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Diagnostic Developments for the DIII-D National Fusion Facility
- Source :
- Fusion Science and Technology. 52:367-374
- Publication Year :
- 2007
- Publisher :
- Informa UK Limited, 2007.
-
Abstract
- The DIII-D National Fusion Facility has long been a center of innovation and development of diagnostics for magnetic fusion devices. The DIII-D device, a moderate size tokamak, with a high flexibility shaping coil set, neutral beam injection (NBI), electron cyclotron heating (ECH) and ion cyclotron heating (ICH), supports a very broad research program in fusion science, including critical aspects related to burning plasmas expected to be encountered in ITER. This scientific program is supported by a large set of diagnostics (approximately 50), which is the product of a highly collaborative program between universities, national laboratories and industry. Although many diagnostic systems are now routinely employed to measure a wide range of plasma parameters, such as temperature, rotation, density and current profiles, there are many areas that are inherently difficult or prohibitively expensive to diagnose. Such areas include the measurements associated with energetic ion populations or with the characterization of plasma flows in the divertor/edge area. In addition, the study of burning plasmas will require the development of new and updated techniques, which need to be developed and tested in existing devices in relevant plasma conditions.
- Subjects :
- Nuclear and High Energy Physics
Tokamak
DIII-D
Plasma parameters
Mechanical Engineering
Divertor
Nuclear engineering
Cyclotron
Plasma
Neutral beam injection
law.invention
Nuclear Energy and Engineering
law
Electromagnetic coil
Environmental science
General Materials Science
Atomic physics
Civil and Structural Engineering
Subjects
Details
- ISSN :
- 19437641 and 15361055
- Volume :
- 52
- Database :
- OpenAIRE
- Journal :
- Fusion Science and Technology
- Accession number :
- edsair.doi...........0f974cad5126499a7a0d63fe5bccf103