High radiation and high density experiments in JT-60U

Kubo, H.; Sakurai, S.; Asakura, N.; Konoshima, S.; Tamai, Hiroshi; Higashijima, S.; Sakasai, A.; Takenaga, H.; Itami, K.; Shimizu, K.; Fujita, T.; Kamada, Y.; Koide, Y.; Shirai, Hiroshi; Sugie, Toshiharu; Nakano, T.; Oyama, N.; Urano, H.; Ishijima, Tatsuo; Hill, K. W.; Ernst, D. R.; Leonard, A.W.; Team, Jt

doi:10.1088/0029-5515/41/2/310

Cited by 55 publications

(45 citation statements)

References 19 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In ELMy H-mode discharges with Ar injection (1.2 MA, 2.5 T, q 95 = 3.4, δ = 0.36), H Hy2 ~ 0.9 has been obtained with a high radiation loss power fraction (~80%) at n e /n GW ~ 0.7 [42]. The H Hy2 -factor is ~50% higher than that without Ar injection (Fig.…”

Section: Characteristics Of Elmy H-mode Pedestalmentioning

confidence: 89%

“…Another approach towards high density is Ar injection into high triangularity H-mode. In this regime, H 89P ~1.7 (H Hy2~1 ) has been achieved up to n e ~ 0.8 n GW [42,43]. Such improvements achieved by high-δ, Arinjection, and pellet injection are all related to the high pedestal temperature discussed in the next section.…”

Section: Extension Of High Confinement Regime Towards High Densitymentioning

confidence: 94%

See 1 more Smart Citation

Profile and Shape Control of the Plasmas with Transport Barriers towards the Sustainment of High Integrated Performance in JT-60U.

Kamada¹

2002

J. Plasma Fusion Res.

View full text Add to dashboard Cite

With the main aim of providing physics basis for ITER and the steady-state tokamak reactor, JT-60U has been optimizing the operational concepts and extending discharge regimes toward the sustainment of high integrated performance. In the two advanced operation regimes, the reversed magnetic shear (RS) and the weak magnetic shear (high-β p ) ELMy H modes characterized by both internal (ITB) and edge (ETB) transport barriers, JT-60U has clarified responses of the transport barrier structure to magnetic shear, heating power, momentum, density, etc. Based on these observations, possible parameter linkages and feedback loops in the core and pedestal regimes have been proposed. With the optimization of profile and shape control, both confinement and high β stability have been enhanced and favorable integrated performance has been achieved. Such operational regimes have been extended to the reactor-relevant regime.

show abstract

Section: Characteristics Of Elmy H-mode Pedestalmentioning

confidence: 89%

Section: Extension Of High Confinement Regime Towards High Densitymentioning

confidence: 94%

Profile and Shape Control of the Plasmas with Transport Barriers towards the Sustainment of High Integrated Performance in JT-60U.

Kamada¹

2002

J. Plasma Fusion Res.

View full text Add to dashboard Cite

show abstract

“…The use of Ar to disperse divertor power flux has also been applied in JT-60U with initial results similar to that of JET -0.5 % of argon made it possible to create plasmas having f R =0.8, f G = 0.65 and H H98(y,2) =1. Even though a half of the radiated power was dissipated in a mantle inside the separatrix, electron and ion temperatures remained high in the centre and also in the pedestal [136,137,138,139]. Good confinement (H H98(y,2) =0.96) at high density ( _ n e /n GW =0.92) with high radiation loss fraction (f R~1 ) has been demonstrated more recently in JT-60U high β p H-mode plasmas by utilizing high-fieldside pellets and Ar injection.…”

Section: Use Of Extrinsic Impurity Radiation To Reduce Divertor Heat mentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

Self Cite

View full text Add to dashboard Cite

Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

show abstract

“…Impurity injection has other advantages also. "Large" tokamaks, such as JET and JT-60U, have demonstrated that good energy confinement is consistent at higher densities under certain impurity injection scenarios [55] [56] [57]. The lower plasma edge temperatures which occur in a radiating mantle, particularly with the ion component, reduces the ion temperatures in the SOL, which in turn, reduces sputtering at the divertor plates from "hot" ions.…”

Section: Radiating Mantle Approachmentioning

confidence: 99%

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

Jardin¹,

Kessel²,

Mau³

et al. 2003

View full text Add to dashboard Cite

The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A ≡ R/a = 4.0, an elongation and triangularity of κ = 2.20, δ = 0.90 (evaluated at the separatrix surface), a toroidal beta of β = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of β N ≡ 100 × β/(I P (MA)/a(m)B(T )) = 5.4. These beta values are chosen to be 10% below the ideal MHD stability limit. The bootstrap-current fraction is f BS ≡ I BS /I P = 0.91. This leads to a design with total plasma current I P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current drive system consists of ICRF/FW for on-axis current drive and a Lower Hybrid system for off-axis. Transport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

show abstract

High radiation and high density experiments in JT-60U

Cited by 55 publications

References 19 publications

Profile and Shape Control of the Plasmas with Transport Barriers towards the Sustainment of High Integrated Performance in JT-60U.

Profile and Shape Control of the Plasmas with Transport Barriers towards the Sustainment of High Integrated Performance in JT-60U.

Chapter 4: Power and particle control

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

Contact Info

Product

Resources

About