Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, Masakazu

doi:10.1016/s0022-3115(00)00165-3

Cited by 9 publications

(11 citation statements)

References 4 publications

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“…9 Insert against gravity by the toroidal field coil structure, and which contains the blanket and divertor systems [259,260]. The basic functions of the vacuum vessel are to provide a suitable vacuum in the plasma chamber, to support the in-vessel components, to assist in the nuclear shielding of the coils and to constitute the first safety boundary.…”

Section: Overview Of Design Features Of Plasma-facing Components For mentioning

confidence: 99%

“…The first-wall design is based on 1 mm thick 10 mm diameter SS cooling tubes with a pitch of ≈ 20 mm, embedded in a Cu alloy layer of ≈ 15 mm thickness (DS-Cu or CuCrZr). The plasma-facing material is a 10 mm thick Be layer [259,260]. The first-wall incorporates the two start-up limiters [261], which need to absorb several MW of power (up to ≈ 8 MWm −2 ) during the limiter phases of each discharge (i.e., ≈ 100 s during plasma start-up and ≈ 100 s during plasma current ramp-down).…”

Section: (January 2001)mentioning

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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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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Section: Overview Of Design Features Of Plasma-facing Components For mentioning

confidence: 99%

Section: (January 2001)mentioning

confidence: 99%

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

et al. 2001

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“…Insert Fig. 9 Insert table 7 Vacuum vessel and blanket: The ITER vacuum boundary is formed by a double-walled vacuum vessel made of stainless-steel SS 316L(N)-IG, ITER Grade: 0.06 to 0.08% nitrogen (60 mm plates for the outer and inner shells, and 40 mm stiffening ribs), which is supported against gravity by the toroidal field coil structure, and which contains the blanket and divertor systems [259,260]. The basic functions of the vacuum vessel are to provide a suitable vacuum in the plasma chamber, to support the in-vessel components, to assist in the nuclear shielding of the coils and to constitute the first safety boundary.…”

Section: Overview Of Design Features Of Plasma-facing Components For mentioning

confidence: 99%

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

Federici¹,

Skinner²,

Brooks³

et al. 2001

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This Max-Planck-Institut für Plasmaphysik (IPP), Garching, Germany, report is intended for internal use. IPP reports express the views of the authors at the time of writing and do not necessarily reflect the opinions of the Max-Planck-Institut für Plasmaphysik or the final opinion of the authors on the subject. Neither the Max-Planck-Institut für Plasmaphysik, nor the Euratom Commission, nor any person acting on behalf of either of these: (1) gives any guarantee as to the accuracy and completeness of the information contained in this report, or that the use of any information, apparatus, method or process disclosed therein may not constitute an infringement of private owned right; or (2) assumes any liability for damage resulting from the use of any information, apparatus, method or process in this report.

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“…In the latter option ( see FIG. 13), FW panels are attached to the shield block using a system of bolts and shear ribs [7].…”

Section: Fw/blanket Designmentioning

confidence: 99%

“…The shield block is cooled preferably with radial flow, where efficient cooling will be achieved in the rear part with an irregular configuration due to intensive grooving. Three configurations of the radial flow cooling have been investigated (a) a front header with co-axial flow, (b) front and rear headers, (c) a middle header with co-axial flow [2,7]. FIG.…”

Section: Fw/blanket Designmentioning

confidence: 99%

ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

Ioki

Dänner

Koizumi³

et al. 2001

Nucl. Fusion

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A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R&D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

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Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Cited by 9 publications

References 4 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

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

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