J P Coad for TFFT 5 November 2009 Task-Force Fusion Technology Status and Main Achievements in 2009 Task Force – Fusion Technology

J P Coad for TFFT 5 November 2009 Tritium in the Tokamak (determine T trapped by PFCs such as tiles and flakes) Tritium Processes and Waste Management (detritiation of PFCs and Water / Detritiation of Waste Materials, Be, MS, Metals etc..) Plasma Facing Components (Erosion, deposition, characterisation of co-deposits) Engineering (Development of new Diagnostics) Neutronics & Safety (Measurement of Dose rates and modeling & Set-up databases of availability and reliability of various components as well as failure rates. Characterisation of hazardous materials, such as dust) Test Beds (AGHS, NB tests) JET TF-FT Outline of the Objectives NB. Main Objective of a FT task is, for instance, the assessment of a diagnostic for JET NOT its implementation in the machine (enhancement) One of the key objectives assigned to EFDA in 1999 was, by taking advantage of the JET facilities, to increase the Research and Development of Fusion Technology with the aim of preparing the way to ITER. For this purpose a dedicated Task Force Fusion Technology (TF-FT) was set up at JET in year Fusion Technology has been focused on six main areas of R&D

J P Coad for TFFT 5 November Tasks per year 132 tasks launched , Total resources ~ 21 m (~2.7 m in 2009) JET FT Overview SCK-CEN CEA IPP FZJ FZK MEdC UKAEA Associations VR ENEA TEKES NB. In 2008 Tasks only under Notifications In 2009 Notifications (~2.6m) and Orders (~110k) For tasks have been approved (~300k Orders)

J P Coad for TFFT 5 November It was confirmed that room temperature is sufficient to regenerate completely the cryopump from hydrogen isotopes. 2. It was confirmed that lighter hydrocarbons are effectively desorbed at 473 K. However, due to the shift reaction taking place inside the impurity processing unit, only the total number of hydrocarbons could be derived. 3. It is expected that, after regeneration at 473 K and moderate vacuum only, there is still a considerable amount of tritiated propane inside the panels. Cryopanel regeneration tests at 288 & 473 K (after Tritium Operation) Tritium Processes and Waste Management (JW2-FT-2.4) These results do not represent a problem for ITER, as ITER establishes the high temperature regeneration step at 475 K step at 10 Pa which is 2-3 orders of magnitude lower than the pressure used at JET during the regeneration of the PCP.

J P Coad for TFFT 5 November 2009 The task aimed at assessing the characteristics of the PCP ITER-relevant cryopanel after tritium contamination. Pumping tests performed at JET and the resulting curves were compared with the corresponding curves taken in 2004, before any tritium operation, when the PCP was freshly activated. Tritium Processes and Waste Management (JW2-FT-2.24) Pumping speed tests (after regeneration) Excellent agreement after 4 years of operation (most of the time exposed to tritiated gases) no indication for any degradation.

J P Coad for TFFT 5 November 2009 Molecular sieves are widely used in all Fusion Plants and JET An experimental program was set up to determine the optimal detritiation conditions for these molecular sieves. The detritiation was done by counter current regeneration at various temperatures using various regeneration media He, Ar, N 2 gas, He saturated with water vapour and hytec (95% Ar / 5% H 2 ) He and N 2 are the best options. Adding water vapour to the regeneration gas increase the slightly the detritiation but at the expense of extra secondary waste (HTO). Tritium Processes and Waste Management (JW5-FT-2.25) J. Braet / K. Dylst (SCK/CEN) Detritiation of Molecular Sieves (MS) Effect of the Temperature

J P Coad for TFFT 5 November 2009 Plasma Facing Components (JW6-FT-3.27) Cross-section of the JET Mk IISRP divertor showing the location of the analysed tiles. Objective: Investigate the 13 C transport in the SOL by analysing a complete poloidal set of divertor tiles (Mk IISRP, ) using IBA, SIMS and TOF-ERDA J. Likonen (TEKES) J.P. Coad (UKAEA) Deposits on tiles 1 and 3 contain high Be/C levels, indicating that vessel temperature did not explain the duplex layer on tiles removed in 2001, since the vessel wall temperature was 200°C throughout operations. 13 C puffed from the outer divertor (between tiles 7 and 8) at the end of C14 campaign was detected mainly near the strike point towards the top of tile 7 and on the apron of tile 8, whereas, deposition at the inner divertor tiles is relatively uniform toroidally. In general the 13 C amount is relatively small elsewhere and most of the puffed 13 C is still unaccounted for.

J P Coad for TFFT 5 November 2009 Up to 90μm of deposit was removed from the tile surface with flash-lamp Total amount of tritium released was 3GBq, ~12% of the total T inventory for the tile, and ~40% from the treated areas. Deuterium is depleted from the treated surface. Desorption to a depth of at least 7μm occurs beyond the depth of material removed from the surface. A low concentration of nickel and other metallic impurities was present in the film, which may accumulate at the surface during photon cleaning and slowdown the tritium release rate. Cross sections showing deposits on (a) untreated and (b) flash-lamp treated regions on Tile 4 Tungsten stripe 13 C injection 20mm Cores cut for SIMS analysis 13 C map on the shadowed area on tile 7. Deposition shows non-uniform coverage of 13 C. Total 13 C amount in this area is 2.2x10 20 atoms. Plasma Facing Components (JW6-FT-3.27) Material transport and erosion/deposition in the JET torus J.P. Coad /J. Likonen (UKAEA/TEKES) NRA D( 3 He,p) 4 He

J P Coad for TFFT 5 November 2009 Plasma Facing Components (JW6-FT-3.32) Objective: Investigate the 13 C transport in the SOL by analysing a complete poloidal set of divertor tiles (Mk IIHD, ) using IBA, SIMS and TOF-ERDA Tiles 1, 3 and 8 exposed during have been analysed by IBA and SIMS. Deposits on tiles 1 and 3 contain high Be/C levels indicating that carbon is removed by chemical erosion. Chemical sputtering does not occur up on the apron of tile 1. Tile 8 appeared clean all over, indicating it was in a net erosion zone. A LB-SRP (5) tile coated with 0.7 and 1.5 µm W stripes was exposed and has been analysed as a test of W-coating durability for the JET-ILW. The 0.7 µm film had been eroded from much of the area, but the 1.5 µm film had been thinned to virtually nothing in places but was still generally protecting the surface (see fig. next slide). J.P. Coad /J. Likonen (UKAEA/TEKES) Cross-section of the JET Mk IIHD divertor showing the location of the analysed tiles apron

J P Coad for TFFT 5 November 2009 LB-SRP tile (5) before (top) and after (bottom) exposure in Material transport and Erosion JW6-FT µm W 1.5 µm W Further analyses were made from MKIISRP outer divertor tiles and SRP tile that were exposed to 13 C puffing at the end of C14 campaign in Tiles 6 and 7 show a relatively uniform deposition (toroidally). In the case of tile 8 there is a strong variation toroidally. SRP tile has highest 13 C amount at the edges of the tile. 13 C distribution on divertor tiles following the 2004 puffing. Experimental results are indicated by symbols and solid line is the result of EDGE2D simulations (J. Strachan, PPPL) Arc Discharge coating by DIARC (Finland)

J P Coad for TFFT 5 November 2009 Plasma Facing Components (JW6-FT-3.33) LIBS spectrum after the 10th laser shot on the W-stripe zone with the deposited layer. H peak disappeared Carbon Be and W peak appear (J.-M.Weulersse, A. Semerok CEA) In situ Laser Induced Breakdown Spectroscopy LIBS spectrum after the first laser shot on the W-stripe zone with the deposited layer. C, H and Be peaks visible

J P Coad for TFFT 5 November 2009 Study of failure modes of 200 µm tungsten coatings The coating: 200 µm VPS W/ CFC with W/20.5μm Re multilayer interlayer (Plansee AG) Thermal tests of VPS-W coating in the electron beam facility JUDITH1 under: (i) Multiple ELM like transient thermal loads (~1000 pulses, 1ms) (ii) Steady State thermal loads (1 pulse, 2s) Bulk W tile W coatings Inner divertor Outer divertor JET ILW full W divertor: VPS-W/CFC and bulk W Objectives: failure modes and damage thresholds of VPS-W coatings under transient thermal loads Plasma Facing Components (JW6-FT-3.35) T. Hirai (FZJ)

J P Coad for TFFT 5 November 2009 N.B. As a result of numbers of observations, ILW Project made a decision to dismiss the VPS-W coating from the Project baseline. Alternatively, thinner PVD-W coating was developed and has been qualified for the Project. (i) ELM-like thermal loads: Surface damages, droplets and peeling of sub-layers of VPS-W layers. Damage threshold: < 0.33 GW/m 2 1ms. Sample dimensions: 100x300 mm 2 (ii) Steady state thermal loads: 113MJ/m 2 2s Melting of the multilayer due to reduction of melting points by alloying. (max temp. 3200°C) Damage threshold < 2825°C (W-Re) and/or 2715°C (W-C). Failure modes and damage thresholds of the VPS coatings Plasma Facing Components JW6-FT-3.35

J P Coad for TFFT 5 November 2009 The task was to make a comparison of the Experimental vs the Calculated dose during a non- operational period e.g. the shutdown of the JET machine ( ). This allowed the bench- marking of the two available 3-D methods (D1S, ENEA and R2S, FZK) for the prediction of the shutdown dose rate. TLD Dedicated measurements have been performed with GM tube outside vessel and TLDs close to the vessel. A satisfying agreement has been obtained between the calculations and measurements in the in-vessel position (within 20%), but both methods (D1S and R2S) underestimate the gamma dose rate measured for the external position (the agreement is within a factor 2 to 3). Neutronics and Safety (JW5-FT-5.20) Shutdown dose rate benchmark experiment at JET R. Villari/U.Fischer (ENEA, FZK) D1S

J P Coad for TFFT 5 November 2009 The discrepancies between calculation and experimental measurements for the external position indicate clearly that there are still deficiencies in the modeling mainly attributed to the uncertainties (related to the material impurities) of the metallic structures surrounding the detector. Modeling inadequacy can generate uncertainties on: the neutron spectra incident on the different components: overestimation of the shielding causes an underestimation of the induced activation; the gamma decay source: lacking structures, components not well described, etc… cause an underestimation of the activated nuclides thus of the decay gammas. Neutron flux measurements during operation and dose measurements after shut-down at the same detector location will be planned in the frame of the follow-up task JW8-FT-5.28 to resolve the severe discrepancies. CATIA 5 drawings including the missing material information (impurities and their relative concentration), are necessary in order to take into account all material information which is currently missing. Neutronic and Safety (JW5-FT-5.20)

J P Coad for TFFT 5 November 2009 Scope of the task was the collection of the work effort data (number of hours*number of persons) during some maintenance tasks foreseen in 2007 JET shutdown, in order to create a first data base useful to assess ORE for ITER maintenance tasks. Neutronics and Safety (JW5-FT-5.23) Monitored Activities KT1-H Spectrometer mirror chamber re-fitting Replacement of KK1 window for diagnostic Radiation monitoring Scaffolding mounting and dismounting Pipe/tube cutting and welding Pipe/tube insulation installation and removal Flange bolted sealing and unsealing Flange lip seal welding Component transportation in and out the maintenance zone Electric cables maintenance A comparison between the WEs JET data and those used in the ITER ORE assessment has been carried out. In average the JET and ITER WEs are of the same order and in most of the cases the values foreseen for ITER are even higher (for example window replacement: ITER 103 p-h, JET 76 p-h). This is because the ITER values are rather conservative, taking into account the uncertainties of the ITER systems layout. (M.T. Porfiri, ENEA)

J P Coad for TFFT 5 November 2009 Conclusions There are now a larger number of tasks on analysis of deposited films from JET(IBA, SIMS, Microprobe, XPS, etc), with more labs prepared to undertake some work with Be. It appears that neutron dose modelling is critically dependent on precise construction details (requiring huge CATIA models), or inadequate. The detritiation/disposal of waste in all forms is going to be a big issue for ITER. Several tasks are making contributions. TFFT is providing the pre-characterisation of JET ITER-Like Wall tiles, and will be responsible for post-mortem analysis Use of the unique JET facilities (AGHS, BeHF, Test-bed) for testing/development has proved very difficult over the years due to conflicts in priorities