DDivertor issues on next step devices,

and validating the Super-X divertor as a promising

solutionM. Kotschenreuther, S. Mahajan, P. Valanju,

Institute for Fusion Studies, University of Texas at Austin

J. Canik, R. Maingi Oak Ridge National Laboratory

B. LaBombardMIT

A. GarafaloGeneral Atomic

Work supported by US DOE-OFES Grant No. DE-FG02-04ER54742

and DOE-ICC Grant No. DE-FG02-04ER54754

Divertor: Limiting Even for ITER 2007 ITER Physics basis:

“The fusion gain in steady state maximizes at low density for

constant N. The limitation on reducing the density in next

generation tokamaks is set by the impact on the divertor.”

The divertor sets the upper bound on the available power density in

steady state mode

Also:

“It should be noted that presently developed advanced scenarios

have not yet provided fully integrated scenarios and several

issues remain to be solved, such as edge compatibility with the

divertor”

• For ITER (which has much lower power density than a reactor)

– Advanced scenarios are already at or beyond the expected limit

of standard divertor heat capacity

ITER-like divertor probably requires unacceptable core radiation in a reactor

• Core fRad ~ 85% qualitatively

different from ITER value

• Required confinement of net power probably too high*

– H89P ~ 4.2-5.2

• Burning plasma: Thermal instability*

– Strongly self heated plasma plus high

fRad probably leads to virulant thermal

instability

• Burning plasma: He build-up & Collapse*

– High core He generation, low core transport likely leads to Helium

collapse for fRad ~ 85%

• CTF’s and hybrids are in between ITER and a reactor

Reactors Core fRad to give same P/R as ITER

ITER (AT mode)

~50%

ARIES RS 83%

ARIES AT 78%

CREST 87%

Slim-CS 86%

EU-C 85%

EU-D 83%

* Kotschenreuther,Valanju and Mahajan, Phys. Plasma July 2007

NEED a Divertor solution for much higher upstream q||

• With the same core radiation fraction as ITER, and ~ 5 times greater P/R,

Reactor q|| is ~ five times greater

• Steady state fusion power optimizes at lower density for current drive and lower edge density for bootstrap

Need divertors which can handle higher q|| and a low n

• Also, ELMs/disruptions in a reactor would be – Several times larger than ITER– Hence, a divertor which can tolerate a much larger

ELM/disruption is highly desirable

Problems at high divertor power/ low density

1. Divertor plate heat flux

– Technological limits of ~ 10 MW/m2 , perhaps less at much higher neutron fluence than ITER

2. Helium pumping

– In simulations, degrades very rapidly with power and lower density

3. Plasma erosion of the plate/plasma impurities

– High plasma plate temperature/low density greatly increases sputtering/reduces prompt redeposition

4. Divertor survival of disruptions/ELMs

• Most Solutions of 1 help 4 but don’t improve (or even degrade) 2 and 3

Need to scope out how bad the integrated problem probably is at DEMO/CTF parameters

Need guidance on how to improve proposed solutions to get integrated solution for the devices we need to build

Research Plan

• Start with relatively inexpensive simulations of innovative divertors for expected conditions of DEMO and CTF (both ST and normal Aspect ratio)

– Examine integrated performance: Heat flux, helium exhaust, erosion, transients

– See where the various innovations stand on all these variables, where improvement is needed

– Examine: Standard divertor, highly tilted plate, Super X-divertor/ X-divertor, Snowflake divertor, liquid metal (Li, SnLi, etc.) and combinations

– Improve edge simulation models as time goes on (better plasma transport models, dust/foam impurity generation, etc.)

• Develop experimental tests of novel divertors on existing tokamaks (US and world)

• Diagnostic upgrades to measure all relevant quantities

• Compare experimental results with simulations

• Testing must emphasize solutions which will probably extrapolate to a reactor- benchmarked simulations will be critical in determining this

SXD: saves NHTX from heat flux menace according to simulation

• SOLPS 2-D calculations (Canik & Maingi)

• With SXD & 30 MW, peak heat flux < 10 MW/m2

• Not possible with standard divertor (peak stays at 30-40 MW/m2)

NHTX Standard Divertor NHTX SXD (Corsica Equilibrium)SOLPS SXDCalculation

Why the integrated divertor problem is harder (1): Erosion/Impurities• Temperature on plate (sputtering) increases rapidly

with higher q|| , prompt re-deposion decreases

(strongly) with lower density

Plate tilting/flux expansion DO NOT

reduce plate temperature or increase plate density

– Temperature at plate is determined by q||, NOT BY Qplate

– Most proposed divertor innovations reduce Qplate, BUT NOT q||

• SXD advantage: Only method of geometrically

reducing q|| at plate (by going to larger R)

– SXD: reduces Tplate by order of magnitude and increases nplate

in simulations

– Expect SXD to greatly reduce erosion and impurities

Why the integrated problem is harder (2): He exhaustITER simulations: with increasing PSOL at fixed nupstream

He exhaust gets worse even faster than heat flux

Qplate ~ P2.5 n-2.4 --- but core He density ~ P3.3 n-5.5

• Most strategies to increase wetted area (plate tilt, flux expansion) probably degrade He exhaust

– Recycled neutral He from plate must transport through the plasma without being ionized to reach the privated region and be pumped out

– Spreading out the wetted surface (at constant Rmajor) reduces He reaching private

region

Higher PSOL together with higher plate tilt/flux expansion is likely to be a

disaster

for He exhaust

• SXD advantages:

– Very large reduction in Tplate/increase in nplate

– For a given wetted area, reduces the distance He must go to get to the private region

– Divertor is closer to outside world, so geometry might allow much better pumping

Simulations are an inexpensive way to look into the future to see

which innovations will extrapolate to an integrated solution

Shortcomings found in simulations can motivate improvements in the concepts where they are weak, and consequent experimental tests

• Uncertain parameters (e.g. SOL width) can be varied within plausible range to indicate the robustness of various solutions

– Simulation models will be improved as other thrusts bear fruit

• Coupling to core simulations can show the degree of improvement in overall performance by enabling

– Lower edge density (higher current drive, bootstrap current=>

higher , E, etc.)

– Lower core radiation with improved confinement, helium exhaust, thermal stability, etc.

Simulations for combinations of ideas

• The divertor problem is so hard that combinations of innovations may be needed

Example: combine novel geometries with fast flowing Liquid Metals (LM) or capillary/pore LM

• Flux expansion can greatly reduce MHD drag effects for fast flowing LMs– SXD reduces drag several times further

• SXD gives much better isolation of the core from evaporated impurities (because of the very long throat)

Simulations can inexpensively examine the complicated physical processes in such combinations to find what works best

• Experimental tests of extrapolatable combinations

Experiments

• Obviously will benchmark, and be interpreted using, simulations

In addition, SOL physics may be clarified by finding how the SOL changes as the magnetic geometry changes– Many innovative divertors vary important parameters far from

a standard divertor (e.g. magnetic shear, line length, B and

q|| at strike point, recycling (Li), etc.)

Finding how the plasma responds could elucidate SOL physics, test theoretical models, etc.

• Diagnostics need to be improved for the integrated problem:– Heat flux, plate temperature/density, plasma radiation, etc.

– Possible He exhaust tests with outboard edge He fueling?

Potential Experimental Tests

NEAR TERM

• Reduced size plasmas on DIII-D with existing hardware

– Snowflake being considered for this year, SXD next

• Testing on MAST

– SXD physics design finalized, engineering design in progress

• Testing on SST

– SXD is planned-possible opportunity for collaboration with diagnostics?

• Tests of X-divertor (flux expansion part of SXD) on NSTX

LONGER TERM, more expensive upgrades

• DIII-D tests with baffling included for SXD

• NSTX with coil/baffling/vacuum vessel modifications for SXD

• NHTX/Vulcan

• CTF (ST-CTF, FDF) and possible future hybrid

• Pure fusion DEMO

Divertor challenge is more than plasma physics: SXD engineering

advantages/disadvantages• Divertor is closer to the outside world, so maintenance might be much easier

• Potentially much easier pumping for same reason

• Divertor plates have substantial neutron shielding (several times less dpa and much lower energy => less He generation)

Could substantially ease divertor material development

Divertor challenge is more than plasma physics: SXD engineering

advantages/disadvantages

• Disadvantage: vacuum chamber has a more complex shape– But calculations up to now show that the SXD DOES NOT

REQUIRE a larger TF coil radius

• Magnetic control requirements need to be refined– Preliminary analysis finds SXD is no harder to control

than standard divertor

– Magnetic geometry response to ELMs, disruptions, etc.

The magnitude of engineering advantages/disadvantages for innovative

divertors (including LMs) needs more thorough system assessment

Opportunity for US leadership

• Focus of other major new experiments in the world is on areas other than Divertor and PMI innovations

EVEN THOUGH THESE ARE THE AREAS WHICH ARE MOST SERIOUSLY IN NEED OF IMPROVEMENT TO MOVE FORWARD IN FUSION

• By emphasizing investigations in these areas, US could establish leadership in a critical domain of fusion research

The severity of the exhaust problem seems to be widely underestimated; simulations could help correct that deficit

• Simulations can demonstrate the need for these innovative ideas in which the US already leads the world (SXD, snowflake, Lithium/Liquid Metal, etc.)

Finally, undertaking concrete experiments to solve this very hard set of problems is bound to cement US leadership

Solution:Super X Divertor: evolution from X-

Divertor

XD/snoflake/tilt plate: expand flux SXD: maximally increase Rdiv

SXD greatly increases: wetted area, line length, margin from sheath

limited regime

SXD: Developed for both copper coil and superconducting

devices

Cu coil devices:

– ST-CTF (Peng) A= 1.5

– Normal A CTF (FDF) A = 3.4

– NHTX A = 1.8

– Other ST-CTFs: A = 1.8, 2.5

Superconducting reactor designs– ARIES-AT A = 4

– Slim- CS A = 2.6

SXD Examples (CORSICA equilibria)With 5% less Amp-m than SD

SXD Examples (CORSICA equilibria)

SXD fits in TF corners - no TF real estate issues

• For NHTX, FDF, and Reactors, SXD does not require larger TF coils

• SXD uses available space (in the corner of TF coils) which is

normally unused

CORSICA Equilibrium for NHTX-SXNHTX (PPPL/ORNL)

TF

SXD design space is smooth and large

SXD is insensitive to plasma changes possibly becausethe long SXD “leg” near the SXD coils is far from plasma current

Gains for SXD equilibria

1 degree plate tilt limit - limits wetted area

• Minimum angle between total B and

plate ~ 1 degree, (generally accepted

limit)

• Limits Aw gains for all flux expanders

– Tilting plate, XD, Snowflake …

• SOL width at midplane (so Asol) is

given

• So only knob left is increasing Rdiv

• Pleasant surprise: SXD can be made

with small changes in standard

poloidal coil positions and currents

(±5%) for a wide range of plasmas

(low-A, high-A..) HPDX - CORSICA Equilibrium

,

Avoiding the Sheath limited regime

• Use analytic criterion for sheath limited regime (modified from Stangeby):

Q||div / ne1.75L||

0.75 = Q|| up (Bdiv / Bup) / ne1.75L||

0.75

– Since flux tube area ~ 1/B, Q||div ~ Q|| up (Bdiv /

Bup)

– Q|| up is determined by upstream SOL width, and

factors independent of the divertor

By increasing line length and decreasing (Bdiv / Bup)

the SXD decreases the criterion by ~ 5 times

Neutron damage to divertor - critical issue

• Tungsten “armor” on a high thermal conductivity actively cooled substrate

– High conductivity substrates (Cu or C) severely deteriorate after only a few dpa

– Reactor walls receive ~ 30-100 dpa

• Only hypothetical high heat flux divertor materials might tolerate ~ 30-100 dpa

– High risk, many decades away with much material development effort

– Slow development would hamstring any high duty cycle DT device (CTF, DEMO)

• SXD: substantial shielding of divertor plates for future CTF, DEMO

Initial MCNP calculations- damage rate reduced by 10-20 times

With SXD, much ITER divertor technology

might by transferred to CTF, DEMO

Further SXD issues to be investigated

• Doesn’t inner divertor now become the problem?

– No. Almost all HPDs are designed as double-null (DN) devices

– In DN, heat flux on inner plate is order of magnitude less than

outer

– SXD does not increase heat flux on inner plate

– So unless SXD decreases heat flux by more than ~5, this is no

problem

• Won’t some instability increase cross-field transport?

– That would be good! It will further increase SXD capability.

– The SXD leg does not go very near the extra X-point, so ergodic

problems (hot spots etc.) are expected to be not bad

• What about pumping, He pumping, etc.

– SXD is better isolated from plasma, lower T, higher n, so higher

neutral pressure

• TF ripple at SXD plates? Solution: shape SXD plates

• Is stable detached operation possible with SXD? Need to

simulate.

Broad interest in implementing SXD• SXD is now the “SD” for high-power density low-A devices (e.g., CTF)

– SXD is now prominently featured in CTF presentations and actual engineering design

work

• PPPL-ORNL-IFS (SOLPS runs) working with SXD for NHTX

– Preliminary results are very encouraging: SXD may be necessary and sufficient for

NHTX

– SXD is now integral part of PPPL plans for NHTX

– Actual design to test SXD (at max Rdiv allowed by vessel/ports) is now underway in

NSTX upgrade

• MAST now has a working group to make SXD a part of MAST upgrade

– A MAST-IFS collaboration has begun

• IPR (India) has formed a group to design SXD on SST (1000 sec. SC tokamak)

– Interestingly, SST heating power is currently limited by divertor - SXD will open

new regimes for SST

• PRC-IFS collaboration now exploring SXD for HL-2A

– Preliminary designs being generated for the device

• Very positive feedback from many at MIT meeting Feb. 08

– Common statement: SXD should be tested on some device as soon as possible

A possible HPDX with SXD

Two possible missions:

• Component Test Facility: short-term,

useful, necessary fusion task (parallel to

ITER) using conservative physics and

technology, or

• Inexpensively develop advanced physics

modes for a pure fusion reactor in an

integrated fusion environment with high

heat fluxes

A 100 MWfusion HPDX (CTF) with SXD

• Goal: get high power density but stay conservatively well within demonstrated physics and technology limits so it can be built soon (at least in principle)

• For small size and coil mass, and easy maintenance:

use Cu coil ST:

A = 1.8 R = 1.35m = 3

7 T on coil Neutron wall load ~ 0.93 MW/m2

• enough room for neutron shield (~0.1 m): centerpost lifetime 1-2 yr

HPDX: dimensionless physics parameters

• Two key parameters: <β>N = (<p>/<B2>)/(I/aB) and H

[confinement enhancement over ITER98H(y,2)]

– No-wall stability limit is βN ~3 for all A

– Use estimated current drive efficiency I neR/Ph= 0.3 x 1020

(<Te>/10kev) A/Wm2

• very close to PPCS studies, ITER analysis

– Use VMEC equilibria with fixed temperature and density

profiles (characteristic of low-collisionality hybrid H-

modes) to get βN , bootstrap current, and fusion power

• Use Sheath limited regime parameter benchmarked to SOLPS

S = Q||u (Bdiv/Bu) /n1.75 L0.75 / 2x10-27 > 1

=> Te > 100 eV

HPDX: conservative design

Take N = 3,

Pfus =100 MW

• Current drive power

needed much lower at

lower density

• Only SXD has S<1 so

plasma cools near

plate

• Need SXD to handle

power exhaust at low

density

• Density ~ 1/3 Greewald

1020

HPDX: a conservative design

HPDX is within

general expectations

for experimentally

accessible plasma

shapes

Low-A allows high k

and we have such HPDX

equilibria

A (still-improving) HPDX Corsica Equilibrium

Iplasma = 11 MA,

Btor = 2.5 T,

Rmajor = 1.35 m,

aminor = 0.75 m,

Aspect A = 1.8

Elong = 3.142

Beta = 19.1%

Max Aw = 9 m2

Bcenter post = 7 T

MCNP shows: neutron

damage to Center post

(with 10 cm steel

cladding), and divertor

plate can be kept

below 2 and 1 dpa/FPY,

so can operate for ~ 1-

2 FPY

Proposed Next Step steady state devices-

much more challenging divertor operation than ITER

Device Heating Power (MW)

Rmajor

(m)

P/R(MW/m)

ITER 120 6.2 19

NHTX 40 1 40

ST-CTF (Peng, Wilson)

~ 50-70 0.75-1.2 40-90

Normal A CTF ~100 2.5 40ARIES-AT, ARIES-RS

390-510 5.2-5.4 74-93

PPCS A,B 990-1246 8.6-9.6 115-130

PPCS C,D 571-792 6.1-7.5 94-106SLIM CS, CREST 645-691 5.4-5.5 117-

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