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Prediction of ITER T retention levels with W PFCs

Prediction of ITER T retention levels with W PFCs. J. Roth, and the SEWG Fuel retention of the EU Task Force on Plasma-Wall Interaction with special contributions from R. Causey, R. Kolasinski, R. P. Doerner, G. Wright, J. Rapp, V. Alimov, O. Ogorodnokova. Overview:

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Prediction of ITER T retention levels with W PFCs

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  1. Prediction of ITER T retention levels with W PFCs J. Roth, and the SEWG Fuel retention of the EU Task Force on Plasma-Wall Interaction with special contributions from R. Causey, R. Kolasinski, R. P. Doerner, G. Wright, J. Rapp, V. Alimov, O. Ogorodnokova • Overview: • Estimates in support of the ITER Design Review • New input data • Special emphasis on W divertor and W walls • Improved estimates for present choice of wall materials

  2. Inventory estimates for ITER • On the basis of the standard scenario for wall fluxes • Present choice of wall materials • Comparison of all-C, all-W and W/Be material choice CFC in divertor dominates tritium retention Be co-deposition needs to be taken into account W not limiting, n-effects must be assessed

  3. Inventory estimates for ITER • On the basis of the standard scenario for wall fluxes • Present choice of wall materials • Comparison of all-C, all-W and W/Be material choice All-CFC reaches T limit in about 25 shots All-W eliminates tritium problem, but n-effects need to be considered For W/Be option co-deposition with Be dominates

  4. New input data • First indication of saturation of implantation/retention inventory in CFC (R. Pugno et al. JAP, accepted 2008). Here emphasis on W • Renewed assessment of D/T fluxes to different wall areas indicate about a factor of 3±2 higher wall fluxes. Previous values from Kukushkin (R. Behrisch et al, JNM 2003) are still within uncertainty (see presentation by A. Kallenbach). • W as divertor material is irradiated at very high flux and fluence. Surface modifications (see presentations by R. Doerner, M. Mayer). Retention data are being produced at PILOT (see presentation by Graham Wright). What are the consequences? • n-irradiation of W creates potential tritium traps in the bulk of the material and may increase the retention. There are two preliminary attempts of modelling (R. Causey and R. Kolasinski, O. Ogorodnikova)

  5. New input data for W • Flux and temperature dependence of retention • Influence of material structure and n-irradiation Normalised to 2x1024/m2

  6. New input data for W • Flux and temperature dependence of retention • Influence of n-irradiation Normalised to 2x1024/m2 • Inventory reduced both, at higher flux and higher temperature • Higher temperature due to higher fluxes

  7. New input data for W • Flux and temperature dependence of retention • Influence of n-irradiation O. Ogorodnikova, J. Roth, M. Mayer, JAP (in print)

  8. New input data for W • Flux and temperature dependence of retention • Influence of n-irradiation • n-irradiation will create additional traps for tritium in the bulk • input parameter for diffusion/trapping codes O. Ogorodnikova, PSI Gifu, 2002

  9. Estimates including flux, temperature and n-irradiation O. Ogorodnikova, ITPA SOL/DIV (2007) R. Causey, R. Kolasinski, piv.com. (2007) • Assumptions: • EU • Ratio of tritium trap density vs. dpa 10-2 • (for Mo 2x10-3, see: M. Eldrup, ICFRM 2007) • Saturation value of 1% at 0.6 dpa • (for Mo 6x10-4, see: M. Eldrup, ICFRM 2007) • US • 0.6% trap concentrations due to n-irradiation • Present at all fluences

  10. Estimates including flux, temperature and n-irradiation O. Ogorodnikova, ITPA SOL/DIV (2007) R. Causey, R. Kolasinski, piv.com. (2007) • Urgently needed: • Correlation of tritium trap density with dpa • (for Mo 2x10-3, see: M. Eldrup, ICFRM 2007) • Experimental investigations on n-irradiated materials • Is trapped tritium in the bulk mobilisable?

  11. Inventory estimates for ITER • On the basis of the standard scenario for wall fluxes • Present choice of wall materials • Reduction of retention in W by a factor of 2 compared to earlier assessment • Co-deposition negligible, even including impurity sputtering • W walls not limiting, n-effects must be assessed

  12. Conclusions • Tritium Inventory • Present choice of materials (CFC/W/Be): • Co-deposition with carbon will dominate (0.5-2 g per shot). • Co-deposition with Be adds to the total inventory (0.1-0.6 g per shot) • Implantation of T in W adds a small contribution and is not limiting • Comparison of proposed material options: • W divertor/Be wall: 350 g T reached after >2500 discharges • Co-deposition with Be dominates • W area with highest inventory are medium-flux divertor plates • all-W machine: 350 g Treached after >10000 discharges • n-irradiation will increase retention • Important Issues: • Is the plasma compatible to W as divertor material? • Where is Be deposited? • How can T be removed from co-deposited Be layers?

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