles such as NBI andICRF heated particles and alpha particles for tokamaks with noncircular flux surfaces. TheNOVA-K code makes use of a kinetic-MHD formalism that includes kinetic effect throughparticle pressure in the momentum equation. The plasma pressure is calculated from theparticle distribution governed by the gyrokinetic equation including finite orbit width andLarmor radius effects. The NOVA-KN code includes particle perturbed pressure into theeigenmode equation and solves it iteratively for the mode structure and eigenfrequency.

Linear high-n MHD + Particles

The HINST code computes space structure and stability of high-n modes such as TAE andballooning modes in the presence of energetic particles such as NBI and ICRF heatedparticles and alpha particles for high beta tokamaks with noncircular flux surfaces. TheHINST code makes use of the Ballooning in poloidal and Fourier in radial directionsformalism to determine the global mode structure and stability. It can be used in two formsto provide the local (1D in ballooning variable) and the global solutions (2D in ballooningand radial variables). It is non-perturbative and includes such effects as particle finite orbitwidth and Larmor radius.

Vacuum and Active Feedback

In the VACUUM code, the magnetic scalar potential is solved from Laplace's equationusing a collocation-Green's function method to calculate the vacuum contribution to dw forthree topologically distinct wall shapes - toroidal, spherical, and segmented. It calculatesthe vacuum response in either a Fourier or a Finite representation in the poloidal angle. Italso reads the output from stability codes and calculates the eddy current pattern on theshells and also simulates the Mirnov loop readings. A thin resistive shell is now an optionand the coding for feedback simulations is being added.

8. Facility

PPPL has designed and fabricated technologically challenging hardware additions to theC-Mod facility where needed to allow extension of the plasma parameter space or diagnosticcapability.

- 43 -


8.1.1. Rework of 4 ICRF transmitters

PPPL RF engineers provided active participation in the rework of C-Mod’s ICRFtransmitters for enhanced reliability and performance. Circuit improvements and sparecomponents were brought from the PPPL ICRF system on TFTR, and our engineers spent weeksworking with their C-Mod counterparts to perform modifications and final checkout.

8.1.2. Fabrication, installation, upgrading of 4-strap ICRF antenna

Bringing our experience from experiments performed on the PLT and TFTR tokamaks,the ICRF group at PPPL designed and fabricated a new 4-strap antenna intended for high-powerICRF heating and current drive on C-Mod. Available port and interior space resulted in anextremely compact, high power density design. Active participation in its operation and theretrofits performed as the result of the detection of operating deficiencies have helped to raise itspower level above 3 MW so far, with no deleterious effects on the plasma.

- 44 -

8.1.3. Ongoing ICRF engineering support

The PPPL RF engineering group provides active support to the C-Mod ICRF (and LH)experiments through close communication with C-Mod engineers, resulting in exchange of ideasand experience, and hands-on participation in hardware modification and transmitter retuning.

8.1.4. Completion of LHCD launcher

- 45 -

Again bringing experience from PPPL experiments, this time LH and LHCD studies onPBX, our RF group designed and fabricated a Lower Hybrid launcher to provide ~2 MW ofdirected wave power for C-Mod experiments. This launcher will be used in conjunction with C-Mod’s 4.6 GHz RF power system, and will be utilized to provide the off-axis current drive crucialto C-Mod’s Advanced Tokamak program.


8.2.1. LHCD launcher #1 installation and commissioning

The commissioning of the new LH system essentially consists in confirming the accuracy ofthe phasing system and determining the best radial position of the front end of the launcher forbest coupling. All this is obtained by monitoring the reflection coefficient in each waveguide andthe total value for various phasings.

- 46 -

These tests will be initially performed at low/medium power but eventually tests at high powerand long pulse will be needed to determine if ponderomotive effects, caused by the waves,modify the plasma in front of the launcher.

In the initial configuration, the fourth arm of the 3 dB splitter will have a short. If the phase ofthe reflected wave in the two connected waveguides is in phase, all the reflected power will returntoward the generator and absorbed. On the contrary, a phase difference will cause some power toenter the fourth arm of the splitter and reflected toward the plasma, possibly changing the phaseof the forward wave.

For this eventuality the coupler is equipped with directional probes which will monitor thepower in the fourth arm. Some experimental time will be devoted to these tests to determine if theshort is acceptable.

If not and a load is needed it has to be a water load as stone loads are incapable to dissipate thepotential power in the fourth arm (remember that the waveguide is very narrow). PPPL alreadytested some possible configurations of this water load and the program can be restarted any time.

8.2.2 Fabrication of LHCD launcher #2

The addition of a second LH launcher will increase the useful power into the plasma byenhancing reliability through reduction in RF power density inside the launcher, as well asallowing additional experimental flexibility. Fabricating and operating experience gained fromthe first launcher will be fed into possible updates to the design of a second launcher, which willthen be fabricated by PPPL and delivered to C-Mod.

8.2.3 Participate in 4-strap ICRF antenna #2

The addition of a second LH launcher will require the freeing up of a C-Mod vacuumport presently in use for one of the two 2-strap ICRF antennas. Replacement of both 2-strapantennas by an additional 4-strap antenna would make the needed port space available, retain thepresent level of ICRF heating power, and double the ICRF power that can be launched as adirected wave. The PPPL ICRF group will participate fully in the design, checkout and operationof this antenna.

8.2.4 Participate in tunable cavities for ICRF transmitters 1 and 2

The PPPL RF engineering group will continue to support the design, fabrication andcheckout of tunable RF cavities which will replace the present fixed-frequency cavities installedon C-Mod’s ICRF transmitters #1 and #2. This modification will provide considerably enhancedexperimental flexibility.

- 47 -

9. Budget

This budget is based on the latest guidance from DoE.

• FY 2003 Baseline budget: Science + Ops + LH $2,771KIncremental: $ 0

• FY 2004 Baseline budget: $2,072KIncremental: Science + Ops $ 450K Faraday screen

Lower Hybrid $ 630K LHCD#2

• FY 2005 Baseline budget: $2,072KIncremental: Science + Ops $ 450K RF cavities

Lower Hybrid $ 630K LHCD#2

- 48 -

10. Manpower

FTE estimates, based on latest funding guidance:

Schilling 1.00 RF, managementScott 1.00 MSE diagnosticBernabei 0.70 LHMikkelsen 0.60 simulations and transport modelingZweben 0.35 GPI turbulence imaging diagnosticWilson 0.30 RFRedi 0.30 ITB modelingHosea 0.25 RFKramer 0.25 reflectometer diagnostic

Engineering and technical support: 0.7 FTE

Additional “free” research support from PPPL Theory, SciDAC, FIRE, NSTX (this runcampaign).

Priority has been given to the Lower Hybrid system startup and experiments.

Incremental budgets for Science + Ops can help by allowing us to increase our researchparticipation on all fronts.

- 49 -

11. Contributors

The following members of the PPPL C-Mod collaboration and other PPPL colleagues havecontributed to this Plan:

Stefano Bernabei (Advanced Tokamak and RF Physics – LHCD)

Joel Hosea (RF Physics and Wave-particle Interactions)

Steve Jardin (Theory and Computation of Macroscopic Stability)

Gerrit Kramer (Reflectometry)

Dale Meade (Burning Plasma Experiments)

David Mikkelsen (Transport)

Cynthia Kieras-Phillips (RF Physics and Particle-wave Interactions - Modeling)

Martha Redi (Transport – ITB Modeling)

Gerd Schilling (Lead)

Steve Scott (Advanced Tokamak - MSE Diagnostic, and Editing)

Daren Stotler (Edge Neutrals Modeling)

Randy Wilson (RF Physics and Wave-particle Interactions)

Stewart Zweben (GPI Edge Turbulence Visualization)

Date post:	17-Oct-2019
Category:	Documents
Upload:	others
View:	4 times
Download:	0 times

Princeton Plasma Physics Laboratory And Alcator C-Mod ... · Princeton Plasma Physics Laboratory...

Documents