Conveners
PWI session
- Sebastijan Brezinsek (Forschungszentrum Jülich)
PWI session
- Kazuaki HANADA
PWI session
- Sebastijan Brezinsek (Forschungszentrum Jülich)
QUEST (Q-shu University experiments with Steady-state Spherical Tokamak) [1] is a medium sized spherical tokamak (ST) (R=0.64m a=0.4m, BT<0.25T @R=0.6m) in Kyushu University. The heating sources for plasma are two RFs which are capable of operation with 28GHz, 350kW, a few second and 8.2 GHz 50kW, CW. Plasma facing walls (PFWs) are mainly composed of stainless steel type 316L and stainless...
Fuel recycling is one of the key issues for long pulse operation for tokamaks. During long pulse operation in tokamaks, the accumulation of fuel particles on the first wall leads to a decreasing of wall pumping capability, and eventually the wall changes to outgassing from pumping due to the accumulation of fuel retention and the increasing of surface temperature. This enhances fuel recycling...
Effective control of fuel recycling and impurity is very key for achievement of long pulse and high-performance plasmas. High recycling and impurity concentration in the plasma would result in usually degradation of plasma confinement, and uncontrollable plasma density and disruptions. Some advanced vacuum and wall conditioning technologies, have been developed and widely used in EAST to...
It is estimated that long-pulse fusion devices may experience rates of net erosion and deposition of solid PFC (Plasma Facing Component) material of 10^3 – 10^5 kg/year, whatever the material used [1]. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘UFOs’...
A fusion reactor based on a stellarator design has the advantage of easier access to long pulse scenarios. In fact, one of the main goals of Wendelstein 7-X (W7-X), the largest advanced stellarator in the world, is to demonstrate the steady-state capabilities of the stellarator line. Therefore, in the recent campaign, a number of experiments were performed in order to prepare long pulse...
Attempts using various particle control knobs have been made at Large Helical Device (LHD) to achieve steady-state plasmas in long pulse discharges. Divertor pumping is an important tool to control plasma density in fusion plasmas. In the divertor region, neutral particles shall be compressed and efficiently pumped out. In the LHD, the development of divertor pumping has been strongly...
The hydrogen isotopes retention and subsequent excessive desorption (dynamic retention) on the plasma facing walls (PFWs) frequently lead to density runaway and plasma termination due to R > 1, where R means recycling ratio. This issue has a significant impact on steady state operation (SSO) of fusion experimental devices. Recently, the use of metallic materials such as tungsten in PFWs on...
The new lower tungsten divertor with horizontal and vertical targets has already been developed and installed in EAST for high-power and long-pulse operation in a full metal wall environment. The flexible magnetic configurations allow the position of lower outer strike point located on either horizontal or vertical target with different divertor configurations. Preliminary experimental results...
One critical issue for achieving stationary high-confinement mode (H-mode) operation of the next-step tokamak fusion reactors is to control the transient heat load induced by large-amplitude edge-localized modes (ELMs) without significant degradation of the plasma performance. Although various small/no ELM H-mode regimes have been achieved on present tokamaks, a detailed physics understanding...
Plasma-facing materials (PFM) for next generation fusion devices like ITER will be submitted to intense fluxes of He and H isotopes (H, D and radioactive T transmutating as He). This is particularly significant for the divertor components, for which tungsten (W) is the first-choice material thanks to its low sputtering yield, low HI retention and high melting point. Plasma-wall interactions...
