Compatibility of the radiating divertor with high performance plasmas in DIII-D

Pétrie, T.W.; Wade, M. R.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hyatt, A.W.; Isler, R.C.; Lasnier, C.J.; Leonard, A.W.; Mahdavi, M. A.; Porter, G. D.; Schaffer, M. J.; Watkins, J.G.; West, W.P.

doi:10.1016/j.jnucmat.2007.01.131

Cited by 40 publications

(57 citation statements)

References 13 publications

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“…during the type I ELM regime (when P rad /P tot < 40% and T ped > 750 eV), a significant leakage of the seeded impurity out of the divertor is observed, as also reported in [14]. Once the pedestal temperature is reduced and the type III ELM is obtained, the impurity gas puff needed to maintain constant radiated power and therefore the regime stationary decreases slowly from N ∼ 6 × 10 22 down to N ∼ 3 × 10 22 e s −1 .…”

Section: Operational Domain Covered During the Experimentssupporting

confidence: 78%

“…In addition, the signals measured by the lines of sight passing through the core (channel nos [5][6][7][8][9][10][11][12][13][14][15][16][17][18][19][20] are increased by almost a factor of 2 on average. This is consistent with the total bulk radiation that is estimated to be P bulk rad ∼ 2 MW without and P bulk rad ∼ 4 MW with type III ELM obtained with N seeding.…”

Section: Plasma Radiationmentioning

confidence: 99%

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Hybrid H-mode scenario with nitrogen seeding and type III ELMs in JET

Corre

Joffrin

Monier‐Garbet

et al. 2008

Plasma Phys. Control. Fusion

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The performance of the 'hybrid' H-mode regime (long pulse operation with high neutron fluency) has been extensively investigated in JET during the [2005][2006][2007] experimental campaign up to normalized pressure β N = 3, toroidal magnetic field B t = 1.7 T, with type I ELMs plasma edge conditions. The optimized external current drive sources, self-generated non-inductive bootstrap current and plasma core stability properties provide a good prospect of achieving a high fusion gain at reduced plasma current for long durations in ITER.One of the remaining issues is the erosion of the divertor target plates associated with the type I ELM regime. regime with a normalized pressure β N = 2.6 (P NBI ∼ 20-22 MW) and a thermal confinement factor of H * 98 (y, 2) ∼ 0.83 has been recently successfully developed on JET with nitrogen seeding. This scenario shows good plasma edge condition (compatible with the future ITER-like wall on JET) and moderate MHD activity. In this paper, we report on the experimental development of the scenario (with plasma current I p = 1.7 MA and magnetic field B t = 1.7 T) and the trade-off between heat load reduction at the target plates and global confinement due to nitrogen seeding and type III ELM working conditions.

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Section: Operational Domain Covered During the Experimentssupporting

confidence: 78%

Section: Plasma Radiationmentioning

confidence: 99%

Hybrid H-mode scenario with nitrogen seeding and type III ELMs in JET

Corre

Joffrin

Monier‐Garbet

et al. 2008

Plasma Phys. Control. Fusion

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“…Reducing power to the divertor via edge radiation has been investigated on a number of devices in ELMy regimes (both Type I and Type III) in the past, but not necessarily while maintaining high confinement. [10,11,12,13,14] …”

Section: Introductionmentioning

confidence: 99%

Power requirements for superior H-mode confinement on Alcator C-Mod: experiments in support of ITER

et al. 2011

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Power requirements for maintaining sufficiently high confinement (i.e., normalized energy confinement time H 98 ≥1) in H-mode and its relation to H-mode threshold power scaling, P th , are of critical importance to ITER. In order to better characterize these power requirements, recent experiments on the Alcator C-Mod tokamak have investigated H-mode properties, including the edge pedestal and global confinement, over a range of input powers near and above P th . In addition, we have examined the compatibility of impurity seeding with high performance operation, and the influence of plasma radiation and its spatial distribution on performance. Experiments were performed at 5.4T at ITER relevant densities, utilizing bulk metal plasma facing surfaces and ion cyclotron range of frequencies waves for auxiliary heating. Input power was scanned both in stationary enhanced D α (EDA) H-modes with no large ELMs and in ELMy H-modes in order to relate the resulting pedestal and confinement to the amount of power flowing into the scrape-off layer, P net , and also to the divertor targets. In both EDA and ELMy H-mode, energy confinement is generally good, with H 98 near unity. As P net is reduced to levels approaching that in L-mode, pedestal temperature diminishes significantly and normalized confinement time drops. By seeding with low-Z impurities such as Ne and N 2 , high total radiated power fractions are possible, along with substantial reductions in divertor heat flux (>4x), all while maintaining H 98 ~ 1. When the power radiated from the confined vs. unconfined plasma is examined, pedestal and confinement properties are clearly seen to be an increasing function of P net , helping to unify the results with those from unseeded H-modes. This provides increased confidence that the power flow across the separatrix is the correct physics basis for ITER extrapolation. The experiments show that P net /P th of one or greater is likely to lead to H 98~1 operation, and also that such a condition can be made compatible with a low-Z radiative impurity solution for reducing divertor heat loads to levels acceptable for ITER.

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“…This technique has been investigated on a number of tokamaks in both type-I and type-III (edgelocalized mode) ELM regimes, though maintaining high normalized confinement has not always been demonstrated [3][4][5][6][7]. The present work reports the results of experiments on enhanced D- (EDA) H-mode plasmas [8] in Alcator C-Mod where the impact to the energy confinement from varying amounts of seeded radiation loss and ICRF input power were investigated.…”

Section: Introductionmentioning

confidence: 99%

Effect of N2, Ne and Ar seeding on Alcator C-Mod H-mode confinement

Reinke

Hughes

Loarte

et al. 2011

Journal of Nuclear Materials

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The mitigation of divertor heat fluxes is an active topic of investigation on existing tokamaks.One approach uses radiation, both inside and outside the last closed flux surface (LCFS), to convert plasma thermal energy, usually directed towards dedicated plasma facing components, to soft x-ray and ultraviolet radiation, spread over a much larger surface area. Recent enhanced D- H-mode experiments on Alcator C-Mod varied the ICRF input power and radiative power losses via impurity seeding to demonstrate that normalized energy confinement depends strongly on the difference between input power and the radiated power inside the LCFS. These investigations also show that when seeded with either Ne or N 2 , a factor of two and higher reduction in outer divertor heat flux is achieved while maintaining H 98,y2 ~ 1.0. Conversely, when seeding with Ar, confinement is limited to H 98,y2 ~ 0.8 for a similar level of exhaust power.

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Compatibility of the radiating divertor with high performance plasmas in DIII-D

Cited by 40 publications

References 13 publications

Hybrid H-mode scenario with nitrogen seeding and type III ELMs in JET

Hybrid H-mode scenario with nitrogen seeding and type III ELMs in JET

Power requirements for superior H-mode confinement on Alcator C-Mod: experiments in support of ITER

Effect of N2, Ne and Ar seeding on Alcator C-Mod H-mode confinement

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