1. 9th Hydrogen Workshop, Salamanca, June 2-3 1 Tritium retention buildup towards pulses in ITER PFCs and dust W.M. SHU, S. Ciattaglia and M. Glugla ITER Organization ITER TF on Tritium inventory Acknowledgement to ITER TF on Tritium inventory

2. 9th Hydrogen Workshop, Salamanca, June 2-3 2 partial-removed lids by FIB observations fully-removed lids by FIB fabrications big blisters small blisters New findings: two kinds of blisters formed on W by low energy D plasma Re-crystallized W; 38 eV, 10 26 D/m 2, around 500 K; W.M.Shu, Appl. Phys. Lett., 92, 211904 (2008). Blistering occurs at W for energy well below the displacement threshold. The lowest energy to produce Frenkel pair is 940 eV for D → W. For the small blisters, internal blister was a hole or pit, but the maximum height against diameter reached 0.7, which is one-order of magnitude greater than that reported before.

3. 9th Hydrogen Workshop, Salamanca, June 2-3 3 crack/void along grain boundary cross-section of a blister crack/void along grain boundary For most cases of big blisters, there was no hollow lid formed, but a crack/void at the grain boundary underneath the blister. New features of blisters By conventional definition, blisters are plastic dome-shaped buildings where a lenticular cavity is included between the blister lid and the bulk material. Re-crystallized W; 38 eV, 10 26 D/m 2, around 500 K; W.M.Shu, Appl. Phys. Lett., 92, 211904 (2008).

4. 9th Hydrogen Workshop, Salamanca, June 2-3 4 Various shapes of big blisters Re-crystallized W; 38 eV, 10 26 D/m 2, around 500 K; W.M.Shu, et al., PSI Conference. (d) (a) (b) (c)

5. 9th Hydrogen Workshop, Salamanca, June 2-3 5 Bursting release and retention ratio Busting release peaks were found in the TDS curve, indicating bursts of some blisters. There is a peak around 500 K. Retention ratio at 775 K and 380-440 K is smaller than 10 -7 and 5×10 -6, respectively. Re-crystallized W; 38 eV, W.M.Shu,et al., PSI Conference. 10 26 D/m 2, 500 K 2×10 26 D/m 2, 400 K

6. 9th Hydrogen Workshop, Salamanca, June 2-3 6 10 26 10 27 Fully-recrystallized W Partially- recrystallized W Annealed W Single W (111) 38 eV D ions, 315 K In comparison with the data of J. Roth et al. (ICFRM 2007) Smaller retention ratio was found in the higher fluence region at lower energy. W.M.Shu,et al., Nucl. Fusion 47 (2007) 201; Phys. Scr T128 (2007) 96.

7. 9th Hydrogen Workshop, Salamanca, June 2-3 7 Calculation by J. Roth et al. (ICFRM 2007) In the calculation, the retention ratio in W was assumed to be around 10 -3, due to their higher energy (200 eV) and lower fluence (max. 10 25 ions/m 2 ). 700 g limit (1000 g (limit in VV) – 120 g (in cryopump) – 180 g (others)) 750 discharges

8. 9th Hydrogen Workshop, Salamanca, June 2-3 8 Assumptions made in this calculation 1. Area, flux and temperature: (1) Divertor (strike points): 3 m 2, 1×10 24 DT atoms/m 2 /s, 775 K (2) Divertor (other target area except strike points): 47 m 2, 1×10 23 DT atoms/m 2 /s, 775 K (3) Divertor (others): 100 m 2, 1×10 22 DT atoms/m 2 /s, 775 K (not considered in [1]) (4) First wall: 700 m 2, 1×10 20 DT atoms/m 2 /s, 380-440 K (750 m 2 in [1]) 2. Retention ratio (retention against fluence) in W PFCs: (1) Divertor (at 775 K): 5×10 -7 [2] 3. Constant retention in Be due to implantation : 7×10 20 DT atoms/m 2 [3] 4. Breading in Be first wall : Tritium inventory I (appm) = 280F - 2350[1 - exp(-0.1F)]; [3] where F(MWa/m2): neutron fluence. 5. Sputtering yield of Be first wall : 4×10 -2 atoms/ions, half is dust [4] 6. Retention ratio of tritium in Be : 4×10 -2 [4] 7. Producing rate of W dust (700 kg in 10 6 s) : 2.3×10 21 atoms/s [5] 8. Retention ratio in W dust : 1×10 -6 [5] [1] J. Roth, et al., “Tritium Inventory in ITER: Laboratory data,” presented at the 1st meeting of ITER DCR 131 (In Vacuum Vessel Tritium Control), Oct.16, 2007. [2] W.M. Shu, et al., Fusion Eng. Des. (in press). [3] R.A. Anderl, et al., J. Nucl. Mater. 273, 1 (1999). [4] GSSR III [5] W.M. Shu and S. Ciattaqlia, internal discussion.

9. 9th Hydrogen Workshop, Salamanca, June 2-3 9 T inventory at case 1: full tungsten divertor The main contribution is from the Be first wall initially, but Be dust will be the controlling factor after 200 seconds. 700 g limit & codeposits

10. 9th Hydrogen Workshop, Salamanca, June 2-3 10 The averaged tritium retention estimated is 0.056 g T/discharge. In the calculation, averaged D-T flux at the first wall was assumed to be 7×10 22 DT atoms/s, the same as that used by Roth. However, Philipps [1] argued that the most recent value of the averaged D-T flux increased to 3-5×10 23 DT atoms/s. If the same assumptions are used, the averaged tritium retention will increase to 0.24-0.4 g T/discharge for the case of large wall flux. [1] V. Philipps, “T–retention from present experiments and further validation,” presented at the 4th meeting of ITER DCR 131 (In-Vacuum Vessel Tritium Control), March 12, 2008. Tritium retention at the case of large wall flux

11. 9th Hydrogen Workshop, Salamanca, June 2-3 11 Baking at 623 K to release major portion of tritium in Be codeposits D/Be (a) and O/Be (b) ratios for deposited material collected on Ta (grey symbols), Mo (dotted symbol) and W (white symbols) deposition probe coupons as a function of coupon temperature. M.J. Baldwin, et al., J. Nucl. Mater. 337- 339, 590 (2005).

12. 9th Hydrogen Workshop, Salamanca, June 2-3 12 Baking at 623 K of divertor after 1750-3000 discharges should be performed to release tritium from the Be dust that is located around divertor region. The DT/Be ratio could decrease from 4×10 -2 to less than 10 -2 after baking. Thus, the averaged tritium retention finally will be 0.06-0.1 g T/discharge if baking is taken into account. Tritium retention after baking at 623 K

13. 9th Hydrogen Workshop, Salamanca, June 2-3 13 1 st D-T year 2 nd D-T year 3 rd D-T year 4 th D-T year 5 th D-T year Equivalent accumulated nominal burn pulses [1] 7501750325057508750 Tritium inventory in Vacuum vessel 50 g110 g200 g350 g530 g [1] Project Integration Document PID, Jan. 2007, ITER Organization, Editor: J. How. Tritium buildup in the first 5 years’ operation ~ 0.06 g T/discharge

14. 9th Hydrogen Workshop, Salamanca, June 2-3 14 Permeation of tritium in CuCrZr (castellation) at 623 K If the transport properties of hydrogen in CuCrZr are the same as that in Cu, tritium permeation through CuCrZr pipes without W armor will reach the steady state within one hour. Permeation flux: =2  DLSP 1/2 /ln(d out /d in ) 3.7×10 -7 g-T/h for inner divertor; 5.7×10 -7 g-T/h for outer divertor; 9.4×10 -7 g-T/h in total. 0.09 mg in 100 h. Graph considers bulk diffusion only, not grain boundary diffusion or leakage.

15. 9th Hydrogen Workshop, Salamanca, June 2-3 15 Permeation of tritium in large SS pipes at 623 K The steady state will be reached in more than one month, and tritium permeation will be negligibly small in 100 hours’ baking. Permeation flux at steady state: =2  DLSP 1/2 /ln(d out /d in ) 1.3×10 -10 g-T/h In 100 hours’ baking: 7×10 -10 g-T in total. Graph considers bulk diffusion only, not grain boundary diffusion or leakage.

16. 9th Hydrogen Workshop, Salamanca, June 2-3 16 Permeation of tritium in small SS pipes at 623 K The steady state will be reached in one day, but tritium permeation will be negligibly small in comparison with that of CuCrZr. Permeation flux at steady state: =2  DLSP 1/2 /ln(d out /d in ) 2.8×10 -10 g-T/h In 100 hours’ baking: 3×10 -8 g-T in total. Graph considers bulk diffusion only, not grain boundary diffusion or leakage.

17. 9th Hydrogen Workshop, Salamanca, June 2-3 17 W divertor is always the major component for T retention. Tritium retention in continuous operation: 2 g /day (9 mg / discharge) 350 days for continuous operation 700 g limit T inventory at case 2: full tungsten PFCs

18. 9th Hydrogen Workshop, Salamanca, June 2-3 18 In comparison with that by J. Roth for the case of Full W PFCs The retained amount calculated by this work is smaller than that by Roth, because of the lower retention ratio in higher fluence region. 700 g limit

19. 9th Hydrogen Workshop, Salamanca, June 2-3 19 In the case of full W divertor and Be first wall, tritium in Be dust (including codeposits) will be the controlling factor after 200 s of discharge. The averaged tritium retention finally will be 0.06-0.1 g T/discharge for the case of large wall flux if baking is taken into account. If baking at 623 K is performed, permeation through CuCrZr pipes located at castellation will be predominant. Considering bulk diffusion only, the total permeation will be 0.09 mg in 100 hours’ baking. In the case of full W FPCs, the major contribution to the inventory is from the divertor, and the averaged tritium retention will be 9 mg/discharge (2 g/day for continuous operation). More accurate calculation should be performed by considering the effects of simultaneous H and He plasma on W blistering and dust producing. Summary

20. 9th Hydrogen Workshop, Salamanca, June 2-3 20 T inventory at case 3: CFC+W divertor 700 g (~700 discharges) & codeposits

21. 9th Hydrogen Workshop, Salamanca, June 2-3 21 Some issues related to in-vessel removal of T by oxidation Some issues related to in-vessel removal of T by oxidation  Highly tritiated water processing  DCR-140  Corrosion  highly tritiated water is very corrosive even to stainless steel due to the radiochemical formation of peroxides and radicals  Radiolysis and tritiated polymer formation  re-deposition and accumulation of tritiated polymers formed in the gas mixture of tritiated water vapour, tritium, CO and CO 2 is unavoidable  Oxidation of Be first wall  tritiated water moisture produced during oxidation may react with beryllium  Wall conditioning  implications for after-oxidation wall conditioning to be evaluated  Increased tritium retention in Be co-deposits  Oxidised Be codeposits are found to retain larger amount of T than pure Be codeposits  Evaluation of safety related issues (such as dust-related) required to determine compatibility with ITER safety requirements