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Power and Particle Control

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ELM(Edge Localized Mode) 1/40

Divertor Heat Load

Can fusion power be handled by surrounding wall?

Power and Particle Control

(2)

Steady-state transport in the Scrape-Off-Layer (SOL)

Total Fusion Power 500 MW

Heating & CD 50 MW

Neutron 400 MW 100 MWα

Plasma

(150 MW)

CoreRadiation

~30 MW

Thermal to SOL

~120 MW

PSOL

~120 MW

< 10 MW/m2

Divertor heat flux

ITER

(3)

3/40

Can fusion power be handled by surrounding wall?

Divertor heat flux

 Peak poloidal heat flux

PSOL

~120 MW

< 10 MW/m2

λq qpol

qpol ~ PSOL / 2 2pRlq

q|| ~ PSOL / 2 2pRlq

B Bq qpol q q

┴,target

Geometry: field pitch,

flux expansion, target plate tilting

 Parallel heat flux

 Target heat flux

Divertor Target plate

q||,u B  q||,target

Upstream L

α

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Divertor Heat Flux Challenge with Short SOL Width

D. Brunner et al., 2017APS-DPP

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5/40

Can fusion power be handled by surrounding wall?

Increasing core radiation Assume fLH = 1.2 for ITER, fLH = 1.1 for EU-DEMO, λq = 5 cm on the target for both

Increase the radiated power fraction in the core

• radiated power limited by the need to stay in H-mode

• PLH should be expressed in Psep = Pheat - Prad

• Psep,min = fLH Psep,LH α ne0.7 Bt0.8 R2

Zohm, H., Physics of divertor power exhaust beyond ITER, 10thITER international school 2019.

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6/40

Can fusion power be handled by surrounding wall?

Divertor heat flux reduction

 increasing core radiation

 Detachment and radiation cooling

 Various configurations

• X, Super X (SX), SnowFlake (SF), Double Null (DN), …

 Liquid Li

Zohm, H., Physics of divertor power exhaust beyond ITER, 10thITER international school 2019.

(7)

7/40

Can fusion power be handled by surrounding wall?

Liquid Lithium: Alternative (liquid) materials may lead to higher allowable Psep

Zohm, H., Physics of divertor power exhaust beyond ITER, 10thITER international school 2019.

Alternative (liquid) materials may lead to higher allowable Psep

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8/40

Erosion of Plasma Facing Component (PFC)

PFC sputtering and chemical erosion

Eckstein et al.

Roth et al., NF 44 (2004) L21

Physical sputtering yield… Chemical erosion yield (D on C)…

Decreases with D ion flux and is sensitive to C target temperature

 W

 W

D

D

D

Increases with projectile energy and mass, while decreasing with target

(PFC) material atomic mass

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9/40

Impurity Influx from Plasma Facing Component (PFC)

Impurity radiation in the core plasma

Q=(fusion power)/(power in); goal for ITER > 10

• Allowed concentration (nW/ne) of ~ 5x10-5

• An injection of a mm diameter droplet of W would lead to radiative collapse in ITER

• Melting should be avoided 

0.0001 0.001 0.01 0.1

0 20 40 60 80

Impurity atomic number

fatal fraction

W Mo

C

D. Whyte, 2005 Jensen PPPL report #1350, 1977

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10/40

Neutron Irradiation on Plasma Facing Component (PFC)

 Tungsten and some 3D graphites have better nuclear damage properties

Operation at high temperature may cause ‘mending’ of neutron damage

L. Snead, J of Nucl. Materials 417 (2011) 629–632

graphite graphite

(11)

11/40

Tritium Retention of Plasma Facing Component (PFC)

 Implantation is the dominant retention process for refractory metals such as W

 Radiation Damage Increases T Retention

 Operating at higher temperatures lowers the retention further

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12/40

Dilution and Radiation for Fusion Operational Space

Both dilution and radiation can be taken into account in determining the size of

operational space for fusion burn through the “Lawson criterion” (burn criterion)

D. Reiter et al., Nucl. Fus. 30 (1990) 2141;

T. Pütterich, EFPW Split, Dec 2014 p12

Higher nT𝜏E and higher T are needed as the He and other impurity concentrations increase.

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13/40

Plasma Facing Component (PFC) Material Selection

0.01 0.1 1 10

10-4 10-3 10-2 10-1 100

physical sputtering

C

W Be

D+

SPUTTERING YIELD (at/ion)

ENERGY (keV)

Wall erosion best with high Z (but no melting of carbon!)

Radiation

Dilution

0.001 0.01 0.1 1

0 100 200 300 400 500 600 700 800

D/C D/BeO D/Be D/Be D/Be D/W D/(W+C) D/(Be+C)

Temperature (C)

D/C D/W D/Be

Tritium retention - an issue for carbon!

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14/40

Can fusion power be handled by surrounding wall?

Plasma Facing Component (PFC) Material selection

 High power fluxes: steady-state and transient

 Strong erosion: sputtering and chemical erosion

 Carbon and Beryllium unusable in reactor

 High-Z tungsten PFC needs improvement

 Severe neutron irradiation

 Tritium retention

 Carbon unusable in reactor Tungsten

Operating at higher temperatures may be a good solution

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