Tritium retention and removal on TFTR

Mueller, D.; Blanchard, W.; Doyle, B.L.; Hosea, J.; Nagy, A.; Pearson, Gavin; Skinner, C. H.; Ulrickson, M.; Wampler, W.R.; Zweben, S. J.

doi:10.1109/fusion.1997.687037

Cited by 8 publications

(10 citation statements)

References 16 publications

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“…59 It was also observed that the fraction of D retained increased with increasing neutral beam heating power [202]; (Fig. 60).…”

Section: Insert Table 13mentioning

confidence: 80%

“…By comparison, He/O GDC performed inside the TFTR vessel resulted in much lower T removal rates [202,749], possibly due to the re-deposition of the reaction products before being pumped out of the tokamak. Laboratory studies have determined this erosion to be a 2-step process: oxidation, followed by particle-induced desorption, with the maximum erosion rate limited by the latter.…”

Section: 62)mentioning

confidence: 94%

“…4.7.2) and have expanded the technical knowledge base for tritium issues in fusion reactors [200]. Table 5 provides a list of key quantities related to tritium in existing tokamaks and a next-step device [201][202][203][204].…”

Section: Other Plasma Disturbances Eg Elmsmentioning

confidence: 99%

“…In JET and TFTR (Sect. 4.7.2) tritium removal took place over several weeks, i.e., considerably longer than the approximately 1000 s cumulative total time of high-power DT discharges [202,204]. This would not be adequate in a fusion facility like ITER where the tritium removal rate will need to be orders of magnitude faster to support the reactor's operational schedule.…”

Section: Tritium Removal From Co-deposited Layersmentioning

confidence: 99%

See 3 more Smart Citations

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

et al. 2001

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The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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“…59 It was also observed that the fraction of D retained increased with increasing neutral beam heating power [202]; (Fig. 60).…”

Section: Insert Table 13mentioning

confidence: 80%

Section: 62)mentioning

confidence: 94%

Section: Other Plasma Disturbances Eg Elmsmentioning

confidence: 99%

Section: Tritium Removal From Co-deposited Layersmentioning

confidence: 99%

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Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

et al. 2001

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“…During the final four years of TFTR operations, 1993 to 1997 approximately 53, 000 Ci of tritium was injected into the TFTR vacuum vessel mainly through neutral beam injection [1,2]. When TFTR finished operation in preparation for final shut down an aggressive three month campaign commenced for the purpose of recovering tritium inventory which was deposited on the internal surfaces of the > 80,000 liter volume vacuum vessel [3].…”

Section: Introductionmentioning

confidence: 99%

In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

Gentile¹,

Skinner²,

Young³

et al. 1999

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-The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), TritiumEngineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time ~ 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting ~ 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and "moist air" purges. Currently ~ 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of ~ 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles.In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were landed at four positions per tile. The geometry for landing the TLDÕs provided measurement at 24 positions per tile. The instrumentation was positioned on the tiles (96 tiles in each bay) by technicians who manipulated the boom and end effector from outside the vacuum vessel port. In addition to obtaining bay L tile measurements, 3 bumper limiter tiles were collected from the vacuum vessel for further analysis. During the bay L measurement campaign it was observed that several of the tiles in the lower third of the bumper limiter exhibited considerable surface flaking.

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In-vessel tritium retention and removal in ITER

Federici

Anderl

Andrew

et al. 1999

Journal of Nuclear Materials

Self Cite

234

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Invited Review Paper -R2 CEI SUN 3 0 O S T ITritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER.We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation.We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER, the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in-situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

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Tritium retention and removal on TFTR

Cited by 8 publications

References 16 publications

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

In-vessel tritium retention and removal in ITER

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