Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F61389021%3A_____%2F21%3A00552289" target="_blank" >RIV/61389021:_____/21:00552289 - isvavai.cz</a>

Výsledek na webu

<a href="https://iopscience.iop.org/article/10.1088/1741-4326/ac1803" target="_blank" >https://iopscience.iop.org/article/10.1088/1741-4326/ac1803</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1088/1741-4326/ac1803" target="_blank" >10.1088/1741-4326/ac1803</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

Popis výsledku v původním jazyce

WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fibre Bragg grating embedded few mm below the surface and flush mounted Langmuir probes. A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions: from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1 cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the 20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine ( 40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW m-2 with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW m-2 steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux.

Název v anglickém jazyce

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

Popis výsledku anglicky

WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fibre Bragg grating embedded few mm below the surface and flush mounted Langmuir probes. A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions: from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1 cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the 20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine ( 40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW m-2 with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW m-2 steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux.

Druh

J_imp - Článek v periodiku v databázi Web of Science

CEP obor

—

OECD FORD obor

10305 - Fluids and plasma physics (including surface physics)

Projekt

—

Návaznosti

I - Institucionalni podpora na dlouhodoby koncepcni rozvoj vyzkumne organizace

Rok uplatnění

2021

Kód důvěrnosti údajů

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Název periodika

Nuclear Fusion

ISSN

0029-5515

e-ISSN

1741-4326

Svazek periodika

61

Číslo periodika v rámci svazku

9

Stát vydavatele periodika

AT - Rakouská republika

Počet stran výsledku

10

Strana od-do

096027

Kód UT WoS článku

000684701900001

EID výsledku v databázi Scopus

2-s2.0-85114022169