Engineering Design and Analysis of the ARIES-CS Power Plant

Raffray, A.R.; El-Guebaly, L.; Malang, S.; Wang, X. R.; Bromberg, L.; Ihli, T.; Merrill, B.J.; Waganer, Lester M.

doi:10.13182/fst08-4

Cited by 54 publications

(36 citation statements)

References 21 publications

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“…with no dedicated PFC armor) or a thin W armor layer had been considered. In the case of a Dual Coolant Lithum Lead (DCLL) blanket [12], the bare wall would be a ferritic steel wall. In any case, for both bare wall or thin W armor cases, the plasma facing material temperature will be fairly high, about 600-800°C.…”

Section: Power Plant Plasma Facing Materialsmentioning

confidence: 99%

“…The proposed concepts range in scale from a plate configuration with characteristic dimension of the order of 1 m [14,15], to the ARIES T-tube configuration with characteristic dimension of the order of 10 cm [12,16], to the EU FZK finger concept with characteristic dimension of the order of 2 cm [17,18]. The trend in moving to smaller-scale units is to minimize the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units (for an example tokamak case, from ~750 to ~110,000 to ~535,000 for the plate, T-tube and finger concepts, respectively) with a corresponding impact on the reliability of the system.…”

Section: Power Plant Divertor Design and Randdmentioning

confidence: 99%

“…The ARIES T-tube concept is a mid size unit, ~15 mm in diameter and ~100 mm long, capable of accommodating at least 10 MW/m 2 [12,16]. As illustrated in INSERT FIGURE 12…”

Section: Us T-tube Divertor Concept For Power Plantmentioning

confidence: 99%

See 2 more Smart Citations

High heat flux components—Readiness to proceed from near term fusion systems to power plants

Raffray¹,

Nygren²,

Whyte³

et al. 2010

Fusion Engineering and Design

122

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A present topic of high interest in magnetic fusion is the "gap" between near-term and longterm concepts for high heat flux components (HHFC), and in particular their application in divertors. This paper focuses on this issue with the aim of characterizing the international status of current HHFC design concepts for ITER and describing the different technologies needed in the designs being developed for fusion power plants. Critical corresponding material and physics aspects are highlighted while evaluating the current readiness level of long-term concepts, identifying the design and R&D gaps, and discussing ways to bridge them.2

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Section: Power Plant Plasma Facing Materialsmentioning

confidence: 99%

Section: Power Plant Divertor Design and Randdmentioning

confidence: 99%

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High heat flux components—Readiness to proceed from near term fusion systems to power plants

Raffray¹,

Nygren²,

Whyte³

et al. 2010

Fusion Engineering and Design

122

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show abstract

“…9, are mechanically attached to the hot shield with mechanical fasteners and coolant manifold0structure via the helium and LiPb coolant piping throughout the power core, as explained in Ref. 9. Fig.…”

Section: Ivf Blanket Module Maintenance Provisionsmentioning

confidence: 99%

ARIES-CS Maintenance System Definition and Analysis

2008

Fusion Sci. Technol

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“…For ignited plasmas, as was assumed for instance in the HELical Advanced Stellarator (HELIAS) studies [11], Q is by definition infinity. The resulting fusion power, for both stellarator and tokamak reactors, lies in the range of 1 to 5 GW [1,12,13]. Besides the confinement, which in H-mode is larger by a factor of two compared to L-mode [9], the magnetic field plays a crucial role for the achievable fusion power amplification.…”

Section: Introductionmentioning

confidence: 99%

Wendelstein 7‐X: An Alternative Route to a Fusion Reactor

Wolf

Team

2009

Contrib. Plasma Phys.

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With the ITER tokamak magnetic confinement fusion is making the decisive step to realize a burning fusion plasma and prepare the basis for a demonstration power plant (DEMO). Despite the advantages of an intrinsically steady state magnetic field and better stability properties, the development of stellarators lags behind by about 1 1/2 device generations, because of the difficulties to realize the desired magnetic field configuration. The goal of the optimized stellarator Wendelstein 7-X is to overcome the principal deficiencies of the stellarator concept and demonstrate its reactor capability, combining sufficiently good thermal and fast ion confinement with reactor relevant β and collisionality under steady state conditions.

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Engineering Design and Analysis of the ARIES-CS Power Plant

Cited by 54 publications

References 21 publications

High heat flux components—Readiness to proceed from near term fusion systems to power plants

High heat flux components—Readiness to proceed from near term fusion systems to power plants

ARIES-CS Maintenance System Definition and Analysis

Wendelstein 7‐X: An Alternative Route to a Fusion Reactor

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