Measurements of tritium retention and removal on the Tokamak Fusion Test Reactor

Skinner, C. H.; Blanchard, W.; Kamperschroer, J.; LaMarche, P. H.; Mueller, D.; Nagy, A.; Scott, S. D.; Ascione, G.; Amarescu, E.; Camp, Roderic Ai; Casey, M.; Collins, J.; Cropper, M.; Gentile, C.; Gibson, Marc A.; Hosea, J.; Kalish, M.; Langford, J.; Langish, S.; Mika, R.; Owens, D. K.; Pearson, Gavin; Raftopoulos, S.; Raucci, R.; Stevenson, T. N.; Halle, A. von; Voorhees, D.; Walters, T.; Winston, Julius

doi:10.1116/1.580224

Cited by 40 publications

(10 citation statements)

References 6 publications

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“…Independent of safety considerations, tritium is limited in supply and, for this reason alone, as much as possible needs to be recovered and re-used during the operation of any next-step device. 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 [414,415] and JET [416,417]. 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 [275,381,418,419].…”

Section: Tritium Retentionmentioning

confidence: 99%

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 Retentionmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…One area of efforts in the ITPA research community to understand and characterize the global gas retention in PFC surfaces has been the investigation of the level of D retention on the sides of tiles (in gaps between PFCs) as well as the processes involved. The literature indicates that a large fraction (15-30%) of overall D/T retained is found on the sides of tiles [97,98]. There are roughly 10 4 tiles in many of the currently operating tokamaks while the ITER design utilizes approximately 10 6 castellated and macrobrush armour tiles.…”

Section: Tritium Retentionmentioning

confidence: 99%

Plasma–surface interaction, scrape-off layer and divertor physics: implications for ITER

et al. 2007

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The work of the ITPA SOL/divertor group is reviewed and implications for ITER discussed. Studies of near SOL gradients have revealed a connection to underlying turbulence models. Analysis of a multi-machine database shows that parallel conduction gradients near the separatrix scale as major radius. New SOL measurements have implicated low-field side transport as driving parallel flows to the inboard side. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry ~10-20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL-larger gaps should reduce surface loads. The predicted divertor power loads for ITER disruptions has been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80% of its thermal energy before the thermal quench (not true for VDEs or ITBs). On the other hand predictions of power loading to surfaces outside the divertor have increased. Disruption mitigation through massive gas puffing has been successful at reducing divertor heat loads but estimates of the effect on the main chamber walls indicate 10s of kG of Be could be melted/mitigation. Estimates of ITER tritium retention have reduced the amount retained/discharge although the uncertainties are large and tritium cleanup may be necessary every few days to weeks. Long-pulse studies have shown that the fraction of injected gas that can be recovered after a discharge decreases with discharge length. The retention rate on the sides of tiles appears to ~ 1-3% of the ion flux to the front surface for C tiles and ~100x less for Mo tiles. T removal techniques are being developed based on surface heating and surface ablation although ITER mixed materials will make T removal more difficult. The use of mixed materials gives rise to a number of potential processes-e.g. reduction of surface melting temperatures (formation of alloys) and reduction of chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

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“…Even though technologies such as thermal oxidation [4], laser techniques [4] and plasma torch [5] showed good results, it is difficulty to think to treat co-deposits in remote areas and in gaps. Plasma glow discharge, which is a well-established cleaning technique, is here proposed as possible method to reach the remote areas of the fusion reactor [6][7][8][9]. Although a good cleaning efficiency was obtained using O 2 , He/O 2 and H 2 /N 2 discharges on a-C:H films [6,10], only few studies were carried out on C-W composite films.…”

Section: Introductionmentioning

confidence: 97%