EU-PWI TF meeting, Ljubljana, 13-15 November 2006 Progress with tritium removal and mitigation...

EU-PWI TF meeting, Ljubljana, 13-15 November 2006 EU Plasma-Wall Interactions Task Force

Progress with tritium removal Progress with tritium removal and mitigation 2005-6and mitigation 2005-6G Counsell1, D Borodin4, P Coad1, J Ferreira8, C Grisolia2, C Hopf3,W Jacob3, A Kirschner4, A Kreter4, K Krieger3, J. Likonen5, A. Litnovsky4,A. Markin3, V Philipps4, J Roth3, M Rubel6, E Salancon3, A. Semerok7,FL Tabares8, C. Tomastik9, A Widdowson1

1EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK2Association EURATOM-CEA, CEA/DSM/DRFC Cadarache, 13108 St.Paul lez Durance, France3Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany4Institut für Plasmaphysik, Forschungszentrum Jülich, Association EURATOM-FZJ5Association EURATOM-TEKES, VTT Processes, PO Box 1608, 02044 VTT, Espoo, Finland6Alfvén Laboratory, Royal Institute of Technology (KTH), Association EURATOM-VR, 100 44 Stockholm, Sweden7CEA Saclay, DEN/DPC/SCP/LILM, Bat. 467,91191Gif sur Yvette, France8Association Euratom/Ciemat. Laboratorio Nacional de Fusión. 28040 Madrid, Spain9Institut für Allgemeine Physik, Vienna University of Technology, Wiedner Hauptstraße 8-10, A-1040 Wien, Austria

2/28

EU Plasma-Wall Interactions Task ForceOutline

A reminder of the problem

Advances in modelling tritium retention with mixed materials

‘New’ challenges – retention in gaps, under layers and in the bulk

Progress with characterising de-tritiation techniques and associated problems

Can we avoid retention in the first place?

Future thoughts

3/28

EU Plasma-Wall Interactions Task Force

• Acceptable ITER operation ~2500 shots before maintenance period

Long term T retention/shot must be< 0.14g/400s shot

Strategies for T removal essential ifCFC targets in DT phase

Removal efficiency must be 80% - 98%

Clear need for T removal schemes

4/28


Preliminary modelling using local erosion & deposition model ERO (still many open questions)

Model assumptions validated against TEXTOR C13 injection experiments

• >80% of wall area in ITER is beryllium

• Eroded Be will transport to divertor (as ions)

modify erosion and co-deposition

Be concentration in plasma plays key role Balance of inc. Be coverage on target (dec.

C erosion) and inc. C erosion due to Be flux For a range of Be conc., T/C and T/Be ratios

0.5g – 6.4gT/400s shot

Be transport will impact C erosion

5/28


Assumptions: Maximum possible stochiometric combination of Be2C formed No chemical erosion for Be2C

Leads to non-linear evolution of relaxation time for reduction in chemical erosion- as observed

Be2C formation potentially significant

ERO modelling of erosion mitigation with Be seeding in PISCES

6/28


• All ITER plasma facing components will be castellated

• >2,000,000 Gaps in ITER (typ. 0.5-1mm x 10mm)

• Increases plasma exposed areas by factor 2 - 5

aC:H co-deposits form in tile gaps

CxHy molecules and radicals form a:C:H co-deposits deep in gaps – how much and how deep is on-going research

CFC tile segments from JET Mk1 divertor, 6mm gaps

Retention in gaps twice that on plasma-facing surfaces (protected from re-erosion)

CFC target (90,000 monoblocks): 50 m2 215 m2

W baffle & dome (1.2M rods): 100 m2 460 m2

Be main wall (300,000 tiles): 680 m2 1290 m2

ITER mock-up

7/28

EU Plasma-Wall Interactions Task ForcePotential for significant T inventory

TZM castellated monoblocks exposed to plasma for 200s in TEXTOR

D retention fall-off dependent on D and Ttile

Dgap ~ 0.4% - 4% of D, between low and high D at Ttile ~200- 260C

Factor 10 decrease in Dgap, 30 Ttile 200C

• Extrapolation to ITER based on D from B2-EIRENE modelling (Kukushkin, 2005): 0.5 – 5gT/400s shot

• Maybe other factors, however:

strong function of gap width carbon source (local or remote) period of exposure

TZM monoblocks from TEXTOR, 0.5mm gaps

8/28


Planar and imbricated castellated tungsten blocks exposed in TEXTOR

Co-deposition significantly reduced for imbricated blocks

Also of note -• Metal intermixing noted in deposits• Some deposits ‘buried’ under W layer

Shaped castellations may help

9/28


N11

ToreSupra:

• 75-85% D retention in short shots (<30s)

• Up to 100% D retention in long shots (>100s)

• Retention in short shots easily recovered by He glow

• Measurements of C erosion suggest co-deposition alone may not explain retention

more than 1 mechanism?

Retention in bulk CFC being considered for high fluence conditions

Lab studies indicate D retention to several m in bulk

D inventory fluence0.5

Calcs. suggest this may initially exceed co-deposition in Tore Supra

D/T retention in CFC bulk

could affect choice of CFC for ITER

• Flux and time dependence needs more study

10/28


• Tritium trapped in aC:D/T co-deposits • Oxidation an obvious candidate for detritiation through the reaction :

aC:D/T + O COx + DTO:D2O:T2O

• In-situ – no need for vessel entry

• Volatile products pumped from vessel

• Several schemes under investigation:

Baking in O2

ECR or ICR -wave plasma in O2 or He/O2 mix

DC Glow discharge cleaning in He/O2 mix

• Studies on-going in both laboratory and tokamak environments and both laboratory produced and tokamak co-deposited films

T-removal through oxidation

11/28


RF Antenna protection tile from TEXTOR180 m thick aC:D co-deposit

O2 baking efficient, but at high Twall

Molecular chemistry – O2 penetrates all regions of deposition but ….

• Low D removal efficiency below ~300C (cf ITER wall bakeout temp 240C)

• High O2 pressure needed for high removal efficiency

• Co-deposit not fully removed – becomes flaky and peels off

Inhibited O penetration and release of volatiles due to carbide formation with impurities? WC and BeC may form in ITER …

• O3/O2 mix effective at <200C and low pressure but damage to bulk CFC seems to be too high

800m

12/28

EU Plasma-Wall Interactions Task ForceO-Plasma – effective at room temp

Erosion rate, E

• increases with Tsurf chemical reactions• and with bias volts collisions

2 step process: surface damage by ion bombardment then chemical erosion

ECR plasma in 100% O2

Products: CO, CO2, H2, H2O

He/O2 mixture:

• E limited by He ion flux at high %O2

E saturates above few %O2

13/28

EU Plasma-Wall Interactions Task ForceHe/O GDC in the tokamak environment

Asdex Upgrade:49h, 25g removed, 7x1018 C-at/s E ~ 1.4x1017 C-at m-2s-1

TEXTOR:3h, 5.2g removed, 2x1019 C-at/s E ~ 5.7x1017 C-at m-2s-1

i.e. 0.075 - 0.3g T/h over 150m2 • CO and CO2 dominant• T2O 30 times higher than He GDC• Production saturates at low O %

No removal in boronised regions:B-coated sample coupons and boron coated co-deposited tiles unaffected- Impact of WC, BeC in ITER?

or from shadowed areas:a:C:H coated samples behind first wall, deep in divertor untouched

• Tokamak and Lab studies less clear on removal from tile gaps

200nm and 350nm soft aC:H deposit on Si coupons in Asdex Upgrade

Arcing and pitting occurred on boron dominated surface layers

Cleaning not uniform on all surfaces

14/28


Hydrocarbon coated elements at base of castellated structure exposed to He/O2 discharge

2 orientations:

• Directly exposed• Facing bottom of the chamber (shadowed from plasma)

Cleaning rate nearly identical in both orientations

Suggests atomic oxygen survives several collisions with the walls.

Cleaning in shadowed areas

• O+/O may penetrate several mm into sufficiently wide gaps

Castellation and tile gap design may be important for ITER

15/28

EU Plasma-Wall Interactions Task ForceImpurities in aC:H reduce efficiency

Fully removed with 6.25 hours lab GDCi~2.5x1018 m-2s-1, 8mbar, 20% O2 in He

Similar E to tokamak GDC

E for tokamak co-deposits up to factor 10 less than for laboratory produced

80% co-deposit eroded during first 20% of plasma exposure

• Tokamak produced aC:H co-deposits on W substrate efficiently cleaned in He/O discharge

effect of impurities in co-deposit building up at surface?

• W, Be will mix with aC:H in ITER

16/28

EU Plasma-Wall Interactions Task ForceCollateral damage and recovery OK

Not all injected O2 is pumped out of the vessel during GDC

• Some O retained in metal oxides

• O retention higher at larger Vbias

opportunity for optimisation

• H2 discharge effective at removing oxides

TEXTOR Recovery: 66h H2 GDC, 0.5h He GDC & boronisation

Asdex Upgrade Recovery: 72h baking at 150C, 10h He GDC & boronisation)

Will recovery extrapolate to BeO?

17/28

EU Plasma-Wall Interactions Task ForceBeO formation a challenge to oxidation

• Beryllium samples pre-oxidised to variety of oxide thicknesses (20-100 nm)

• Samples exposed to hydrogen plasma for 5 - 6 hours at 300 °C - 600 °C

• Residual oxide layer thickness always ~25 nm

• Thick layers are reduced but thin layers actually increase

Development of an equilibrium state with the plasma

• Oxide thickness depends on the plasma conditions, especially

oxygen impurities

Complex (yet to be determined) consequences for formation and impact of oxygen based de-tritiation schemes

18/28


N2 injection into Asdex Upgrade sub-divertor

Factor 5 reduction in aC:H net co-deposition rate No significant N retention Effect not seen with Ar (laboaratory studies)

• ‘Scavenging’ proposed as one mechanism– moping-up of reactive radical pre-cursors

• But also alternative explanations -

• Synergistic interaction of H and N at surface – peaks at ~ 75%:25%

• Erosion rates high in H2/N2 plasmas

E up to 1m/hour for lab deposits

in ECR plasma

less than O, but not optimised

Alternative chemistry may have a role

19/28


Without CH4:

Chemical sputtering dominates HCN molecules are formed and released

With CH4:

N ions release nitrogenated compounds from surface – react with cracked CH4

C2H2 dominates, little HCN

Role of atomic N as yet unknown

Alternative chemistry may have a role

• Cryo-trapping assisted mass spectroscopy

• Quantitative analysis of compounds formed during discharge cleaning

• Pre–deposited hydrocarbon films exposed to N2/H2 and CH4/N2/H2 plasma glow

20/28

EU Plasma-Wall Interactions Task ForceT-removal through ‘photonic cleaning’

• aC:H co-deposits have poor thermal conductivity compared to substrates (CFC, Be, W)

• Surface heat flux leads to rapid temperature rise in co-deposit ablation or chemical ‘bond-breaking’

• Two ‘photonic cleaning’ schemes under investigation:

LASER Flash-lamp

• Requires vessel access, but can operate in high magnetic fields and in vacuuo, inert gas or atmospheric conditions

• Studies on-going in both laboratory and tokamak environments and both laboratory produced and tokamak co-deposited films

21/28


100mm2 in 2s with 20W Ytterbium fibre laser (1060nm, 120ns, 20kHz), 2J/cm2 on 250m spot @ 40cm

0.03 - 0.3g T/h over 150m2

• No difference between active and inert gas environment

Laser cleaning of TEXTOR tile• Energy density threshold for removal

• Threshold factor 5 lower for co-deposit compared to graphite

selective removal

22/28


Galvo-scanning fibre laser developed for use in JET Be handling facility

Laser cleaning in JET BeHF

Tritium release only accounts for ~10% of tritium in cleaned region

micro-particulate?

• >200 mm of co-deposit easily removed by single scan at low scan rate (40s) and pitch

• At fastest rate (4s), not all deposit removed even with 4 passes

23/28


JET 2004 trial showed engineering feasibility of flash-lamp technology

Flash-lamp cleaning of tritiated aC:H

• Trials now conducted using flash-lamp in JET berylium handling facility

• Aim to clean thick, tritiated co-deposit from inner divertor CFC tile

Photon flux from 500J, 140s flash-lamp

3.6MW Rep. rate 5Hz Focused using semi-elliptical cavity – Footprint ~30cm2 @ 30mm

375MWm-2, 6J/cm2

• Three positions treated (with varying co-deposit thickness and tritiation)

• Tritium release monitored and tile sent for SIMS/IBA/SEM analysis

24/28

EU Plasma-Wall Interactions Task Force1st demonstration of T-removal

• Total T release ~9g.

• Decreasing efficiency with number of pulses

• 40% of T inventory & 70-90 m co-deposit, removed (off gas & SEM)

0.075g T/h over 150m2

Untreated Treated

7m de-tritiation at surface of treated zone

Consistent with FE calcs of bulk heating above 700K

Build-up of Ni at surface explanation for roll-over of tritium release/pulse? (similar

results for Be on other treated tiles)

25/28

EU Plasma-Wall Interactions Task ForceCould prevention be the best cure?

• Hot liner added to PSI-2 linear plasma device

• CH4/H2 added to H2 plasma column

Cold liner – CH4 decomposes to form co-deposted layers on liner walls

Hot liner – co-deposit re-eroded to form CH4, C2H2 and heavier species

- ~100% C pumped through duct or returned to main chamber

Relies on presence of sufficient atomic hydrogen (C2H2 production saturates at high CH4 flows)

26/28

EU Plasma-Wall Interactions Task ForceThe year that was ….

Significant advances during 2005-6 but also growing recognition of new challenges -

Improvements in modelling of retention and co-deposition, including with beryllium fluxes

Importance of retention in gaps, castellations and bulk CFC now better understood and characterised

More machines prepared to run with oxidising plasmas but difficulties of shadowed regions, mixed materials and oxide formation now clear

Further studies into alternative chemistry (apart from oxygen)

Technological approaches continue to develop and first trials with tritiated samples

A growing understanding that ITER may require a combination of techniques to operate within the inventory limit

27/28


• No single T-removal scheme likely to be sufficient – let’s not close any doors

• Integration of different schemes on different timescales will probably be required – the ‘good housekeeping’ approach

Big ‘toolbox’ needed for ITER

Example of T-removal integrated into ITER operating schedule Extrapolated from predicted/measured T-removal rates allowing for future optimisation

28/28


Characterise the impact of Beryllium Carbide formation in ITER relevant conditions

CFC bulk retention – is it real? Need for confirmation of the effect and improved characterisation (impact of surface temperature, morphology etc.)

Tile gaps – how does deposition in gaps scale the ITER? Resolve issue of shadowing on plasma removal schemes (e.g. how many wall collisions can the chemically ‘active’ atoms survive)

Demonstrate impact of repetitive oxidising plasmas and plasma recovery on Beryllium surfaces

Explore further the impact of surface temperature on deposition in ITER relevant conditions

Begin to develop a serious scheme for operating ITER with a CFC/Be/Wmaterials mix in the DT phase

Start to characterise retention in an all metal ITER – is it a problem or not?

Key issues for the near future

How can the EU-PWIand the SEWG help?

29/28


Not a comprehensive list! -

G Counsell et al 33rd EPS, 11th PFMCD Borodin et al 11th PFMCP Coad et al 17th PSIJ Ferreira et al 11th PFMCC Grisolia, et al 24th SOFT, 17th PSIC Hopf, et al 17th PSIW Jacob, et al 17th PSIA Kirschner, et al 17th PSIA Kreter, et al 17th PSIK Krieger, et al 17th PSIJ. Likonen, et al 17th PSIA. Litnovsky, et al 11th PFMCA Markin, et al 11th PFMCV Philipps, et al 17th PSIJ Roth, et al 17th PSIM Rubel, et al 17th PSI, 21st IAEAE Salancon, et al 17th PSIA. Semerok, et al 21st IAEAFL Tabares et al 17th PSI, 21st IAEAC. Tomastik, et al 11th PFMCA Widdowson, et al 17th PSI

Many presentations this year …..

EU-PWI TF meeting, Ljubljana, 13-15 November 2006 Progress with tritium removal and mitigation...

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