Tritium inventory in the first wall of JET

Peacock, Alan T.; Andrew, P.; Brennan, D. P.; Coad, J.P.; Hemmerich, H; Knipe, S.; Penzhorn, R.-D.; Pick, M.

doi:10.1016/s0920-3796(00)00340-9

Cited by 31 publications

(32 citation statements)

References 2 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%

See 1 more Smart Citation

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

279

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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%

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

279

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“…Figure 1a shows the D/C ratio obtained during two systematic investigations of carbon codeposition in laboratory-based experiments [9,10]. Also shown in the figure are measurements of D/C values found during post-mortem analysis of codeposits produced over long run campaigns in several confinement devices [11,12,13].…”

Section: Retention In Carbon Codepositsmentioning

confidence: 99%

Codeposition of deuterium with ITER materials

Doerner¹,

Baldwin²,

Temmerman³

et al. 2009

Nucl. Fusion

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Abstract:The levels of retention in codeposited layers of each of the three ITER materials (C, Be and W) are compared.Scaling laws, based on the conditions during the codeposition process (surface temperature, incident particle energy and ratio of the depositing fluxes), are presented to allow prediction of expected retention under ITER conditions. Retention in carbon codeposits scales inversely with incident particle energy, whereas in the metallic codeposits the retention level scales proportional to increasing particle energy. The differing scaling of retention with incident particle energy provides insight into which material may impact the global retention in ITER depending on where it may form codeposits. In addition to the amount of retention, the release behavior of tritium from codeposits will influence the tritium accumulation rate within ITER. The thermal release behavior of T (or D) from codeposits can be used to evaluate the effectiveness of baking at different temperatures as a means of tritium removal. Finally, the desorption kinetics from Be and W codeposits are contrasted. In the case of W codeposits, the duration of the baking cycle is important in determining the removal efficiency, whereas with Be codeposited layers, the maximum achievable bake temperature plays the leading role in determining removal efficiency.

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“…The tritium inventory increased a factor-of-two faster than expected peaking at 11.5 g T with more than half of the tritium on site trapped in the vessel. Films were formed with high (~ 0.8) D/C ratio on the divertor louvers leading to subsequent flaking and accumulation in the subdivertor [18]. The retention rate during the DTE1 campaign was 40%.…”

Section: Tritium Retentionmentioning

confidence: 99%

Is carbon a realistic choice for ITER's divertor?

Skinner¹,

Federici²

2006

Phys. Scr.

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Tritium retention by codeposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of codeposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives.

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Tritium inventory in the first wall of JET

Cited by 31 publications

References 2 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Codeposition of deuterium with ITER materials

Is carbon a realistic choice for ITER's divertor?

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