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Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas
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Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas J.B. Liu a , H.Y. Guo a,b,∗ , L. Wang a,c,∗ , G.S. Xu a , H.Q. Wang a , S.C. Liu a , W. Feng a , J.C. Xu a , G.Z. Deng a , Q.Q. Yang d , B.L. Ling a a
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China General Atomics, PO Box 85608, San Diego, CA 92186, USA c School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024, China d College of Science, Donghua University, Shanghai 201620, China b
h i g h l i g h t s • The work demonstrated that both divertor particle and heat flux widths in EAST L-mode plasmas exhibit a strong negative dependence on plasma current Ip (or poloidal field Bp ).
• Moreover, it is also demonstrated that the inverse Ip (or Bp ) scaling is independent of divertor configurations (LSN, DN and USN), which is good news for tokamaks operated in DN configurations.
• The particular analysis on the heat flux profiles measured by LFS reciprocating LPs also demonstrates the strong inverse scaling on Ip (or Bp ). • The heat and particle flux fall-off widths measured by divertor LPs show a approximately linear dependence, indicating the divertor heat flux profiles are dominated by the particle flux profiles in EAST L-mode plasmas.
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Article history: Received 4 December 2014 Received in revised form 19 June 2015 Accepted 20 June 2015 Available online xxx Keywords: Heat and particle flux widths Divertor configurations L-mode EAST
a b s t r a c t The scalings of divertor heat and particle flux widths have been systematically performed on the Experimental Advanced Superconducting Tokamak (EAST) in radio-frequency (RF) heated L-mode plasma regime under various divertor configurations. The widths were calculated from the measurements of divertor Langmuir probe (LP) arrays and the reciprocating LP diagnostic on the low-field side. A strong inverse scaling of particle and heat flux widths with plasma current Ip (equivalently the poloidal field Bp ) has been demonstrated. The measurements of divertor LPs show that the power decay length exhibits q,div = 4.97Ip −0.94 (q,div = 1.46Bp −1.15 ), in good agreement with the particle decay length js = 4.74Ip −1.02 (js = 1.34Bp −1.21 ). Similar trend measurements have also been demonstrated by the reciprocating LPs. In addition, the scaling with Ip (Bp ) appears to be insensitive of the divertor configurations in EAST. The study of heat and particle flux width scaling in EAST is useful for the extrapolation to the future tokamaks such as ITER. © 2015 Elsevier B.V. All rights reserved.
1. Introduction For many present tokamaks and the next-generation fusion devices, the steady-state distribution of heat and particles flux to plasma material surface is of great importance in determining the lifetime of plasma facing components (PFCs), especially the divertor targets. The heat is transferred to scrape-off-layer (SOL) across the
∗ Corresponding authors at: Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China. E-mail addresses: [email protected] (H.Y. Guo), [email protected] (L. Wang).
magnetic separatrix and eventually deposits onto PFCs. There are many issues related to the high power and particles flux deposition onto divertor targets, such as the power and particles exhaust, edge recycling, impurity influx, and the limit of peak heat flux. The particle and power deposition widths on the divertor target plates are determined by the competition between the transports processes parallel (//) and perpendicular (⊥) to the magnetic field (B) in the SOL [1–3]. Although, the target peak heat load will also be influenced by the power dissipation into the divertor private flux region, the power decay length in the SOL plays a very important role in tokamaks. The prediction of the power footprint width in tokamaks, is one of the most important tasks for the plasma
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physics in the SOL. It has been extensively studied very recently on NSTX [4], DIII-D [5], Alcator C-Mod [6], ASDEX Upgrade and JET [3,7], MAST [8] and so on. A key finding is that the plasma current Ip (related to the poloidal field at the outer mid-plane Bp ) has a strong inverse relationship with the SOL width, pointing to a common transport dynamics and much narrower SOL width for ITER than the previously estimated [2]. The result is consistent with the heuristic drift-model and the critical gradient model proposed by Goldston [9] and Makowski [5] separately, which was also derived from other conventional and spherical tokamaks. In the 2010 and 2012 campaigns of Experimental Advanced Superconducting Tokamak (EAST), dedicated particle and power behaviors, including the divertor particle and power flux deposition widths, in RF-heated experimental conditions under various divertor configurations (i.e., lower single null (LSN), double null (DN) and upper single null (USN)) were studied. The scaling of heat flux footprint widths during H-mode phase has been reported in [10]. In this work, to further study and complete the physical understanding, we mainly concentrated on the width scaling in RF-heating L-mode plasmas on EAST. By applying the same methodology as in [11,12], we will present the detailed study on the divertor heat and particle flux widths scaling with Ip (or Bp ). The rest of this paper is organized as follows. The EAST tokamak and the experimental setup for the divertor heat and particle flux width scaling study are briefly introduced in Section 2. Section 3 shows the detailed experimental results and discussion. Finally, the summary is given in Section 4.
2. Experimental setup and heat/particle flux diagnostics EAST is a fully superconducting tokamak to demonstrate highpower, long-pulse operations with ITER-like configuration and heating scheme. The major and minor radii of EAST are R ∼ 1.9 m and a ∼ 0.45 m, respectively. The designed maximum plasma current and toroidal field are Ip = 1 MA and Bt = 3.5 T, which is planned to be extended to Ip = 1.5 MA and Bt = 4 T by further reducing the temperature of the superconducting magnets. With a flexible poloidal field control system, the machine can be operated in SN and DN divertor configurations and enables a periodically switch between different divertor configurations during a long-pulse discharge. Moreover, it also has different heating and current drive systems, including both Lower Hybrid Current Drive (LHCD) and Ion Cyclotron Resonance Heating (ICRH) and Neutral Beam Injection (NBI). In the 2010 campaign, all the plasma facing components in EAST were graphite tiles. With the upper and lower divertor targets remaining graphite, the low-field side (LFS) and high-field side (HFS) tiles of the main chamber were upgraded to molybdenum in the 2012 campaign [10]. In the 2014 commissioning campaign, the upper divertor was successfully upgraded into ITER-like, actively water-cooled tungsten monoblock structure, the lower divertor keeps graphite material and the main chamber was still molybdenum [13]. The dedicated scaling experiments reported in this paper were all performed with graphite divertors of full top-down symmetry, which can provide useful information on some physics and engineering issues for ITER. In this work, the heat and particle flux widths were mainly obtained by divertor triple Langmuir probe (LP) arrays and the fast reciprocating LPs located at the LFS mid-plane. The divertor LPs embedded in the target plates can evaluate the divertor particle and heat flux widths simultaneously. The detailed information of divertor LP arrays on EAST before the 2014 campaign was reported in [14], the upper arrays of which have been successfully upgraded in ITER-like, water-cooled, tungsten monoblock divertor in 2014 [15]. The measurements of the profiles of upstream edge plasma parameters and the SOL flow velocity can be provided by the
Fig. 1. Poloidal layout of divertor Lanmuir probe (LP) arrays and LFS mid-plane reciprocating LPs in EAST. LO(I) – lower outboard (inboard) divertor, UO (I) – upper outboard (inboard) divertor. The reciprocating LP head shown in the right hand is the one employed in the dedicated experiments.
fast reciprocating LP system at the outer mid-plane [16]. Fig. 1 shows the poloidal layout of both divertor and LFS reciprocating LPs related to this work. Total 74 groups of triple divertor LPs were distributed in the upper and lower divertor target plates, with the spatial and temporal resolutions being 10–15 mm poloidally and 0.02 ms, respectively. The particle flux profiles can be directly derived from the ion saturation current densities (js ) of diverter LPs, according to ion = nt Cst = js /e, where ion, nt and Cst are the ion flux incident to probe, electron density and ion sound speed at the target plates respectively. The ion sound speed Cst = 2Tt /mi with Tt the electron temperature, assuming Ti = Te . The parallel heat fluxes are calculated using the standard sheath model, as q = nt Cst Tt , where the sheath heat transition factor is assumed to be = 7 [17]. To reduce the uncertainty of divertor LPs measurements, the data selected to generate each single profile were averaged over a time interval, i.e., 20 ms-averaged LP data for single profile. On EAST, the fast reciprocating LPs system can provide accurate measurements with a maximum velocity of 2 ms−1 . The probe tips can scan the upstream edge plasma region by driving horizontally with the data sampling rate being of 2 MHz. The dedicated experiments were carried out in LHW heated L-mode deuterium discharges under various divertor configurations by varying the plasma current on EAST, with the toroidal field Bt = 1.7–2 T, auxiliary heating power ∼1 MW, the line average electron density, ne = 1.8–3 × 1019 m−3 . The divertor conditions involved in this work were all in attached regimes. Fig. 2 shows three typical discharges in different divertor configurations (LSN USN and DN) investigated in this study, in which dRsep = RL − RU is defined as the physical radial separation of the X-points’ flux surfaces, with RL and RU being the lower and upper X-point radii mapped to the outer mid-plane, respectively. For each divertor configuration, other key plasma parameters of the dedicated shots remain similar, except changing the plasma current. For the target heat and particle flux widths, the experimental method published by the T. Eich in [3] is more accurate and commonly accepted by the current devices. We have also done the fitted curve with the function for the full target profile [3]. The calculated results of q are similar to the simple exponential decay. The regression on plasma current is approximate in L-mode scenarios.
Please cite this article in press as: J.B. Liu, et al., Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.06.098
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this paper are all for the outboard targets, since the behavior of the inboard target is much different between LSN/USN and DN configurations. In addition, the behavior of the inboard target was also affected by the secondary separatrix in SN configurations [17] with different dRsep values. 3.1. Divertor particle flux fall-off width (js ) scaling in LSN, DN and USN
Fig. 2. Three representative discharges in the USN (a), DN (b) and LSN (c) divertor configurations investigated in this study. The time traces shown are plasma current (black lines) and dRsep (red lines). (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
Here all the data come from the LPs, and just have 20 available data points per profile. The available data point at the left side of the peak heat flux is less and sparse. In addition, the result measured by the reciprocating LPs is not suitable of the function for the full target profiles. In this paper, the simple exponential fit may be better. The exponential fit using the least square fitting method in the SOL side was made to obtain a 1/e decay length, i.e., in the form of A = A0 exp[− (R − R0 )/], where R0 is the major radius of the maximum heat/particle flux mapped to the outer mid-plane by using the EFIT magnetic equilibrium reconstruction code. 3. Experimental results and discussion In EAST, by using the divertor LP arrays embedded in the target plates, the divertor heat and particle flux profiles can be obtained simultaneously. As a fundamental passive diagnostic, divertor LPs can provide very useful information for the physics study [19,20]. To obtain the experimental data for the scaling of the divertor heat and particle flux widths, a series of dedicated experiments were conducted over a wide range of plasma current Ip (equivalent Bp ) between 0.3 and 0.8 MA. To better compare the difference between L-mode plasmas and improving the prediction of the heat and particle flux widths for next step devices such as ITER, the L-mode phases were all chosen before the L-H transition with relatively smooth line-averaged densities in each discharge. As aforementioned, to reduce the uncertainty inherent in LP evaluations, all the data to generate a single profile were averaged over a time interval, i.e. averaged during a 20 ms interval for each profile, which accommodates the data of 100 time points in 2010 and 1000 time points in 2012, respectively. Note that the divertor LP data presented in
For the three different divertor configurations (LSN, DN and USN) in EAST, the plasma current scans were carried out under similar other conditions. Fig. 3 shows the particle flux fall-off widths, extracted from the ion saturation current density (js ) profiles at the outboard divertor target during L-mode phase, versus the plasma current Ip (a), and the poloidal field Bp (b) respectively. By using the EFIT magnetic equilibrium reconstruction code, all the data were mapped to the LFS mid-plane. The scattering points represent the experimental data measured by divertor LPs, while the bold lines express the simple fittings to the profiles in the SOL side. As can be seen, the simple e-folding regressions of js on Ip follows the form of js = AIp −˛ (or Bp in the form of js = ABp −˛ ). The divertor LPs measurements clearly show that the SOL width has a strong negative dependence on Ip (or Bp ) in L-mode scenarios. The regression on Ip (or Bp ) by the divertor LP data yield js = Ip −1.02 (or js = Bp −1.21 ), which points to a scaling law of js ∝ Ip −1 approximately. The trend of scaling is nearly consistent with the experimental observations in EAST during the H-mode scenarios with edge localized modes (ELMs) [10]. What is truly noticeable is that the inverse Ip (or Bp ) scaling in L-mode appears to be independent of the plasma configurations by combining EAST experimental results in different divertor configurations. The results suggest that the particle flux fall-off width at the outboard targets is independent of the length of divertor-to-divertor field lines [8]. 3.2. Divertor heat flux fall-off width (q ) scaling in LSN, DN and USN The parallel heat fluxes are calculated using the standard sheath model as shown in Section 2. The SOL power decay length (q ) is commonly obtained by the measurements of heat flux profiles at the outer divertor targets. Fig. 4 illustrates the exponential fall-off widths of parallel heat flux profiles (q// ) versus the plasma current Ip (a) and the poloidal field Bp (b) for L-mode scenarios in various divertor configurations as in Fig. 3. All the profiles from divertor LPs were mapped to the LFS mid-plane by using the EFIT magnetic equilibrium reconstruction code. In addition, the fitting decay lengths of upstream SOL heat flux profiles measured by LFS reciprocating LPs are shown in Fig. 4. The calculated results of LFS reciprocating LPs will be introduced in Section 3.4 and shown in Fig. 7. The experimental results of the two diagnostics independently show that the heat flux width q on SOL also has a strong negative dependence on Ip (or Bp ) in L-mode scenarios. The regression of q on Ip (or Bp ) based on the divertor LP measurements gives q,div = 4.97Ip −0.94 (or q,div = 1.46Bp −1.15 ) in a simple e-folding method. Compared with the scaling using divertor LPs, the result obtained by reciprocating LPs q,reci.-LP = 6.14Ip −0.57 (or q,reci.-LP = 2.55Bp −0.76 ) shows a weaker dependence on Ip (or Bp ). The systematic uncertainties of the divertor probes and reciprocating probes have been done as the systematic error for both measurements for SOL widths. As shown in Fig. 4, the error bar of divertor probes is about ±2∼4 mm, while for reciprocating probes it is about ±1.3∼2 mm. Note that the exponential regression trend coefficient by reciprocating LPs in L-mode is both about half of the trend coefficient during H-mode scenarios (q = 3.97Ip −1.04 or q = 0.73Bp −1.46 ) [10] for both Ip and Bp scalings, which may probably be due to the confinement performance of
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Fig. 3. The ion saturation current density (js ) fall-off width mapped to the outer mid-plane with Ip (a) and Bp (b) for L-mode in the LSN (green inverted triangles), DN (blue stars) and USN (black triangles) divertor configurations. The solid lines are the corresponding scaling fitting curves, js versus Ip , and js versus Bp (the poloidal field at LFS mid-plane around separatrix). (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
upstream SOL is different from that of divertor target plates in Land H-mode scenarios separately. In addition, it may also be due to the effective number of reciprocating LP data points is limited. Further discussion of reciprocating LP measurements will be shown in detail later in Section 3.4. Both of the diagnostic measurements exhibit a strong negative dependence on the plasma current and poloidal field. The heat flux width scaling is consistent with the particles flux results in Section 3.1 and the results in EAST ELMy H-mode scenarios [10], both of which have similar scaling trend of other multi machines [3–8]. In accordance with the results of particle flux fall-off width scaling the EAST experimental results, by the integration of LSN, DN and USN, indicate that the inverse Ip (or Bp ) scaling in L-mode scenarios also appears to be insensitive of divertor configuration. This is similar to the results in EAST H-mode scenarios, which is also consistent with the experimental results on Alcator C-Mod [6] and MAST [18]. It is beneficial to tokamak operation with the DN
configuration since the heat flux width on divertor target does not narrow with respect to the SN configuration [8]. 3.3. Relation between divertor heat and particle flux fall-off widths Fig. 5 illustrates the examples of ion saturation current density js (a) and parallel heat flux q// (b) profiles at the upper outboard (UO) divertor targets during a series of sequent L-mode discharges with different plasma current (Ip = 0.3, 0.4 and 0.5 MA) in DN configurations. For heat (or particle) flux profiles in the SOL, the exponential fall-off width fitting equation is shown in Section 2. It is also shown that both the particle and heat flux fall-off widths decrease strongly with increasing Ip (or Bp ). In addition, it is shown in Fig. 5(a) and (b) that the divertor heat flux decay length is similar to the falloff width of corresponding particle flux fall-off width. To further understand this observation, in Fig. 6, we compared the widths for
Fig. 4. The variation of the parallel heat flux (q|| ) fall-off width q with different plasma current Ip (a), and the poloidal field Bp (b) at the outer mid-plane separately. All the data were mapped to the outer mid-plane in LSN (green inverted triangles), DN (blue stars and magenta circle) and USN (black triangles) divertor configurations. The red solid and purple dotted lines represent the scaling fitting curves to the divertor and reciprocating LP data, respectively. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
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Fig. 5. The examples of ion saturation current density js (a) and the parallel heat flux q|| (b) profiles measured at the upper outboard (UO) divertor targets with simple e-folding fittings during L-mode phases. The red bold curves are the fitted particle and heat flux profiles. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
all the dataset obtained by divertor LPs in the dedicated experiments. The relation between the heat and particle fluxes for EAST is found by linear least squares fitting to be: q = 0.89(±0.159)js +0.88( ± 1.728). The comparison shows that the divertor heat flux profiles are mainly dominated by the divertor particle flux profiles in the L-mode scenarios on EAST, nearly consistent with the results of EAST H-mode scenarios with type-III ELMs. For EAST, the SOL widths in L-mode are similar to the results in type-III H-mode, which is different from the scaling results of
multi machines [7]. The main reason may be that the scaling in EAST with respect to multi machines was carried out in a different H-mode operation scenario, i.e., EAST in type-III ELMy H-mode (around the L-H transition power threshold), while multi machines in type-I ELMy H-mode or EDA H-mode (with high heating power). The difference in edge stability for the two regimes may be one potential reason for the 2× factor [10].
3.4. Upstream heat flux fall-off width measured by LFS reciprocating LPs
Fig. 6. Comparison of heat and particle flux fall-off widths simultaneously obtained by divertor LPs in L-mode phases with different divertor configurations on EAST.
During plasma discharges in tokamak, it is very difficult for the reciprocating LPs to plunge close enough to the separatrix to obtain useful information, which may induce large perturbations to the plasma or result in a disruptions eventually [10,16], especially during H-mode phases. This section presents a analysis on the L-mode heat flux width measured by the LFS reciprocating LPs on EAST. Fig. 7 shows the heat flux profiles of the upstream SOL in three dedicated successive DN discharges at Ip = 0.3–0.5 MA. The parallel heat fluxes are also calculated using the standard sheath model, as shown in Section 2 for divertor LPs. The experimental data captured by the reciprocating LPs are illustrated in black lines, and the red lines represent the simple exponential decay fittings to the whole profiles in the SOL. Note that the decay length q,reci-LP is slightly less than the measurements of divertor LPs, within diagnostic uncertainty. There were only three effective shots with reciprocating LP data in the dedicated experiments, but the diagnostic still provided a number of important information on the scaling of heat flux width. With respect to the H-mode phases [10], the reciprocating LPs in L-mode phases were inserted more deeply, i.e., reached ∼2 mm close to the separatrix. The results of heat flux width calculated by reciprocating LP also demonstrate the inverse relationship between heat flux fall-off width and Ip (or Bp ) during the L-mode scenarios, in good agreement with the divertor LP measurements.
Please cite this article in press as: J.B. Liu, et al., Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.06.098
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configurations, since the divertor heat flux widths do not narrow with respect to the SN case. The particular analysis on the heat flux profiles measured by LFS reciprocating LPs also demonstrates the strong inverse scaling on Ip (or Bp ). The heat and particle flux falloff widths measured by divertor LPs show a approximately linear dependence, indicating the divertor heat flux profiles are dominated by the particle flux profiles in EAST L-mode plasmas. The trend of heat and particle flux width scaling in EAST L-mode regime is consistent with the scaling in EAST ELMy H-mode scenario and the joint scaling obtained on multi machines. Further investigation will be carried out in L- and H-mode regimes with NBI dominant heating in EAST, also the relation between the dissipation width into divertor private flux region and the divertor geometry structures will be studied in the future. Acknowledgments We would like to acknowledge the support and contributions from the rest of the EAST team. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contract Nos. 2013GB107003, 2010GB104001, and National Natural Science Foundation of China under Grant Nos. 10990212, 11021565, 11105177, 11321092 and 11175209. This work was also supported by Scientific Research Grant of Hefei Science Center of CAS under contract 2015SRG-HSC001 as well as the Thousand Talent Plan of China. References Fig. 7. The heat flux profiles measured by reciprocating LPs in three successive DN discharges at Ip = 0.3–0.5 MA. The red solid curves are the exponential fittings, also shown are the fitted SOL power decay lengths of reciprocating LP data. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
4. Summary In summary, significant progress has been made in the scaling of outboard divertor heat and particle flux widths with the plasma current Ip (equivalently the poloidal field Bp ) on EAST, using divertor LP and LFS reciprocating LP diagnostics, in RFheated L-mode regime under various divertor configurations. It is demonstrated that both divertor particle and heat flux widths in EAST L-mode plasmas exhibit a strong negative dependence on Ip (or Bp ), i.e., js = Ip −1.02 (or js = Bp −1.21 ), q,div = 4.97Ip −0.94 (or q,div = 1.46Bp −1.15 ), which provide more information on the prediction of particle and heat flux deposition width for ITER. Moreover, it is also demonstrated in EAST L-mode plasmas that the inverse Ip (or Bp ) scaling is independent of divertor configurations (LSN, DN and USN), which is good news for tokamaks operated in DN
[1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13]
[14] [15]
[16] [17] [18] [19] [20]
J.-W. Ahn, et al., Plasma Phys. Controlled Fusion 48 (2006) 1077. A.W. Leonard, et al., J. Nucl. Mater. 266–269 (1999) 109. T. Eich, et al., Phys. Rev. Lett. 107 (2011) 210501. T.K. Gray, et al., J. Nucl. Mater. 415 (2011) S360. M.A. Makowski, et al., Phys. Plasmas 19 (2012) 056122. B. LaBombard, et al., Phys. Plasmas 18 (2011) 056104. A. Scarabosio, et al., J. Nucl. Mater. 438 (2013) S426. J.R. Harrison, et al., J. Nucl. Mater. 438 (2013) S375. R.J. Goldston, et al., Nucl. Fusion 52 (2012) 013009. L. Wang, et al., Nucl. Fusion 54 (2014) 114002. L. Wang, et al., Plasma Sci. Technol. 13 (2011) 435. A. Loarte, et al., J. Nucl. Mater. 266–269 (1999) 587. B.N. Wan, et al., Advance in H-mode physics for long pulse operation on EAST (OV3-3), in: 25th IAEA Fusion Energy Conference, October 13–18, St. Petersburg, Russia, 2014. T.F. Ming, et al., Fusion Eng. Des. 84 (2009) 57. J.C. Xu, L. Wang, et al., Upgrade of divertor Langmuir probe diagnostic in actively water-cooled, ITER-like tungsten monoblock divertor on the EAST superconducting tokamak, Rev. Sci. Instrum. (2015) (submitted for publication). W. Zhang, et al., Rev. Sci. Instrum. 81 (2010) 113501. L. Wang, et al., Nucl. Fusion 52 (2012) 063024. A.J. Thornton, et al., Plasma Phys. Controlled Fusion 56 (2014) 055008. J. Li, H.Y. Guo, et al., Nat. Phys. 9 (2013) 817. Y.F. Liang, et al., Phys. Rev. Lett. 110 (2013) 235002.
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Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas J.B. Liu a , H.Y. Guo a,b,∗ , L. Wang a,c,∗ , G.S. Xu a , H.Q. Wang a , S.C. Liu a , W. Feng a , J.C. Xu a , G.Z. Deng a , Q.Q. Yang d , B.L. Ling a a
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China General Atomics, PO Box 85608, San Diego, CA 92186, USA c School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024, China d College of Science, Donghua University, Shanghai 201620, China b
h i g h l i g h t s • The work demonstrated that both divertor particle and heat flux widths in EAST L-mode plasmas exhibit a strong negative dependence on plasma current Ip (or poloidal field Bp ).
• Moreover, it is also demonstrated that the inverse Ip (or Bp ) scaling is independent of divertor configurations (LSN, DN and USN), which is good news for tokamaks operated in DN configurations.
• The particular analysis on the heat flux profiles measured by LFS reciprocating LPs also demonstrates the strong inverse scaling on Ip (or Bp ). • The heat and particle flux fall-off widths measured by divertor LPs show a approximately linear dependence, indicating the divertor heat flux profiles are dominated by the particle flux profiles in EAST L-mode plasmas.
a r t i c l e
i n f o
Article history: Received 4 December 2014 Received in revised form 19 June 2015 Accepted 20 June 2015 Available online xxx Keywords: Heat and particle flux widths Divertor configurations L-mode EAST
a b s t r a c t The scalings of divertor heat and particle flux widths have been systematically performed on the Experimental Advanced Superconducting Tokamak (EAST) in radio-frequency (RF) heated L-mode plasma regime under various divertor configurations. The widths were calculated from the measurements of divertor Langmuir probe (LP) arrays and the reciprocating LP diagnostic on the low-field side. A strong inverse scaling of particle and heat flux widths with plasma current Ip (equivalently the poloidal field Bp ) has been demonstrated. The measurements of divertor LPs show that the power decay length exhibits q,div = 4.97Ip −0.94 (q,div = 1.46Bp −1.15 ), in good agreement with the particle decay length js = 4.74Ip −1.02 (js = 1.34Bp −1.21 ). Similar trend measurements have also been demonstrated by the reciprocating LPs. In addition, the scaling with Ip (Bp ) appears to be insensitive of the divertor configurations in EAST. The study of heat and particle flux width scaling in EAST is useful for the extrapolation to the future tokamaks such as ITER. © 2015 Elsevier B.V. All rights reserved.
1. Introduction For many present tokamaks and the next-generation fusion devices, the steady-state distribution of heat and particles flux to plasma material surface is of great importance in determining the lifetime of plasma facing components (PFCs), especially the divertor targets. The heat is transferred to scrape-off-layer (SOL) across the
∗ Corresponding authors at: Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China. E-mail addresses: [email protected] (H.Y. Guo), [email protected] (L. Wang).
magnetic separatrix and eventually deposits onto PFCs. There are many issues related to the high power and particles flux deposition onto divertor targets, such as the power and particles exhaust, edge recycling, impurity influx, and the limit of peak heat flux. The particle and power deposition widths on the divertor target plates are determined by the competition between the transports processes parallel (//) and perpendicular (⊥) to the magnetic field (B) in the SOL [1–3]. Although, the target peak heat load will also be influenced by the power dissipation into the divertor private flux region, the power decay length in the SOL plays a very important role in tokamaks. The prediction of the power footprint width in tokamaks, is one of the most important tasks for the plasma
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physics in the SOL. It has been extensively studied very recently on NSTX [4], DIII-D [5], Alcator C-Mod [6], ASDEX Upgrade and JET [3,7], MAST [8] and so on. A key finding is that the plasma current Ip (related to the poloidal field at the outer mid-plane Bp ) has a strong inverse relationship with the SOL width, pointing to a common transport dynamics and much narrower SOL width for ITER than the previously estimated [2]. The result is consistent with the heuristic drift-model and the critical gradient model proposed by Goldston [9] and Makowski [5] separately, which was also derived from other conventional and spherical tokamaks. In the 2010 and 2012 campaigns of Experimental Advanced Superconducting Tokamak (EAST), dedicated particle and power behaviors, including the divertor particle and power flux deposition widths, in RF-heated experimental conditions under various divertor configurations (i.e., lower single null (LSN), double null (DN) and upper single null (USN)) were studied. The scaling of heat flux footprint widths during H-mode phase has been reported in [10]. In this work, to further study and complete the physical understanding, we mainly concentrated on the width scaling in RF-heating L-mode plasmas on EAST. By applying the same methodology as in [11,12], we will present the detailed study on the divertor heat and particle flux widths scaling with Ip (or Bp ). The rest of this paper is organized as follows. The EAST tokamak and the experimental setup for the divertor heat and particle flux width scaling study are briefly introduced in Section 2. Section 3 shows the detailed experimental results and discussion. Finally, the summary is given in Section 4.
2. Experimental setup and heat/particle flux diagnostics EAST is a fully superconducting tokamak to demonstrate highpower, long-pulse operations with ITER-like configuration and heating scheme. The major and minor radii of EAST are R ∼ 1.9 m and a ∼ 0.45 m, respectively. The designed maximum plasma current and toroidal field are Ip = 1 MA and Bt = 3.5 T, which is planned to be extended to Ip = 1.5 MA and Bt = 4 T by further reducing the temperature of the superconducting magnets. With a flexible poloidal field control system, the machine can be operated in SN and DN divertor configurations and enables a periodically switch between different divertor configurations during a long-pulse discharge. Moreover, it also has different heating and current drive systems, including both Lower Hybrid Current Drive (LHCD) and Ion Cyclotron Resonance Heating (ICRH) and Neutral Beam Injection (NBI). In the 2010 campaign, all the plasma facing components in EAST were graphite tiles. With the upper and lower divertor targets remaining graphite, the low-field side (LFS) and high-field side (HFS) tiles of the main chamber were upgraded to molybdenum in the 2012 campaign [10]. In the 2014 commissioning campaign, the upper divertor was successfully upgraded into ITER-like, actively water-cooled tungsten monoblock structure, the lower divertor keeps graphite material and the main chamber was still molybdenum [13]. The dedicated scaling experiments reported in this paper were all performed with graphite divertors of full top-down symmetry, which can provide useful information on some physics and engineering issues for ITER. In this work, the heat and particle flux widths were mainly obtained by divertor triple Langmuir probe (LP) arrays and the fast reciprocating LPs located at the LFS mid-plane. The divertor LPs embedded in the target plates can evaluate the divertor particle and heat flux widths simultaneously. The detailed information of divertor LP arrays on EAST before the 2014 campaign was reported in [14], the upper arrays of which have been successfully upgraded in ITER-like, water-cooled, tungsten monoblock divertor in 2014 [15]. The measurements of the profiles of upstream edge plasma parameters and the SOL flow velocity can be provided by the
Fig. 1. Poloidal layout of divertor Lanmuir probe (LP) arrays and LFS mid-plane reciprocating LPs in EAST. LO(I) – lower outboard (inboard) divertor, UO (I) – upper outboard (inboard) divertor. The reciprocating LP head shown in the right hand is the one employed in the dedicated experiments.
fast reciprocating LP system at the outer mid-plane [16]. Fig. 1 shows the poloidal layout of both divertor and LFS reciprocating LPs related to this work. Total 74 groups of triple divertor LPs were distributed in the upper and lower divertor target plates, with the spatial and temporal resolutions being 10–15 mm poloidally and 0.02 ms, respectively. The particle flux profiles can be directly derived from the ion saturation current densities (js ) of diverter LPs, according to ion = nt Cst = js /e, where ion, nt and Cst are the ion flux incident to probe, electron density and ion sound speed at the target plates respectively. The ion sound speed Cst = 2Tt /mi with Tt the electron temperature, assuming Ti = Te . The parallel heat fluxes are calculated using the standard sheath model, as q = nt Cst Tt , where the sheath heat transition factor is assumed to be = 7 [17]. To reduce the uncertainty of divertor LPs measurements, the data selected to generate each single profile were averaged over a time interval, i.e., 20 ms-averaged LP data for single profile. On EAST, the fast reciprocating LPs system can provide accurate measurements with a maximum velocity of 2 ms−1 . The probe tips can scan the upstream edge plasma region by driving horizontally with the data sampling rate being of 2 MHz. The dedicated experiments were carried out in LHW heated L-mode deuterium discharges under various divertor configurations by varying the plasma current on EAST, with the toroidal field Bt = 1.7–2 T, auxiliary heating power ∼1 MW, the line average electron density, ne = 1.8–3 × 1019 m−3 . The divertor conditions involved in this work were all in attached regimes. Fig. 2 shows three typical discharges in different divertor configurations (LSN USN and DN) investigated in this study, in which dRsep = RL − RU is defined as the physical radial separation of the X-points’ flux surfaces, with RL and RU being the lower and upper X-point radii mapped to the outer mid-plane, respectively. For each divertor configuration, other key plasma parameters of the dedicated shots remain similar, except changing the plasma current. For the target heat and particle flux widths, the experimental method published by the T. Eich in [3] is more accurate and commonly accepted by the current devices. We have also done the fitted curve with the function for the full target profile [3]. The calculated results of q are similar to the simple exponential decay. The regression on plasma current is approximate in L-mode scenarios.
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this paper are all for the outboard targets, since the behavior of the inboard target is much different between LSN/USN and DN configurations. In addition, the behavior of the inboard target was also affected by the secondary separatrix in SN configurations [17] with different dRsep values. 3.1. Divertor particle flux fall-off width (js ) scaling in LSN, DN and USN
Fig. 2. Three representative discharges in the USN (a), DN (b) and LSN (c) divertor configurations investigated in this study. The time traces shown are plasma current (black lines) and dRsep (red lines). (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
Here all the data come from the LPs, and just have 20 available data points per profile. The available data point at the left side of the peak heat flux is less and sparse. In addition, the result measured by the reciprocating LPs is not suitable of the function for the full target profiles. In this paper, the simple exponential fit may be better. The exponential fit using the least square fitting method in the SOL side was made to obtain a 1/e decay length, i.e., in the form of A = A0 exp[− (R − R0 )/], where R0 is the major radius of the maximum heat/particle flux mapped to the outer mid-plane by using the EFIT magnetic equilibrium reconstruction code. 3. Experimental results and discussion In EAST, by using the divertor LP arrays embedded in the target plates, the divertor heat and particle flux profiles can be obtained simultaneously. As a fundamental passive diagnostic, divertor LPs can provide very useful information for the physics study [19,20]. To obtain the experimental data for the scaling of the divertor heat and particle flux widths, a series of dedicated experiments were conducted over a wide range of plasma current Ip (equivalent Bp ) between 0.3 and 0.8 MA. To better compare the difference between L-mode plasmas and improving the prediction of the heat and particle flux widths for next step devices such as ITER, the L-mode phases were all chosen before the L-H transition with relatively smooth line-averaged densities in each discharge. As aforementioned, to reduce the uncertainty inherent in LP evaluations, all the data to generate a single profile were averaged over a time interval, i.e. averaged during a 20 ms interval for each profile, which accommodates the data of 100 time points in 2010 and 1000 time points in 2012, respectively. Note that the divertor LP data presented in
For the three different divertor configurations (LSN, DN and USN) in EAST, the plasma current scans were carried out under similar other conditions. Fig. 3 shows the particle flux fall-off widths, extracted from the ion saturation current density (js ) profiles at the outboard divertor target during L-mode phase, versus the plasma current Ip (a), and the poloidal field Bp (b) respectively. By using the EFIT magnetic equilibrium reconstruction code, all the data were mapped to the LFS mid-plane. The scattering points represent the experimental data measured by divertor LPs, while the bold lines express the simple fittings to the profiles in the SOL side. As can be seen, the simple e-folding regressions of js on Ip follows the form of js = AIp −˛ (or Bp in the form of js = ABp −˛ ). The divertor LPs measurements clearly show that the SOL width has a strong negative dependence on Ip (or Bp ) in L-mode scenarios. The regression on Ip (or Bp ) by the divertor LP data yield js = Ip −1.02 (or js = Bp −1.21 ), which points to a scaling law of js ∝ Ip −1 approximately. The trend of scaling is nearly consistent with the experimental observations in EAST during the H-mode scenarios with edge localized modes (ELMs) [10]. What is truly noticeable is that the inverse Ip (or Bp ) scaling in L-mode appears to be independent of the plasma configurations by combining EAST experimental results in different divertor configurations. The results suggest that the particle flux fall-off width at the outboard targets is independent of the length of divertor-to-divertor field lines [8]. 3.2. Divertor heat flux fall-off width (q ) scaling in LSN, DN and USN The parallel heat fluxes are calculated using the standard sheath model as shown in Section 2. The SOL power decay length (q ) is commonly obtained by the measurements of heat flux profiles at the outer divertor targets. Fig. 4 illustrates the exponential fall-off widths of parallel heat flux profiles (q// ) versus the plasma current Ip (a) and the poloidal field Bp (b) for L-mode scenarios in various divertor configurations as in Fig. 3. All the profiles from divertor LPs were mapped to the LFS mid-plane by using the EFIT magnetic equilibrium reconstruction code. In addition, the fitting decay lengths of upstream SOL heat flux profiles measured by LFS reciprocating LPs are shown in Fig. 4. The calculated results of LFS reciprocating LPs will be introduced in Section 3.4 and shown in Fig. 7. The experimental results of the two diagnostics independently show that the heat flux width q on SOL also has a strong negative dependence on Ip (or Bp ) in L-mode scenarios. The regression of q on Ip (or Bp ) based on the divertor LP measurements gives q,div = 4.97Ip −0.94 (or q,div = 1.46Bp −1.15 ) in a simple e-folding method. Compared with the scaling using divertor LPs, the result obtained by reciprocating LPs q,reci.-LP = 6.14Ip −0.57 (or q,reci.-LP = 2.55Bp −0.76 ) shows a weaker dependence on Ip (or Bp ). The systematic uncertainties of the divertor probes and reciprocating probes have been done as the systematic error for both measurements for SOL widths. As shown in Fig. 4, the error bar of divertor probes is about ±2∼4 mm, while for reciprocating probes it is about ±1.3∼2 mm. Note that the exponential regression trend coefficient by reciprocating LPs in L-mode is both about half of the trend coefficient during H-mode scenarios (q = 3.97Ip −1.04 or q = 0.73Bp −1.46 ) [10] for both Ip and Bp scalings, which may probably be due to the confinement performance of
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Fig. 3. The ion saturation current density (js ) fall-off width mapped to the outer mid-plane with Ip (a) and Bp (b) for L-mode in the LSN (green inverted triangles), DN (blue stars) and USN (black triangles) divertor configurations. The solid lines are the corresponding scaling fitting curves, js versus Ip , and js versus Bp (the poloidal field at LFS mid-plane around separatrix). (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
upstream SOL is different from that of divertor target plates in Land H-mode scenarios separately. In addition, it may also be due to the effective number of reciprocating LP data points is limited. Further discussion of reciprocating LP measurements will be shown in detail later in Section 3.4. Both of the diagnostic measurements exhibit a strong negative dependence on the plasma current and poloidal field. The heat flux width scaling is consistent with the particles flux results in Section 3.1 and the results in EAST ELMy H-mode scenarios [10], both of which have similar scaling trend of other multi machines [3–8]. In accordance with the results of particle flux fall-off width scaling the EAST experimental results, by the integration of LSN, DN and USN, indicate that the inverse Ip (or Bp ) scaling in L-mode scenarios also appears to be insensitive of divertor configuration. This is similar to the results in EAST H-mode scenarios, which is also consistent with the experimental results on Alcator C-Mod [6] and MAST [18]. It is beneficial to tokamak operation with the DN
configuration since the heat flux width on divertor target does not narrow with respect to the SN configuration [8]. 3.3. Relation between divertor heat and particle flux fall-off widths Fig. 5 illustrates the examples of ion saturation current density js (a) and parallel heat flux q// (b) profiles at the upper outboard (UO) divertor targets during a series of sequent L-mode discharges with different plasma current (Ip = 0.3, 0.4 and 0.5 MA) in DN configurations. For heat (or particle) flux profiles in the SOL, the exponential fall-off width fitting equation is shown in Section 2. It is also shown that both the particle and heat flux fall-off widths decrease strongly with increasing Ip (or Bp ). In addition, it is shown in Fig. 5(a) and (b) that the divertor heat flux decay length is similar to the falloff width of corresponding particle flux fall-off width. To further understand this observation, in Fig. 6, we compared the widths for
Fig. 4. The variation of the parallel heat flux (q|| ) fall-off width q with different plasma current Ip (a), and the poloidal field Bp (b) at the outer mid-plane separately. All the data were mapped to the outer mid-plane in LSN (green inverted triangles), DN (blue stars and magenta circle) and USN (black triangles) divertor configurations. The red solid and purple dotted lines represent the scaling fitting curves to the divertor and reciprocating LP data, respectively. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
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Fig. 5. The examples of ion saturation current density js (a) and the parallel heat flux q|| (b) profiles measured at the upper outboard (UO) divertor targets with simple e-folding fittings during L-mode phases. The red bold curves are the fitted particle and heat flux profiles. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
all the dataset obtained by divertor LPs in the dedicated experiments. The relation between the heat and particle fluxes for EAST is found by linear least squares fitting to be: q = 0.89(±0.159)js +0.88( ± 1.728). The comparison shows that the divertor heat flux profiles are mainly dominated by the divertor particle flux profiles in the L-mode scenarios on EAST, nearly consistent with the results of EAST H-mode scenarios with type-III ELMs. For EAST, the SOL widths in L-mode are similar to the results in type-III H-mode, which is different from the scaling results of
multi machines [7]. The main reason may be that the scaling in EAST with respect to multi machines was carried out in a different H-mode operation scenario, i.e., EAST in type-III ELMy H-mode (around the L-H transition power threshold), while multi machines in type-I ELMy H-mode or EDA H-mode (with high heating power). The difference in edge stability for the two regimes may be one potential reason for the 2× factor [10].
3.4. Upstream heat flux fall-off width measured by LFS reciprocating LPs
Fig. 6. Comparison of heat and particle flux fall-off widths simultaneously obtained by divertor LPs in L-mode phases with different divertor configurations on EAST.
During plasma discharges in tokamak, it is very difficult for the reciprocating LPs to plunge close enough to the separatrix to obtain useful information, which may induce large perturbations to the plasma or result in a disruptions eventually [10,16], especially during H-mode phases. This section presents a analysis on the L-mode heat flux width measured by the LFS reciprocating LPs on EAST. Fig. 7 shows the heat flux profiles of the upstream SOL in three dedicated successive DN discharges at Ip = 0.3–0.5 MA. The parallel heat fluxes are also calculated using the standard sheath model, as shown in Section 2 for divertor LPs. The experimental data captured by the reciprocating LPs are illustrated in black lines, and the red lines represent the simple exponential decay fittings to the whole profiles in the SOL. Note that the decay length q,reci-LP is slightly less than the measurements of divertor LPs, within diagnostic uncertainty. There were only three effective shots with reciprocating LP data in the dedicated experiments, but the diagnostic still provided a number of important information on the scaling of heat flux width. With respect to the H-mode phases [10], the reciprocating LPs in L-mode phases were inserted more deeply, i.e., reached ∼2 mm close to the separatrix. The results of heat flux width calculated by reciprocating LP also demonstrate the inverse relationship between heat flux fall-off width and Ip (or Bp ) during the L-mode scenarios, in good agreement with the divertor LP measurements.
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configurations, since the divertor heat flux widths do not narrow with respect to the SN case. The particular analysis on the heat flux profiles measured by LFS reciprocating LPs also demonstrates the strong inverse scaling on Ip (or Bp ). The heat and particle flux falloff widths measured by divertor LPs show a approximately linear dependence, indicating the divertor heat flux profiles are dominated by the particle flux profiles in EAST L-mode plasmas. The trend of heat and particle flux width scaling in EAST L-mode regime is consistent with the scaling in EAST ELMy H-mode scenario and the joint scaling obtained on multi machines. Further investigation will be carried out in L- and H-mode regimes with NBI dominant heating in EAST, also the relation between the dissipation width into divertor private flux region and the divertor geometry structures will be studied in the future. Acknowledgments We would like to acknowledge the support and contributions from the rest of the EAST team. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contract Nos. 2013GB107003, 2010GB104001, and National Natural Science Foundation of China under Grant Nos. 10990212, 11021565, 11105177, 11321092 and 11175209. This work was also supported by Scientific Research Grant of Hefei Science Center of CAS under contract 2015SRG-HSC001 as well as the Thousand Talent Plan of China. References Fig. 7. The heat flux profiles measured by reciprocating LPs in three successive DN discharges at Ip = 0.3–0.5 MA. The red solid curves are the exponential fittings, also shown are the fitted SOL power decay lengths of reciprocating LP data. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
4. Summary In summary, significant progress has been made in the scaling of outboard divertor heat and particle flux widths with the plasma current Ip (equivalently the poloidal field Bp ) on EAST, using divertor LP and LFS reciprocating LP diagnostics, in RFheated L-mode regime under various divertor configurations. It is demonstrated that both divertor particle and heat flux widths in EAST L-mode plasmas exhibit a strong negative dependence on Ip (or Bp ), i.e., js = Ip −1.02 (or js = Bp −1.21 ), q,div = 4.97Ip −0.94 (or q,div = 1.46Bp −1.15 ), which provide more information on the prediction of particle and heat flux deposition width for ITER. Moreover, it is also demonstrated in EAST L-mode plasmas that the inverse Ip (or Bp ) scaling is independent of divertor configurations (LSN, DN and USN), which is good news for tokamaks operated in DN
[1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13]
[14] [15]
[16] [17] [18] [19] [20]
J.-W. Ahn, et al., Plasma Phys. Controlled Fusion 48 (2006) 1077. A.W. Leonard, et al., J. Nucl. Mater. 266–269 (1999) 109. T. Eich, et al., Phys. Rev. Lett. 107 (2011) 210501. T.K. Gray, et al., J. Nucl. Mater. 415 (2011) S360. M.A. Makowski, et al., Phys. Plasmas 19 (2012) 056122. B. LaBombard, et al., Phys. Plasmas 18 (2011) 056104. A. Scarabosio, et al., J. Nucl. Mater. 438 (2013) S426. J.R. Harrison, et al., J. Nucl. Mater. 438 (2013) S375. R.J. Goldston, et al., Nucl. Fusion 52 (2012) 013009. L. Wang, et al., Nucl. Fusion 54 (2014) 114002. L. Wang, et al., Plasma Sci. Technol. 13 (2011) 435. A. Loarte, et al., J. Nucl. Mater. 266–269 (1999) 587. B.N. Wan, et al., Advance in H-mode physics for long pulse operation on EAST (OV3-3), in: 25th IAEA Fusion Energy Conference, October 13–18, St. Petersburg, Russia, 2014. T.F. Ming, et al., Fusion Eng. Des. 84 (2009) 57. J.C. Xu, L. Wang, et al., Upgrade of divertor Langmuir probe diagnostic in actively water-cooled, ITER-like tungsten monoblock divertor on the EAST superconducting tokamak, Rev. Sci. Instrum. (2015) (submitted for publication). W. Zhang, et al., Rev. Sci. Instrum. 81 (2010) 113501. L. Wang, et al., Nucl. Fusion 52 (2012) 063024. A.J. Thornton, et al., Plasma Phys. Controlled Fusion 56 (2014) 055008. J. Li, H.Y. Guo, et al., Nat. Phys. 9 (2013) 817. Y.F. Liang, et al., Phys. Rev. Lett. 110 (2013) 235002.
Please cite this article in press as: J.B. Liu, et al., Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.06.098