- Email: [email protected]
First study of EAST divertor by impurities puffing
Journal of Nuclear Materials 415 (2011) S345–S348
Contents lists available at ScienceDirect
Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat
First study of EAST divertor by impurities puffing Z.W. Wu a,⇑, H.Y. Guo a,b, Y.J. Chen a, D.S. Wang c, G.N. Luo a, X. Gao a, W. Gao a, L. Zhang a, W.Q. Zha a, D. Kato d, I. Murakami d, W. Zhang a, J.H. Yang a, J.H. Wu a, S.Z. Zhu a, J. Wu a, H. Chen a, L.Q. Hu a, B.N. Wan a, J. Li a a
Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031, China Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, CA 92688-7010, USA c Department of Modern Physics, University of Science and Technology of China, Hefei, Anhui 230026, China d National Institute for Fusion Science, Toki, Gifu 509-5292, Japan b
a r t i c l e
i n f o
Article history: Available online 21 December 2010
a b s t r a c t A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in–out heat flux distribution asymmetry is improved. Ó 2010 Elsevier B.V. All rights reserved.
1. Introduction
2. Experimental setup
The Experimental Advanced Superconducting Tokamak (EAST) [1,2] was built to achieve high power and long pulse operation for investigating reactor-relevant issues under steady-state conditions for the next step device, such as ITER [3]. As in ITER, impurity control, particle and power exhaust are the key issues for achieving high power, long pulse operations in EAST with envisioned pulse length up to 1000 s. Divertor is the main region for plasma-wall-interactions in fusion devices. It is of great importance to understand impurity transport and divertor screening for the control of impurity source and radiation in the plasma boundary without significantly cooling or diluting the fusion fuel in the confined plasma. The basic functions of the divertor are to control the core plasma impurity content, remove the helium ash and particle and power exhaust, and achieve better confinement (H-mode) [4], etc. The plasma impinges onto the target plates, causing the wall material to be released as impurities either by physical or chemical sputtering. The neutral impurities are ionized and distributed in the plasma by various transport mechanisms. This paper presents the first experimental study of the EAST divertor by seeding CH4 and high-Z impurity (Ar) in the ohmic-mode plasmas with the double null divertor configuration on EAST.
The EAST superconducting tokamak with major radius of 1.7 m and minor radius of 0.4 m has a very flexible poloidal field control system, which accommodates a wide range of magnetic configurations including the limiter, single null and double null divertor configurations [5,6]. EAST has been upgraded with a full graphite wall in the spring of 2008. The inertial cooled tiles are bolted to the water-cooled divertor plates and vessel wall. The tiles are made of the doped graphite named GBST1308 (1%B4C, 2.5%Si, 7.5%Ti) with SiC coating. The maximum power flux to the divertor target plates must be maintained below 2 MW/m2 to ensure long pulse operations. Thus, a large fraction of power from the core plasma transported into the SOL must be lost by volumetric process, such as radiation and charge exchange, before reaching the divertor target plates along the field lines. Impurities such as argon, neon and CH4 have been introduced into divertor from the divertor gas fuelling system in order to alleviate the power flux to divertor target plates by an increase of edge impurity radiations. In the same time, the basic divertor plasma characteristics and divertor screening for impurities have also been studied in the divertor plasmas on EAST. A new multipurpose gas fuelling system has been installed for the latest divertor physics campaign, allowing gas introduction from various localized poloidal positions such as inner and outer mid-plane, inner and outer divertors, and dome for both upper and lower divertors, as shown in Fig. 1. The steady-state separatrix of double null configuration in Shot No. 11668 is also drawn in this figure. Various gases including D2, CH4 and Ar have been used to control divertor asymmetry, and to study gas fuelling efficiency, divertor screening for impurities and radiative divertor effects. In
⇑ Corresponding author. E-mail address: [email protected] (Z.W. Wu). 0022-3115/$ - see front matter Ó 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2010.12.040
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Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
Fig. 1. A new gas puffing system on EAST to allow gas puffing from various poloidal locations. 1, 2, 3 and 4-outer mid-plane jets, 5-inner mid-plane jet, 6, 7 and 8-upper divertor jets, 9, 10 and 11-lower divertor jets, 12-cryopump. The steady-state separatrix of double null configuration in Shot No. 11668 is also drawn in this figure.
addition, an extensive set of divertor and edge plasma diagnostics has been installed and upgraded in EAST to facilitate these studies, including the key divertor/edge probes and spectroscopy diagnostics. There is 222 divertor target embedded graphite probes, which can be configured as 74 triple probes or single probes, and two sets of reciprocating probes inserted from window A and window E at the mid-plane, as shown in Fig. 2. There is 18-channel Da and 18-channel CII/CIII viewing the lower outboard divertor from the top of the machine, two arrays of 35-channel Da viewing the inner target and dome surfaces of both upper and lower divertors from the outer mid-plane through the in-vessel reflection mirrors, one UV and Visible spectrometer with CCD (200–750 nm) viewing the lower-outer divertor, two sets of monochromators for CIII and OII lines and one interference filter for Ha monitor in the outer mid-plane, and 8-channel visible bremsstrahlung for Z-effective, as shown in Fig. 3. These diagnostics can provide useful information such as electron temperature, electron density, ion saturation current, Mach numbers, particle and impurity fluxes etc. in the divertor and SOL region. 3. Results and discussion 3.1. Basic divertor plasma characteristics Fig. 4 shows the basic divertor plasma behaviors for a typical ohmic discharge of Shot No. 11668 with Ip = 250 kA, BT = 2 T and
Fig. 3. View geometry of divertor and edge spectroscopy diagnostics.
ne ramps up from 1.0 to 3.0 1019 m3 from 3.5 s to 6.5 s. The discharge duration is about 7 s and the plasma is in stable double null (DN) divertor configuration from 3.5 s to 6.5 s. The directions of toroidal magnetic field and plasma current both are anticlockwise. Detachment can be achieved in Ohmic discharges by ramping up plasma density. As line-averaged density, ne increases, ion saturation current, Js, obtained from the divertor target Langmuir probe near strike points (UI_11, UO_12), first increases, then starts to roll over as ne further increases, finally decreases at sufficiently high densities as the plasma enters the detached regime. And ion saturation current, Js, obtained from the divertor target Langmuir probe far from strike points (UI_07, UO_05) increases latterly than the ones near strike points. Detachment starts near the strike point, then expands further out along the divertor target, as plasma density rises. Divertor neutral fluxes, as indicated by Da (Da_19c), keep e , but shows a tendency to saturate during increasing with n detachment. In addition, there is a strong in–out divertor asymmetry in particle and heat fluxes, as seen by the target probes. The electron density net obtained from the divertor target Langmuir probe in upper_outer divertor (UO_12 and UO_05) is much higher than the ones from upper_inner divertor (UI_11 and UI_07). The electron temperature Tet is just the reverse. Note that for the particular discharge shown in Fig. 4, the ion B rB drift is directed toward the upper divertor, and only the upper divertor measurements are shown. Usually the divertor plasma starts to de e exceeds 1.5 1019 m3. The detachment occurs first tach when n at inner target, followed by the detachment at the outer target at a e , i.e., about 2 1019 m3. When the detachment arose, higher n both Js and Tet are significantly reduced with Tet 5 eV near the separatrix. 3.2. Divertor screening for carbon
Fig. 2. Distribution of divertor target probes and reciprocating probe diagnostics.
Carbon is the dominant impurity in most of full graphite wall tokamaks such as EAST. Methane screening experiments have been conducted to quantify the ability of the SOL/divertor to ionize carbon and transport it to the divertor, preventing core plasma contamination in different plasma confinement regimes [7–9]. A series of experiments has recently been carried out under different plasma conditions to investigate the source distribution and divertor screening efficiency for the intrinsic carbon impurity on EAST, by injecting the pure methane with active divertor pumping. The
Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
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Fig. 4. Time evolution of an Ohmic double null discharge during a density ramp-up of Shot No. 11668. ne – line-averaged electron density, Da_19c – the radiation intensity of Da line corresponded to the ion saturation Js_UI_11 from upper-inner divertor, Js_UI_11 (UO_12) – ion saturation current near the upper-inner (upper-outer) strike points, Js_UI_07 (UO_05) – ion saturation current far from the upper-inner (upper-outer) strike points, net and Tet – electron density and temperature corresponded to the ion saturation Js_UI_11 (07) and Js_UO_12 (05).
experiment was primarily focused on ohmic discharges with Ip = 250 kA, BT = 2T and ne = 1.0, 1.5 and 2.0 1019 m3 three kinds of density platform discharges with discharge duration of about 7 s. The directions of toroidal magnetic field is clockwise and plasma current is anticlockwise. The carbon screening was measured by injecting CH4 at 4.1 s for 0.09 s from various poloidal locations, i.e., inner mid-plane, inner divertor target, outer divertor target and dome for different electron densities, i.e., 1.0, 1.5 and 2.0 1019 m3. Fig. 5 shows the time evolution of CIII measured at the LFS inner wall near the lower-outer divertor under different operating conditions. It is found that for the double null diverted configuration, strong screening occurs when the methane or methane mixture is injected at the inner mid-plane, or the outer target with little
changes in CIII intensity following the injection of CH4. In contrast, injection from the inner target, especially the dome leads to a large increase in CIII intensity, indicating weak divertor screening for carbon impurity from the inner target and private flux region. Inner mid-plane puffing has the best divertor screening, possibly attributed to the strong flow in the inner SOL [10]. The strong divertor screening for the outer divertor puffing may arise from the higher electron temperature so that neutrals can be ionized in the presheath region close to the target plate where strong flow is also present. The divertor screening for impurities exhibits a strong dependence on divertor plasma operation regimes with weak screening efficiency for the detached plasma regime (ne 2.0 1019 m3), but improving for the attached regime (ne 1.0–1.5 1019 m3), as also shown in Fig. 5. As can be seen,
Fig. 5. Changes in CIII intensity with CH4, puffing at outer target, inner target, private flux region (dome), and inner mid-plane, in DN divertor plasmas at 3.63 1021 particles/s, starting at 4.1 s for 0.09 s.
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Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
3.4. Discussion To further study and actively control divertor plasma, a large number of new divertor diagnostics have been installed in the Spring of 2010, including an IR camera, an array of bolometer detectors and a pressure gauge under the lower-outer divertor target plate. In order to achieve high power, long pulse operations, we will develop an active feedback control system for the ion saturation current to maintain the divertor plasma under partial detachment conditions. In addition, the ‘puff-and-pump’ technique [15] will be investigated to further reduce the leakage of injected high-Z impurities into the core plasma. This will be helpful and useful for us to improve this study. L-mode and H-mode divertor target plasma will be focused on next step study in EAST. 4. Summary and conclusions
19
3
Fig. 6. Time traces of an Argon seeded discharge with ne = 1.5 10 m in DN divertor plasma of Shot No. 12924. ð3:32 1021 particles=sÞ at 5 s over 0.15 s), (a) plasma current, (b) central line-averaged density, (c) central electron temperature, (d) radiation power (measured by XUV bolometer), (e) divertor Da intensity, (f) ion saturation current of the inner and outer strike points, (g) electron temperature of the inner and outer strike points, (h) peak heat flux of the inner and outer strike points.
C2+ influx is slightly increased with D = 0.18–0.27, where D is the relative change in C2+ influx, when CH4 is injected into the plasma at 4.1 s in the attached DN divertor plasmas. In contrast, as the plasma detaches at high densities, C2+ influx exhibits a strong increase with D = 0.32–1.62, which suggests decreasing divertor/ SOL screening efficiency at high density detached regime. However, the inner mid-plane puffing always shows strong screening, independent of the divertor plasma conditions. 3.3. Effect of Ar injection Extrinsic impurity seeding has been used in many tokamaks such as DIII-D [11], JET [12], ASDEX-Upgrade [13] and NSTX [14] to achieve partial or complete divertor detachment with strongly radiative divertor plasmas, which has also been considered as a candidate for the control of target heat load to acceptable levels in ITER. We have conducted the first radiative divertor experiments on EAST in April, 2009, by injecting the highly radiating recycling impurity Ar and its mixture with deuterium, i.e., 5.7% Ar in D2 with active divertor pumping in DN divertor ohmic discharges. Fig. 6 shows the results with 5.7% Ar in D2 being injected into the outer divertor chamber. The peak heat flux at the outer target plate is significantly reduced (over 50% reduction), by increasing radiated power in the divertor. The divertor plasma temperature and sheath potential are also reduced. This is accompanied by the strong increase main plasma density and strongly radiating plasma region located between the outer divertor separatrix strike point and the X-point. It is observed that the ion saturation current and peak heat flux of the outer strike points is almost reduced to the level of inner strike points after Ar puffing, hence improving the in–out heat flux distribution asymmetry. However, it is 2–5 times larger than the values of inner strike points before puffing. And the electron temperature of the inner and outer strike points is also reduced to 6 eV near the electron temperature in detachment regime after Ar puffing.
Detachment of the divertor plasma has been achieved by density ramp-up in EAST. Impurity screening experiments have been conducted by methane puffing in the different conditions. Strong screening occurs when the methane is injected at the inner midplane, or the outer divertor target with little changes in CIII intensity from LFS inner wall near the lower-outer divertor following the injection of CH4. In contrast, injection from the inner target, especially the dome leads to a large increase in CIII intensity, indicating weak divertor screening for impurities from the inner target and private flux region. The divertor screening for impurities exhibits a strong dependence on divertor plasma operation regimes with weak screening efficiency for the higher density, detached regime, but improving for the low and intermediate densities, attached regime. In addition, radiative divertor experiments have been conducted in EAST, by injecting the highly radiating recycling impurity Ar (5.7% Ar in D2) with active divertor pumping. The injection of mixed Ar into the outer divertor chamber significantly reduces the peak heat flux at the outer target plate, hence improving the in–out heat flux distribution asymmetry. This shows an effective means to control heat fluxes at divertor target plates for future high power, long pulse operation. Acknowledgements The authors would like to acknowledge the support from the rest of the EAST Team. This work is funded by National Nature Science Foundation of China under Contract Nos. 10475078 and 10728510, and also partly supported by JSPS-CAS Core-University Program in the Field of Plasma and Nuclear Fusion. Reference [1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] [15]
Y.X. Wan et al., Nucl. Fusion 40 (2000) 1057. P.D. Weng et al., Fusion Eng. Des. 75–79 (2005) 143. R. Aymar et al., Nucl. Fusion 41 (2001) 1301. The ASDEX Team, Nucl. Fusion 29 (1989) 1959. X. Gao et al., Phys. Lett. A 372 (2008) 2286. B.N. Wan et al., Plasma Sci. Technol. 9 (2007) 125. G.M. McCracken et al., Nucl. Fusion 33 (1993) 1409. M.F. Stamp et al., J. Nucl. Mater. 266–269 (1999) 685. J.D. Strachan et al., J. Nucl. Mater. 290–293 (2001) 972. M.R. Wade et al., J. Nucl. Mater. 266–269 (1999) 44. M.E. Fenstermacher et al., Plasma Phys. Controlled Fusion 41 (1999) A345. J. Rapp et al., J. Nucl. Mater. 337–339 (2005) 826. J. Neuhauser et al., Plasma Phys. Controlled Fusion 37 (1995) A37. V.A. Soukhanovskii et al., J. Nucl. Mater. 363–365 (2007) 432. S. Ishida, JT-60 Team, Nucl. Fusion 39 (1999) 1211.
Contents lists available at ScienceDirect
Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat
First study of EAST divertor by impurities puffing Z.W. Wu a,⇑, H.Y. Guo a,b, Y.J. Chen a, D.S. Wang c, G.N. Luo a, X. Gao a, W. Gao a, L. Zhang a, W.Q. Zha a, D. Kato d, I. Murakami d, W. Zhang a, J.H. Yang a, J.H. Wu a, S.Z. Zhu a, J. Wu a, H. Chen a, L.Q. Hu a, B.N. Wan a, J. Li a a
Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031, China Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, CA 92688-7010, USA c Department of Modern Physics, University of Science and Technology of China, Hefei, Anhui 230026, China d National Institute for Fusion Science, Toki, Gifu 509-5292, Japan b
a r t i c l e
i n f o
Article history: Available online 21 December 2010
a b s t r a c t A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in–out heat flux distribution asymmetry is improved. Ó 2010 Elsevier B.V. All rights reserved.
1. Introduction
2. Experimental setup
The Experimental Advanced Superconducting Tokamak (EAST) [1,2] was built to achieve high power and long pulse operation for investigating reactor-relevant issues under steady-state conditions for the next step device, such as ITER [3]. As in ITER, impurity control, particle and power exhaust are the key issues for achieving high power, long pulse operations in EAST with envisioned pulse length up to 1000 s. Divertor is the main region for plasma-wall-interactions in fusion devices. It is of great importance to understand impurity transport and divertor screening for the control of impurity source and radiation in the plasma boundary without significantly cooling or diluting the fusion fuel in the confined plasma. The basic functions of the divertor are to control the core plasma impurity content, remove the helium ash and particle and power exhaust, and achieve better confinement (H-mode) [4], etc. The plasma impinges onto the target plates, causing the wall material to be released as impurities either by physical or chemical sputtering. The neutral impurities are ionized and distributed in the plasma by various transport mechanisms. This paper presents the first experimental study of the EAST divertor by seeding CH4 and high-Z impurity (Ar) in the ohmic-mode plasmas with the double null divertor configuration on EAST.
The EAST superconducting tokamak with major radius of 1.7 m and minor radius of 0.4 m has a very flexible poloidal field control system, which accommodates a wide range of magnetic configurations including the limiter, single null and double null divertor configurations [5,6]. EAST has been upgraded with a full graphite wall in the spring of 2008. The inertial cooled tiles are bolted to the water-cooled divertor plates and vessel wall. The tiles are made of the doped graphite named GBST1308 (1%B4C, 2.5%Si, 7.5%Ti) with SiC coating. The maximum power flux to the divertor target plates must be maintained below 2 MW/m2 to ensure long pulse operations. Thus, a large fraction of power from the core plasma transported into the SOL must be lost by volumetric process, such as radiation and charge exchange, before reaching the divertor target plates along the field lines. Impurities such as argon, neon and CH4 have been introduced into divertor from the divertor gas fuelling system in order to alleviate the power flux to divertor target plates by an increase of edge impurity radiations. In the same time, the basic divertor plasma characteristics and divertor screening for impurities have also been studied in the divertor plasmas on EAST. A new multipurpose gas fuelling system has been installed for the latest divertor physics campaign, allowing gas introduction from various localized poloidal positions such as inner and outer mid-plane, inner and outer divertors, and dome for both upper and lower divertors, as shown in Fig. 1. The steady-state separatrix of double null configuration in Shot No. 11668 is also drawn in this figure. Various gases including D2, CH4 and Ar have been used to control divertor asymmetry, and to study gas fuelling efficiency, divertor screening for impurities and radiative divertor effects. In
⇑ Corresponding author. E-mail address: [email protected] (Z.W. Wu). 0022-3115/$ - see front matter Ó 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2010.12.040
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Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
Fig. 1. A new gas puffing system on EAST to allow gas puffing from various poloidal locations. 1, 2, 3 and 4-outer mid-plane jets, 5-inner mid-plane jet, 6, 7 and 8-upper divertor jets, 9, 10 and 11-lower divertor jets, 12-cryopump. The steady-state separatrix of double null configuration in Shot No. 11668 is also drawn in this figure.
addition, an extensive set of divertor and edge plasma diagnostics has been installed and upgraded in EAST to facilitate these studies, including the key divertor/edge probes and spectroscopy diagnostics. There is 222 divertor target embedded graphite probes, which can be configured as 74 triple probes or single probes, and two sets of reciprocating probes inserted from window A and window E at the mid-plane, as shown in Fig. 2. There is 18-channel Da and 18-channel CII/CIII viewing the lower outboard divertor from the top of the machine, two arrays of 35-channel Da viewing the inner target and dome surfaces of both upper and lower divertors from the outer mid-plane through the in-vessel reflection mirrors, one UV and Visible spectrometer with CCD (200–750 nm) viewing the lower-outer divertor, two sets of monochromators for CIII and OII lines and one interference filter for Ha monitor in the outer mid-plane, and 8-channel visible bremsstrahlung for Z-effective, as shown in Fig. 3. These diagnostics can provide useful information such as electron temperature, electron density, ion saturation current, Mach numbers, particle and impurity fluxes etc. in the divertor and SOL region. 3. Results and discussion 3.1. Basic divertor plasma characteristics Fig. 4 shows the basic divertor plasma behaviors for a typical ohmic discharge of Shot No. 11668 with Ip = 250 kA, BT = 2 T and
Fig. 3. View geometry of divertor and edge spectroscopy diagnostics.
ne ramps up from 1.0 to 3.0 1019 m3 from 3.5 s to 6.5 s. The discharge duration is about 7 s and the plasma is in stable double null (DN) divertor configuration from 3.5 s to 6.5 s. The directions of toroidal magnetic field and plasma current both are anticlockwise. Detachment can be achieved in Ohmic discharges by ramping up plasma density. As line-averaged density, ne increases, ion saturation current, Js, obtained from the divertor target Langmuir probe near strike points (UI_11, UO_12), first increases, then starts to roll over as ne further increases, finally decreases at sufficiently high densities as the plasma enters the detached regime. And ion saturation current, Js, obtained from the divertor target Langmuir probe far from strike points (UI_07, UO_05) increases latterly than the ones near strike points. Detachment starts near the strike point, then expands further out along the divertor target, as plasma density rises. Divertor neutral fluxes, as indicated by Da (Da_19c), keep e , but shows a tendency to saturate during increasing with n detachment. In addition, there is a strong in–out divertor asymmetry in particle and heat fluxes, as seen by the target probes. The electron density net obtained from the divertor target Langmuir probe in upper_outer divertor (UO_12 and UO_05) is much higher than the ones from upper_inner divertor (UI_11 and UI_07). The electron temperature Tet is just the reverse. Note that for the particular discharge shown in Fig. 4, the ion B rB drift is directed toward the upper divertor, and only the upper divertor measurements are shown. Usually the divertor plasma starts to de e exceeds 1.5 1019 m3. The detachment occurs first tach when n at inner target, followed by the detachment at the outer target at a e , i.e., about 2 1019 m3. When the detachment arose, higher n both Js and Tet are significantly reduced with Tet 5 eV near the separatrix. 3.2. Divertor screening for carbon
Fig. 2. Distribution of divertor target probes and reciprocating probe diagnostics.
Carbon is the dominant impurity in most of full graphite wall tokamaks such as EAST. Methane screening experiments have been conducted to quantify the ability of the SOL/divertor to ionize carbon and transport it to the divertor, preventing core plasma contamination in different plasma confinement regimes [7–9]. A series of experiments has recently been carried out under different plasma conditions to investigate the source distribution and divertor screening efficiency for the intrinsic carbon impurity on EAST, by injecting the pure methane with active divertor pumping. The
Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
S347
Fig. 4. Time evolution of an Ohmic double null discharge during a density ramp-up of Shot No. 11668. ne – line-averaged electron density, Da_19c – the radiation intensity of Da line corresponded to the ion saturation Js_UI_11 from upper-inner divertor, Js_UI_11 (UO_12) – ion saturation current near the upper-inner (upper-outer) strike points, Js_UI_07 (UO_05) – ion saturation current far from the upper-inner (upper-outer) strike points, net and Tet – electron density and temperature corresponded to the ion saturation Js_UI_11 (07) and Js_UO_12 (05).
experiment was primarily focused on ohmic discharges with Ip = 250 kA, BT = 2T and ne = 1.0, 1.5 and 2.0 1019 m3 three kinds of density platform discharges with discharge duration of about 7 s. The directions of toroidal magnetic field is clockwise and plasma current is anticlockwise. The carbon screening was measured by injecting CH4 at 4.1 s for 0.09 s from various poloidal locations, i.e., inner mid-plane, inner divertor target, outer divertor target and dome for different electron densities, i.e., 1.0, 1.5 and 2.0 1019 m3. Fig. 5 shows the time evolution of CIII measured at the LFS inner wall near the lower-outer divertor under different operating conditions. It is found that for the double null diverted configuration, strong screening occurs when the methane or methane mixture is injected at the inner mid-plane, or the outer target with little
changes in CIII intensity following the injection of CH4. In contrast, injection from the inner target, especially the dome leads to a large increase in CIII intensity, indicating weak divertor screening for carbon impurity from the inner target and private flux region. Inner mid-plane puffing has the best divertor screening, possibly attributed to the strong flow in the inner SOL [10]. The strong divertor screening for the outer divertor puffing may arise from the higher electron temperature so that neutrals can be ionized in the presheath region close to the target plate where strong flow is also present. The divertor screening for impurities exhibits a strong dependence on divertor plasma operation regimes with weak screening efficiency for the detached plasma regime (ne 2.0 1019 m3), but improving for the attached regime (ne 1.0–1.5 1019 m3), as also shown in Fig. 5. As can be seen,
Fig. 5. Changes in CIII intensity with CH4, puffing at outer target, inner target, private flux region (dome), and inner mid-plane, in DN divertor plasmas at 3.63 1021 particles/s, starting at 4.1 s for 0.09 s.
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Z.W. Wu et al. / Journal of Nuclear Materials 415 (2011) S345–S348
3.4. Discussion To further study and actively control divertor plasma, a large number of new divertor diagnostics have been installed in the Spring of 2010, including an IR camera, an array of bolometer detectors and a pressure gauge under the lower-outer divertor target plate. In order to achieve high power, long pulse operations, we will develop an active feedback control system for the ion saturation current to maintain the divertor plasma under partial detachment conditions. In addition, the ‘puff-and-pump’ technique [15] will be investigated to further reduce the leakage of injected high-Z impurities into the core plasma. This will be helpful and useful for us to improve this study. L-mode and H-mode divertor target plasma will be focused on next step study in EAST. 4. Summary and conclusions
19
3
Fig. 6. Time traces of an Argon seeded discharge with ne = 1.5 10 m in DN divertor plasma of Shot No. 12924. ð3:32 1021 particles=sÞ at 5 s over 0.15 s), (a) plasma current, (b) central line-averaged density, (c) central electron temperature, (d) radiation power (measured by XUV bolometer), (e) divertor Da intensity, (f) ion saturation current of the inner and outer strike points, (g) electron temperature of the inner and outer strike points, (h) peak heat flux of the inner and outer strike points.
C2+ influx is slightly increased with D = 0.18–0.27, where D is the relative change in C2+ influx, when CH4 is injected into the plasma at 4.1 s in the attached DN divertor plasmas. In contrast, as the plasma detaches at high densities, C2+ influx exhibits a strong increase with D = 0.32–1.62, which suggests decreasing divertor/ SOL screening efficiency at high density detached regime. However, the inner mid-plane puffing always shows strong screening, independent of the divertor plasma conditions. 3.3. Effect of Ar injection Extrinsic impurity seeding has been used in many tokamaks such as DIII-D [11], JET [12], ASDEX-Upgrade [13] and NSTX [14] to achieve partial or complete divertor detachment with strongly radiative divertor plasmas, which has also been considered as a candidate for the control of target heat load to acceptable levels in ITER. We have conducted the first radiative divertor experiments on EAST in April, 2009, by injecting the highly radiating recycling impurity Ar and its mixture with deuterium, i.e., 5.7% Ar in D2 with active divertor pumping in DN divertor ohmic discharges. Fig. 6 shows the results with 5.7% Ar in D2 being injected into the outer divertor chamber. The peak heat flux at the outer target plate is significantly reduced (over 50% reduction), by increasing radiated power in the divertor. The divertor plasma temperature and sheath potential are also reduced. This is accompanied by the strong increase main plasma density and strongly radiating plasma region located between the outer divertor separatrix strike point and the X-point. It is observed that the ion saturation current and peak heat flux of the outer strike points is almost reduced to the level of inner strike points after Ar puffing, hence improving the in–out heat flux distribution asymmetry. However, it is 2–5 times larger than the values of inner strike points before puffing. And the electron temperature of the inner and outer strike points is also reduced to 6 eV near the electron temperature in detachment regime after Ar puffing.
Detachment of the divertor plasma has been achieved by density ramp-up in EAST. Impurity screening experiments have been conducted by methane puffing in the different conditions. Strong screening occurs when the methane is injected at the inner midplane, or the outer divertor target with little changes in CIII intensity from LFS inner wall near the lower-outer divertor following the injection of CH4. In contrast, injection from the inner target, especially the dome leads to a large increase in CIII intensity, indicating weak divertor screening for impurities from the inner target and private flux region. The divertor screening for impurities exhibits a strong dependence on divertor plasma operation regimes with weak screening efficiency for the higher density, detached regime, but improving for the low and intermediate densities, attached regime. In addition, radiative divertor experiments have been conducted in EAST, by injecting the highly radiating recycling impurity Ar (5.7% Ar in D2) with active divertor pumping. The injection of mixed Ar into the outer divertor chamber significantly reduces the peak heat flux at the outer target plate, hence improving the in–out heat flux distribution asymmetry. This shows an effective means to control heat fluxes at divertor target plates for future high power, long pulse operation. Acknowledgements The authors would like to acknowledge the support from the rest of the EAST Team. This work is funded by National Nature Science Foundation of China under Contract Nos. 10475078 and 10728510, and also partly supported by JSPS-CAS Core-University Program in the Field of Plasma and Nuclear Fusion. Reference [1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] [15]
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