Tritium Issues in Next Step Devices

Skinner, C.H.; Federici, G.

doi:10.2172/788204

Cited by 13 publications

(12 citation statements)

References 15 publications

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“…The presently foreseen methods to remove tritium from ITER are reviewed in [419,415,417] and can be roughly divided into four categories : a) Plasma cleaning. On areas in direct contact with the plasma, T retention can be reduced by isotope exchange during deuterium plasma operation or desorption by plasma heating of the material.…”

Section: Tritium Removal Methodsmentioning

confidence: 99%

“…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%

“…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

Loarte¹,

et al. 2007

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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: Tritium Removal Methodsmentioning

confidence: 99%

Section: Database On Fuel Retention In Present Fusion Devicesmentioning

confidence: 99%

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Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…Co-deposition of carbon with hydrogen is known to lead to significant retention as was observed in the TFTR [5] and JET [6] tokamaks. Tokamak co-deposition studies involving Be and W are unavailable but work conducted in plasma devices offers some insight into the retention properties of Be and W co-deposited layers.…”

Section: Introductionmentioning

confidence: 93%

An empirical scaling for deuterium retention in co-deposited beryllium layers

et al. 2008

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Abstract.Different mechanisms will contribute to tritium retention in ITER, amongst which co-deposition with materials from the plasma-facing components may be the main contributor. A systematic study of the influence of the deposition conditions (substrate temperature, deposition rate, energy of the incident particles) on the deuterium retention in co-deposited beryllium layers has been carried out in PISCES-B. The mechanism by which deuterium co-deposits with beryllium appears to be a combination of co-deposition and implantation, with a decreased retention for increased deposition rate and an increased retention for increased incident deuterium particles energy. A scaling equation is developed, providing a method to predict the retention in Be codeposits formed in PISCES-B as a function of the layer formation conditions. Using this equation, previously published data on retention in Be co-deposits are re-examined and relatively good agreement is found with the prediction of the scaling equation.

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“…Tritium fuel has been successfully used in the tokamak fusion test reactor (TFTR) and the joint european torus (JET) producing 10 and 16 MW of fusion power respectively [19,20]. A large fraction of tritium was retained during DT plasma operations in TFTR and JET by codeposition with eroded carbon and by isotope exchange with previously retained deuterium [21,22]. When the tritium in-vessel inventory approached the administrative safety limit, it was removed by extensive campaigns involving several weeks of glow discharge cleaning and deuterium operation.…”

Section: Key Issues and Randd Prioritymentioning

confidence: 99%