# Presentation: 007193

## First Wall Issues:ITER to DEMO

- D.G. Whyte,  B. Lipschultz
- Plasma Science & Fusion Center, MIT, Cambridge USA
- Greenwald Panel, FESAC Meeting
- August 8, 2007

## First-Wall & PSI becoming increasingly important and difficult as we move from present tokamaks  ITER  demonstration fusion power plants

| Issue / Parameter | Present Tokamaks | ITER | DEMO | Consequences |
| --- | --- | --- | --- | --- |
| Quiescent energy exhaust
GJ / day | ~ 10 | 3,000 | 60,000 |  active cooling
 max. tile thickness ~ 10 mm |
| Transient energy exhaust from plasma instabilities 
T~ MJ / Awall(m2) / (1 ms)1/2  | ~ 2 | 15 | 60 |  require high Tmelt/ablate
 limit? ~ 60 for C and W 
 surface distortion |
| Yearly neutron damage in plasma-facing materials
displacements per atom | ~ 0 | ~ 0.5 | 20 |  evolving material properties: thermal conductivity & swelling |
| Max. gross material removal rate with 1% erosion yield
(mm / operational-year) | < 1 | 300 | 3000 |  must redeposit locally
 limits lifetime
 produces films |
| Tritium consumption
(g / day) | < 0.02 | 20 | 1000 |  Tritium retention in materials and recovery |

## Upstream peak power exhaust: SOL Width ~ R leads to q//,max ~ 2-3 GW/m2 in reactorDaunting engineering task, but can we access physics in present devices?

**q****//,sep**
**(W/m****2****)**

**10****8**

**10****9**

** ****q****//,sep**** ********1.43x10****4**** q********0.86 **** q****cyl**** ****1.03**

**10****9**

**10****8**

- ITER

- ARIES-AT

**P****SOL****=0.7 (P********+P****aux****)**

*Whyte et al. ITPA 07 Garching*

## Tritium/fuel control and retention represents the single largest step between present devices/ITER and a DEMO

## High materials temperatures seem mandatory to control fuel retention and anneal neutron damage:What temperature will be required?

| Film Location | Tmax
(K) | (H+D) / C |
| --- | --- | --- |
| Front-face | >1000 | 0.01 |
| Behind tile | 420 | ~ 1 |

**JT-60U tokamak ****(Tanabe et al. ITPA Toronto 06)**

- Thermal conductivity restored
- by high-T annealing

*L. Snead, T. Burchell,*
*J. Nucl. Mater. ****224**** (1995) 222.*

## Both long pulse AND high temperatures required to reach true particle equilibrium in first wall.

- ITER
- ambient T

**Seconds required**
**To permeate**
**1 cm thick****armour**

**10****8**

**10****6**

**10****4**

*G.M. Wright, Ph.D. thesis, U. Wisc 2007*

## Summary

- The requirements for a sustainable fuel cycle and wall viability push us strongly to investigate first walls with much higher ambient temperatures
- Tritium recovery, suppression of retention in codeposits.
- Walls and particles in equilibrium
- Annealing of neutron damage
- Technological requirement: gas cooling + peak heat load removal
- A key physics / design issue will be the selection of the ambient temperature that provides the appropriate trade-offs between temperature limits (e.g. heating from transients), desired high power density, and high temperature benefits such as tritium recovery.
- Exponential T dependences demand experimental demonstration / testing.
- “Hot-wall” tokamak operation would investigate a physics, operational and technological path that seems vital to fusion’s success, but which no one else, including ITER, is pursuing.
- Must be coupled with a concentrated effort on PSI science and diagnosis + sustainment physics.

## Additional Materials

## Cross-device study that showed no dependence of q with R, also revealed a scaling of q ~ P0.5 that results in favorable extrapolation to ITER & reactors

- ARIES-AT:  qpeak ~ 6 MW/m2

## Target heat removal is the highest priority in edge design, but the strategy forward is confused by the lack of consistent or compelling empirical scalings

- “All than can be strongly concludedfrom Table 1 [scalings for SOL heat flux] is that
- there is a need for improved experimental
- measurements and a theory-oriented
- approach for making extrapolations
- for the target heat flux..”
*Tokamak Physics Update: Power and Particle**Control*, A. Loarte, et al. Nucl. Fusion **47** (2007) S203.

## As Kallenbach1, we find T ~ R. Further regression  analysis shows  T is invariant with PSOL and insensitive to other global and separatrix parameters

*1 Kallenbach,et al. JNM 337-339 (2005) 381. *

- T ~ R1.1±0.1 q*0.15±0.2 PSOL0.01±0.05 nsep0.2±0.06

- T (m)

- The invariance with PSOL counter-indicates the scaling expected from   conduction

*Whyte et al. ITPA 07 Garching*

## The separatrix seems to exhibit a critical gradient scale length set by R

- Te (eV)

- Rsep / Ro

- AUG
- DIII-D
- JET
- C-Mod
- JT-60U

*Whyte et al. ITPA 07 Garching*

## The scaling gives (P/R)2/7 as the figure-of-merit to set upstream separatrix Te

**The scaling gives (P/R)****2/7**** as the figure-of-merit to set upstream separatrix T****e**

- Consistent with original derivation by Lackner1, who argued Te /Eatom should be matched for edge similarity if SOL width scaled as R.
- Upstream atomic rates vary weakly with T, e.g. D ionization rates match within few percent between present devices & ITER..
- Upstream atomic physics constraint can probably be dropped from similarity requirements, as it is in core.

*1 **Comm. Plasma Phys. Controlled Fusion ****15**** (1994) 359.*

**T****e.sep**** (eV)**

** ****T****e,sep**** ********1.09 (P****SOL****/R)****0.28********0.02 **** q****cyl****0.6********0.07**

- ITER

*Whyte et al. ITPA 07 Garching*

## “Archeological” deposition measurements:Tokamak plasmas effectively net “transfer” carbon from one location of the wall to another

- 13-C isotope tracer experiments support idea that C is transferred from main-wall “limiters” to inboard divertor
- Controlling mechanisms of main-wall erosion sources, long-range transport and deposition balance are not well understood.

- 13CH4

*McLean et al APS 04*

*Whyte, Stangeby et al. IVC 07 Stockholm*

## Operational consequences of 1021 C/s global transfer rate demonstrate necessity for high ambient temperature to control Tritium retention in carbon films.

| Ambient wall temperature | T/C | # Pulses to reach 350 g T limit in ITER* | # days to T-limit in DEMO 1 | # days lifetime
for “limiters” in DEMO 2 |
| --- | --- | --- | --- | --- |
| ~ 400 K | ~ 0.3 | 500 | ~ 6 | ~240 |
| > 1000 K | ~ 0.01 | >104 | ~ 230 |  |

- 1 Assumes 1 kg safety limit
- 2 Assumes 20 m2 limiter surface

- * ITER is water-cooled
- Tambient ~ 400K

*Whyte et al. IVC07 Stockholm*

## 14 MeV neutron-induced damage set lifetime limits for graphitic fusion materials 

*T. Burchell, J. Nucl. Mater. ****179-181**** (1991) 205. *

- Displacements per atom

- 0

- 20

- 40

- 0

- % Dimension Change

- 3

**End of**
**lifetime**

**Neutron induced shrink/swell in N3M graphite **