Integrated core–SOL–divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario

Zagórski, R.; Voitsekhovitch, I.; Ivanova‐Stanik, I.; Köchl, F.; Belo, P.; Fable, E.; García, J.; Garzotti, L.; Hobirk, J.; Hogeweij, G. M. D.; Joffrin, E.; Litaudon, X.; Polevoi, A.; Telesca, G.; Contributors, Jet

doi:10.1088/0029-5515/55/5/053032

Cited by 6 publications

(5 citation statements)

References 27 publications

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“…separatrix density, radial anomalous transport). In line with our earlier simulations for JET-ILW [20,21] and ITER [22], in the case of JT-60SA it has also been found that the increased radial…”

Section: Results For the W Divertorsupporting

confidence: 90%

“…separatrix density, radial anomalous transport). In line with our earlier simulations for JET-ILW [20] and ITER [22] also in the case of JT-60SA it has been found that the increased radial diffusion (D SOL = 1 m 2 /s, open symbols in Fig.5) improves the impurity retention in the SOL leading to the increased divertor radiation losses and easier access of detachment. Similar effect on the results has the increased edge density.…”

Section: Results For W Divertorsupporting

confidence: 89%

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Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

et al. 2017

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JT-60SA reference design scenarios at high (#3) and low (#2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for standard carbon wall and the full W have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr) on the radiation losses and plasma parameters. For the scenario #3 in carbon environment the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (#2), the carbon and seeding impurity radiation does not e?ectively reduce power to the targets. In this case only increase of neither average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that in case of tungsten divertor the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the carbon divertor. For the high density case with neon seeding operation in full detachment mode is observed. Ar seems to be an optimal choice for the low density high power scenario #2, showing wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semi-detached conditions in the divertor.

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Section: Results For the W Divertorsupporting

confidence: 90%

Section: Results For W Divertorsupporting

confidence: 89%

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

et al. 2017

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“…The COREDIV code, which solves self-consistently the radial 1D plasma energy and particle transport in the core region and 2D multi-fluid transport in the SOL region, can be used for this purpose. The code package has already been successfully used to benchmark against numerous discharges in various existing tokamaks [20][21][22][23] and to make predictions for future fusion devices [24][25][26][27]. In the following, the COREDIV model related to current study is briefly introduced.…”

Section: Modelling Methodsmentioning

confidence: 99%

“…The core plasma profiles with a range of pedestal densities are simulated by the physics-based transport solver implemented in the OMFIT (one modeling framework for integrated tasks) framework [19]. Since f burnup depends strongly on both core and SOL plasma conditions, numerical analysis of f burnup is performed by using the core-SOL coupling COREDIV code [20][21][22][23][24][25][26][27]. By adjusting the transport coefficients, COREDIV is used to match the core profiles obtained by OMFIT, and the resulting heat and particle fluxes across the boundary are used as input to simulate the SOL region.…”

Section: Introductionmentioning

confidence: 99%

Evaluation of tritium burnup fraction for CFETR scenarios with core-edge coupling simulations

et al. 2020

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A key mission for the next-step fusion tokamak device China Fusion Engineering Test Reactor (CFETR) is demonstrating tritium self-sufficiency, which requires a sufficiently high tritium burnup fraction (fburnup) in order to match a practically achievable tritium breeding ratio (TBR) with the blanket design constraints. Core-edge coupling simulations are performed to investigate the dependence of fburnup on different controlling parameters for CFETR scenarios. Core plasma profiles with a range of pedestal densities are simulated by consistent iterative calculations of equilibrium, transport, auxiliary heating and current drives within the OMFIT framework. The core-SOL integrated COREDIV code is then used to evaluate fburnup with the OMFIT modelled core plasma parameters as input. According to the simulations, fburnup can be effectively increased with a higher pedestal density on account of the fusion power increasing faster than the fueling source required to maintain steady-state. Higher can also increase the fburnup due to increase of fuel recycling. Deeper pellet fueling deposition and lower ratio of particle to thermal diffusivities D/χ can both increase the effective particle confinement time and thus fburnup. However, the effect of helium and other impurities (Ar and W) is shown to reduce fburnup for comparable impurity and main ion transport. Based on our analysis, using present pellet fueling technology, achieving fburnup > 3% for CFETR will be very challenging. This is a lower limit for the required TBR (>1) to match the achievable TBR for tritium self-sufficiency. Furthermore, our study suggests that if fueling can penetrate deeper than r/a < 0.8 under optimistic conditions, the required burnup fraction could be attainable. The modelling results thus provide important suggestions and implications for the optimization of CFETR scenarios and development of advanced fueling systems.

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“…In the case of the COREDIV simulations, we have already performed a number of simulations to investigate the influence of the SOL radial transport on simulation results. It has been done for present day experiments (JET[jet- [7,17], ASDEX-U [18]) and for future devices (ITER [19], DEMO [20]). The main conclusion from these simulations is that the stronger radial transport in the SOL helps to keep the impurities in the SOL and leads to the increased impurity radiation in the SOL.…”

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confidence: 99%

Numerical analyses of JT-60SA scenarios with the COREDIV code

et al. 2015

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JT-60SA design scenarios have been analyzed with the help of the self-consistent core-edge COREDIV code, with the aim to assess the influence of impurities on the plasma parameters and tokamak performance. In particular, the reduction of divertor target power load due to radiation of sputtered and externally seeded impurities has been investigated. For all scenarios considered, the gradual replacement of carbon by low Z seeding impurity (N, Ne) is observed as the gas influx increases. For high auxiliary power and low density scenarios, the carbon and seeding impurity radiation does not effectively reduce power to plate. Consequently, results with very high

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Integrated core–SOL–divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario

Cited by 6 publications

References 27 publications

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

Evaluation of tritium burnup fraction for CFETR scenarios with core-edge coupling simulations

Numerical analyses of JT-60SA scenarios with the COREDIV code

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