27th IAEA Fusion Energy Conference - IAEA CN-258

Session

FIP/3 DEMO & Advanced Technology

FIP/3

25 Oct 2018, 16:40

Mahatma Mandir Conference Centre

122. Overview of the DEMO Design-Staged Approach in Europe

Dr Gianfranco Federici (Eurofusion Consortium)

25/10/2018, 16:40

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

This paper describes the status of the DEMO design activities performed in Europe and discusses the impact of some of the key requirements (such as electrical output, tritium self-sufficiency), the main constraints (e.g., those deriving from the tokamak-side and the Balance-of-Plant (BoP) / Power Conversion System (PCS)) on the systems design solutions and on the overall plant architecture....

127. Development of physics and engineering designs for Japan's DEMO concept

Dr Yoshiteru Sakamoto (National Institutes for Quantum and Radiological Science and Technology)

25/10/2018, 17:00

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

Recent design progress of Japan's DEMO is presented regarding the engineering and physics conceptual design of a steady-state DEMO with a major radius of 8 m class and fusion power of 1.5 GW level. The design concept of divertor is similar to that of ITER. By considering the neutron irradiation environment, a Cu-alloy cooling pipe is used only in the large heat flux region, while a RAFM steel...

372. Novel Radio Frequency Current Drive Systems for Fusion Plasma Sustainment on DIII-D

Dr Gregory Wallace (MIT Plasma Science and Fusion Center)

25/10/2018, 17:20

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

The DIII-D National Fusion Facility is advancing the science and technology of steady-state fusion plasma sustainment through the implementation of two first-of-a-kind radio frequency current drive systems: the "helicon" or fast wave in the lower hybrid range of frequencies (LHRF), and high field side (HFS) launch of the lower hybrid slow wave. Using existing DIII-D discharges, we have...

154. Impact of High Field & High Confinement on L-mode-Edge Negative Triangularity Tokamak (NTT) Reactor

Dr Mitsuru Kikuchi (QST)

25/10/2018, 17:40

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

“NTT” is a unique reactor concept based on “power-handling-first“ philosophy by locating long-leg (~2.7m) divertor at outboard side with negative triangularity δ<0 and making flux tube expansion to maximize heat exhaust surfaces (grazing angle ~2°).
Our previous design (Ip=21MA, A=Rp/ap=3, Rp=9m, HH=1.12, Bt=5.86T) uses standard magnet design based on the wedge support and maximum field is...

29. Amelioration of plasma-material interactions and improvement of plasma performance with a flowing liquid Li limiter and Li conditioning on EAST & Experiments on FTU with a liquid tin limiter

Dr Rajesh Maingi (Princeton Plasma Physics Laboratory)

25/10/2018, 18:00

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

A. Wall conditioning and ELM control has played a crucial role in enabling access to record long H-mode pulses in EAST. Here we present new results where 1) a 2nd generation (2G) flowing liquid lithium (Li) limiter was inserted into the EAST midplane and used to mitigate plasma-materials interactions (PMI); 2) Li powder was injected to eliminate ELMs in upper-single null (USN) configuration...

406. Experiments on FTU with a liquid tin limiter

Dr Rajesh Maingi (Princeton Plasma Physics Laboratory)

25/10/2018, 18:10

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

In this paper we report experimental results obtained, for the first time in the world, in a tokamak with a liquid tin limiter (TLL). The FTU TLL was realized by using a molybdenum tube covered with Capillary Porous System (CPS) made by stripes of tungsten felt filled with tin. The TLL can be cooled by flowing air and atomized water inside a copper pipe inserted in the molybdenum tube. To test...

732. Development of a Lithium Vapor Box Divertor for Controlled Plasma Detachment

Robert Goldston (Princeton Plasma Physics Laboratory)

25/10/2018, 18:20

FIP - Fusion Engineering, Integration and Power Plant Design

Oral

A lithium vapor-box configuration [1] has been proposed to provide volumetric radiative dissipation in the divertor region of tokamak plasmas. While recent experiments have achieved continuous vapor shielding in close proximity to a lithium coated target in Magnum-PSI [2], this approach seeks to provide controlled detachment far from the divertor target, in a lithium vapor cloud maintained...