Integrated Modelling of ITER Scenarios with ECCD


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1 Integrated Modelling of ITER Scenarios with ECCD J.F. Artaud, V. Basiuk, J. Garcia, G. Giruzzi*, F. Imbeaux, M. Schneider Association EuratomCEA sur la Fusion, CEA/DSM/DRFC, CEA/Cadarache, St. PaullezDurance, France * The main application of EC waves in ITER will be the control of NTM by means of the top launchers. However, virtually every ITER scenario would profit from appropriate use of ECCD by means of the equatorial and/or the top launchers. In particular, hybrid and steadystate scenarios require offaxis current drive in order to either keep the safety factor above 1 or produce negative shear in a large part of the plasma crosssection. In this type of applications, alignment of the current sources and selfconsistency of current and temperature profiles are critical issues, which can only be addressed by integrated modelling. To this end, the CRONOS suite of codes has been applied to the simulation of these scenarios. The ECCD module of CRONOS includes toroidal raytracing and linear computation of the driven current, which is generally adequate for ITER parameters. This paper gives a short description of the CRONOS suite of codes, followed by results of simulations of ITER hybrid scenarios assisted by ECCD. Use of ECCD in steadystate scenarios is also addressed. It is shown that ECCD at midradius could be a key ingredient for establishing a noninductive discharge with large bootstrap fraction and controlled Internal Transport Barrier. 1. Introduction Electron Cyclotron (EC) waves are the most versatile heating and current drive system available on ITER. Despite the wide variety of applications for which this system has been designed, its potential has not yet been fully investigated. Owing to the good localisation properties of the driven current and power deposition profiles, the impact of the use of EC waves on various ITER scenarios can only be predicted by integrated modelling codes [1], in which the 1D evolutions of the various heat and current sources are selfconsistently evaluated with the profiles of the main plasma quantities. If the reference scenario (elmy H mode) has been defined on the basis of global confinement scaling laws and a wide experimental data base, the long pulse scenarios (hybrid and steadystate) rely on the possibility of producing particular safety factor profiles [2,3] (q flat and above unity for the hybrid scenario, and q reversed and above 2 for the steadystate scenario). To this end, the localised currents provided by EC waves could prove useful or even indispensable, not only for MHD control (the primary application on ITER), but also for realtime controlled profile tailoring. In this paper, integrated modelling studies of ITER hybrid and steadystate scenarios in which EC waves play an important role are reported. These studies are performed by the CRONOS suite of codes [4], which will be shortly described in Sec. 2. Results for the hybrid and for the steadystate scenarios are presented in Secs. 3 and 4, respectively. Conclusions and recommendations for the final definition of the ITER ECRH system, based on these results, are given in Sec The CRONOS suite of codes The suite of codes CRONOS [4] solves the transport equations for various plasma fluid quantities (current, energy, matter, momentum). This is done in one dimension (the magnetic flux coordinate associated with the minor radius) selfconsistently with magnetic equilibrium which is calculated by means of the HELENA module [5] or by analytical
2 formulae based on the moments of the equilibrium equation [6,7], whenever a faster simulation is needed. The neoclassical terms, and in particular the bootstrap current, which is the dominant current source in advanced regimes, are determined using the NCLASS [8] code. The sources are computed by external modules coupled with the main transport equations. The Neutral Beam Current Drive is calculated by means of the SINBAD module [9,10], coupled to either an analytical solution of the appropriate kinetic equation for the fast ion distribution function, or to the orbit following MonteCarlo code SPOT [11]. PION [12] is used for Ion Cyclotron Resonance Heating. The ECCD source is computed by the REMA code, including toroidal raytracing [13] and a linear formula for the driven current [14], which is generally adequate for ITER parameters. The alpha power deposition profile is evaluated from the fusion reactivity given by the BoschHale formulae [15]. The core plasma line and bremsstrahlung radiation are computed with a model based on coronal equilibrium [16]. The synchrotron radiation loss is computed by means of EXATEC [17]. Two classes of models for the heat transport are available in CRONOS: first principles models, based on the linear growth rates of the various instabilities which are the source of plasma turbulence, as the gyrolandaufluid model GLF23 [18], as well as more empirical models, based on global scaling laws, as Kiauto [19]. In the simulations of the hybrid scenario, the GLF23 model has been applied in the plasma core, however, since the pedestal can not be obtained by means of this model, the Kiauto scaling has been used to impose given pedestal height and width. On the other hand, for the simulation of the steadystate scenarios, a simpler adhoc model has been used, allowing both a pedestal and the formation of an ITB whenever the magnetic shear becomes negative (see Ref. [17] for details). Finally, the density profiles are prescribed and fixed during the time evolution, and the helium concentration is obtained by solving a purely diffusive equation and setting the diffusion coefficient in order to impose τ He /τ E =5, where τ He is the helium confinement time and τ E is the energy confinement time. The CRONOS code has been validated by interpretative simulations and comparison with experimental results on various machines (Tore Supra, JET, DIIID), as well as by FIG. 1: Time evolution of computed quantities: plasma current, central density and Q factor (top left); additional powers (bottom left); bootstrap, noninductive current and Greenwald fractions (top right); normalised beta, 4l i, H factor and Z eff (bottom right). comparison with complex simulations of ITER reference and hybrid scenarios performed by other codes of the same type [20]. 3. ITER hybrid scenario Hybrid scenarios [21] are characterised by q > 1 and flat in the central region, i.e., for normalised radius ρ < In present day experiments this condition is probably realised because of a complex interplay between MHD activity and safety factor profile [22], but the extrapolation of such mechanisms to ITER is uncertain. Therefore, it is important to characterise the safety factor profile in these projected scenarios by integrated modelling simulations, and to identify the possible actuators that could be used to enforce the desired q profile shape. The main parameters adopted for these simulations are those of Ref. [20], i.e.,
3 an ITER plasma of current I p = 12 MA and central density n e (0) = m 3, with prescribed and fixed profile. The additional heating is provided by ICRH (20 MW, minority heating, 2 nd T harmonic) and NBI (33 MW at 1 MeV, offaxis injection). The heat transport model used is GLF23 with fixed pedestal temperature. The simulation has been run for 1200 s, in order to obtain a stationary q profile. The results are shown in Fig. 1 (global quantities vs time) and Fig. 2 (profiles at the end of the simulation). FIG. 2: Profiles of computed quantities, from left to right: temperatures and density; power sources; current density and current sources; safety factor at three different times. It appears that this type of discharge, despite its good performance, does not attain the main goal of having q > 1 in the stationary phase. Use of offaxis ECCD is then investigated for this purpose. In this case, the wave launch configuration providing the largest amount of (moderately) offaxis current is required: therefore, the equatorial launcher is used, rather than the top one, which is optimised for NTM stabilisation [23]. A good compromise between offaxis power deposition (at 1/3 of the minor radius) and current drive efficiency (~ 0.2 A W m 2 ) is found for toroidal angles = 32, 38 and 40 for the top, middle and bottom rows of mirrors of the equatorial launcher, respectively. A CRONOS simulation has been performed with the same parameters as the previous one, but with the addition of 20 MW of ECCD in the conditions of Fig. 3. The main FIG. 3: Raytracing results, for three beams simulating wave result is a little but launch from the equatorial launcher: poloidal projections of the ray essential change in the trajectories (left); toroidal projections (middle); power deposition profiles (right). final q profile, which is now flatter in the central part and practically > 1, as shown in Fig. 4 (left). Since 20 MW have been added, the price to FIG. 4: Simulated q profiles at t = 1200 s with and without ECCD (left); time evolution of the corresponding Q (top right); time evolution of the central q (bottom right). FIG. 5: Simulated j and q profiles at t = 1200 s for three different EC power depositions, obtained by varying the toroidal angles. a) is the case of Figs. 3,4.
4 be paid is of course a decrease of Q (the ratio of the fusion and additional powers), which drops from 8.7 to 6.5, as also shown in Fig. 4 (right). Other combinations of toroidal injection angles have been tried, in order to obtain the most offaxis driven current (curves b of Fig. 5, obtained for the angles 37, 43, 45 ), and for a broader driven current (curves c of Fig. 5, obtained for the angles 45, 36, 42 ). It appears that the case of Figs. 3,4 (curves a of Fig. 5), is the closest one to the optimum, owing to the fact that it maximizes the total driven current. In conclusion, the EC power available and the good CD efficiency properties of the equatorial launcher provide an effective tool for modifying the q profile around ρ 0.4. Although small, such a modification can be instrumental in forcing a rather flat q > 1 profile, to help the establishment and control of the hybrid scenario. We recall that these simulations do not include any model of the MHD effect on the qprofile, which could be an important issue for the hybrid scenario [22]. 4. ITER Steadystate scenario Although less ambitious in terms of global fusion performance (Q 5), steadystate scenarios combine a number of challenges. The very long pulses ( s), required for significant neutron fluence and the associated material testing, can only be realised if the loop voltage is practically zero. This can always be attained for sufficiently high current drive power, but the Q 5 condition limits the total auxiliary power that can be used. The simultaneous constraints on fusion performance and loop voltage can only be satisfied for extremely high bootstrap current fractions (significantly higher than 50 %), which, in turn, are more likely to be obtained in the presence of an Internal Transport Barrier (ITB). In ITER, ITBs would be associated to negative magnetic shear rather than to rotation shear. This implies that the control of the current density profile is essential to sustain ITBs for a long time, but this is notoriously difficult when the bootstrap fraction is the dominant contribution (current alignment problem). Although various scenarios have been considered for steadystate operation on ITER [13,2426], to our best FIG. 6: Time evolution of computed quantities for the steadystate scenario knowledge, no steady sustainment of ITB for times of the order of 3000 s, with the power available on ITER (P NBI < 33 MW, P IC < 20 MW, P EC < 20 MW, P LH < 20 MW) has been documented in simulations so far. FIG. 7: Profiles of computed quantities for the steadystate scenario, from left to right: temperatures and density; power sources; current density and current sources; safety factor at three different times.
5 In order to avoid shrinking or erosion of the ITB, a method is needed to control the dominant current component, i.e., the bootstrap current, which is in turn essentially related to the dominant heating source, i.e., the alpha heating. The main result of our simulations is that ECCD at ρ ~ 0.5 is an effective control tool, if the power is high enough and if the driven current is sufficiently localised (Δρ ~ 0.1). This property of ECCD is confirmed by experimental observations in DIIID [27]. In order to alleviate the current alignment problem, the current source needed to reach the zeroloop voltage regime has to be located as far from the ITB as possible. The current driven by NBI is generally located inside ρ = 0.5 and, in our simulations, invariably deteriorates the ITB on time scales 500 s. On the other hand, LHCD is generally found well in the outer half of the plasma profile [28], where the effect on the ITB stability is minimised. Therefore, NBI is not used in this scenario, which is obtained using P IC 20 MW, P EC 20 MW, P LH 20 MW. The simulations are performed for parameters typical of the projected steadystate scenarios [2], but somewhat lower current (8 MA). The simple transport model of Refs. [24,25,17] is used, i.e., χ i = χ e = χ i,neo + k(1+3ρ 2 )F(s), where F is a shear function (vanishing for negative magnetic shear) and for the constant k a more conservative value is used, i.e., k = 0.4. The pedestal temperature is fixed at a modest value, T ped 3 kev. For the LHCD, a fixed Gaussian profile centred at ρ = 0.7 is used, with a CD efficiency γ CD AW 1 m 2. The optimum EC wave deposition (i.e., at ρ ~ 0.5, Δρ ~ 0.1) is obtained using the Upper Steering Mirrors of the Top Launcher [23]. 20 MW are launched at φ tor = 20 (fixed) and φ pol = 67, although 13 MW only are in principle connected to those mirrors. The results of the simulations are shown in Figs. 6 and 7. It appears that the region of minimum q and the associated ITB are locked at ρ ~ 0.5, where the ECdriven current peaks, with practically no change in the last 1000 s of evolution. A plasma of good global performance is obtained: Q = 6, H = 1.6 (improvement factor with respect to the ITER Physics Basis scaling law), β N = 2.7, f G = 0.95, f bs > 0.7. However, 5 % of the plasma current is still ohmically driven (corresponding to V loop ~ 2 mv and a flux consumption in the flattop phase of 3 Wb) and the plasma is above the nowall stability limit (β N > 4l i ), owing to the flatness of the current density profile. The possibility of obtaining a scenario with similar properties but using a combination of top and equatorial launcher (i.e., corresponding to the present allocation of power to the ITER ECCD launchers) is now under investigation. 5. Conclusions The results of integrated modelling with the CRONOS code provide compelling evidence of the potential of ECCD for establishing and controlling advanced scenarios in ITER. In particular, offaxis ECCD using the equatorial launcher (which provides the maximum CD efficiency) could help to obtain the q profile required for the hybrid scenario (above unity and flat for ρ < 0.5). Note that the theoretical possibility of obtaining this type of q profiles with ICRH and NBCD only has not been demonstrated so far. LHCD could in principle fulfil this task [19], but the required power is likely to be higher than the 20 MW allocated, and at the present state of the ITER project LHCD will only be available in the second phase of operation. The EC power available (~ 20 MW) is in principle adequate but marginal and the possibility of a power upgrade of the equatorial launcher should be considered. A new steadystate scenario has been developed, using RF power only, in which ECCD localised at midradius plays a key role in preventing the loss of the ITB. This scenario could provide an interesting possibility for the phase 2 of the ITER programme, nevertheless, the present ECCD system is far from being optimised for current drive at mid
6 radius. In fact, there is a gap between the operational spaces of the equatorial and of the top launchers exactly at that location [23], and only part of the power is available to drive current there. The main result of this study is that current drive at midradius should be considered as one of the main functions of the ITER ECCD system, therefore it is urgent to analyse the possibility of optimisation or power upgrade of the system for this function. Acknowledgements: fruitful discussions in the framework of the ITPA SteadyState Operation Group are gratefully acknowledged. References [1] HOULBERG, W.A. et al., Nuclear Fusion 45 (2005) [2] SIPS, A.C.C., et al., Plasma Phys. Control. Fusion 47 (2005) A19. [3] SHIMADA, M. et al., Nuclear Fusion 44 (2004) 350. [4] BASIUK, V. et al., Nuclear Fusion 43 (2003) 822. [5] HUYSMANS, G.T.A., et al., CP90 Conf. Comp. Physics (World Scientific 1991) 371. [6] DE BLANK, H., Fusion Sci. Tech. 49 (2006) 111. [7] LAO, L.L., et al., Phys. Fluids 24 (1981) [8] HOULBERG, W.A., et al., Phys. Plasmas 4 (1997) [9] FENG, Y., et al., Comp. Phys. Comm. 88 (1995) 161. [10] WOLLE, B., et al., Plasma Phys. Control. Fusion 36 (1994) [11] SCHNEIDER, M., et al., Plasma Phys. Control. Fusion 47 (2005) [12] ERIKSSON, L.G. et al., Nucl. Fusion 33 (1993) [13] KRIVENSKI, V. et al., Nuclear Fusion 25 (1985) 127. [14] LINLIU,Y.R., et al., Phys. Plasmas 10 (2003) [15] BOSCH, H.S., et al., Nucl. Fusion 32 (1992) 611. [16] POST, D.E., et al., Atomic Data and Nuclear Tables 20 (1977) 397. [17] ALBAJAR, F., et al., Nucl. Fusion 45 (2005) 642. [18] KINSEY, J.E., et al., Phys. Plasmas 12 (2005) [19] ARTAUD, J.F., et al., 32nd EPS Conf. on Plasma Phys. ECA 29C (2005) P1.035; IMBEAUX, F., et al., Plasma Phys. Control. Fusion 47 (2005) B179. [20] KESSEL, C.E., et al., 21st IAEA Fusion Energy Conf. (Chengdu, 2006) IAEACN IT/P17. [21] LUCE, T., et al., Nucl. Fusion 43 (2003) 321. [22] CHU, M.S., et al., Nucl. Fusion 47 (2007) 434. [23] HENDERSON, M., et al., Proc. of the 14 th Joint Workshop on ECE and ECRH (Santorini 2006), p [24] POLEVOI, A.R., et al.,19th IAEA Fusion Energy Conf. (Lyon, 2002) file CT/P08. [25] POLEVOI, A.R., et al., Nucl. Fusion 45 (2005) [26] MURAKAMI, M., et al., Nucl. Fusion 45 (2005) [27] MURAKAMI, M., et al., Phys. Rev. Lett. 90 (2003) [28] BONOLI, P.T., et al., 21st IAEA Fusion Energy Conf. (Chengdu, 2006) IAEACN IT/P12.
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