Latest Results: Task IV, FED paper, etc. Investigators –IFS: M. Kotschenreuther, P. Valanju, L. Zheng –LNLL: T. Rognlien.

Latest Results:Task IV, FED paper, etc.

Investigators

– IFS: M. Kotschenreuther, P. Valanju, L. Zheng

– LNLL: T. Rognlien

1. Effect of fast flowing LM wall on plasma MHD modes

2. Summary of submitted FED paper: Reactor Implications of convective SOL transport

• In collaboration with Tom Rognlien; use 2-D simulations (UEDGE) & IFS neutral code NUT

3. ICC funding grant awarded for APEX developed concept: “field line extraction divertor”

– Develop coil designs for reactors and retrofit experimental tests

4. The importance of an APEX-like program in the future

Four Areas Considered

• Plasma MHD code has been completed (AEGIS)• Has been benchmarked against GATO for circular cross

section plasmas• Code works for non-circular plasmas, but coupling to the

vacuum/resistive wall has only been completed for– circular cross section– large aspect ratio plasmas

• Effect of a fast flowing wall on pressure driven kink modes in a high beta plasma (circular) presented here

• Even though circular, these results are a large improvement in realism over previous analysis

Effect of a Fast Flowing LM Wall on Plasma MHD Modes (RWMs)

• Previous analysis by Zakharov, Kotschenreuther: – Low betaN, not high betaN like a reactor– Different instability drive mechanism from reactor cases:

plasma current rather than plasma pressure– Different plasma current profile from reactor cases– Most analysis also circular– Only qualitatively similar to reactor cases

• New Results:• Realistic high bootstrap fraction cases considered• Reactor relevant high betaN from 3 to 6 considered

– ARIES RS has betaN = 4.8 – Significant plasma profile optimization

• But still only circular

Recall Previous Results for RWMs

Results for RWM with Realistic Plasma Profiles, fast flowing LM

• Stabilization for betaN= 3 :– For 2 cm Li, flow > 20 m/s– For 2 cm Sn, flow ~ 50 m/s

• Flow velocity increases sharply with higher betaN

• For betaN = 4-5 (ARIES ) flow > 100 m/s (even for separation = 0)

• PRELIMINARY RESULT: realistic LM flow velocity will not stabilize desirable cases with high beta and high bootstrap fraction

• BUT: Elongated Plasmas might give better results

BetaN = 3.1 case

0

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40

60

80

100

0 1 2 3 4 5 6Separation Li from plasma (cm)

Li

Flo

w V

elo

city

m/s

n=1

n=2

n=3

1. Higher Elongation (close fitting metal shell-fast flow not needed)2. High temperature edge boundary condition (even for low

elongation): using Li edge or extraction divertor

• We have examined case 2) here for circular plasmas, with high plasma edge temperature:

• RWMs are found to be very severe– cannot be stabilized even for Li flows > 100 m/s (2 cm

thickness)– Stabilization of RWM by plams flow (controlled loss) still a

possibility- code development to include plasma flow is mostly complete

• Code modifications to include high elongation are in progress

Two Routes to High Beta discussed in APEX

• Controlled loss: MHD stabilization by producing plasma rotation

– Plasma MHD stability from flow: verified in experiments

• Have modified MHD code to include rotation; almost complete

loss concept requires liquid surface for erosion control, but is more flexible than fast LMs:

– Works with flibe, slow moving LMs– May only require very limited wall coverage with a liquid – May only require a thin liquid wetted surface for

erosion/pumping, rather than a fast stream

• Controlled alpha loss solves multiple problems:– Control of thermal runaway in plasma AT modes with internal

transport barriers– Alphas hit wall with Mev energy so liquid pumping is much

more efficient

loss to Stabilize RWMs

EFFECTS of CONVECTIVE BLOB TRANSPORT on

REACTORS

• Subject of submitted FED paper: summary here• In collaboration with Tom Rognlien, Prashant

Valanju (IFS)

Our Physics Understanding of SOL Transport has Changed

• Pioneering experiments at C-Mod and DIII-D:

large convective transport of plasma blobs

• Theoretical investigations: blobs of plasma should

rapidly convect to the main chamber wall

• IFS-LLNL collaboration: investigation of potential

REACTOR effects of convection for the first time:

effects on solid walls (W) and flibe, SnLi, Sn

Serious Effects of Blob Convective Transport

• First Wall Erosion

– Concern for enhanced erosion mentioned in the literature, but heretofore not estimated for reactors

– Serious implications found here for W

– Chamber walls with low Z sputtering may be required for reactor feasibility (flibe, SnLi)

• Helium Pumping

– Not discussed in literature, but: far SOL transport disproportionately effects helium removal--work in progress

2-D Simulations using UEDGE • Present state of the art: use empirical diffusion coefficients

for reactor simulation• But convection appears essential in far SOL• We use empirically motivated convection model similar to

that on present experiments to estimate reactor effects

Distance to wall

Con

vect

ive

Vel

ocit

y m

/s

100

DIII-D Simulation (Pigarov)

DIII-D Simulation (Pigarov)

IFS Reactor Simulation

Several Convection Profiles Examined• Convection profile varied so that the simulated SOL density profile

for a reactor matches SOL profile characteristics found in experiments (literature for DIII-D, C-mod, ASDEX UG)

• Quantitative profile characteristics checked with experiments:1. Ratio of density in the SOL (d/a=0.04) to separatrix density2. Density scale length at d/a = 0.043. Convective flux at d/a = 0.04

• Four convection profiles tried:A) 10 m/s to 100 m/s (found to give BEST FIT to data)B) 10 m/s to 50 m/s (reduce convection near wall)C) 5 m/s to 100 m/s (reduce convection near plasma)D) 0 m/s to 0 m/s (no convection)

• Case A gives best match: reducing the convection (B and C) results

in a poorer match to characteristics 1,2 and 3

Why Wall Erosion Estimates Based on Models w/o Convection are Likely to be Low

• Standard SOL transport model: constant diffusion only• Probably underestimates plasma-chamber interaction by ~ 30

Withconv

Noconv

Density Ratio Comparison

• Find: a strong relationship between the density ratio in SOL and the Greenwald ratio

• For high Greenwald ratio (like reactors), density in the far SOL is high => STRONG WALL INTERACTION

• Case A is most consistent with experiments

• Case D without convection does not match experiments

0

0.1

0.2

0.3

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0 0.2 0.4 0.6 0.8 1

Normalized Density

dens

ity ra

tio

L-modeexpts

H-modeexptsUEDGEreactor B

C

D

A

Density Ratio = nsep/nd/a=0.04

Wall Flux Comparison

• Experimentally estimated plasma flux to the wall has a large scatter

• Flux trends are described by the expression of Labombard: L=1021(ne/1020)2

• Case A most similar to data

• Case D without convection: under-estimates flux by nearly two orders of magnitude 0.01

0.1

1

10

0 0.2 0.4 0.6 0.8 1

Normalized Density

Flux

/

L

L-mode exptsH-mode exptsUEDGE reactor

AB

C

D

SOL Density Scale Length Comparison

• SOL density decay is slow in experiments, and tends to be flatter at higher density

• Case A most consistent• Case D without

convection: density profile does not match experiments 0.1

1

10

100

0 0.5 1

Normalized Density

a / le

ngth

L-mode expts H-mode exptsUEDGE reactor

D

C

B

A

Kinetic Neutral Code NUT evaluates the hot CX neutral flux to the wall

• Wall erosion is dominated by hot CX neutral flux to the wall

• The fluid treatment in UEDGE cannot evaluate this

• Thus: the kinetic neutral code NUT is used

– NUT: benchmarked against experimental data on TEXT & C-MOD

• The plasma profiles and neutral source found by UEDGE are input into

NUT

• NUT computed the energy distribution of CX neutrals back the the wall

• The sputtering coefficient is integrated over the CX neutral

distribution to obtain the wall sputtering

With Realistic Convection: Strong First Wall Erosion

• Most sputtering-resistant material: Tungsten

• Without convection: 0.17 mm/yr

• With convection: 0.61 mm/yr

– estimate small prompt re-deposition

• My question for engineers: is this

structural erosion serious?

Other Consequences of High Tungsten Erosion

• Large dust generation

– ITER: ~10% of sputtered material forms micron dust

– With convection: ~ 340 kg/yr dust after 2 years

– LOVA dosage marginally exceeds no evacuation limit (even

with 99% filter, -adapting analysis of Merril et. al.)

• Plasma Impurities

– C-Mod has high-Z wall: H-mode screening factors 1-10%

– ASDEX ~ 1% UEDGE ~ 10%

– This range of screening can have unacceptable consequences:

– H-mode ignition precluded due to radiation for ~ 0.5

1.0 % penetration

Implications for Liquids• Flibe, SnLi, Sn considered

– Obviously dust, structural erosion are not issues• For low Z PFCs (Flibe, SnLi):

– Plasma: much more tolerant of Low Z impurity • Acceptable screening factor found to be ~ 5 % • Recall experimental values are ~ 1 - 10 %

– Sn walls• High Z: acceptable concentration slightly higher than

W, but sputtering also slightly higher• Required screening factor ~ same as W: very

worrisome

Conclusions• Better physics understanding of SOL transport

required: could be show-stopper– Plasma-wall interaction: structural erosion, dust– Impurity transport and core plasma contamination

• Alternative concepts to W wall may be required:– Low-Z liquid walls

• Low Z => acceptable plasma impurity level • Continually replenish wall => no structural erosion, dust

– Extraction divertor • Low density SOL operation to minimize SOL convection

Beyond FED paper:• Erosion near edges of protrusions and cavities

– Near corners, projections: blobs will dump plasma much more strongly

– Recycled neutral source many times higher => Local erosion rates several times higher (?)

• Wall next to ICRF antennas, and antenna itself

• Wall near blanket test modules which are inset by ~cms

• Assuming ITER edge is the same as previous calculation– 10,000 shots, 400 sec => ~ 3 mm Be erosion for flat wall

– Several times higher (?) near protrusions and cavities

• Experiments find that an ELM causes a massive plasma blob with effects similar to the continuous blobs here

Blob Implications for ITER• PFC erosion appears to have been underestimated for ITER and

reactors

• More intensive investigations of plasma –wall interactions required– To set a relevant ITER research program

– To ensure adequate ITER operation

• Beyond research, new design solutions for ITER may be required – W wall gives acceptable erosion for low duty cycle ITER

– BUT WHAT ABOUT PLASMA IMPURITIES?

• New reactor relevant design solutions could be needed:– Low-Z liquid wetted wall near edges?

– Low Z wetted projection into SOL to intercept blobs before they reach the main chamber?

– Beyond ITER – field line extraction divertor to run in SOL regime with low blob transport?

Future Work• We are considering 3-D NUT calculations to examine hot CX caused

erosion near edges (with model SOL profiles)

– Don’t know how to interface with ITER design, etc

• IFS will attempt to develop better models of SOL turbulence

– More physics based models of blob erosion

– Better physics basis for extrapolation to reactors

– Experiments give strong indications that a low density, high T edge

would have much less blob transport

– Solution to blob transport problems: Li walls, extraction divertor ??

• Investigate convection effects on He exhaust using UEDGE

– He path from the core to exhaust is mainly in the far SOL

– How seriously is He exhaust degraded by strong SOL convection?

Have Received ICC Grant to Design Field Line Extraction Divertor

• Use design/optimization tools developed for NCSX Compact Stellarator

– Highly sophisticated algorithms optimize coils for 3-D magnetic fields

– Optimizations can include an arbitrary number of engineering and physics properties

– After modification, tools will enable optimal coil designs to be developed for extraction divertors in many scenarios:

• Reactor scenarios

• Retro-fits of existing devices to test the concept (NSTX, Pegasus, CDX-U, others?)

• Why these tools?

– Battle-tested: greatly simplified NCSX coils

– Range from initial “filament” to full engineering design

– NCSX tools developed after much effort- allowing extended divertor design to be done with relatively small incremental cost to DOE

External Divertor Coil Design/Optimization

• Similarity with NCSX Compact Stellarator coil design:

– 3-D magnetic field requires optimized coils

– Manual optimization has proved principle, but gives complex design

– Design parameter space multi-dimensional

– Both engineering and physics target functionals are required

– Target functionals can be quantified

– Targets are non-linear functions of input parameters

– Possibility of many isolated optima

– Sophisticated optimization methods are needed

Optimization Targets• Engineering targets (new and already in NCSX tools):

– Clearance of extracted field lines from coils

– Currents, (heating for Cu coils)

– Stresses

– Geometry: Curvature, Torsion, manufacturability

– Effects of finite cross-section (filament -> real coils)

– Feeders: geometry, stresses, and ripples

– Ease of assembly and replacement

– Neutron fluxes on coils

• Physics targets:

– Ripple in plasma

– Plasma recycling

– Plasma shaping: elongation, triangularity

Why Liquid Surface Investigations Should be Continued

• In the past, APEX justified liquid surface work for long term objectives– A cheaper fusion reactor

• But liquid surfaces may be needed even for lowest order feasibility– Erosion of divertor plates due to ELMS

– Structural erosion of main chamber- especially near edges

– Radiation collapse of the core plasma for W walls (reactor)

– Dust generation and environmental/regulatory acceptability

• Conventional strategy is to find a plasma operating regime which is compatible with all the constraints of solid walls, as well as

– High plasma radiation fractions

– Low disruptivity (antithetical with high radiation fractions)

– Good confinement and high beta (antithetical with low temperature edge)

• We should admit that a viable plasma operating regime may not exist, and that the engineering constraints may need to be alleviated through novel approaches- using concepts developed in APEX

The Value of an APEX-like Program in the Present Environment

• Perhaps instead of positioning APEX to develop futuristic reactors, it should be positioned as providing novel engineering solutions to zero’th order feasibility issues

• The value in APEX is in providing options to avoid technological/physics dead ends which may be too difficult to solve

• This value could be substantial even for next generation burning plasma devices

• To make this value more apparent, it might be useful to write an analysis – Of conventional divertor/first wall options and their engineering

risks

– How those risks could lead to much wasted time/funds

– How broader technological options examined by APEX could provide crucial alternatives

Why APEX-like Investigations Should be Continued – ITER Relevance?

• Even for ITER, thin wetted liquid surfaces could greatly improve the chances for success by:– Eliminating erosion concerns from ELMS in the divertor– Eliminating erosion concerns on the main chamber first wall without

introducing the plasma contamination concerns of W• A wetted divertor application requires a thermo-hydraulic &

MHD analysis, the expertise for which resides in the chamber technology area

• A wetted section of the first: also chamber analysis• It appears to me that a compelling case can be made for liquid

surface work within the chamber technology area, even without the goal of a high neutron wall load– Liquid surfaces (perhaps thin wetted surfaces) for erosion/impurity

control may be needed even for basic feasibility• ITER may reveal serious limitations of a conventional divertor

– A “35 year demo” may require unconventional concepts considered in APEX (e.g. field line extraction divertor, liquid divertor)