TRITIUM REMOVAL BY LASER HEATING C.H. Skinner, C. A. Gentile, G. Guttadora, A. Carpe, S. Langish, K.M. Young, M. Nishi (b) , W. Shu (b) , N. Bekris (c) (a) Princeton Plasma Physics Lab, Princeton NJ 08543, USA (b) Tritium Engineering Laboratory, JAERI, Ibaraki, Japan (c) Tritium Laboratory, Karlsruhe, FRG. Tritium removal from plasma facing components is a serious challenge facing next step magnetic fusion devices that use carbon plasma facing components. The long term tritium inventory for ITER-FEAT is limited to about 350 g, mainly due to safety considerations. It is potentially possible that the inventory limit could be reached after a few weeks operation, requiring tritium removal before plasma operations can continue. Techniques for tritium removal have been demonstrated in the laboratory, and on tokamaks but they are slow and generally involve oxidation which will decondition the vessel walls (requiring additional time devoted wall conditioning) and generate undesirably large quantities of HTO. A novel laser heating technique has recently been used to remove tritium from carbon tiles that had been exposed to tritium plasmas in TFTR. A continuous wave Nd laser operates at powers up to 300 watts. The beam is directed by galvonometer driven scanning mirrors and focussed on the tile surface. The surface temperature is measured by an optical pyrometer. The tritium released is measured by a ionization chamber and surface tritium measured by an open walled ion chamber. Any changes in the laser irradiated surface are monitored with a microscope. To date tritium has been released in air and argon atmospheres and surface temperatures up to 2,300 C have been achieved. We will present measurements of the removal of tritium as a Presented at 6th International Conference on Tritium Science and Technology, Tsukuba, Japan Nov. 11-16th
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TRITIUM REMOVAL BY LASER HEATING C.H. Skinner, C. A. Gentile, G. Guttadora, A. Carpe, S. Langish, K.M. Young, M. Nishi (b), W. Shu (b), N. Bekris (c) (a)
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TRITIUM REMOVAL BY LASER HEATINGC.H. Skinner, C. A. Gentile, G. Guttadora, A. Carpe,
S. Langish, K.M. Young, M. Nishi(b), W. Shu(b), N. Bekris(c)
Tritium removal from plasma facing components is a serious challenge facing next step magnetic fusion devices that use carbon plasma facing components. The long term tritium inventory for ITER-FEAT is limited to about 350 g, mainly due to safety considerations. It is potentially possible that the inventory limit could be reached after a few weeks operation, requiring tritium removal before plasma operations can continue. Techniques for tritium removal have been demonstrated in the laboratory, and on tokamaks but they are slow and generally involve oxidation which will decondition the vessel walls (requiring additional time devoted wall conditioning) and generate undesirably large quantities of HTO.
A novel laser heating technique has recently been used to remove tritium from carbon tiles that had been exposed to tritium plasmas in TFTR. A continuous wave Nd laser operates at powers up to 300 watts. The beam is directed by galvonometer driven scanning mirrors and focussed on the tile surface. The surface temperature is measured by an optical pyrometer. The tritium released is measured by a ionization chamber and surface tritium measured by an open walled ion chamber. Any changes in the laser irradiated surface are monitored with a microscope. To date tritium has been released in air and argon atmospheres and surface temperatures up to 2,300 C have been achieved. We will present measurements of the removal of tritium as a function of the laser intensity, and scan rate. Potential implementation of this method in a next step fusion device will be discussed.
Presented at 6th International Conference on Tritium Science and Technology, Tsukuba, Japan Nov. 11-16th
• Next decade offers prospect of construction of next-step DT burning tokamak(s).
• Plasma material interactions will scale up orders of magnitude with increase in stored energy and pulse duration (bigger change than core plasma parameters).
• Tritium retention in machines with carbon plasma facing components will become significant constraint in plasma operations.
• Techniques for rapid efficient removal of tritium are needed.
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0 50 100 150 200
Number of ITER pulses (~400 s/each)
10 g-T/pulse
5 g-T/pulse
1 g-T/pulse
JET DTE-1equivalent
ITER-FEAT Modelling predictions
Brooks et al.: 2-5 g-T/pulse
Precautionary operating limit for
mobilisable in-vessel T inventory
ITER plans to install CFC divertor with option to switch to more reactor relevant all-W armoured targets prior to D-T operation.
Change depends on:
• frequency and severity of disruptions,
• success achieved in mitigating the effects of T co-deposition.
ITER
• In TFTR, several weeks were needed for tritium removal after only 10-15 min of cumulative DT plasmas
– Future reactors with carbon plasma facing components need T removal rate >> retention rate
• Heating is proven method to release tritium but heating vacuum vessel to required temperatures (~350 C) is expensive.
• Present candidate process involves oxidation, requiring lengthy machine re-conditioning and expensive DTO processing.
Tritium measured by FemptoTech ion chamber in closed loop system
Surface tritium on cube surface measured with open wall ion chamber,
Chamber contamination measured with swabs,
CUBEthermo-coupleVacuum Pump
Barytronpressure
gauge
Ar/air input
circulatingpump
flowmeter
60 micronfilter
Ion Chamber
15 micronfilter
Nd laser
Differential AtmosphericTritium Sampler (DAT)used for one expt.
Pdcatalyst
mol.sieve
mol.sieve
HTO trapped HTtrapped
Irradiate left 1/2 of cube KC22 1E in air atmosphere, pump & purge,then fill with Ar, and irradiate right 1/2. Laser 242 w, 50 mm/s, soft focus.
First measure tritium with ion chamber in closed loop, then add molecular sieve envelopes in DAT to determine HTO/HT mix
AtmospheremCi
(Femptotech)mCi HTO
(DAT)mCi HT(DAT)
air 8.3 4.8 0.7
argon 6.9 0.99 not measured
Not possible to measure HT in DAT without oxygen.
RGA shows Ar @ 99%, - possibly trapped H2O in tiles
- Note: not an issue with tiles inside operating tokamak.
Femptotech electrically calibrated by manufacturer only
Swabs show ~ microCi contamination of chamber.
~ 20 ms heating to > ~ 1500 C gives good tritium release with minimal change in surface (yellow area)
- how much tritium is left behind ?
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0 1000 2000degrees C
Tritium release vs. temperature
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Ci
duration > 500 C (msec)
Tritium release vs. scan speed
Cube 6E Laser Bake in air
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1 11 21 31 41 51 61 71 81 91
time < 45 mins >
mCi
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degrees C
Temp. ->
T released230 - 450 C < -
laser power 40 w, 80 w , 100 w, 80 w
• First Nd laser scan...
• Then Nd laser bake:
– rotate cube so that cut side faces laser, remove 1 lens to defocus laser. 100 w stationary laser beam heats cube to > 400 C for 40 mins in air to oxidise codeposit.
Two experiments so far:
• 1) ‘soft’ focus: – 46 % of total tritium released by scan
with almost no effect on surface
• 2) ‘hard focus’ - cube @ focal plane– 84% of total tritium released with
minor changes on surface.
Conclude:major part of co-deposited tritium can be released by scanning laser.
• Time needed to scan ?– 30 MJ required to heat top 100µm of 50
m2 area. - corresponds to output of 3kW laser for only 3 hours !
• Nd laser can be coupled via fiberoptic
• Potential for oxygen free tritium release in operating tokamak
– avoid deconditioning plasma facing surfaces
– avoid HTO generation(HTO is 10,000x more hazardous than T2 and very expensive to reprocess)
• Tritium removal by laser heating demonstrated.– no oxygen to decondition PFC’s – no HTO to process
• Method scalable to next-step device• Further optimization planned
* * * BONUS from Nd laser work * * *
Heating by continuous wave laser mimics heat loads in
transient off-normal events in tokamaks.
Opens new technique for studying key issues for
next step devices:
erosion by brittle destruction.
particulate (dust) generation.
Preprint: PPPL reports 3603, 3604 available from http://www.pppl.gov/pub_report/