9th hydrogen workshop, salamanca, june 2-3 1 tritium retention buildup towards pulses in iter pfcs...

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.


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).


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)


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

Rele

as

e r

ate

(1

01

7 D2/m

2 /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

Rele

as

e r

ate

(1

017 D

2/m

2 /s)

Temperature (K)

300 400 500 600 700 800

1019

1020

1021

1022

De

ute

riu

m r

eta

ined

(D

/m2 )

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


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.


0,1 1 10 100 1000 10000 100000 10000001x1021

1x1022

1x1023

1x1024

1x1025

1x1026

1x1027

2,5E-4 0,0025 0,025 0,25 2,5 25 250 2500

Total retention

800m2 ITER mix

CFC implantation

3m2, 1x1024/m2s

47m2, 1x1023/m2s750 K

Be co-depositionDIVIMP calcul.D/Be=0.05

Be implantation

700m2, 1x1020/m2s

C co-depositionfull ERO calcul.S=0D/C=0.4

number of 400s ITER dischargesR

eta

ine

d a

mou

nt (

ato

ms)

Time (s)

W implantation

100m2, 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


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.


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

Total

Be dust

W divertor

Be (implantation)

Be (transmutation)

W dustRet

ain

ed

tri

tiu

m (

ato

ms

)

Discharge time (s)

Number of 400 s ITER discharge 2.5 25 250 2500 25000

700 g limit

& codeposits


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


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).


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


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


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

ati

ve p

erm

eati

on

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.


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

ati

ve p

erm

eati

on

Time (s)

Time (day)



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

ati

ve p

erm

eati

on

Time (s)

Time (day)



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

ain

ed

am

ou

nt

(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


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

ain

ed

am

ou

nt

(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


• 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


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

Tri

tiu

m r

ete

nti

on

(at

om

s)

Time (s)

Number of 400 s ITER discharge 2.5 25 250 2500 25000

700 g (~700 discharges)

& codeposits


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

9th hydrogen workshop, salamanca, june 2-3 1 tritium retention buildup towards pulses in iter pfcs...

Documents