Tritium experience in large tokamaks: Application to ITER

Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D.; Coad, J.P.; Federici, G.; Haange, R.

doi:10.1088/0029-5515/39/2/410

Cited by 33 publications

(29 citation statements)

References 43 publications

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“…Such a tile structure is thought to be the best solution to ensure the thermo-mechanical durability and integrity of materials under high heat flux loads, especially when considering the use of metals (W and Be) [6]. Although deposition in gaps between PFC tiles has been studied in several machines [7][8][9], little is known about the fuel retention in narrow grooves. Until recently, the JET Mk-I divertor had been the only large-scale castellated structure used in fusion experiments.…”

Section: Introductionmentioning

confidence: 99%

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Rubel

Coad

Pitts

2007

Journal of Nuclear Materials

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The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 · 10 15 cm À2 . Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.

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Section: Introductionmentioning

confidence: 99%

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Rubel

Coad

Pitts

2007

Journal of Nuclear Materials

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“…It also raises issues of collateral damage on in-vessel components. The product of the oxidation is tritiated water that requires expensive reprocessing [3]. A further complication is that the temperature required is beyond the range of pressurized water cooling.…”

Section: Tritium Removal By a Scanning Lasermentioning

confidence: 99%

“…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 [1,2]. This experience together with focussed laboratory studies, has illuminated the challenges [3]. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition [4,5].…”

Section: Introductionmentioning

confidence: 99%

Tritium Issues in Next Step Devices

Skinner

Federici

2002

Advanced Diagnostics for Magnetic and Inertial Fusion

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“…The safe and routine operation with tritium on TFTR was an important milestone for the fusion program and provided experience for future ignition machines (Skinner et al 1998a). …”

Section: Tftr Devicementioning

confidence: 99%

“…In addition, in a commercial reactor operating almost continuously, retention fractions <0.1% are required (Skinner et al 1998a). …”

Section: Tftr Devicementioning

confidence: 99%

Results from D-T Experiments on TFTR and Implications for Achieving an Ignited Plasma

Hawryluk¹,

Blanchard²,

Batha³

1998

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Progress in the performance of tokamak devices has enabled not only the production of significant bursts of fusion energy from deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET) but, more importantly, the initial study of the physics of burning magnetically confined plasmas. As a result of the worldwide research on tokamaks, the scientific and technical issues for achieving an ignited plasma are better understood and the remaining questions more clearly defined. The principal research topics which have been studied on TFTR are transport, magnetohydrodynamic stability, and energetic particle confinement. The integration of separate solutions to problems in each of these research areas has also been of major interest. Although significant advances, such as the reduction of turbulent transport by means of internal transport barriers, identification of the theoretically predicted bootstrap current, and the study of the confinement of energetic fusion alphaparticles have been made, interesting and important scientific and technical issues remain for achieving a magnetic fusion energy reactor.In this paper, the implications of the TFTR experiments for overcoming these remaining issues will be discussed.

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Tritium experience in large tokamaks: Application to ITER

Cited by 33 publications

References 43 publications

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Tritium Issues in Next Step Devices

Results from D-T Experiments on TFTR and Implications for Achieving an Ignited Plasma

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