FW/Blanket and vacuum vessel for RTO/RC ITER

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, Masakazu

doi:10.1016/s0920-3796(00)00284-2

Cited by 12 publications

(12 citation statements)

References 3 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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“…1, 2) [1]. However, the blanket modules will be supported directly by the VV [2], and the blanket cooling manifolds will be mounted on the plasma-side surface of the VV inner wall. The inner and outer shells are both 60 mm plates and the stiffening ribs 40 mm plate (see Table. 1).…”

Section: Vacuum Vessel 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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“…The latest design of the next fusion reactor includes the use of titanium alloys in specific parts of the structure. In particular, the (a + b) alloy Ti 6 Al 4 V has been proposed to be used in flexible supports between the ITER blanket modules and the vacuum vessel [6,7]. One of the most important technical issues for the titanium alloys as fusion structural material is the influence of the transmutant helium production on the irradiation-related mechanical property changes.…”

Section: Introductionmentioning

confidence: 99%