Challenges to Radiative Divertor/Mantle Operations in Advanced, Steady-State Scenarios

Transcription

1 PSFC/RR Challenges to Radiative Divertor/Mantle Operations in Advanced, Steady-State Scenarios M.L. Reinke Oak Ridge Institute for Science Education, Oak Ridge, TN, USA, Post-Doctoral Research Fellow at the MIT Plasma Science and Fusion Center, Cambridge, MA, USA November, 2013 Plasma Science and Fusion Center Massachusetts Institute of Technology Cambridge MA USA This work supported by US DoE contracts DE-FC02-99ER54512 and in part by an appointment to the US DOE Fusion Energy Postdoctoral Research Program administered by ORISE. Reproduction, translation, publication, use and disposal, in whole or in part, by or for the United States government is permitted. ***Submitted to the proceedings of the 7 th Operation of Magnetic Fusion Devices*** IAEA Technical Meeting on Steady State

2 I/10: Challenges to Radiative Divertor/Mantle Operations in Advanced, Steady-State Scenarios M.L. Reinke Oak Ridge Institute for Science Education, Oak Ridge, TN, USA Managing the heat exhaust problem is well recognized to be a major challenge in transforming present successes in magnetic confinement fusion experiments to demonstration of costeffective, steady-state power generation from fusion. One approach is to convert plasma thermal energy, normally directed to isolated surfaces, to isotropic photon emission, distributing exhaust power over a large surface area. Successful demonstrations of this technique on existing short pulse devices is discussed, along with the inherent limitations; the collapse of core confinement with excessive radiation from the bulk plasma and restrictions to dissipation in the divertor volume. Feedback control of impurity seeding is reviewed, showing recent examples from tokamaks. For steady-state devices, additional constraints on divertor scenarios are driven by long-term plasma material interaction effects such as net erosion limits, forcing detached plasma operation where both heat and particle fluxes are substantially reduced at divertor surfaces. The instability of these detachment layers in standard X-point divertors with impurity seeding is discussed. Achieving these steady-state, high performance scenarios also restricts the divertor solution by requiring it be compatible with heating and current-drive actuators, and enhanced core confinement regimes with high bootstrap current. These complications in convincingly scaling these concepts to a reactor are outlined giving suggested avenues of continued research on existing and planned devices. 1

3 Section 1: Introduction Magnetic confinement fusion (MCF) offers the possibility of providing a substantial amount of the world s electrical power with minimal environmental impact using fuels that are plentiful and distributed world-wide. To achieve this goal using the tokamak concept, advancements in physics and engineering are required to translate present successes into a functioning demonstration reactor. While definitions of this reactor vary, domestic programs around the world have discussed a DEMO device that moves beyond ITER to demonstrate the feasibility of commercial power generation from MCF [Kwon 2008, Tanaka 2008, Neilson 2012]. There is unanimous agreement that finding a way to deal with the tremendous, > 500 MW, heat exhaust is major challenge, and nearly all concepts require achieving steady-state, or pulse lengths of order a year. Both of these are far beyond any current device, as well as planned operations for ITER. One exhaust solution which has been empirically demonstrated is to use non-fuel ions (reactants or products), to convert plasma thermal energy to photon emission in wavelength bands from the soft x-ray to the visible. This works to distribute power more isotropically over the reactor walls in contrast to having it conducted to surfaces in narrow channels in the boundary plasma, resulting heat fluxes much higher than material limits, ~ 10 MW/m 2 [Raffray 2010]. These so-called radiative scenarios are expected to form an important part of the heat exhaust strategy in ITER [Loarte 2007]. While the concept is straight-forward, application of radiative exhaust is not trivial. If not properly controlled, impurities seeded into the plasma can easily remove too much energy or dilute the fuel, both reducing the efficiency, and thus economic viability, of a commercial fusion reactor. This document outlines suggested steps to translate operation of radiative scenarios established on pulsed tokamaks to longer pulsed devices, focusing on DEMO. It is assumed that such a device will be developed along the direction of ITER, using a conventional x-point divertor with tungsten plasma facing materials. Similar reviews of radiative scenarios have been published [Fundamenski 2009, Kotschenreuther 2007] and are part of the ITER Physics basis [Loarte 2007], some reaching different conclusions than those presented here. The primary outcome of this review, is the identification of four high-level goals which are accessible to research on existing devices and those presently under construction. Completion of these tasks will increase confidence in moving forward with current divertor designs and aid in 2

4 the integrated modeling of plasmas with radiative mantles and divertors. If significant difficulties or uncertainties are encountered, increased investment in research on alternative divertor concepts should be considered. These tasks, not in order of importance are, a) in plasmas with core internal transport barriers, demonstrate the capability of to drive significant bootstrap current, f bs > 70%, while maintaining a stable radiative mantle, P RAD,CORE /P LOSS > 80 %, and active control of high-z radial impurity transport b) empirically establish the required net power crossing the seperatrix, P NET = P LOSS - P RAD,CORE, and its scaling, possibly P L-H, to simultaneously maintain reactor-like conditions; H 98 ~ 1, β N > 3, in regimes where ELMs are mitigated or eliminated and main-ion purity, n i /n e > 0.9, is achieved c) determine the maximum stationary P RAD,DIV using recycling impurity seeding, and demonstrate active control of divertor detachment to achieve the zero-erosion state, T e < T sp,thr, in the absence of ELMs, while maintaining high core confinement and core ion purity, n i /n e > 0.9 d) continue engineering and technology advancements to demonstrate compatibility of heating and current drive actuators with radiative scenarios, focusing on ICRF operation without wall conditioning and lower hybrid current drive in radiative divertor plasmas The rest of this document discusses the background behind these goals and is organized as follows. Section 2 briefly reviews examples of impurity seeding in short-pulse, inductive devices, focusing on results which demonstrate the impact of core radiation on performance metrics, helping to determine the impact of radiative mantles. Examples of radiative divertors that have demonstrated substantial heat flux reduction while maintaining high confinement are also included. Section 3 discusses the concepts introduced in Section 2 in relation to operating a steady-state DEMO-like device. Section 4 discusses the goals listed above and possible avenues for investigation on existing or near-term experiments. 3

5 Section 2: Review of Radiative Operations on Short Pulse Tokamaks This section describes work completed on existing short pulse, inductive tokamaks where radiation has been used to exhaust a large fraction of the input power, both from the core, P RAD,CORE, or from the divertor, P RAD,DIV, focusing on H-mode plasmas with good energy confinement. Generally, these are the necessary but not sufficient examples that show the tokamak heat exhaust problem can be attempted on long-pulse devices. While nearly all tokamaks have run experiments with radiative mantle or divertors, demonstrating simultaneous achievement of high core confinement, 0.9 < H 98 < 1.0, at low divertor heat loading, P ODIV /P LOSS < 20% (or P RAD,TOT /P LOSS > 80%) is less ubiquitous. Section 2.1: Demonstrating the Radiative Mantle In many reactor scenarios, the core plasma volume is not fully utilized for neutron production. The region outside of the core, r/a ~ 0.5 and inside of the pedestal, r/a ~ 0.9, is typically dominated by turbulent transport where temperature profiles are stiff, R/L T ~ constant. An attractive use of this plasma volume would be to remove power via volumetric photon emission, creating the so-called radiative mantle, rather than have the heat transported across the plasma boundary, into the divertor. Because of the elevated electron temperatures, a few kev in current devices and above 10 kev in burning plasmas, higher-z impurities, Ar, Kr, Xe, etc., are necessary in order to still have bound electrons. Atomic physics models of partially-ionized high-z atoms are still being validated, impacting the reliability of using isolated line-emission to quantitatively constrain impurity transport modeling. Despite this, properties integrated over charge-states such as the T e -dependent radiative efficiency, L Z (T e ), are known well enough to confirm that radiative mantles can be used without impacting dilution. Example curves of L Z are shown in [Kallenbach JNM 2011] where the radiated power density (MW/m 3 ) is computed via ε z =f z n 2 e L z. For a total (DEMO-like) plasma volume at 10 3 m -3 with n e ~ 1x10 20 m -3, to radiate ~500 MW requires, on average, f z L z ~ 5x10-35 Wm 3. Impurities such as Xe have L z < [Wm 3 ] for T e ranges expected within DEMO mantles, requiring f z = n z /n e > Ionization states of Z ~ 40 should be expected, making the ion fraction n i /n e =1-Zf z, range from n i /n e < Note this doesn t include effects from main-ion bremsstrahlung or even higher-z impurities such as tungsten, both which can increase ε z at little to no change in n i /n e. This perturbation is 4

6 relatively small compared to Z eff =Z(Z-1)f z which can be of order unity when using radiative seeding at levels which can perturb power balance. The impact of this radiation layer is important to understand in order to include its effect in integrated modeling codes. No empirical scaling of the energy confinement-times, such as τ E,89 or τ E,98, include radiated power, yet L-mode and H-mode plasmas have demonstrated differences when strong core power loss via radiation is applied. In L-mode plasmas, τ E has been observed to be independent of the radiated power fraction, P RAD,CORE /P LOSS, up to values approaching unity [Hill 1999, Greenwald 2007], with P LOSS = P IN dw/dt. This is presumable due stiff energy transport where heat flux is increased at nearly fixed gradient scale length, R/L T, with radiation simply reducing the Q e necessary for turbulence to transport radially. A confirmation of this in radiative plasmas, where Q e,rad +Q e,turb is observed to be constant at fixed R/L T would be a useful for mantle scenario design. Impurity seeding into L-mode plasmas has in some cases led to increased energy confinement time, H 89 increasing from to unity to approximately 1.5, resulting in the radiative-improved (RI) mode (see [Jackson 2002] and references therein). This has not been considered for a viable reactor scenario, but points to interesting physics that may be relevant such as Z eff or dilution stabilization of turbulent transport. For H-mode plasmas, the impact of a radiative mantle on energy confinement is connected to sustaining the edge transport barrier (ETB), the pedestal, which requires a minimum level of power. In regimes without internal transport barriers, global confinement (β N, H 98 ) are well correlated to the edge, with the idea that stuff profiles are raised by the ETB boundary condition, hence the pedestal moniker. In recent Alcator C-Mod experiments, the effect of radiative mantles on H-mode confinement was directly studied [Loarte 2011]. In EDA H-mode plasmas, the input ICRF power was scanned during discharges, along with shot-to-shot scans of impurity seeding rate. Ne, N 2 and Ar seeding were used in separate, repeatable discharges, and compared alongside unseeded plasmas (which in reality had varying levels of intrinsic molybdenum emission). When restricting to plasmas with seperatrix density below 50% of the line-averaged density, H 98 was shown to respond linearly with P NET H 98 = P NET /P LH. Here the P NET =P LOSS -P RAD,CORE is the net power crossing the seperatrix onto open field lines, and P LH is the cross-machine scaling of the L-H power threshold [Martin ] computed using the density during the steady H-mode phase. Core confinement remained well correlated to pedestal 5

7 temperature, implying an increase in P RAD,CORE simply led to a colder pedestal. Obtaining H 98 ~ 1 at P NET /P LH ~ 1 is favorable which places reduced burden on the divertor for heat exhaust, although the C-Mod EDA H-mode regime is not reactor relevant [Greenwald ]. For ELMy H-mode plasmas, this behavior can be masked by the impact of radiation on the ELM frequency and regime. In JET plasmas seeded with either Ne or N 2, the ELM frequency varied non-montonically with the radiated power fraction, F rad, measured between ELMs, transitioning between Type-I, compound and Type-III ELMs. Over 0.15 < F rad < 0.60, the pedestal temperature dropped monotonically, while pedestal pressure only dropped after F rad > 0.3 [Beurskens 2008]. Earlier results using argon seeding showed consistent results with confinement degradation tied to the pedestal, with seeded and unseeded reference shots lying on the same trend of pedestal to total stored energy [Monier-Garbet 2005]. In those plasmas, H 98 dropped as P RAD /P LOSS, with H 98 > 0.9 maintained up to P RAD /P LOSS ~ 0.7, beyond which Type-III ELMs were observed. C-Mod ELMy H-mode plasmas were shown to have a similar scaling with P NET as the EDA H-modes discussed above [Hughes 2011]. Comparisons of radiative regimes between devices have only recently been started. Coordinated under the role of the ITPA Integrated Operational Scenarios group, the core confinement versus was compared between Alcator C-Mod, ASDEX-Upgrade and JET (CFC) and summarized in [Kallenbach IAEA 2012]. This work found H 98 =0.98(P NET /P TH ) 0.15, much weaker scaling than found directly on C-Mod, but still a favorable result as H 98 ~ 1 is found at P NET /P TH ~ 1. This results suggests radiative mantles can be useful to remove any input or self-generated power above ~P TH. Further confirmation of this observation is suggested, especially in ELM-mitigated regimes. Section 2.2: Demonstrating the Radiative Divertor The power necessary to sustain the H-mode is transported across the seperatrix, and then must be either be removed by radiation or charge-exchange in the boundary plasma, or conducted along open magnetic field lines to material surfaces. Recent work to characterize the area of deposition has suggested that in cases without dissipation, power is deposited in a very narrow layer which does not scale with device size [Eich 2012]. While questions remain on extrapolating this 6

8 result to a burning plasma [Whyte 2013], the result implies a need for strong dissipation to reduce the peak heat flux to within material limits. The introduction of intrinsic or extrinsic impurities into a divertor plasma enhances the dissipation of energy, driving down the temperature. Unlike in the core, T e values in the divertor favor the use of low-z, partially ionized impurities. In the sheath-limited and high-recycling divertor regimes, pressure remains constant along flux tubes from the divertor surface moving towards the confined plasma. In the high recycling regime, dissipation is sufficient to drop the temperature, requiring increased density along the field line to maintain a constant pressure. At sufficiently low temperature, friction, recombination and charge-exchange processes can remove momentum, impacting pressure balance along a field line. This can lead to detachment where pressure and temperature at material surfaces are much less than measurements on the same field lines further from the plate. A more thorough description of divertor detachment can be found in [Matthews 1995] and [Stangeby ]. The effect of detachment has generally been shown to have negative consequences on core plasma performance. Examples from JET [Fundamenski 2009], DIII-D [Fenstermacher 2007] and Alcator C-Mod [Goetz 1996] show the strong radiation layer moving from the divertor volume into the confined plasmas, resulting in a drop in stored energy. This represents the unchecked evolution of the dissipation, and cases exist of sufficient radiative removal of exhaust power to significantly drop heat flux while maintaining core energy confinement at 0.9 < H 98 < 1, and have been shown in ASDEX-U [Kallenbach ], JET [Giroud ] and Alcator C-Mod [Loarte 2011]. Infrared imaging or Langmuir probe data are used estimate the heat flux profiles, and changes due to seeding examined. In ASDEX-U, core argon seeding and divertor nitrogen seeding have been combined to demonstrate enhanced radiative loss plasmas where nearly over 85% of the ~23 MW of input power were converted to radiation with core performance, H 98 ~ 1.0 and β N ~3.0, was maintained. For C-Mod, reduction of the divertor heat flux to within ITER targets, P O-DIV /P IN < 20%, were achieved with all extrinsic seeding but only Ne and N 2 were able to achieve this while maintaining H 98 > 0.9. When using argon to reach similar low levels of heat flux, radiation in the core degraded pedestal temperature. Similarly, in JT-60U H-mode plasmas with main-chamber argon seeding, P RAD,TOT /P IN > 80% was achievable only for H 98 < 0.85, and was correlated with the transition to Type-III ELMy H-mode [Asakura 7

9 2009]. In JET (CFC) both Ne and N 2 seeding were also able to reduce the peak heat flux at the outer divertor by an order of magnitude and approach detachment while keeping H 98 > 0.9. Increases in seeding which transitioned plasmas from Type-I to Type-III ELMs were observed to have reduced confinement, consistent with earlier work [Monier-Garbet 2005]. Similar experiments in JET with a tungsten divertor are currently being analyzed and will present an important result for the tokamak community, comparing high-z walls with low-z seeding with earlier work with a carbon divertor. Within these experiments, effects due to neutral deuterium fueling have also been shown to be important, both in adding additional dissipation as well as impacting impurity transport. Core levels of impurities used for divertor heat exhaust varied between experiments, and generally require more detailed analysis to separate from measurements such as Z eff which may also include intrinsic impurities. Section 3: Radiative Operations on Steady-State Tokamaks The distinction of what makes a plasma steady-state depends on the physics of interest. Energy transport and current profile diffusion occur on time-scales much shorter than it takes for wall conditioning techniques to degrade. Long-pulse, devices which are sustained for several hours, may themselves be insufficient to access plasma material interactions which evolve on the order weeks or months [Whyte 2011]. In this sense, steady-state is defined here relative to the ~1 year, continuous operation timescale generally accepted to be required for economic operation of a fusion reactor [Zohm 2010]. This section describes how existing radiative mantle and divertor operations are affected by moving from pulsed, inductive devices to steady-state noninductive devices. The pulsed DEMO reactor [Zohm 2010], which includes some inductive current drive is not explicitly discussed, but many concepts are still relevant. Section 3.1: Controlling the Zero Erosion Divertor In present devices, much effort has been spent on reduction of the heat flux to be within limits of present technology, < 10 MW/m 2, but when moving to steady-state, an additional constraint on the particle flux to surfaces is created by erosion. At sufficient, material-dependent temperatures, charged particle impacts on surface will result in a removal or sputtering of the plasma facing component (PFC). In reactor devices, higher density and longer pulse lengths 8

10 would result in 10 5 higher material removal rates than present experiments and 10 2 higher than expected for ITER [Stangeby 2011]. Thus, it is argued that the only reliable solution is to turn off sputtering by reducing the temperature at the divertor target below the sputtering threshold, T sp,thr. As in reducing heat flux, impurities seeded into the divertor can play a large role in reducing T e < T sp,thr, but because energy is gained by the sheath potential, multiple charged impurity ions impact at higher energies, making T sp,thr a function of n z and ionization state. Note that material erosion and transport in the main chamber [Pitts 2005, Matthews 2005] is also important to consider, but radiative operation is not expected to impact that process. To achieve this zero-erosion operation will require the divertor to operate at or near the detached divertor regime. Empirically this has been linked to reductions in energy confinement, with estimates from parallel transport modeling describing a thermal front instability which causes the region of low-t e to move from the plate toward the confined plasma [Hutchinson 1994, Krasheninnikov 1999]. Time-dependent, 1-D simulations of coupled mass, momentum and heat transport show this effect can be fast, with front speeds on order of 1 km/s [Nakazawa 2000, Nakamura 2011]. These simulations also indicate possible means to stabilize the front evolution by cross-field heat and particle exchange on open field lines. More advanced divertor simulations using B2-EIRENE show that stable solutions near detachment are possible only when radiation transport is included [Kotov 2012]. These models suggest that active control of the detachment front may be possible by using direct heating in the boundary or divertor. By providing localized heating on flux tubes that are detached from the plate, the thermal front could be stabilized at an intermediate location, preventing the high-density cold plasma layer from approaching the confined plasma. In present devices, this effect may already be at play in the form of ELMs and heat pulses through the boundary from sawteeth at frequencies of order 100 Hz. Detailed investigation of the 2D, time-evolving structure of low temperature divertor plasmas will allow for confirmation of these effects and offer new observers for more advanced real-time control of divertor detachment. Section 3.2: Feedback Control of Seeded Impurities In the radiative H-mode experiments discussed in Section 2, the preferred gas for divertor seeding has been nitrogen, primarily chosen for its maximum radiative efficiency at low temperatures, T e ~ 10 ev. Demonstration of feedback control of N 2 gas puffing was shown 9

11 during initial detachment research [Lipschultz 1997, Kallenbach 1995], and present operation of ASDEX-U with tungsten PFCs requiring the use of seeding control in high powered operation. In initial feedback work with full tungsten PFCs, the observer was thermo-electric current through a tile at outer divertor strike point, measured in real-time by the voltage drop across a shunt resistor. Good correlation was found between the current and temperature measured using SOL Langmuir probes [Kallenbach ]. More recent work has used multichord resistive bolometer data, a more ubiquitous diagnostic, to simultaneously control core argon puffing and divertor nitrogen puffing [Kallenbach 2012]. Feedback control of JT-60U seeding was also accomplished using bolometry [Asakura 2009]. In JET (CFC) hybrid H- mode plasmas, control of divertor nitrogen seeding was demonstrated using observation of VUV emission, the nm Li-like N V line, at the inner strike point [Maddison 2011]. Upcoming operation will focus on real-time control of low-z seeding in JET ITER-like wall plasmas [Horton 2013]. For recent Alcator C-Mod impurity seeding, feedback control was not implemented. A monotonic drop in core line emission from intrinsic molybdenum was observed with increased low-z seeding, initially dropping P RAD,CORE. A minimum in radiated power was observed with increased radiated power in the core resulting from increased seeded impurities [Reinke 2012]. This suggests a multiple observer system would be required in order to distinguish changes in intrinsic and extrinsic impurities. For high-z radiative mantles using extrinsic seeding, argon has been widely used while experience with krypton and xenon in H- mode plasmas is limited. Krypton seeding has recently been used in extensions of the twoimpurity feedback control research on ASDEX-U [Kallenbach 2013] and xenon has been used in radiative H-mode studies at C-Mod, neither demonstrating long-term effects on machine operation. Further effort on characterizing storage of high-z noble gases in metallic PFCs is suggested. When moving to steady-state, D-T discharges, the use of nitrogen has additional complications. Empirically, low-z seeded operation with nitrogen requires nearly an order of magnitude higher mass flux than using neon. Unlike noble gases, nitrogen is only partially recycling and will chemically react to form nitride layers and ammonia [Neuwirth 2012]. Moving to high temperature plasma facing components, as expected in DEMO [Tillack], may ameliorate the impact of ammonia on fuel retention, but it has not be thoroughly studied. Near term, the tritium processing technique used in JET are incompatible with nitrogen, and upcoming D-T operation 10

12 currently under discussion will not allow its use for heat flux control. Thus, there is some risk of incompatibility for continuing radiative scenarios developed designed exclusively for nitrogen. Qualitatively similar results have been found using neon and argon seeding as discussed in Section 2, but the detailed feedback control will be different due to its higher recycling. While radiative efficiency of Ne and Ar are similar to nitrogen, they peak at different temperatures [Kallenbach ], possibly impacting solutions for active detachment control. Section 3.3: Compatibility of Radiative Operations with Heating & Current Drive Actuators In present inductive and short-pulse superconducting devices, wall conditioning techniques can be used to ameliorate influx from plasma facing components for short durations. Removal of these layers via plasma material interaction occurs on a time scales between a handful of discharges on inductive devices and wall equilibration or erosion timescales in steady-state devices. While their presence cannot be assumed during sustained flattop, they may remain relevant for ramp-up or used as a means for learning during initial short-pulse operation on new devices. An important impact of the lack of conditioning is on the use of ion cyclotron range of frequency (ICRF) heating. For machines with metallic PFCs, influx of impurities during ICRF heating has led to excessive core radiation in plasmas without conditioning. While this concept may seem compatible with the radiative mantle, the efficiency of using ICRF for heating or current-drive is reduced, and the complexity increased by linking to an actuator for radiative heat exhaust. On Alcator C-Mod, electron cyclotron discharges seeded with diborane are used to coat PFCs with a thin layer, referred to as boronization. This process is required for operation of steady, ICRF-heated H-modes with high normalized confinement, H 98 ~ 1 [Lipschultz 2006]. In a fixed target plasma, degradation of this conditioning process is tied to the integral number of Joules injected by an ICRF antenna [Wukitch 2009], and operation with seeding has been demonstrated to have little impact [Reinke 2012]. Seeding has been shown to improve the stability of ICRF operation, reducing the variance but not the mean in the time-evolving loading through an as of yet unidentified mechanism. This has enabled steady operation closer to the same voltage safety limits, reducing trips and enabling ~10% higher input energy [Reinke 2012]. Investigations into the underlying mechanism of ICRF impurity generation is a longstanding problem, and one of the main theories is the role of enhanced supptering at PFCs due to 11

13 rectifying RF fields increasing the sheath potential (see [Myra ] for a discussion of this process). To try and mitigate the effect by reducing the parallel electric field, E, along field lines passing an ICRF structure, C-Mod implemented a rotated antenna design which made ICRF straps perpendicular to the magnetic field [Garrett 2012]. Recent results have shown a reduction in molybdenum in H-mode plasmas using this field aligned antenna relative to plasmas heated with a conventional antenna [Wukitch 2012]. Despite this advantage, initial boronization of C-Mod was still required, but may be linked to the role of conditioning in lowering the hydrogen fraction for efficient minority heating. In ASDEX-U with full tungsten PFCs, unrestricted operation of ICRF initially required boronization to prevent strong influx of tungsten [Gruber 2009]. A substantial fraction of this impurity source was identified to come from outboard limiter structures, where sputtering enhanced by rectified sheath potentials was also invoked [Bobkov , Bobkov ]. Modifications to the ICRF antenna straps and surrounding structures were shown to improve compatibility with unboronized plasmas. Coating antenna limiters with thick layers of boron resulted in a clear reduction in core tungsten by nearly 50% when compared to ICRF heating using antennas with tungsten limiters [Bobkov ]. Design of a new 3-strap antenna made to reduce E by reducing RF image currents in limiters is underway, and is planned for upcoming ASDEX-U operation. Examining the impact of a radiative mantle and divertor on auxiliary current drive, two primary concepts, electron cyclotron (ECCD) and neutral beam (NBCD) current drive do not appear to be strongly perturbed. Increases in Z eff by strong radiative mantle will impact these processes through the conductivity and, for NBCD, the beam deposition profile, but the underlying physics is known. Care must be taken when solving the parallel neoclassical transport in cases with perturbing, Z 2 n z /n i > 1, high-z impurities as poloidal variations in their density are much easier to access than main-ions [Reinke 2013] and impact flux-surface averaged conductivity through the ion-impurity friction. Parameterized approaches to find the impurity impact on conductivity [Sauter ] are more relevant for Z eff ~ 1 from low-z ions, and codes such as NEO which solve the multi-species parallel transport should be included in self-consistent scenario modeling. Lower hybrid current drive (LHCD) has also been demonstrated as an important tool for noninductive current drive on present devices, and is expected to play a role in ITER hybrid 12

14 scenarios. Unlike electron cyclotron or neutral beam power which can propagate through a large vacuum layer, LHRF structures must remain close to maintain efficient coupling, which can be perturbed by changes to the edge density. This forms the concern for compatibility of LHCD with radiative divertors which must maintain high pressure in the divertor for dissipation, with pressure balance on open field lines resulting in higher density and colder temperatures upstream towards launcher than in attached divertors. Substantial work on LHCD physics has been in limited tokamaks such as Tore-Supra, FTU, and Alcator-C where compatibility with radiative divertors cannot be explored, and little work on this topic has been completed in divertor tokamaks, making predictions for ITER or DEMO uncertain. In present devices, the efficiency of LHCD has been shown to drop off strongly as line-averaged core density is increased in both diverted [Wallace 2012] and limited configurations [Goniche ]. While the underlying cause has not been reliably demonstrated, absorption in the SOL [Wallace 2012] and parametric decay instabilities [Baek 2013] have been proposed as possible explanations, both of which have been observed empirically. It has been argued that maintaining a strong singlepass absorption regime is crucial for avoiding both of these mechanisms, leading to situation where at high density and low collisionality is required inside the pedestal. This regime is inaccessible in present devices, suggesting that ITER will be the first device to be able to demonstrate efficient LHCD with a radiative divertor. Technology changes could allow this to be explored in present scale devices by moving to high-field side or off-midplane launch [Podpaly 2011], allowing for changes in radial LH wave propagation and avoiding interactions in boundary plasmas. Basic testing of novel concepts is strongly encouraged in small-scale devices which offer low-cost means to demonstrate high-impact changes to conventional tokamak designs [LaBombard 2013]. An additional note on compatibility with radiative mantles comes from the additional heating of LHRF launcher structures. While radial power flux at the wall will always be dominated by neutrons in a burning plasma, their deposition is volumetric relative to the SXR/VUV emission from photons being absorbed at the surface. Particular concern may arise from the mantle radiation from high-z impurities in plasmas where rotation is sufficient, v φ /v th,z ~ 1, to allow centrifugal force to concentrate impurity density on the outboard midplane. Again, this physical effect is well known and simply needs to be included in scenario modeling. 13

15 Section 3.4: Compatibility of Radiative Mantle with Internal Transport Barrier In commercial reactors, maintaining a low externally-sustained current drive fraction, f CD < 0.3, has been shown to be important for maintaining low circulating power, and thus economic efficiency [Zohm 2010]. Self-generated, bootstrap currents can be sustained by pressure gradients as described by parallel neoclassical transport [Sauter ], J bs ~ p 0.5 ln n e ψ ln T e ψ ln T i ψ enabling an attractive operating regime where current is maintained simultaneously with high pressure, and thus high fusion reactivity, in the core plasma. To generate substantial bootstrap current, internal transport barriers (ITBs) have been formed on a number of devices, where the mid-radius turbulent transport is suppressed by a mix of ExB and magnetic shear [Wolf 2003]. Work on JT-60U has demonstrated high bootstrap current phases, f bs > 70%, sustained for multiple current relaxation times [Ide 2005]. Along with suppression of main-ion transport to form the ITB, the impurity transport is also suppressed, leading to strong accumulation inside the barrier. In JET plasmas, the impurity convection has been observed to approach neoclassical predictions with, v neo D neo = Z z Z i 1 n i n i r H 1 T i T i r and 0.2 < H < 0.5 [Dux 2004]. Here v neo /D neo is the ratio of impurity convection and diffusion with positive values resulting in peaked profiles. This presents difficulties for operating at high bootstrap fraction with a radiative mantle, since electron density peaking efficiently drives current and inward convection of impurities. In JT-60U high-β p ITBs, on-axis ECH was demonstrated to reduce core accumulation of argon, but reduced electron density peaking at the same time, consistent with this model. In reversed shear ITB plasmas, the ECH heating shown to be insufficient to substantially modify the argon or electron density profile [Takenaga 2003]. In Alcator C-Mod, ITB plasmas are generated by off-axis ICRF heating, where reduced peaking in the ion temperature profile is through to reduce ion temperature gradient (ITG) growth rates to levels where intrinsic ExB shear is sufficient to stabilize the mode [Fiore 2012]. In these plasmas, impurity and electron density are observed to increase monotonically, leading to 14

16 eventual collapse, unless on-axis ICRF heating is applied. Measurements and modeling indicate the destabilization of density gradient driven trapped electron mode (TEM) turbulence is responsible for this increased transport [Ernst 2004, Ernst ]. Results from ITB interactions with radiative mantles exemplify the interlinked aspects of transport and currentdrive highlighted in [Luce 2011], with reliable means of maximizing f bs while avoiding impurity accumulation still to be demonstrated. 3.5 Compatibility of Radiative Divertor with ELM Mitigation Strategies In present devices, sustained Type-I ELMs are not a concern for maintaining PFC integrity over their lifetime. As outlined earlier, for steady-state scenarios reaching zero-erosion is critical, and stabilizing ELMs is an important part [Pitts 2005]. Recent work outlined the approaches ITER is planning to take [Lang 2013], and techniques for DEMO can be expected to be developed along similar lines. For active control, ELM triggering via pellet injection (not discussed here) and ELM stabilization with the use of 3D magnetic fields are suggested as viable for ITER, with research on the impact of radiative divertor operations an important area of continued research. When non-axisymmetric fields from external coils add a radial field component and are phased properly relative to the field line geometry, a resonant magnetic perturbation (RMP) is applied which has been demonstrated to suppress ELMs on multiple devices [Evans 2013]. Although the mechanism is still under investigation, the effect is to prevent the pedestal from evolving uncontrollably towards the peeling-ballooning limit which would otherwise result in an ELM. This necessarily results in a degraded pedestal relative to a typical ELMy H-mode, although the optimization of 3D field has not yet been attempted. Initial degradations of performance metrics can be made up by increasing heating power without recovering ELMs [Evans 2008]. This works to increases the P NET /P LH, placing an additional burden on the radiative divertor, and makes integration of ELM mitigation with estimates of necessary power to maintain the core solution important. Compatibility of radiative divertor with RMP ELM mitigation has only recently begun. Divertor seeding of Ar into DIII-D H-mode plasmas show a 50-80% increase in core argon content compared to plasmas without using RMP [Petrie 2011]. Observed modification of the magnetic field structure in the region of the x-point [Shafer ] may be leading to enhanced penetration of divertor seeded impurities into the core, and the compatibility 15

17 with detachment control needs to be explored. Additionally, the 3D-field works to make the boundary plasma non-axisymmetric, and toroidally non-uniform heat flux patterns have been predicted [Schmitz ]. In order to prevent non-uniform erosion, the phase of the 3D field must adjusted so that the perturbation rotates in the lab frame, with a few Hz operation indicated as sufficient for ITER [Loarte 2013]. Section 4: Discussion/Conclusions In Section 1, tasks are listed which compiled physics or operational scenarios goals which have been argued to be important throughout Sections 2 and 3 for increasing the confidence in steady state reactors. In some cases the underlying physics can be explored in present or near-term devices using inductive and non-inductive operations and this section outlines possible avenues for exploration. Operation with radiative mantles can certainly be explored in present devices. A more complete demonstration of pedestal response to P NET =P IN -P RAD,CORE -dw/dt in ELM-less regimes is critical to establish the divertor boundary condition. The QH-mode and RMP H-mode can be explored on DIII-D plasmas that approach the ITER collisionality and should be combined with core radiative seeding and high-z impurity transport studies. DIII-D is an ideal experiment, as the lack of high-z PFCs allows dedicated studies of the core impurity transport physics independent of interaction with the wall. Additionally, if I-mode is to be established as a credible reactor scenario, a proper normalization for confinement versus P NET and continued cross-machine scaling is required. The technique of on-axis impurity control in ITB plasmas, mirrors the use of on-axis heating to control impurity accumulation in JET [Valisa 2011] and AUG [Dux 2003] in conventional H-modes. In AUG, small, < 1 MW, of core ECH is shown to suppress core accumulation of tungsten, but also has effects on the peaking of the electron density and rotation profile [McDermott 2011]. A better understanding, leading to predictive capability, of these effects will help identify the turbulent transport mechanisms at work. Selective suppression of anomalous transport responsible for heat and particle transport would aid in maximizing bootstrap current and reducing core high-z impurity peaking. Heating in early current ramps enables transient weak or reversed magnetic shear, where basic physics of impurity transport in 16

18 internal transport barriers is accessible. For steady-state integration, the JT-60SA device will be an ideal testing ground for this research, continuing on from critical research on advanced scenarios discussed above [Kamada 2011]. To provide a convincing answer as to the compatibility of ICRF with steady-state reactors, results from multiple devices must show similar success. Comparing ASDEX-U and Alcator C- Mod indicates modification of antenna limiter structures may be insufficient, as thick boron coatings or boron-nitride antenna limiters in C-Mod were insufficient to permit high-quality unboronized H-mode operation. Initial results from ICRF operation in JET s ITER-like wall show more similarities with C-Mod results, with impurity sources away from the ICRF limiters responsible for core high-z contamination [Mayoral 2012]. ICRF-heated H-mode experiments have recently been demonstrated on the EAST tokamak [Zhang 2013], and unique contributions to the study of impurity source evolution can be expected from that device. It is sufficiently clear that engineering and technology changes in ICRF antenna design are being widely pursued as means to ameliorate the long-standing problem of impurity contamination. Continuing this research on multiple devices is necessary to demonstrate antenna concepts, and it would also help to define boundary conditions for acceptable, weakly perturbing ICRF operation. For example, at C-Mod, a compact torus with all metal PFCs and high-performance H-modes that are entirely RF-heated, successfully running non-boronized H-modes might be too stringent of a test for demonstration of ICRF operation in a DEMO. Lastly, there are no first-principles calculations that thoroughly demonstrate that the conventional x-point divertor is or is not sufficient for a steady-state fusion reactor. The ability to model divertor plasmas near detachment has demonstrated qualitative and quantitative shortcomings that remain unresolved [Wischmeier 2012]. Current flexibility allowed in numerical modeling needs to be constrained via measurements using new diagnostics of the volumetric divertor properties. Improved diagnoses and improved modeling capabilities will reduce, but certainly not eliminating risk in DEMO scenario planning. To help further mitigate this risk, alternate concepts should be tested on current or near-term devices, such as the Super-X [Valanju 2009] and snowflake divertor [Ryutov 2007]. These, or other concepts, are envisioned as a possible manifestation of the Divertor Test Tokamak discussed in the EFDA Fusion Roadmap [Romanelli 2013]. 17

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