Critical Gaps between Tokamak Physics and Nuclear Science › FNST › FNST-PFC-2-6-Aug-2010 › ... · (Step 1: Identifying critical gaps) (Step 2: Options to fill the critical gaps…initiated)

(Step 1: Identifying critical gaps)(Step 2: Options to fill the critical gaps…initiated)

(Step 3: Success…not yet)

Clement P.C. WongGeneral Atomics

Fusion Nuclear Science and Technology Annual MeetingAugust 2, 2010 Rice Conference Room, UCLA

Critical Gaps between Tokamak Physics and Nuclear Science

The beautiful picture of fusion developmentInexhaustible clean energy for the world

FusionPowerReactor

DEMO

DIII-D, CMODNSTX, JET, EASTJT-60, KSTAR

FNSF

The beautiful picture of fusion development:Inexhaustible clean energy from MFE fusion power reactor

FusionPowerReactor

DEMO

DIII-D, CMODNSTX, JET, EASTJT-60, KSTAR ITER

FNSFIFMIF

For MFE we will need ITER and FNSF+IFMIF

40-60

3

80-100

150Under DT reaction, PFM and structural material damages from neutron dpa+He generation and DT He+ will become major challenges

~0 dpaDD neutrons

We can see the obvious critical issue from dpa effects

40-60

3

80-100

150

~0 dpaDD neutrons

We need to provide technical connections between major devices

ITER has helped to focus world fusion resourcesIt has also identified critical issues when extended to DEMO

?

ReNeW Workshop was the First Step for Systems Interface activities

18 Thrusts were Identified

5 Critical areas were selected

• Tritium supply, burnup fraction, tritium processing and TBR

• Divertor configuration and peak heat flux reduction

• Transient events: disruptions, ELMs and runaway electrons

• First wall heat flux and design

• Plasma surface material

Critical areas that will have major impacts to the feasibility of fusion power

Fast pace review on the first 3 areas

Tritium supply, burnup fraction, tritium processing and TBR







12

Tritium Breeding Blankets must be developed in the near term to solve the serious issue of external tritium supply

• A Successful ITER will exhaust most of the world supply of tritium, but 5-10 kg will be needed to start DEMO

• Any future long pulse burning plasma device (including ITER Extended Phase) will need tritium breeding technology

• The availability and cost of external tritium supply is a serious issue for FNSF development

• Engineering development and reliability growth stages must be done in a small fusion power device; only fusion break-in stage can be done in ITER

Tritium Consumption in Fusion is HUGE! Unprecedented!

55.8 kg per 1000 MW fusion power per year

Production & Cost:

CANDU Reactors: 27 kg from over 40 years, $30M/kg (current)

Fission reactors: 2–3 kg/year/reactor,

$84M-$130M/kg (per DOE Inspector General*)

*www.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdf

We cannot wait very long for blanket developmentM. Abdou, UCLA, 2007

0

5

10

15

20

25

30

1995 2000 2005 2010 2015 2020 2025 2030 2035 2040 2045

Year

Pro

ject

ed O

nta

rio

(O

PG

) Tr

itiu

m In

ven

tory

(kg

) CANDU Supply

w/o Fusion

1000 MW Fusion10% Avail,

TBR 0.0 ITER-FEAT(2004 start)

http://www.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdf



Impact of T burn-up fraction in plasma on start-up T inventory for new power plant

L. El-Guebaly, UW

Implications of tritium burn-up fraction for ITER ~ 0.3%

A power reactor consumes ~ 0.5 kg per day, and if ttp is ~ 24 hours like TSTA, then the tritium inventory in the fuel storage will be > 160 kg!! Totally unacceptable. If ttp is reduced

to 4 hours, it will be ~ 27 kg. Still too high!!

A power reactor with the same as ITER would be unacceptable!

ITER is designed with a burn-up fraction of 0.3%Low tritium burn-up fraction will need large tritium startup inventoryThe key is to optimize burn-up fraction and tritium processing, ttp

M. Abdou, UCLA, 2007

Divertor configuration and peak heat flux reduction







Most divertors are designed to a peak heat flux of 10 MW/m2

M. Kotschenreuther, ARIES Workshop, San Diego, CA, May 20-21 2010

Peak heat flux for DEMO is very uncertainNew divertor configurations should be considered

D. Ryutov, ARIES Workshop, San Diego, CA, May 20-21 2010

Transient events: disruptions, ELMs and runaway electrons







However even with prefect disruption mitigation, a power reactor will need to be designed to withstand a few unforeseeable events

First wall heat flux and design







ReNeW PFC panel generated unexpected surprises

• ITER First wall panels are designed to

1 MW/m2 and 5 MW/m2 heat flux

while steady state radiation is only 0.5 MW/m2

• Why???

R. Nygren SNL, 2010

R. Nygren SNL, 2010

Correct question: Can physicists manage without such a requirements ?

Main chamber ELM loads• Clearly present in higher

triangularity configurations

DIII-D

#13

82

19

Before ELM

During ELM

IR TV

DIII-DSecondary

strike

See J. G. Watkins, Poster P2-66, Tuesday68193, 57 s

JET

R. A. Pitts et al., APS 2007

R.A. Pitts, 19 PSI conference May 2010, San Diego

ITER Thermal Load Specifications: resumé

Start-up:q|| ~ 25 MWm-2, lq|| ~ 5.0 cmSeveral seconds

Confinement transientsq|| ~ 250 MWm-2, ~2-3 secs

Start-up and rampdown:q|| ~ 40 MWm-2, lq|| > 1.2 cmSeveral seconds

VDE (up):q|| ~ 70-270 MJm-2, lq|| > 3.0 cmt = 1.5-3.0 ms

VDE (down):q|| ~ 90-300 MJm-2, lq|| > 3.0 cm

Steady state:q|| ~ 8 MWm-2, lq|| > 4.0 cmq|| ~ 24 MWm-2, lq|| > 2.5 cm (ELMs)

Disruptionsq|| ~ 45-120 MJm-2, lq|| > 20 cm

t = 3.0-6.0 ms

Radiation:SS: 0.5 MWm-2

(photon+CX)

DisruptionsTQ: ~0.5 MJm-2

t ~ 1 ms (mitigated)CQ: ~0.9 MJm-2

t ~ 10 ms

R.A. Pitts, 19 PSI conference May 2010, San Diego

Very high parallel heat fluxes

ELMs Mitigation

Extension to FNSF

• How can we accommodate necessary RMP coils while satisfying TBR, thermal performance and components lifetime requirements?

Note:

• Other ELMs control methods are being investigated, e.g. Low field side (LFS) pellet injection

• High performance ELMs free plasma (e.g. QH mode) may be the right approach but open issues remain

Plasma surface material







Surface Material is a Key Item for Fusion Development

Surface material is critically important to next generation tokamak devices:

• Plasma performance is affected by transport of impurities

• Surface heat removal, tritium co-deposition and inventory will have impacts on material selection for devices beyond ITER

• Radiation effects from neutrons and edge alphas, material design limits and component lifetimes will have to be taken into consideration

C and Be will not be suitable for the next generation devices and DEMO

due to surface erosion and radiation damage. Presently W is the

preferred choice, but feasibility issues have been identified

DIII-D JET-ILWC-Mod AUG EAST ITER FNSF DEMO

Surface material options

(High neutron and edge alpha fluence)

C Mo W Be/W/C C/W Be/W/C ? ?

Plasma Surface Materials and Effects from Helium

1. Transmuted helium*

2. He ions, generated from DT reaction

* Radiation impacts from high energy neutrons in addition to dpa effects

Low energy He+ irradiation in plasma simulator NAGDIS

H bubble and hole formation on W surface @ > 10 eV

D. Nishijima, Journal of Nuclear Materials, 329 (2007) 1029-1033

Damages to W First Wall have also been Projected from He+

When exposed to He at high temperature, W

surface showed growth of W nano-structure from the bottom; the thickness increases with plasma exposure time

Baldwin and Doerner, Nuclear Fusion 48 (2008) 1-5

Significant Issues Projected for W-surface Operation

Independent of Alloy Development

ITER disruption loading:

10-30 MJ/m2 for 0.1 to 3 ms

Irreversible surface material damage

M. Rödig, Int. HHFC workshop, UCSD Dec. 2009

We cannot eliminate un-predicted

disruptions even if disruption detection

and mitigation work perfectly

W fuzz most likely will not be observed in Tokamaks, but it can still have major impacts as a form of surface erosion

Plasma Facing Material Design and Selection

Requirements for Next Generation Devices

1. Withstand damage from DT generated He

2. Withstand transient events like ELMs and disruptions

Additional critical requirements:

Physics performance:

3. Material suitable for high performance plasma operation

4. Suitable for edge radiation to reduce maximum heat flux at the divertor

5. Low physical and chemical erosion rate

Engineering performance:

6. Transmit high heat flux for high thermal efficiency conversion

7. Minimum tritium inventory

8. Minimum negative effect to tritium breeding performance

9. Low activation materials

10. Replenish damaged surface material suitable for steady stateoperation and long lifetime

11. Match materials temperature design requirements

12. Withstand high neutron fluence at high temperature

A Possible Si filled W-surface concept could Satisfy most

Requirements

The concept: Si-filled W-surface (#3,#4,#9)

• Thin Si surface could protect the W surface

from He damage (#1)

• Exposed W will have a low erosion rate (#5)

• ~2mm thick W with indents could transmit high heat flux,

thus retaining high effective kth of W layer, necessary for

DEMO (#6,#8,#11)

• Enough Si is provided to withstand ELMs and a few disruptions

(modeling showed vaporized Si ~10 μm/disruption including vapor

shielding effect) “W-Tmelt@ 3410°C, Si-Tmelt@1412°C, Si-Tboil@ 2480°C” (#2)

• Should be able to control tritium inventory at temperature ~1000 °C (#7)

• Suitable real time siliconization could be used to replenish Si when and

where needed (#10)

(Satisfying requirements #12 TBD)

W-buttons

W-buttons

filled with Si

Si filled W-buttons W-buttons with

1 mm dia. indentsLoaded DiMES sample

2 Si-W, 3 graphite, 2 W buttons

Initial Results of Transient Tolerant Si-filled W-buttons

Sample exposed

To 4 LSN dischargesAfter one additional disruption,

but not fully thermally loaded

Exposed in

DIII-D lower divertor

Shot 14261-14264Shot 142706

Si-W Buttons Exposure Observations

• As expected Si on the W button surface got removed during normal

discharges easily via sputtering or vaporization

• Favorable result shows much of the Si is retained in the indents even

under relatively high heat and particle flux

• Retained Si could demonstrate the vapor shielding effect to protect

the W-button surface from melting

• Initial results show expected results in the performance of Si-filled W-

surface to fulfill its function, much more development and testing will

be needed

Conclusions

From ITER to FNSF critical gaps have been identified between Tokamak Physics and Nuclear Science

and they can only be resolved with close interactions between physics, material, technology and design communities

Examples are:

• Tritium supply, burnup fraction, tritium processing and TBR• Divertor configuration and peak heat flux reduction• Transient events: disruptions, ELMs and runaway electrons• First wall heat flux and design• Plasma surface material

Conclusions-Solutions


Higher T burnup fraction >>0.3% is needed, furthermore a net tritium producing device like FNSF is needed before the tritium supply runs out


New divertor configuration and with radiation to maintain acceptable peak heat flux is needed for a robust FNSF design with design margin


Different schemes of radiation are needed to mitigate damaging peak power flux impacts to surface material


Radiation to spread the peak heat flux and ELM-free operation, like QH mode are needed


Si-filled W-surface, which uses radiation to mitigate surface material damage is a possible transient tolerance approach and should be developed

Conclusions-Radiation is the key

We will need FNSF soon, and radiation is the key to control the damaging transient events. Si-filled W-surface design is proposed as a possible PFM for steady state operation of FNSF and should be organized for more systematic studies.


Higher T burnup fraction >>0.3% is needed, furthermore a net tritium producing device like FNSF is needed before the tritium supply runs out


New divertor configuration and with radiation to maintain acceptable peak heat flux is needed for a robust FNSF design with design margin


Different schemes of radiation are needed to mitigate damaging peak power flux impacts to surface material


Radiation to spread the peak heat flux and ELM-free operation, like QH mode are needed


Si-filled W-surface, which uses radiation to mitigate surface material damage is a possible transient tolerance approach and should be developed

Documents

Critical Gaps between Tokamak Physics and Nuclear Science › FNST › FNST-PFC-2-6-Aug-2010 › ... · (Step 1: Identifying critical gaps) (Step 2: Options to fill the critical gaps…initiated)