But ExB flows can adversely affect screening on HFS for single-null, with favorable grad-B.

1. Analysis of Long-Leg divertor performance

ExaminingInnova,veDivertorandMainChamberOp,onsforaNa,onalDivertorTestTokamakB. LaBombard1, M. Umansky2, D. Brunner1, A. Kuang1, E. Marmar1, G. Wallace1, D. Whyte1, S. Wukitch1 [1] MIT-PSFC, [2] LLNL

https://www.burningplasma.org/resources/ref/Workshops2015/PMI/PMI_fullreport_21Aug2015.pdf

Clear consensus emerged from 2015 FES Community Planning Workshop on Plasma-Materials Interactions for a National Divertor Test Tokamak

Cross-Cutting Initiative #4: “Develop integrated plasma-material solutions in a purpose-built Divertor Test Tokamak”

“In the judgment of the panel, an experiment with both DEMO-relevant upstream heat flux and pressure, together with high divertor and first-wall configuration and material flexibility, will be needed to proceed to DEMO with scientific confidence.”

A new facility for boundary and divertor research is needed ...

“Recently a new high-power-density divertor test tokamak facility has been analyzed in the United States [8]. It features long divertor legs and a flexible poloidal field configuration, along with flexibility in gas dynamics and in the use of solid and liquid plasma-facing materials. At the time of ReNeW, two concepts for a boundary and divertor physics machine were under consideration [9,10] both with high power, long pulses and hot first walls. This new short-pulse concept adds considerably to the range of options available for consideration. A national working group should be established to develop options for a United States-led divertor test tokamak. The European Roadmap argues that control of boundary and divertor physics is “probably the main challenge towards the realization of magnetic confinement fusion.” The United States has the opportunity to be the world leader in this area, and should seize the opportunity.”

8 Labombard, B. et al., Nucl. Fusion 55, 053020 (2015). ADX 9 Goldston, R.J. et al., IAEA Fusion Energy Conference 2008, FT-P3-12. NHTX 10 Olynyk, G.M. et al., Fusion Eng. Design 87, 224 (2012). Vulcan

... and a national working group should be established to develop options for a US-led “DTT”.

UEDGE analysis of X-point Target Divertor (XPTD)1,2 Passively-stable, fully-detached divertor is obtained

over a factor of ~10 variation in exhaust power •  Upgrade of UEDGE allows simulation of secondary X-points in divertor legs

[3] B. LaBombard et al., Nucl. Fusion 55, 053020, 2015

Fig.2 : Outer leg of XPTD from UEDGE simulations. As exhaust power is increased, a stable radiation/detachment front moves down the divertor leg, accommodating a factor of ~10 variation in exhaust power.

Fig.1 : UEDGE grid for a XPTD configuration based on the ADX divertor test tokamak3

A C-Mod like case with ADX parameters is used for comparison

7%

•  Two.configuraEons.–.XPTD.and.SVPD.•  Same.underlying.magneEc.geometry,.physics.model,.boundary.condiEons,.etc..•  In.SVPD.the.legs.cut.short.to.roughly.match.C#Mod.verEcal.plate.configuraEon.

Standard Vertical Plate Divertor (SVPD)

X-Point Target Divertor (XPTD)

R [m]

Z [m]

R [m]

P1/2 = 1.8 MW P1/2 = 1.4 MW P1/2 = 1.0 MW Prad [MW/m3]

R [m] R [m] R [m]

Z [m

]

1% C0.0

2.5

5.0

7.5

10.0

•  The combined effects of long leg, neutral interactions and secondary divertor x-point lead to a factor 10 enhancement in the peak power handling and operational power window for XPTD compared to SVPD.

•  Power exhaust window for attaining stable detachment was explored for XPTD compared to other concepts: standard vertical plate (SVPD), long vertical leg (LVLD), and super-X divertors (SXD).

Near-Balanced Double Null plasmas exhibit good impurity screening on the high-field side (HFS) SOL while providing precise SOL profile control.

IAEA FEC2016 EX/P3-6: LaBombard et al., “Plasma profiles and impurity screening behavior of the high-field side scrape-off layer...”

HFS density drops by two orders of magnitude in 6 mm

-5 0 5 10 15 20

-5 0 5 10 15 20

Density

Electron Pressure

10

100

1020

eV

m-3

0.1

1.0

1020

m-3

-5 0 5 10 15 20ρ, Distance into SOL (mm)

High field side SOL – very sharp profiles, controlled by magnetic topology

Low-field side SOL – broad shoulders

Good HFS impurity screening in Balanced Double Null ...

... despite very sharp profiles Double Null

Put R

F an

tenn

as h

ere

Supports idea to locate RF actuators on HFS to mitigate PMI4 …

LFS

HFS

Unfavorable drift direction – HFS, LFS similar. I-mode

Factor of ~2 better screening on HFS vs. LFS

0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

Rel

ativ

e P

enet

rati

on F

acto

r

-15 -10 -5 0 5 10 15SSEP: X-point separation, mapped to OMP (mm)

Double NullLower Null Upper Null

0.80 MA, RevBHFS LFS N2 Puff Location

EDA H-mode Favorable drift direction – poor HFS screening

0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

Rel

ativ

e P

enet

rati

on F

acto

r

-15 -10 -5 0 5 10 15SSEP: X-point separation, mapped to OMP (mm)

Double NullLower Null Upper Null

0.55 MAHFS LFS N2 Puff Location

0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

Rel

ativ

e P

enet

rati

on F

acto

r

-15 -10 -5 0 5 10 15SSEP: X-point separation, mapped to OMP (mm)

Double NullLower Null Upper Null

0.80 MA0.80 MA, Rev B0.55 MA

1.1 MA HFS LFS N2 Puff Location

LFS puff

L-mode

HFS puff

Maximum Electron Temperature at Target Platesas a Function of Scrape-off Layer Power

T max

[eV]

P1/2 [MW]

SXD

XPTD

SVPD

LVLD

Onset ofCore PlasmaX-point MARFE

4

2

3Detachment Power Window for XPTD

Results from recent research are helping to inform options for a National Divertor Test Tokamak.

Mission Elements for a NDTT (based on the ADX idea):

(1)  Divertor Test Tokamak – develop and demonstrate advanced divertor and first-wall PMI solutions at the power densities and pressures anticipated for a reactor.

(2)  RF Sustainment Test Tokamak – develop and demonstrate advanced RF technologies that scale to efficient, low PMI, SS current drive and heating in a reactor.

(3)  Integrated Core-Divertor-Actuator Test Tokamak - develop/test/demonstrate high core performance (e.g., H98, pressure) in reactor-relevant regimes (e.g., coupled e-ions, low external torque), using only the class of actuators available to a reactor (e.g., no neutral beams)

symmetry plane

SVPD

XPTD

SXDLVLD

symmetry plane

Z [m

]

R [m] R [m]

Z [m

]

1

3

2

4

Modeled cases based on geometry & parameters from ADX design:

§  MHD equilibrium (5.4 tesla, 1 MA) §  Fully recycling material surfaces §  1% carbon impurity radiation §  SOL profiles projected to ADX for a low

density case, nsep ~ 5x1019/m3

§  Fixed values of D and χe,i§  Power into lower half-domain, P1/2,

varied 0.2 – 4 MW §  λq~ 0.86 mm => q//,omp ~ 5 GW/m2

3. High-field side SOL impurity screening and plasma profile measurements – support for HFS RF actuators

Initial design concept [Nucl. Fusion 55 (2015) 053020]

Inside Launch LHCD

Inside Launch ICRF

Demountable TF Magnet

Vertically Extended Vacuum Vessel

Outside Launch ICRF

Internal Divertor PF Coils

Advanced Magnetic Divertors

ADX

4. Implementation of Long-Leg Divertor with SC PF coils in reactor

2. Heat flux sharing experiments in Near Double-Null

Factor of ~2.5 better screening on HFS vs. LFS in DN

•  Direct external control of plasma conditions at RF actuator interface (gap, flux balance)

•  Quiescent SOL; thin SOL; no ‘blobs’ – reduced wave interactions

•  No ELM load, runaway e-, energetic ion orbit loss •  Low neutral pressure – increased RF voltage •  RF-generated fast e- drift away from launcher •  Reduce neutron flux on HFS above and below

midplane [M. Youssef et al., FED 83 (2008) 1661]

Brunner APS Contributed Talk NO4.00014: “Divertor conditions near double null in Alcator C-Mod”

IAEA FEC2016 TH/P6-32: Umansky, “Assessment of X-point target divertor configuration for power handling and detachment front control”

In Near Double-Null, ~90% of the power into the SOL exhausts to the Upper + Lower Outer Divertor Targets

I-mode EDA H-mode L-mode

X-point Target Divertor has been implemented in ARC v2 design.

Output of MIT NSE Course 22.63 Reactor Design Class, Spring 2016 – paper in progress.

•  Superconducting PF coil set, fully shielded to neutrons by FLiBe tank

•  Acceptable coil currents •  (~50 to 70% of Ip), current

densities and forces •  Identical plasma cross

section, shaping and TF coil size as original ARC

•  TBR of original ARC (1.08) is maintained

•  Neutron damage on outer divertor targets greatly reduced

Message: Demountable TF + FLiBe immersion blanket allows implementation of advanced divertors with neutron-shielded PF coil set and acceptable PF currents.

Demountable TF Coil

Demountable TF Coil

[1] Umansky et al., FEC2016, TH/P6-32. [2] Umansky APS Invited Talk JI3.00003: “Attainment of stable, fully- detached plasma state in innovative divertor configurations”

BT ~ 6.5 tesla Ip ~ 1.5 MA Paux ~ 14 MW q//,omp ~ 12 GW/m2

A high-field, compact tokamak is a compelling option for a NDTT.

ARC employs HFS LHCD for sustainment.

[4] Reinke, APS Invited Talk, GI2.00005 : Experimental pathways to understand and avoid high-Z impurity contamination from ICRF heating in tokamaks”

… while greatly enhancing RF wave physics5

−4 −3 −2 −1 0 1 2 3 4SSep [mm]

0.00.10.20.30.40.50.60.70.80.91.0

Fraction of total power flux L-mode,0.8 MA,6.7e+19m , FWD-B

Inner Lower

Outer Lower

Outer UpperInner Upper

-3

−4 −3 −2 −1 0 1 2 3 4SSep [mm]

0.10.20.30.40.50.60.70.80.91.0

Fraction of total power flux I-mode,0.8 MA,9.0e+19m , REV-B

Inner Lower

Outer Lower

Outer Upper

-3

Inner Upper

−4 −3 −2 −1 0 1 2 3 4SSep [mm]

0.10.20.30.40.50.60.70.80.91.0

Fraction of total power flux H-mode,0.8 MA,2.5e+20m , FWD-B

Inner Lower

Outer Lower

Outer Upper

-3

Inner Upper

Favorable grad-B Unfavorable grad-B Unfavorable grad-B Favorable grad-B

[5] IAEA FEC2016 TH/5-1: Bonoli et al., “Novel Reactor Relevant RF Actuator Schemes for the Lower Hybrid and the Ion Cyclotron Range of Frequencies.”