tritium retention buildup towards pulses in iter pfcs and dust
DESCRIPTION
Tritium retention buildup towards pulses in ITER PFCs and dust. W.M. SHU , S. Ciattaglia and M. Glugla ITER Organization Acknowledgement to ITER TF on Tritium inventory. fully-removed lids by FIB fabrications. big blisters. small blisters. partial-removed lids by FIB observations. - PowerPoint PPT PresentationTRANSCRIPT
9th Hydrogen Workshop, Salamanca, June 2-31
Tritium retention buildup
towards pulses in ITER PFCs and dust
W.M. SHU, S. Ciattaglia and M. GluglaITER Organization
Acknowledgement to ITER TF on Tritium inventoryITER TF on Tritium inventory
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 blistersformed on W by low energy D plasma
Re-crystallized W; 38 eV, 1026 D/m2, 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.
9th Hydrogen Workshop, Salamanca, June 2-3 3
crack/void along grain boundary
cross-section of a blister
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, 1026 D/m2, around 500 K; W.M.Shu, Appl. Phys. Lett., 92, 211904 (2008).
9th Hydrogen Workshop, Salamanca, June 2-3 4
Various shapes of big blistersVarious shapes of big blistersRe-crystallized W; 38 eV, 1026 D/m2, around 500 K; W.M.Shu, et al., PSI Conference.
(d)
(a) (b)
(c)
9th Hydrogen Workshop, Salamanca, June 2-3 5
Bursting release and retention ratioBursting release and retention ratio
0
1
2
3
4
5
6
7
300 400 500 600 700 800 900 1000 1100
Rel
ease
rate
(10
17 D
2/m2 /s
)
Temperature (K)
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.
0
2
4
6
8
10
12
14
16
300 400 500 600 700 800 900 1000
Rel
ease
rate
(10
17 D
2/m2 /s
)
Temperature (K)
300 400 500 600 700 800
1019
1020
1021
1022
Deu
teriu
m re
tain
ed (
D/m
2 )
Exposure temperature (K)
= 1x1026 D/m2 (TDS) = 1x1026 D/m2 (NRA, 0-7 m)
38 eV D plasma W1026 D/m2, 500 K
2×1026 D/m2, 400 K
9th Hydrogen Workshop, Salamanca, June 2-3 6
1026 1027
Fully-recrystallized W
Partially- recrystallized W
Annealed W
Single W (111)
38 eV D ions, 315 K
In comparison with the data of In comparison with the data of J. Roth J. Roth et al. et al. (ICFRM 2007)(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.
9th Hydrogen Workshop, Salamanca, June 2-3 7
0,1 1 10 100 1000 10000 100000 10000001x1021
1x1022
1x1023
1x1024
1x1025
1x1026
1x10272,5E-4 0,0025 0,025 0,25 2,5 25 250 2500
Total retention800m2 ITER mix
CFC implantation3m2, 1x1024/m2s47m2, 1x1023/m2s750 K
Be co-depositionDIVIMP calcul.D/Be=0.05
Be implantation700m2, 1x1020/m2s
C co-depositionfull ERO calcul.S=0D/C=0.4
number of 400s ITER dischargesR
etai
ned
amou
nt (a
tom
s)
Time (s)
W implantation100m2, 1x1020/m2s
350 g T limit
Calculation by J. Roth Calculation by J. Roth et al.et al. (ICFRM 2007) (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. 1025 ions/m2).
700 g limit (1000 g (limit in VV) – 120 g (in cryopump) – 180 g (others))
750 discharges
9th Hydrogen Workshop, Salamanca, June 2-3 8
Assumptions made in this calculationAssumptions made in this calculation
1. Area, flux and temperature:(1) Divertor (strike points): 3 m2, 1×1024 DT atoms/m2/s, 775 K(2) Divertor (other target area except strike points): 47 m2, 1×1023 DT atoms/m2/s, 775 K(3) Divertor (others): 100 m2, 1×1022 DT atoms/m2/s, 775 K (not considered in [1])(4) First wall: 700 m2, 1×1020 DT atoms/m2/s, 380-440 K (750 m2 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: 7×1020 DT atoms/m2 [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 106 s): 2.3×1021 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.
9th Hydrogen Workshop, Salamanca, June 2-3 9
T inventory at case 1: full tungsten divertorT 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.
1014
1016
1018
1020
1022
1024
1026
10-1 100 101 102 103 104 105 106 107
TotalBe dustW divertorBe (implantation)Be (transmutation)W dustR
etai
ned
triti
um (a
tom
s)
Discharge time (s)
Number of 400 s ITER discharge 2.5 25 250 2500 25000
700 g limit
& codeposits
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 7×1022 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 3-5×1023 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 fluxTritium retention at the case of large wall flux
9th Hydrogen Workshop, Salamanca, June 2-3 11
Baking at 623 K to release major portion of Baking at 623 K to release major portion of tritium in Be codepositstritium 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).
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 KTritium retention after baking at 623 K
9th Hydrogen Workshop, Salamanca, June 2-3 13
1st D-T year
2nd D-T year
3rd D-T year
4th D-T year
5th D-T year
Equivalent accumulated nominal
burn pulses [1]750 1750 3250 5750 8750
Tritium inventory in Vacuum vessel
50 g 110 g 200 g 350 g 530 g
[1] Project Integration Document PID, Jan. 2007, ITER Organization, Editor: J. How.
Tritium buildup in the first 5 Tritium buildup in the first 5 years’ operationyears’ operation
~ 0.06 g T/discharge
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Permeation of tritium in CuCrZr (castellation) at 623 KPermeation of tritium in CuCrZr (castellation) at 623 K
0
0.2
0.4
0.6
0.8
1
0 5x102 1x103 1.5x103 2x103
Rel
ativ
e pe
rmea
tion
Time (s)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:=2DLSP1/2/ln(dout/din)
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.
9th Hydrogen Workshop, Salamanca, June 2-3 15
Permeation of tritium in large SS pipes at 623 KPermeation 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 fluxat steady state:
=2DLSP1/2/ln(dout/din) 1.3×10-10 g-T/h
In 100 hours’ baking:7×10-10 g-T
in total.0
0.2
0.4
0.6
0.8
1
0 5x105 1x106 1.5x106 2x106 2.5x106 3x106
0 4 8 12 16 20 24 28 32
Rel
ativ
e pe
rmea
tion
Time (s)
Time (day)
Graph considers bulk diffusion only, not grain boundary diffusion or leakage.
9th Hydrogen Workshop, Salamanca, June 2-3 16
Permeation of tritium in small SS pipes at 623 KPermeation 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 fluxat steady state:
=2DLSP1/2/ln(dout/din) 2.8×10-10 g-T/h
In 100 hours’ baking:3×10-8 g-T
in total.0
0.2
0.4
0.6
0.8
1
0 2x104 4x104 6x104 8x104 1x105
0 0.2 0.4 0.6 0.8 1
Rel
ativ
e pe
rmea
tion
Time (s)
Time (day)
Graph considers bulk diffusion only, not grain boundary diffusion or leakage.
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)
1017
1019
1021
1023
1025
1027
10-1 100 101 102 103 104 105 106
W divertorW First wallW dust
Ret
aine
d am
ount
(ato
ms)
Time (s)
350 g T limit
Number of 400 s ITER discharges 2.5 25 250 2500
350 days for continuous operation
700 g limit
T inventory at case 2: full tungsten PFCsT inventory at case 2: full tungsten PFCs
9th Hydrogen Workshop, Salamanca, June 2-3 18
In comparison with that by J. Roth In comparison with that by J. Roth for the case of Full W PFCsfor 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.
1021
1022
1023
1024
1025
1026
1027
10-1 100 101 102 103 104 105 106
Ret
aine
d am
ount
(ato
ms)
Time (s)
350 g T limit
Number of 400 s ITER discharges 2.5 25 250 2500
By J. Roth
This work
700 g limit
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• 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.
SummarySummary
9th Hydrogen Workshop, Salamanca, June 2-3 20
T inventory at case 3: CFC+W divertorT inventory at case 3: CFC+W divertor
1013
1015
1017
1019
1021
1023
1025
1027
10-1 100 101 102 103 104 105 106 107
TotalC codepositsBe dustBe (implantation)Be (transmutation)W divertorW dust
Triti
um re
tent
ion
(ato
ms)
Time (s)
Number of 400 s ITER discharge 2.5 25 250 2500 25000
700 g (~700 discharges)
& codeposits
9th Hydrogen Workshop, Salamanca, June 2-3 21
Some issues related to in-vessel Some issues related to in-vessel removal of T by oxidationremoval of T by oxidation
Highly tritiated water processing Highly tritiated water processing DCR-140 DCR-140 Corrosion Corrosion highly tritiated water is very corrosive even to stainless steel highly tritiated water is very corrosive even to stainless steel
due to the radiochemical formation of peroxides and radicalsdue to the radiochemical formation of peroxides and radicals Radiolysis and tritiated polymer formation Radiolysis and tritiated polymer formation re-deposition and re-deposition and accumulation of tritiated polymers formed in the gas mixture of tritiated accumulation of tritiated polymers formed in the gas mixture of tritiated water vapour, tritium, CO and COwater vapour, tritium, CO and CO22 is unavoidable is unavoidable Oxidation of Be first wall Oxidation of Be first wall tritiated water moisture produced during tritiated water moisture produced during oxidation may react with berylliumoxidation may react with beryllium Wall conditioning Wall conditioning implications for after-oxidation wall conditioning to be implications for after-oxidation wall conditioning to be
evaluatedevaluated Increased tritium retention in Be co-deposits Increased tritium retention in Be co-deposits Oxidised Be codeposits Oxidised Be codeposits
are found to retain larger amount of T than pure Be codepositsare found to retain larger amount of T than pure Be codeposits Evaluation of safety related issues (such as dust-related) required to Evaluation of safety related issues (such as dust-related) required to determine compatibility with ITER safety requirementsdetermine compatibility with ITER safety requirements