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m (number of ports was wrong (see Ref. http://dx.doi.org/10.1088/0029-5515/53/12/126001 ))
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[[File:W7X.png|400px|thumb|right|W7-X model]]  
[[File:W7X.png|400px|thumb|right|W7-X model]]  
Wendelstein 7-X (W7-X) is an experimental stellarator currently being built in Greifswald, Germany by the [[:Wikipedia:Max-Planck-Institut_f%C3%BCr_Plasmaphysik|Max-Planck-Institut für Plasmaphysik]] (IPP). W7-X is an [[Stellarator optimization|optimized stellarator]], i.e. the magnetic field has been tailored to meet several physical optimization criteria.
Wendelstein 7-X (W7-X) is an experimental stellarator currently being operated in Greifswald, Germany by the [[:Wikipedia:Max-Planck-Institut_f%C3%BCr_Plasmaphysik|Max-Planck-Institut für Plasmaphysik]] (IPP). W7-X is an [[Stellarator optimization|optimized stellarator]], i.e. the magnetic field has been tailored to meet several physical optimization criteria.
 
== Background ==
 
The Wendelstein 7-X (W7-X) is a fusion research device that uses the [[stellarator]] concept to confine plasma in a magnetic field. As of 2023, it is the world's largest stellarator device, and its main goal is to demonstrate the feasibility of continuous operation in a fusion power plant. The W7-X is located in Greifswald, Germany, and is an independent partner project of the Max-Planck Institute for Plasma Physics with the University of Greifswald.
 
The W7-X is based on a five-field-period Helias configuration and consists of 50 non-planar and 20 planar superconducting magnetic coils. These coils induce a magnetic field that prevents the plasma from colliding with the reactor walls, allowing it to be confined for longer periods. The plasma vessel, built of 20 parts, is on the inside and adjusted to the complex shape of the magnetic field, and it has 254 ports for plasma heating and observation diagnostics. The heating system includes high power gyrotrons for electron cyclotron resonance heating (ECRH), which can deliver up to 15 MW of heating to the plasma. Additionally, neutral beam injection and ion cyclotron resonance heating (ICRH) systems are also available for physics operation.
 
The W7-X has undergone several operational phases, each with its specific objectives. During Operational Phase 1 (OP1.1), the W7-X produced helium plasma for about 0.1 seconds, followed by hydrogen plasma with gradually increasing discharge power and duration. More than 2,000 pulses were conducted before the shutdown. Five poloidal graphite limiters served as the main plasma-facing components during this first campaign (instead of the divertor modules). Experimental observations confirmed 3D modeling predictions that showed heat and particle flux deposition patterns on the limiters in clear correlation with the lengths of the open magnetic field lines in the plasma boundary.
 
Operational Phase 1 continued (OP1.2) in 2017 to test the uncooled divertor. During the last experiments of 2018, the density reached 2 × 10<sup>20</sup> particles/m<sup>3</sup> at a temperature of 20 million degrees. With good plasma values, long-lasting plasmas with long discharge times of 100 seconds were obtained, and energy content exceeded 1 megajoule. In 2018, a record ion temperature of about 40 million degrees, a density of 0.8 × 10<sup>20</sup> particles/m<sup>3</sup>, and a confinement time of 0.2 seconds yielded a record fusion product of 6 × 10<sup>26</sup> degree-seconds per cubic metre.
 
The limiter and divertor experimental campaigns aimed to test plasma-wall interactions in the complex 3D geometry of the W7-X. The limiter campaign presented the first-time characterization of the limiter heat loads, confirming major geometry effects of the connection length on heat transport predicted by 3D modeling. The divertor experimental campaign tested the [[Island Divertor|island divertor]] concept, which plays a central role in the device mission to demonstrate reactor relevant plasma confinement for steady-state time scales of up to 30 minutes in the high-performance campaign (OP2). During the first campaign with the inertially cooled island divertor, a large step in the experimental qualification of this divertor concept was made.


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* [https://www.ipp.mpg.de/16900/w7x Wendelstein 7-X]
* [https://www.ipp.mpg.de/16900/w7x Wendelstein 7-X]
* [[Island Divertor]]


== References ==
== References ==
Line 59: Line 72:
* L. Wegener, ''Status of Wendelstein 7-X construction'', [[doi:10.1016/j.fusengdes.2009.01.106|Fusion Engineering and Design '''84''', 2-6 (2009) 106-112]]
* L. Wegener, ''Status of Wendelstein 7-X construction'', [[doi:10.1016/j.fusengdes.2009.01.106|Fusion Engineering and Design '''84''', 2-6 (2009) 106-112]]
* H.-S. Bosch et al, ''Construction of Wendelstein 7-X; Engineering a Steady-State Stellarator'', [[doi:10.1109/TPS.2009.2036918|IEEE Trans. Plasma Science '''38''', 3 (2010) 265]]
* H.-S. Bosch et al, ''Construction of Wendelstein 7-X; Engineering a Steady-State Stellarator'', [[doi:10.1109/TPS.2009.2036918|IEEE Trans. Plasma Science '''38''', 3 (2010) 265]]
* H.-S. Bosch et al, ''Technical challenges in the construction of the steady-state stellarator Wendelstein 7-X'', [[doi:10.1088/0029-5515/53/12/126001|Nucl. Fusion '''53''' (2013) 126001]]
* D. Clery, ''Feature: The bizarre reactor that might save nuclear fusion'', [[doi:10.1126/science.aad4746|Science, 21 October 2015]]
* O. Neubauer et al, ''Diagnostic setup for investigation of plasma wall interactions at Wendelstein 7-X'', [[doi:10.1016/j.fusengdes.2015.06.102|Fusion Engineering and Design '''96-97''' (2015) 891-894]]
* T. Sunn Pedersen et al, ''Plans for the first plasma operation of Wendelstein 7-X'', [[doi:10.1088/0029-5515/55/12/126001|Nucl. Fusion '''55''' (2015) 126001]]
* M. Krychowiak et al, ''Overview of diagnostic performance and results for the first operation phase in Wendelstein 7-X'', [[doi:10.1063/1.4964376|Review of Scientific Instruments '''87''' (2016) 11D304]]
* R.C. Wolf et al, ''Major results from the first plasma campaign of the Wendelstein 7-X stellarator'', [[doi:10.1088/1741-4326/aa770d|Nucl. Fusion '''57''' (2017) 102020]]
* D. Hathiramani et al, ''Upgrades of edge, divertor and scrape-off layer diagnostics of W7‐X for OP1.2'', [[doi:10.1016/j.fusengdes.2018.02.013|Fusion Engineering and Design '''136A''' (2018) 304-308]]
* A. Dinklage et al, ''Magnetic configuration effects on the Wendelstein 7-X stellarator'', [[doi:10.1038/s41567-018-0141-9|Nature Phys '''14''' (2018) 855–860]]
* R.C. Wolf et al, ''Performance of Wendelstein 7-X stellarator plasmas during the first divertor operation phase'', [[doi:10.1063/1.5098761|Physics of Plasmas '''26''' (2019) 082504]]


[[Category:Toroidal confinement devices]]
[[Category:Toroidal confinement devices]]

Latest revision as of 07:37, 10 April 2023

W7-X model

Wendelstein 7-X (W7-X) is an experimental stellarator currently being operated in Greifswald, Germany by the Max-Planck-Institut für Plasmaphysik (IPP). W7-X is an optimized stellarator, i.e. the magnetic field has been tailored to meet several physical optimization criteria.

Background

The Wendelstein 7-X (W7-X) is a fusion research device that uses the stellarator concept to confine plasma in a magnetic field. As of 2023, it is the world's largest stellarator device, and its main goal is to demonstrate the feasibility of continuous operation in a fusion power plant. The W7-X is located in Greifswald, Germany, and is an independent partner project of the Max-Planck Institute for Plasma Physics with the University of Greifswald.

The W7-X is based on a five-field-period Helias configuration and consists of 50 non-planar and 20 planar superconducting magnetic coils. These coils induce a magnetic field that prevents the plasma from colliding with the reactor walls, allowing it to be confined for longer periods. The plasma vessel, built of 20 parts, is on the inside and adjusted to the complex shape of the magnetic field, and it has 254 ports for plasma heating and observation diagnostics. The heating system includes high power gyrotrons for electron cyclotron resonance heating (ECRH), which can deliver up to 15 MW of heating to the plasma. Additionally, neutral beam injection and ion cyclotron resonance heating (ICRH) systems are also available for physics operation.

The W7-X has undergone several operational phases, each with its specific objectives. During Operational Phase 1 (OP1.1), the W7-X produced helium plasma for about 0.1 seconds, followed by hydrogen plasma with gradually increasing discharge power and duration. More than 2,000 pulses were conducted before the shutdown. Five poloidal graphite limiters served as the main plasma-facing components during this first campaign (instead of the divertor modules). Experimental observations confirmed 3D modeling predictions that showed heat and particle flux deposition patterns on the limiters in clear correlation with the lengths of the open magnetic field lines in the plasma boundary.

Operational Phase 1 continued (OP1.2) in 2017 to test the uncooled divertor. During the last experiments of 2018, the density reached 2 × 1020 particles/m3 at a temperature of 20 million degrees. With good plasma values, long-lasting plasmas with long discharge times of 100 seconds were obtained, and energy content exceeded 1 megajoule. In 2018, a record ion temperature of about 40 million degrees, a density of 0.8 × 1020 particles/m3, and a confinement time of 0.2 seconds yielded a record fusion product of 6 × 1026 degree-seconds per cubic metre.

The limiter and divertor experimental campaigns aimed to test plasma-wall interactions in the complex 3D geometry of the W7-X. The limiter campaign presented the first-time characterization of the limiter heat loads, confirming major geometry effects of the connection length on heat transport predicted by 3D modeling. The divertor experimental campaign tested the island divertor concept, which plays a central role in the device mission to demonstrate reactor relevant plasma confinement for steady-state time scales of up to 30 minutes in the high-performance campaign (OP2). During the first campaign with the inertially cooled island divertor, a large step in the experimental qualification of this divertor concept was made.

Parameter Value Unit
Major radius, R0: 5.5 m
Minor radius, a: 0.53 m
Plasma volume, V: 30 m3
Non-planar coils: 50
Planar coils: 20
Number of ports: 254
Rotational transform, ι/2π: 5/6-5/4
Magnetic field on axis, B0: <3 T
Stored energy, W: 600 MJ
Heating power, P: 15-30 MW
Pulse length: 30 min
Machine height: 4.5 m
Machine diameter: 16 m
Machine mass: 725 t

Optimization criteria

See also

References

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