Implementation and verification of a HELIAS module for the systems code PROCESS

Fusion Science and Technology

Dinklage

et al. 2015

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In fusion power plant studies a high confinement improvement with respect to empirical scalings is often assumed in order to design compact machines. In this work the limits of such a confinement enhancement is studied for helical-axis advanced stellarators (HELIAS).As a first exercise, the well-established power balance approach is used to investigate the impact of confinement enhancement (in terms of the ISS04 renormalisation factor) on the required size of HELIAS power plants. It is found that both a lower (0.5) and an upper limit (1.5 -1.7) exists for which, respectively, ignition is no longer possible or further confinement enhancement irrelevant due to physics limits.In the second part of the work a predictive neoclassical transport model is introduced and employed in order to determine a selfconsistent confinement time based on transport modelling. It is found that the confinement enhancement with respect to the ISS04 scaling decreases in comparison to W7-X as the device is scaled to reactor size dropping from ∼ 2.5 to 1.2 -1.3. This behaviour is explained with underlying scaling relations and transport effects. The results from both models are consistent and important for future HELIAS systems studies.

Section: Introductionmentioning

confidence: 99%

Limits of Confinement Enhancement for Stellarators

Fusion Science and Technology

Dinklage

et al. 2015

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“…The implementation of the proposed stellarator module to the systems code PROCESS and its verification is described in detail in a separate work [10]. In the mentioned work the HELIAS models are tested with respect to W7-X and, exploiting the generality of the models, with respect to a tokamak DEMO reference case.…”

Section: Discussionmentioning

confidence: 99%

“…The work is summarised and the results are discussed in Section 4. A detailed verification study of the HELIAS module will be presented in a separate work [10] where the models have been implemented in the systems code PROCESS.…”

Section: Introductionmentioning

confidence: 99%

HELIAS module development for systems codes

Fusion Engineering and Design

Dinklage

et al. 2015

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a b s t r a c tIn order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations.This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work.This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

“…Another important aspect of a fusion power plant is the controlled exhaust of energy and particles through the divertor. The model of the island divertor concept consists of a geometrical description including cross-field diffusion and radiation in the SOL and around the X-point [4,9,22]. In order to model the island divertor, a set of assumptions is needed: For the SOL a perpendicular heat diffusion coefficient of χ = 1.5m 2 /s has been chosen from experimental experience.…”

Section: Iss04mentioning

confidence: 99%

“…These models were implemented in the systems code PROCESS [8] which is a well-established, partly modular, European tokamak systems code which has gained maturity through years of applications. After the successful implementation, a verification study was carried out, in detail described in [9]. W7-X was modeled within the stellarator representation of PROCESS and compared to the real machine parameters.…”

Section: Introductionmentioning

confidence: 99%

System Code Analysis of HELIAS-Type Fusion Reactor and Economic Comparison With Tokamaks

Torrisi

IEEE Trans. Plasma Sci.

et al. 2016

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Systems codes are commonly employed for the analysis and conceptual design of fusion reactors. For the helicalaxis advanced stellarator (HELIAS) line a new set of systems code models have been developed to account for the stellaratorspecific 3D aspects. The models have recently been implemented in the systems code PROCESS and verified with respect to different test cases. After having established confidence in the stellarator models, systems studies were carried out for the 5-field period HELIAS case to define the accessible reactor design window. In the multidimensional physics and engineering parameter space sensitivity studies are carried out for the reactor regime to ascertain tradeoffs between different parameters and costs. Exemplary design points are analysed in more detail using the plasma operation contour approach which, for example, can be used to determine the optimum start-up path to ignition. Finally, with a common set of non-device-specific models, the PROCESS framework allows a direct comparison of tokamaks and stellarators. Although the 5-period HELIAS is a larger machine in terms of major radius, the required mass for both concepts is comparable leading to similar construction costs.