Tritium retention and clean-up in JET

Andrew, P.; Brennan, P.D.; Coad, J.P.; Ehrenberg, J.; Gadeberg, M.; Gibson, A.; Hillis, D.L.; How, J.; Jarvis, O.N.; Jensen, H.S.; Lässer, R.; Marcus, F.B.; Monk, R.D.; Morgan, P. D.; Orchard, J.; Peacock, Alan T.; Pick, M.; Rossi, A.; Schild, P.; Schunke, B.; Stork, D.; Pearce, R.

doi:10.1016/s0920-3796(99)00084-8

Cited by 154 publications

(85 citation statements)

References 20 publications

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“…In general disagreements by factors of 3 or larger are found between post-mortem analysis and fuel retention determination from gas balance of deuterium and hydrogen discharges due to the difficulties of the latter technique. The agreement is, however, much better for tritium experiments due to the better diagnostic possibilities of the tritium gas balance compared to deuterium/hydrogen [419,415,416]. …”

Section: Database On Fuel Retention In Present Fusion Devicesmentioning

confidence: 99%

“…The recycling of hydrogenic fuel and its retention in plasma facing-materials considered for next step devices has been a traditional research topic in all fusion devices but the research in this field has become particularly active after the use of considerable amounts of tritium and its incomplete recovery in the full D-T campaigns in TFTR [413,414] and JET [415,416]. A more detailed description of these observations, the mechanisms of T retention and the techniques used to remove fuel from PFCs in fusion devices can be found in [274,380,417,418].…”

Section: Introductionmentioning

confidence: 99%

“…A more detailed description of these observations, the mechanisms of T retention and the techniques used to remove fuel from PFCs in fusion devices can be found in [274,380,417,418]. [415,419,420,18] and foreseen for ITER.…”

Section: Introductionmentioning

confidence: 99%

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Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

280

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Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

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Section: Database On Fuel Retention In Present Fusion Devicesmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

280

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“…Given the difficulties in fuel removal [3][4][5], there is a concern that the radioactive fuel will be accumulated in the gaps of such a castellation [6][7][8][9]. Dedicated studies addressing the performance of castellated structures under plasma impact are ongoing on several tokamaks [10][11][12].…”

Section: Introductionmentioning

confidence: 99%

Castellated structures for ITER: Differences of impurity deposition and fuel accumulation in the toroidal and poloidal gaps

Litnovsky¹,

Philipps²,

Wienhold

et al. 2009

Journal of Nuclear Materials

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Castellation is foreseen for the first wall and divertor area in ITER. The concern of the fuel accumulation and impurity deposition in the gaps of castellated structures calls for dedicated studies.Recently, a tungsten castellated limiter with rectangular and roof-like shaped cells was exposed to the SOL plasmas in TEXTOR. After exposure, roughly two times less fuel was found in the gaps between the shaped cells whereas the difference in carbon deposition was less pronounced. Up to 70 at. % of tungsten was found intermixed in the deposited layers in the gaps. The metal fraction in the deposit decreases rapidly with a depth of the gap. Modeling of carbon deposition in poloidal gaps has provided a qualitative agreement with an experiment. Misalignment of toroidal gaps with respect to the magnetic field lines has led to the significant anisotropy of C and D distribution in the gaps.

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“…Results of the JET deuterium-tritium campaign (DTE1) [Keilhacker_NF99] revealed that significant amounts of tritium accumulated in areas shadowed from the plasma [Andrew_FED99], which underlined the need to better understand the mechanisms leading to fuel retention in fusion devices. Tritium-rich carbon films were found at the water-cooled louver at the entrance of the pumping duct in the JET inner divertor, which connects tritium retention to the issue of long-range carbon migration in fusion devices operating with carbon plasma-facing components (PFCs).…”

Section: Introductionmentioning

confidence: 99%

Scrape-off layer transport and deposition studies in DIII-D

Groth

Allen

Boedo³

et al. 2007

Physics of Plasmas

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Trace 13 CH 4 injection experiments into the main scrape-off layer of low density L-mode and high-density H-mode plasmas have been performed in the DIII-D tokamak[Luxon_NF02] to mimic the transport and deposition of carbon arising from a main chamber sputtering source. These experiments indicated entrainment of the injected carbon in plasma flow in the main SOL, and transport toward the inner divertor. Ex-situ surface analysis showed enhanced 13 C surface concentration at the corner formed by the divertor floor and the angled target plate of the inner divertor in L-mode; in H-mode, both at the corner and along the surface bounding the private flux region inboard of the outer strike point. Interpretative modeling was made consistent with these experimental results by imposing a parallel carbon ion flow in the main SOL toward the inner target, and a radial pinch toward the separatrix. Predictive modeling carried out to better understand the underlying plasma transport processes suggests that the deuterium flow in the main SOL is related to the degree of detachment of the inner divertor leg. These simulations show that carbon ions are entrained with the deuteron flow in the main SOL via frictional coupling, but higher charge state carbon ions may be suspended upstream of the inner M. Groth et al.Scrape-off layer transport and deposition studies in DIII-D GA-A25625, DRAFT 3 APS/DPP Invited ms 2 divertor X-point region due to balance of the friction force and the ion temperature gradient.

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Tritium retention and clean-up in JET

Cited by 154 publications

References 20 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Castellated structures for ITER: Differences of impurity deposition and fuel accumulation in the toroidal and poloidal gaps

Scrape-off layer transport and deposition studies in DIII-D

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