# Presentation: 486301

## Outline

- ITER JWS Garching

**Outline**
- Highlights of PFC design & material selection
- Erosion during transient heat loads
- Remaining open questions
- Summary

**Materials Selection and **
**Engineering Design of ITER PFCs**

**2****nd**** SOL and Divertor Physics ITPA Meeting, **
- Ioffe Institute, St. Petersburg - July 14-17, 2003

## Rationale for material selectionPresent choice: 3 different materials

- Rationale for material selection**Present choice: 3 different materials**

|  | Advantages | Drawbacks |
| --- | --- | --- |
| Be | Good O2 gettering capability
~ tokamak practice (mainly JET);
low Z;
repairability by plasma-spraying;
established joining technology;
low tritium inventory. | Inadequate for the divertor (Tm<1300 ̊C).
Be dust on ‘hot’ surfaces -> H production  due to reactivity with steam during accidents.
Compatibility with Type I ELM loads disruptions or ‘mitigated’ disruptions. |
| C | Good power handling, thermal shock and thermal fatigue resistance;
does not melt (but sublimes); 
low radiative power losses with influx to plasma  due to low Z;
well established joining technology;
broad tokamak operation experience. | Chemical erosion --> erosion lifetime (~few thousand pulses).
Tritium codeposition --> needs adequate recovery methods. |
| W | low physical sputtering yield and high sputtering threshold energy;
no chemical sputtering in H-plasma;
low tritium inventory;
repairability by plasma-spraying;
well established joining technology. | Plasma compatibility-->favourable operation experience from ASDEX-Upgrade, dust radiological hazard. 
Near strike points, its operational lifetime, has still uncertainties due to melt layer loss during disruptions and ‘large’ ELMs. |

*ITER*

## Divertor design loads/lifetime

**Nominal surface heat loads****	****q = 10 MW/m****2****, ******** = 400-500 sec,  N = 3000 cycles ****(1****st ****div.)**
**Transient heat load****	****	****q = 20 MW/m****2****, ******** = 10 sec,  N = 300 cycles (i.e., 10%)****	**
**Disruption heat load****	****	****Q = 10-100 MJ/m****2****, ******** = 0.1********10 ms, N = 300 event (i.e., 10%) ****	**
**Ion flux parameters****	****	****J = 10****21****-10****24**** m****-2****s****-1****,   E = 10-100 eV****	**
**Neutron load****	****	****	****J****(E>0.1MeV)********10****18**** m****-2****s****-1****, D = ~0.1 dpa****	**
**Other factors****	****	****	****electromagnetic loads (P******** 4 M****Pa****), H environment, etc.**

**Code simulation of surface heat flux on divertor**

*ITER*

**W/m****2**

## Design assembly/maintenance

- The divertor comprises 54 cassettes installed in ITER via 3 equi-spaced handling ports.
- Each cassette consists of a cassette body onto which are mounted 3 PFCs.
- These PFCs can be exchanged in hot cell in order to refurbish or to change the geometry of the divertor.
- Several complete exchanges are foreseen during the life of ITER.

- The cassettes are accurately positioned in the vessel such that each PFCs is aligned within ±2mm with respect to the PFCs on adjacent cassettes.

*ITER*

## Initial operation strategy

- Maintain option to switch to an all-W divertor, prior or during T-operation. This option will be considered if
- we do not succeed in mitigating the effects of T co-deposition (need to be to be determined during D-phase):
- Design mitigation/temperature tailoring.
- Mixed-material effects (=>_**need to use existing tokamaks**_).
- we fail  in developing reliable and effective techniques of in-situ tritium removal, which need to be demonstrated/tested in tokamaks.
- there is substantial progress in mitigating               heat loads during disruptions and ELMs.

**CFC**

**W**

**Initial operation strategy**
- Current strategy is to initially install CFC on the targets

*ITER*

- Urgent need for development PMI diagnostics to be tested validated in existing tokamaks and during early operation with D in ITER.

## First-wall design loads/lifetime

- Total power		Q  690 MW
- Surface heat loads	q = 0.25 MW/m2 (avg.) 0.5 max.
-                                          = 400-500 sec,  N = 30000 cycles
- Disruption heat load	None.
- VDEs: 		q = 60 MW/m2,  = 300 ms
- 				N=300 cycles (1%),
- Neutron load		0.56 MW/m2 (avg.) / 0.78 (max)
- 				D = ~1 dpa.

*ITER*

- 412 blanket modules attached to the vessel.
- ~4.5 t/module RH constraint.
- steel shielding block
- separate first wall panels.

## Design assembly/maintenanceSeparate first-wall to minimise the operational waste

- Design assembly/maintenanceSeparate first-wall to minimise the operational waste

- Each blanket module is connected to a permanent water cooling manifold by two pipes.
- The FW part repairable and/or replaceable in hot-cell.
- The modules are maintained by a special remotely driven in-vessel transporter inserted through the equatorial port.

*ITER*

## Separate first-wall conceptsThe FW has 4 or 6 separate panels depending on the option chosen for FW attachment. 

**Option A:** attachment with bolts and small shear ribs to support EM loads and to prevent sliding due to thermal expansion.
**Option B:** central beam attachment connected to the shield block on the rear side.

- Separate first-wall conceptsThe FW has 4 or 6 separate panels depending on the option chosen for FW attachment.

*ITER*

## Remote handling classificationbased on need for scheduled and unscheduled maintenance and modifications, likelihood for maintenance, and impact on operations and availability

- Remote handling classificationbased on need for scheduled and unscheduled maintenance and modifications, likelihood for maintenance, and impact on operations and availability

**Divertor is Class 1: **
- Requires scheduled maintenance or replacement.
- Component design and RH equipment and procedures, optimised to ensure task completion within a minimum time.
- Feasibility of maintenance tasks demonstrated with R&D during EDA.
- Demonstration using real components during initial assembly prior to active phase of operation is highly desirable.
**Blanket is Class 2:**
- Do not require scheduled but likely unscheduled or very infrequent maintenance.
- Components are designed for full remote repair or replacement but “minimisation of repair is subordinate to consideration on nuclear performance and reliability”.
- Feasibility of maintenance tasks partially demonstrated with R&D during EDA.
- Demonstration using real components during initial assembly prior to active phase of operation is desirable.

*ITER*

## Maintenance time estimatesIn-vessel components are removed from the VV by 4 equatorial ports, 3 divertor ports and (?) upper ports (EC and diagnostics). 

**Divertor cassette refurbishment**
- 18 cassettes --> 1 RH port =>_**2 months**_ (7 working day/week, 2 (8 hrs) wk shifts/day, 1 cask transporter (8 hrs/shift/day).
- 3 RH ports in series ==> _**6 months**_
- Replacement of 1 single faulty cassette _**≤ 2 months**_!!
**Blanket maintenance**
- Replacement of some shield modules is likely due to local damage.
- Replacement of the full first wall is presently not anticipated, but should be feasible.
- 1 blanket module:		_**25 days**_
- 1 toroidal row:		_**32-151 days**_ *
- All blanket modules:	_**276-916 days**_ *
- * Depends on # of deployed IVT.

- Current design policy: small number of spare parts.

- Maintenance time estimatesIn-vessel components are removed from the VV by 4 equatorial ports, 3 divertor ports and (?) upper ports (EC and diagnostics).

*ITER*

## Critical PFC design/operation issues

**Critical PFC design/operation issues**

*ITER*

**Heat loads and erosion during type I ELMs**
- Ongoing vigorous ITPA effort, EU PWI task force
**Heat loads and erosion during thermal quench disruptions and VDEs**
- Ongoing vigorous ITPA effort, EU PWI task force
**Hydrocarbon transport/T codeposition in remote areas and removal**
- Surveys in tokamaks (EU PWI task force)
- Identify source and sinks.
- Deposition patterns and dependence on operation parameters.
- Composition of exhaust.
- Laboratory simulations (EFDA, EU PWI task force)
- Sticking of radicals.
- Effects of temperature, H/C ratios, etc.
- Mixing of materials (C/Be)

## Heat loads and erosion during type I ELMs

**Heat loads and ****erosion during type I ELMs**
- Tolerable ELMs in ITER set by materials: TplateELM < physical limits (evap., melting).

*ITER*

- Critical parameters are:
- (1) energy loss from pedestal,
- (2) fraction reaching the divertor,
- (3) wetted area,
- (4) duration/shape of ELM heat pulse.
- Knowledge of these quantities still uncertain.

**Temp. excursions during ELMs=> no ratcheting!!!**

- Triang. ELMs, 0.3 ms, 1 MJ/m2
- (1) BOL: 20 mm CFC, 10 mm W.
- (2) EOL: 2 mm CFC, 2 mm W.

**Only near-surface ≤400μm**
**=> Steep Temp. gradients**

## Tolerable ELM size

- Erosion lifetime shorter if one use a statistical evaluation of ELM parameters.
- More inclined divertor target performs better.
- Compatibility of inter-ELM plasmas with irregular W surface remains an issue.
- Additional macroscopic erosion mechanisms due  to high frequency pulsing and high temperature excursions localised in the near-surface (<400μm).

*ITER*

| Armour material
For W assume 50% melt loss. | CFC | W |
| --- | --- | --- |
| Baseline divertor design
 All ELMs w. same parameters
 Statistical evaluation | 
9 MJ
4 MJ | 
10 MJ
3 MJ
 |
| More inclined divertor target
 All ELMs w. same parameters
 Statistical evaluation | 
15 MJ
6 MJ | 
17 MJ
5 MJ
 |

- A large number of ELMs >1 MJ/m2 cannot be tolerated

*ITER*

- In ITER: Wped ~ 100 MJ, nped ~8x1019 m-3,
- Tped~3.5 keV, w~10 cm =>low collisionality (ped ~0.04)
- Scaling to ITER - WELM/Wped:
- **********ped ********15%-20% --> 15-20 MJ.**
******//********10%-15% --> 10-15MJ.**
** ****n****ped****/n****GW ******** **** 4%-5%   -->  4 -5 MJ.**
- 50-70% of energy is deposited in divertor;
- Wetted area: 4.5 -9 m2 with modest (~50%) broadening;
- impact time > // μsIR/ // =1.5-3.1.

## A non-negligible fraction of the ELM energy from the main plasma reaches the main chamber wall

- Where does the Rest of WELM go ?
- Toroidal Asymmetries (probably No)
- Main Chamber (probably Yes)
- Transiently enhanced PRADELM (probably No)

- WELMdiv ~ 50 – 80 % of WELMdia

- JET – Type I ELM

*ITER*

- Despite narrow @ELM  WELMdiv/WELM ~ 0.6

## Effects of type I ELMs at the main chamber wall 

*ITER*

- Effects of type I ELMs at the main chamber wall

- Expect interaction with protruding surfaces (~1 m2)
- ==>  tolerable only 1-2 MJ depending on duration (if ≥ 1 ms no melting of W)

## Thermal Quench Disruptions

- If the JET results extrapolate to ITER then disruptions would not damage a W target.
- However, it is not known at the moment where and by what processes the missing and thermal an magnetic energies are deposited in the main chamber.
- If this energy deposition is not sufficiently uniform, then additional damage to main chamber components might be expected.

- Evolution of the surface temperature near
- and far from strike points in a Wth=5.6 MJ

**Thermal Quench Disruptions**
- Only a fraction of the energy reaches the divertor and is distributed rather uniformly.

*ITER*

**G. Matthews et al., 19th IAEA FEC 2002, Lyon**
**To appear in Nucl. Fusion.**

## ITER thermal quench specifications

*ITER*

**ITER thermal quench specifications**
- Need to be revisited on the basis of the new findings  (ongoing)

- If energy deposited in the divertor during disruptions is < 40% of the thermal energy with a broadening of the order 20-30 times, energy density at the target remains below the melting threshold for W.
- Some shallow melting can nevertheless take place some times.

- Case 1

- Case 2

- Concern remains on whether generation during ELMs/disruptions of surface irregularities in tungsten due to melting, and in CFC due to brittle destruction, might form hot spots during normal operation.

## Effects of ‘mitigated’ disruptionsAllowed time scale for energy dissipation (see talk of D. Whyte) 

- The thermal energy density in ITER (plasma stored thermal energy/ wall surface area) is ~350 MJ/800m2 = 0.45 MJ/m2.
- This energy density sets the minimum time in which the plasma thermal energy can be radiatively dissipated to the wall before melting/ablation occurs.

- The allowed time scales for ITER  w. Be first wall is closed to the limits set by the MHD time scale.
- Assuming uniform (spatial and temporal) dissipation tlim≥0.5 ms.
- Assuming a PF of 1.5-2 tlim≥2 ms.

*ITER*

**Effects of ‘mitigated’ disruptions**Allowed time scale for energy dissipation (see talk of D. Whyte)** **

*G. Federici, G. Strohmayer 3/2003*

**Melting as a function **
**of deposited energy**

## Parametric analyses of VDE effectsVDE parameters: 60 MJ·m-2, 300 ms, 1% UncertaintiesMelting - comparison Be and W

*ITER*

**Parametric analyses of VDE effects****VDE parameters: 60 MJ·m****-2****, 300 ms, 1%**** ****Uncertainties**Melting - comparison Be and W

| 10 mm |  | 2 mm |
| --- | --- | --- |
|  |  |  |

- W

- Be

- <10 MJ/m2 t≥500 ms avoid melting

- <10 MJ/m2 t>500 ms avoid melting

- <10 MJ/m2 t>100 ms avoid melting

- <10 MJ/m2 t>100 ms avoid melting

- If t=500 ms no melting up to 70 MJ/m2

- If t=500 ms no melting up to 30 MJ/m2

## Parametric analyses of VDE effectsTemp. excursions at Cu interface  - comparison Be and W

*ITER*

**Parametric analyses of VDE effects**Temp. excursions at Cu interface  - comparison Be and W

| 10 mm |  | 2 mm |
| --- | --- | --- |
|  |  |  |

- W

- Be

- TCu melting

- TCu melting

- Problems!!
- We need at least
- 5 mm W

## Future Work/ Open Questions3-5 years before starting PFC material procurement for ITER Current efforts: EU task force on PWI / ITPA work

- Type I ELM and disruption heat loads - energy loss, energy to divertor and wall, duration, broadening, impurities, etc.; (ITPA, JET TF, EU PWI TF)
- Mitigation of disruptions (ITPA, JET TF, EU PWI TF);
- Effect of mitigated disruptions on Be wall (ITPA);
- Material damage due to ELMs/disruptions (Russia, EU PWI TF);
- Erosion - mixed material effects (PISCES-B/EFDA)
- Explain co-deposition in existing machines (ITPA, JET & EU PWI TFs);
- Mitigation of co-deposition (temperature tailoring, reactive species (N2));
- Monitor C*x*H*y* deposits on cryopumps (JET, AUG);
- Carbon removal - e.g. baking with oxygen  - temperature, hold time (???);
- Dust generation - volume, location, BET, mobilisation.
- Plasma interaction with irregular and/or  molten W surfaces.
**Measurements/diagnostics**

**Future Work/ Open Questions**3-5 years before starting PFC material procurement for ITER** **Current efforts:** **EU task force on PWI / ITPA work

*ITER*

## SummaryOpportunities and challenges

- We take a big risk if we construct ITER without first testing/validating proposed material mix in an existing tokamak.
- We should reduce to a minimum the extent that ITER has to be a PMI experiment and explore all these effects before hand in existing machines.
- In case we retain CFC armour and use it during T operation a reliable method to remove the codeposted layers and control T-uptake is required.
- Type-I ELMs are still challenging for the ITER divertor and design optimisation is ongoing.  Power deposition on the first-wall is still uncertain.
- IF JET divertor results extrapolate to ITER, the large majority of disruptions would not lead to melting a W target. Compatibility of plasma with irregular W surfaces and macroscopic C erosion and impact on operation/performance require investigations.
- ITER divertor design has enough flexibility.  Feasibility of maintenance has been demonstrated. For the first-wall only infrequent maintenance is anticipated and feasibility demonstration is needed.

**Summary**Opportunities and challenges

*ITER*