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Blanket–plasma interaction in tokamaks
Fusion Engineering and Design 81 (2006) 1589–1598
Blanket–plasma interaction in tokamaks Implication from JT-60U, JFT-2M and reactor studies M. Kikuchi a,∗ , S. Nishio a , G. Kurita a , K. Tsuzuki a , M. Bakhtiari b , H. Kawashima a , H. Takenaga a , Y. Kusama a , K. Tobita a a b
Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Muko-yama 801-1, Naka, Ibaraki, Japan University of Wisconsin-Madison, 735 Engineering Research Bldg., 1500 Engineering Physics, Madison, WI 53706, USA Received 7 March 2005; received in revised form 5 September 2005; accepted 5 September 2005 Available online 10 January 2006
Abstract Optimization of blanket–plasma interaction is one of important subjects of tokamak reactor design. This paper summarizes key physics R&D results of blanket–plasma interactions in tokamak system based on JT-60U, JFT-2M and DEMO studies at JAERI. Conflicting requirement between wall stabilization and disruption tolerance with respect to the electromagnetic coupling between plasma-structure seems to have a solution with somewhat distant wall location while minimization of the distance between plasma and stabilizing wall is critically important for high beta operation. Effect of ferromagnetism on plasma stability is well studied in JFT-2M and theoretically giving positive prospect for use of reduced activation ferritic steel as a first wall material. Enhancement of ion/neutral 1st wall interaction with 2nd SOL and saturated wall is discussed as an important R&D element for DEMO and existence of high-energy neutrals in JT-60U was mentioned. © 2005 Elsevier B.V. All rights reserved. Keywords: Blanket; Plasma; Disruption; Wall stabilization; Ferromagnetism; Wall saturation
1. Introduction Optimization of blanket–plasma interaction is one of major design issues of economically viable fusion DEMO. Tokamak without wall stabilization of ideal MHD modes has a beta limit, so-called Troyon limit βt = βN Ip /(ap Bt ) where βt , βN , Ip , ap , Bt are volume average toroidal beta, normalized beta value (βN ∼ 4li : ∗
Corresponding author. E-mail address: [email protected] (M. Kikuchi).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.09.048
li is internal inductance), plasma current, plasma minor radius, toroidal magnetic field, respectively. Wall stabilization of ideal MHD modes through strong coupling between plasma and stabilizing shell provides an important opportunity to increase attainable beta in a tokamak while strong plasma–blanket coupling may lead to difficulty for disruption tolerance of blanket. Therefore, it is important to clarify the prospect of disruption mitigation and to find a solution consistent with wall stabilization for DEMO and commercial reactors. Also, adoption of reduced activation ferritic
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2. Disruption as a controllable plasma–blanket interaction
assume “no disruption” since any magnetic configuration has an operational limit and there is non-zero probability to reach its operational limit for example by a flake of surrounding material into the plasma and to induce radiative collapse. Any man-made system has possibility of troubles and it is fundamentally difficult to say “no trouble for this system”. Thus, the key point for this question on “disruption” is not to eliminate “disruption” but to reduce the frequency of disruption less than the frequency of off-normal event acceptable for commercial use. In order to categorize “disruption” as “trouble”, machine must withstand electromagnetic (EM) force and thermal loads by rare disruptions. Acceptable frequency of off-normal event could be discussed from past experiences of power plants and 1 event/2 year is proposed as a typical number in reference [7].
2.1. Disruption as an off-normal event in DEMO and commercial reactor
2.2. Development of disruption mitigation techniques
Firstly, we would like to consider the condition of disruption in DEMO and future commercial reactors. DEMO as a final step before the commercialization of fusion power must demonstrate reliable operation up to a year at least towards the end of its operational life. It is a long-standing question whether “disruption-free” can be realized to improve operational reliability and to relax the design requirement of blanket system against disruption force and thermal loads. It is unrealistic to
2.2.1. Neutral point for VDE, Killer Pellet for thermal quench and runaway electron suppression Disruption phenomena is illustrated in Fig. 1 showing four important processes, namely thermal quench, current quench, runaway electrons and VDE. Possibility of “thermal quench” exists for all magnetic confinement systems by radiative or MHD collapses while tokamak including ST has additional effects due to EM force induced by current quench.
steel (RAF) as a primary candidate structural material for first wall brings concerns on the effect of ferromagnetism for plasma performance. Avoidance of energetic neutrals/ions impinging to first wall is also a critical issue to assure the long life of blanket. Tokamak devices at JAERI such as JT-60U [1] and JFT-2M [2] addresses above physics issues to optimize blanket–plasma interaction such as disruption mitigation [3] and wall stabilization for high beta steady-state operations [4], and wall-plasma/neutral interactions. Reactor system studies at JAERI have progressed remarkably in recent years to have many implications to a direction of fusion plasma researches [5,6].
Fig. 1. Schematic figure of physical processes of disruption is shown. Thermal quench is an abrupt loss of thermal energy and is followed by plasma current termination. Under some conditions, fraction of plasma current is kept due to the formation of runaway electron current tail. During these processes, plasma may have a vertical movement, namely “Vertical Displacement Event (VDE)” and this movement induces poloidal current named “Halo Current” to induce strong electromagnetic force on structural components contacting VDE plasma.
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Fig. 2. Waveform of plasma current, electron temperature during forced plasma shut down induced by Ar + H2 , H2 and Neon pellet injection. Demonstration of plasma shut down by Ar + H2 injection as compared with “Killer Pellet” for current termination. Intensity of hard X-ray (measure of runaway electron population) and the heat flux to the divertor plate are plotted for various combinations of gas injection showing that runaway population and heat flux are minimum for Kr + H2 injection [14].
The vertical displacement event (VDE) is known to induce halo current that would damage the in-vessel components. Control technique for VDE is developed in JT-60U by locating the center of the plasma current at so-called “neutral point” as predicted by numerical calculation [8] and JT-60U experiments [9]. The neutral point is defined as a location where attractive forces due to eddy currents induced by dIp /dt become zero in both radial and vertical direction. The major passive element of the in-vessel components of tokamak fusion reactor is the stabilizing shell behind the breeding blanket and it should be designed so that center of plasma current is located at the neutral point. Thermal quench may lead a large convection/ conduction heat flux preferentially to divertor plates. Key point to soften this phenomenon is to reduce heat flux to the divertor plates through radiation power deposition to the first wall before the thermal quench, for example, by impurity pellet injection. This concept was demonstrated in JT-60U [10] and was called “Killer Pellet”. Thermal quench and associated increase of plasma resistance may lead to a production of runaway electrons. Various methods to suppress runaway electrons such as excitation of MHD [11], and lowering q surface to less than 2 [12] were successfully demonstrated. A collision-less loss mechanism due to magnetic perturbation was analyzed and found to be effective to suppress runaway electrons [13].
Recent works on disruption mitigation in JT-60U focused on the prediction of disruption through neural network [14] and also development of mitigation technique for thermal quench and runaway electron [15]. Although “Killer Pellet” is an important tool for mitigation of thermal load to the divertor, it might excite runaway electrons since it does not provide enough electron density and its drag force. Injection of mixture of inert gas and hydrogen provides another opportunity for a suppression of thermal load to the divertor plates and generation of runaway electron. In particular, Kr + H2 gas injection was quite effective in terminating the discharges before thermal quench without runaway electron generation and with low divertor heat load as shown in Fig. 2 [15]. If we adopt this kind of mitigation technique for tokamak fusion reactor, the first wall structure must be strong enough against thermal shock from this radiation. Despite such successful development of disruption mitigation techniques, it is unlikely to increase the current quench time much longer than the expected values of ITER scaling (dIp /dt ∼ S ∼ 27 ms for ITER: S is cross-sectional area of plasma). 2.3. Blanket design withstanding EM force of current quench Most important force on blanket structure due to current quench is an EM force by an induced current as shown in Fig. 3. This force produces a
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Fig. 3. Two types of blanket structures to withstand EM forces in tokamak reactor. Left hand side shows blanket concept of SSTR [16] where FGM insulator is used to eliminate loop current around the blanket box. Right hand side shows recent blanket concept [6] allowing current flowing in the blanket box where induced eddy current loop is shown by “→” and EM force on blanket box is shown by “ ”.
torque to twist the blanket box. The torque M is roughly proportional to Ip ·Bt ·d·t where Ip , Bt , d and t are plasma current, toroidal field at the blanket, toroidal width of blanket, radial thickness of blanket, respectively. There are two ways of blanket design withstanding this EM force. First one is to make a cut for this current loop by a FGM (Functionally Graduated Material: insulator) at the back of blanket box placed vertically adopted for SSTR design [16] as shown in Fig. 3 (left). This design allows easier structural design with keys and separation of tritium breeding blanket to fixed blanket and replaceable blanket (double-layer blanket to reduce radioactive-waste). The feasibility of FGM as a part of blanket structure becomes an important R&D issue for this design. Another approach is to adopt single layer blankets without FGM that are fixed to back plate firmly by using welding and thick keys [6]. It is reported that blanket design withstanding torque of 2 MNm would be possible expected for a current quench time of 30 ms close to a value expected in ITER. For a low aspect ratio DEMO concept [5], difference of toroidal field at low and high field sides allows three times longer toroidal width d for low field side blankets which might be sufficient to achieve overall TBR > 1 since low field side blankets share major part of TBR.
3. Plasma performance enhancement through optimization of blanket–plasma interaction 3.1. High normalized beta plasma sustainment in JT-60U long pulse experiments and needs for high beta steady-state operation above the no-wall limit ITER is an important step towards the technical feasibility of fusion energy. In tokamak confinement geometry, beta limit under no-wall condition is governed by so-called Troyon scaling (βt = βN Ip /aBt : βN ∼ 2.8). Technical objectives of ITER could be achieved at the plasma beta below the Troyon limit. And the long sustainment of ITER relevant high normalized beta was successfully demonstrated in JT-60U by extending the machine capability from 15 to 65 s (high heating power is limited to 30 s). Fig. 4 shows recent typical waveform of long sustainment of high βN discharge much longer than the current diffusion time and the progress of JT-60U experiments in (βN , duration) plane with target regimes of ITER inductive and steady-state operation. Long sustainment of high beta close to ITER advanced operation was successfully demonstrated. However, economically viable fusion reactor needs steady-state operation at higher βN above the ITER advanced operation. The mass power density (MPD)
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Fig. 4. Long sustainment of ITER relevant high βN discharges longer than current diffusion time in JT-60U. High normalized beta βN = 2.3 was sustained for 22.3 s with feedback control of plasma stored energy. Lower figure shows sustained βN as a function of duration with ITER inductive and advanced operation regimes.
defined by MPD = fusion power/total reactor weight is one of major figure of merit of economical potential of fusion power station. The MPD will increase with fusion power density pf = Pf /V for a fixed reactor geometry. Therefore, it is important to increase the fusion power density by raising the plasma beta. To achieve high mass power density (MPD > 100 kW/tonnes), the high normalized beta value βN of 3.5–5.5 is typically required although βN is not a single parameter to determine the MPD. To achieve such high βN , wall stabilization or stabilization of resistive wall mode is required and the optimization of stabilizing shell for DEMO is crucially important. 3.2. Optimization of blanket–plasma interaction for wall stabilization of MHD modes Normalized beta at ideal MHD stability limit is plotted as a function of wall position for two plasma aspect ratios (A = R/a = 2.5 and 3.0) as shown in Fig. 5. Critical normalized beta βN is slightly higher for lower aspect ratio for fixed safety factor, elongation and triangularity. As described in Section 2.3, blanket modules must be segmented toroidally to withstand EM force during the current quench. This segmentation reduces the
stabilizing effect for low n ideal MHD modes. And the wall stabilization against low n ideal modes comes from a shell structure behind the blanket modules. A noble shell structure “Saddle Shell” behind the blanket is proposed [6] using the “Saddle Loop” concept [18] to improve coupling of shell to low n global MHD modes. This shell consists of parallel plates and side plates as shown in Fig. 6 that can produce the eddy current pattern quite effectively close to the no-cut case. It should be noted that stabilizing shell position might be greater than 1.3 and smaller than rw /a = 1.4 if we adopt stabilizing shell behind the blanket module with thickness of 0.6 m and some space between first wall and plasma surface. Critical βN is a strong function of normalized wall radius rw /a as shown in Fig. 5. A big difference in critical βN is expected between rw /a = 1.3 and 1.4. Therefore, it is important to optimize the blanket structure to minimize normalized wall radius rw /a. It may be concluded that fat plasma may be favorable for a fixed blanket thickness and plasma volume to have lower rw /a value. On the other hand, too compact reactor leads to higher rw /a due to smaller plasma minor radius a, and optimization of device size should take this (rw /a varies with device size) into account.
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Fig. 5. Reactor core configuration showing plasma, blanket and shell geometry is shown in left. With segmented blanket, blanket does not have a strong shell effect. And the wall stabilization comes from stabilizing shell behind the blanket. Critical βN for A = 2.5 (triangles) and A = 3.0 (squares) with n = 1 and 2 mode as a function of ideal wall location rw /a in right. Expected shell location is between 1.3 and 1.4 and is shown by 2 lines. Negative shear plasma with qmin = 2.4 in the parabolic pressure profile, and the fixed plasma elongation at κ95 = 1.8 are assumed.
It should be noted also that the stability limit depends on current and pressure profiles. It is difficult to control current and pressure profiles (especially pressure profile) to match the optimized ones. Therefore, an operational beta should have a reasonable margin from the beta limit calculated for “optimum profile” to achieve low probability of off-normal events. To evaluate such reactor-relevant high beta steady-state operation in collision-less regime, National Centralized Tokamak (NCT: JT-60 steady-state high beta device) program is planned and ready for implementation [17].
3.3. Optimal use of ferritic wall for fast particle confinement and MHD stability There are number of candidates for blanket structural materials such as reduced activation ferritic steel (RAF), Vanadium alloy and SiC/SiC composite. Among them, RAF is a most likely candidate of blanket structural material for first generation tokamak reactor with its excellent properties against neutron irradiation and its manufacturability. However, magnetization (µ/µ0 > l) provides two blanket–plasma interactions,
Fig. 6. Basic assembly of “Saddle Shell” is shown in the right. Left figure shows eddy current flow pattern and its assembly proposed in [6]. Side plates are quite effective in compensating “cut” effect.
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Fig. 7. Ripple reduction geometry using ferritic steel in JFT-2M. Ferritic plates are placed outside of vacuum vessel of JFT-2M as shown in left. (a–c) Right three figures show IRTV view without plasma, temperature rise measured with IRTV without ferritic plates, and that with ferritic plates. Most significant change between two cases is seen as a reduction of temperature rise in the ripple trapped loss region. But temperature rise by banana drift ion loss is also reduced.
namely effect of error field on plasma performance and effect of attractive force on MHD stability. Leakage field caused by magnetization can be used to compensate toroidal field ripple as demonstrated in JFT-2M experiments (Fig. 7) [19]. Care must be taken so that low n (for example, n = 1) error field due to miss-alignment of blanket is small enough not to trigger locked modes as well as higher harmonic modes produced by toroidal non-uniformity of plasma-facing blanket modules at port section [20]. High beta experiments have been done in JFT-2M using a full coverage ferritic steel in JFT-2M and have successfully demonstrated DEMO-relevant high βN plasma as shown in Fig. 8 [21]. Effect of the ferromagnetic wall on resistive tearing modes was studied in the JFT-2M tokamak and it was concluded that destabilization effect due to the
attraction of the perturbed magnetic field by non-unity permeability is very small compared with the stabilizing effect of the conducting wall [22]. Effect of the ferromagnetic wall on resistive wall modes was also studied numerically for an application to NCT device as a follow-on device of JT-60U [4]. Left hand side of Fig. 9 shows growth rate normalized by a poloidal Alfven transit time as a function of normalized beta βN for two permeability constants µ = µ0 and 2µ0 . Case with µ = 2µ0 corresponds to saturated magnetization. Normalized growth rate increases with βN and sharp increase happens at some βN where transition of RWM to ideal kink mode occurs. The βN value at this transition region decreases by 10% with the factor 2 increase in permeability constant µ. The right hand side of Fig. 9 shows normalized growth rate and real frequency as a function of normalized wall radius
Fig. 8. Plasma equilibrium geometries in JFT-2M to test effect of full coverage of ferritic steel inside the vacuum vessel in the left figure. Achieved normalized beta as a function of various normalized wall radius rw /a before and after full coverage of ferritic steel. Reactor relevant high βN was achieved.
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Fig. 9. Numerical calculation of normalized growth rate as a function of normalized beta βN (left figure) and normalized growth rate and real frequency showing RWM stability window with plasma rotation with and w/o ferromagnetism (right figure).
rw /a for resistive wall modes under the plasma rotation (V = 0.5Vpa ) [23]. The RWM is unstable (γ > 0) when wall is close to the plasma (smaller rw /a) and becomes stable from some rw /a value until the wall location of ideal kink unstable region. This basic characteristics does not change by a change in the permeability constant µ although real frequency and growth rate are modified by the ferromagnetism as shown in the right hand side of Fig. 9 [4]. In summary, plasma compatibility test of RAF was successfully done in JFT-2M and theoretical studies gives positive prospects for the usage of RAF as a first wall material if careful design were made to minimize the error field.
4. Neutral and ion flux control
tion near the 1st wall. And the control of such 2nd SOL is important for future steady-state tokamak reactors. 4.2. Energy and flux distribution of neutrals under saturated wall in JT-60U Recent long pulse experiments in JT-60U showed a sign of wall saturation that will be common to steady-state tokamak reactors. Evaluation of energy spectrum of neutral flux to the 1st wall is important to evaluate possible 1st wall damage due to energetic neutral. Fig. 11 shows estimated average energy and neutral flux to the 1st wall in JT-60U under the saturated wall condition (E44020) using DEGAS [25]. Average energy of neutrals at the 1st wall is 100 eV and flux of a few ×l019 particles/s/m2 . Fig. 11 also shows the distribution of particle fluxes for energy range of 1–1000 eV. It was found that there
In this section, we address behavior of energetic neutrals/ions incident to first wall. Influx of energetic neutrals/ions to first wall should be sufficiently low to keep sputtering erosion low enough. 4.1. Extension of SOL plasma by 2nd SOL Decay of SOL plasma towards the 1st wall is important to assure the reliability of 1st wall for long pulse operation in a fusion reactor. Second SOL with much longer decay length than 1st SOL was observed in JT60U divertor [24]. Fig. 10 shows typical example of 2nd SOL observed in JT-60U for two densities. Stronger 2nd SOL appears for higher density case. The mechanism for the 2nd SOL is not yet identified but the effective perpendicular particle diffusivity seems to be enhanced. The existence of 2nd SOL would increase the plasma density and hence plasma–blanket interac-
Fig. 10. Electron density as a function of distance from separatrix for two average bulk plasma densities. Vertical dashed line shows location of separatrix. Change of the decay length of density is clearly seen and two SOLs (first and second) are seen.
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Fig. 11. DEGAS2 mesh geometry (left figure), average energy and D neutral flux to wall as a function of poloidal distance L (middle figure), energy distribution of neutral flux to the wall (right figure) calculated with DEGAS2 code for long pulse saturated wall experiment in JT-60U.
is non-negligible population of high-energy (a few 100 eV–1 keV) neutrals. Existence of such energetic neutrals is of concern for future long pulse operation of fusion reactor and left for future studies.
in this field. We also thank to Dr. M. Nagami for valuable comments.
References 5. Summary Progresses of understanding of blanket–plasma interaction from JT-60U, JFT-2M and reactor studies are summarized. Philosophical discussion of disruption was made to treat disruption as off-normal event. And basic design guideline to compromise disruption tolerance and requirement for wall stabilization is described based on status of development of disruption control techniques. It is shown that there seems to be a solution to withstand disruption force for a blanket while keeping the wall stabilization against the kink-ballooning modes with somewhat distant wall location. Compatibility of ferritic first wall to high performance plasma is demonstrated in JFT-2M and also theoretical works show that this ferromagnetism can be manageable in tokamak fusion reactor. Importance of assessment of 2nd SOL and warm neutrals around the main plasma is pointed out for a long-lived first wall.
Acknowledgements Authors appreciate Dr. M. Seki and Dr. H. Ninomiya for giving us an opportunity to summarize our activity
[1] S. Ide and the JT-60 Team, Overview of JT-60U progress towards steady-state advanced tokamak, 20th IAEA Fusion Energy Conference, paper IAEA-CN-116/OV1-1. [2] K. Tsuzuki, K. Kamiya, K. Shinohara, M. Bakhtiari, H. Ogawa, G. Kurita, et al., Compatibility of reduced activation ferritic steel wall with high performance plasma on JFT-2M, 20th IAEA fusion energy conference, paper IAE A-CN-116/EX/P2-33. [3] ITER physics basis chapter 3: MHD stability, operational limits and disruptions, Nucl. Fusion, 39 (1999) 2251–2389. [4] G. Kurita, J. Bialek, T. Tuda, M. Azumi, S. Ishida, G.A. Navratil, et al., 20th IAEA fusion energy conference, paper IAEA-CN116/FTYP7-7. [5] K. Tobita, S. Nishio, M. Enoeda, M. Sato, T. Isono, S. Sakurai, et al., Fusion DEMO plant design at JAERI, these proceedings. [6] S. Nishio, J. Ohmori, T. Kuroda, K. Tobita, M. Enoeda, D. Tsuru, et al., Consideration on blanket structure for fusion DEMO plant at JAERI, these proceedings. [7] The subcommittee of the fusion council for fusion development strategy, Report on the technical feasibility of fusion energy and extension of the fusion program and basic supporting researches, May 17, 2000 (in Japanese; English translation is available at http://www.naka.jaeri.go.jp/SCI/). [8] Y. Nakamura, R. Yoshino, Y. Neyatani, T. Tsunematsu, M. Azumi, N. Pomphrey, et al., Mechanism of vertical displacement events in JT-60U disruptive discharges, Nucl. Fusion 36 (1996) 643–656. [9] R. Yoshino, Y. Nakamura, Y. Neyatani, et al., Avoidance of VDEs during plasma current quench in JT-60U, Nucl. Fusion 36 (1996) 295–307. [10] R. Yoshino, Y. Neyatani, N. Isei, Y. Koide, Y. Kawano, A. Tanga, et al., Plasma physics and controlled nuclear fusion research
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Blanket–plasma interaction in tokamaks Implication from JT-60U, JFT-2M and reactor studies M. Kikuchi a,∗ , S. Nishio a , G. Kurita a , K. Tsuzuki a , M. Bakhtiari b , H. Kawashima a , H. Takenaga a , Y. Kusama a , K. Tobita a a b
Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Muko-yama 801-1, Naka, Ibaraki, Japan University of Wisconsin-Madison, 735 Engineering Research Bldg., 1500 Engineering Physics, Madison, WI 53706, USA Received 7 March 2005; received in revised form 5 September 2005; accepted 5 September 2005 Available online 10 January 2006
Abstract Optimization of blanket–plasma interaction is one of important subjects of tokamak reactor design. This paper summarizes key physics R&D results of blanket–plasma interactions in tokamak system based on JT-60U, JFT-2M and DEMO studies at JAERI. Conflicting requirement between wall stabilization and disruption tolerance with respect to the electromagnetic coupling between plasma-structure seems to have a solution with somewhat distant wall location while minimization of the distance between plasma and stabilizing wall is critically important for high beta operation. Effect of ferromagnetism on plasma stability is well studied in JFT-2M and theoretically giving positive prospect for use of reduced activation ferritic steel as a first wall material. Enhancement of ion/neutral 1st wall interaction with 2nd SOL and saturated wall is discussed as an important R&D element for DEMO and existence of high-energy neutrals in JT-60U was mentioned. © 2005 Elsevier B.V. All rights reserved. Keywords: Blanket; Plasma; Disruption; Wall stabilization; Ferromagnetism; Wall saturation
1. Introduction Optimization of blanket–plasma interaction is one of major design issues of economically viable fusion DEMO. Tokamak without wall stabilization of ideal MHD modes has a beta limit, so-called Troyon limit βt = βN Ip /(ap Bt ) where βt , βN , Ip , ap , Bt are volume average toroidal beta, normalized beta value (βN ∼ 4li : ∗
Corresponding author. E-mail address: [email protected] (M. Kikuchi).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.09.048
li is internal inductance), plasma current, plasma minor radius, toroidal magnetic field, respectively. Wall stabilization of ideal MHD modes through strong coupling between plasma and stabilizing shell provides an important opportunity to increase attainable beta in a tokamak while strong plasma–blanket coupling may lead to difficulty for disruption tolerance of blanket. Therefore, it is important to clarify the prospect of disruption mitigation and to find a solution consistent with wall stabilization for DEMO and commercial reactors. Also, adoption of reduced activation ferritic
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2. Disruption as a controllable plasma–blanket interaction
assume “no disruption” since any magnetic configuration has an operational limit and there is non-zero probability to reach its operational limit for example by a flake of surrounding material into the plasma and to induce radiative collapse. Any man-made system has possibility of troubles and it is fundamentally difficult to say “no trouble for this system”. Thus, the key point for this question on “disruption” is not to eliminate “disruption” but to reduce the frequency of disruption less than the frequency of off-normal event acceptable for commercial use. In order to categorize “disruption” as “trouble”, machine must withstand electromagnetic (EM) force and thermal loads by rare disruptions. Acceptable frequency of off-normal event could be discussed from past experiences of power plants and 1 event/2 year is proposed as a typical number in reference [7].
2.1. Disruption as an off-normal event in DEMO and commercial reactor
2.2. Development of disruption mitigation techniques
Firstly, we would like to consider the condition of disruption in DEMO and future commercial reactors. DEMO as a final step before the commercialization of fusion power must demonstrate reliable operation up to a year at least towards the end of its operational life. It is a long-standing question whether “disruption-free” can be realized to improve operational reliability and to relax the design requirement of blanket system against disruption force and thermal loads. It is unrealistic to
2.2.1. Neutral point for VDE, Killer Pellet for thermal quench and runaway electron suppression Disruption phenomena is illustrated in Fig. 1 showing four important processes, namely thermal quench, current quench, runaway electrons and VDE. Possibility of “thermal quench” exists for all magnetic confinement systems by radiative or MHD collapses while tokamak including ST has additional effects due to EM force induced by current quench.
steel (RAF) as a primary candidate structural material for first wall brings concerns on the effect of ferromagnetism for plasma performance. Avoidance of energetic neutrals/ions impinging to first wall is also a critical issue to assure the long life of blanket. Tokamak devices at JAERI such as JT-60U [1] and JFT-2M [2] addresses above physics issues to optimize blanket–plasma interaction such as disruption mitigation [3] and wall stabilization for high beta steady-state operations [4], and wall-plasma/neutral interactions. Reactor system studies at JAERI have progressed remarkably in recent years to have many implications to a direction of fusion plasma researches [5,6].
Fig. 1. Schematic figure of physical processes of disruption is shown. Thermal quench is an abrupt loss of thermal energy and is followed by plasma current termination. Under some conditions, fraction of plasma current is kept due to the formation of runaway electron current tail. During these processes, plasma may have a vertical movement, namely “Vertical Displacement Event (VDE)” and this movement induces poloidal current named “Halo Current” to induce strong electromagnetic force on structural components contacting VDE plasma.
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Fig. 2. Waveform of plasma current, electron temperature during forced plasma shut down induced by Ar + H2 , H2 and Neon pellet injection. Demonstration of plasma shut down by Ar + H2 injection as compared with “Killer Pellet” for current termination. Intensity of hard X-ray (measure of runaway electron population) and the heat flux to the divertor plate are plotted for various combinations of gas injection showing that runaway population and heat flux are minimum for Kr + H2 injection [14].
The vertical displacement event (VDE) is known to induce halo current that would damage the in-vessel components. Control technique for VDE is developed in JT-60U by locating the center of the plasma current at so-called “neutral point” as predicted by numerical calculation [8] and JT-60U experiments [9]. The neutral point is defined as a location where attractive forces due to eddy currents induced by dIp /dt become zero in both radial and vertical direction. The major passive element of the in-vessel components of tokamak fusion reactor is the stabilizing shell behind the breeding blanket and it should be designed so that center of plasma current is located at the neutral point. Thermal quench may lead a large convection/ conduction heat flux preferentially to divertor plates. Key point to soften this phenomenon is to reduce heat flux to the divertor plates through radiation power deposition to the first wall before the thermal quench, for example, by impurity pellet injection. This concept was demonstrated in JT-60U [10] and was called “Killer Pellet”. Thermal quench and associated increase of plasma resistance may lead to a production of runaway electrons. Various methods to suppress runaway electrons such as excitation of MHD [11], and lowering q surface to less than 2 [12] were successfully demonstrated. A collision-less loss mechanism due to magnetic perturbation was analyzed and found to be effective to suppress runaway electrons [13].
Recent works on disruption mitigation in JT-60U focused on the prediction of disruption through neural network [14] and also development of mitigation technique for thermal quench and runaway electron [15]. Although “Killer Pellet” is an important tool for mitigation of thermal load to the divertor, it might excite runaway electrons since it does not provide enough electron density and its drag force. Injection of mixture of inert gas and hydrogen provides another opportunity for a suppression of thermal load to the divertor plates and generation of runaway electron. In particular, Kr + H2 gas injection was quite effective in terminating the discharges before thermal quench without runaway electron generation and with low divertor heat load as shown in Fig. 2 [15]. If we adopt this kind of mitigation technique for tokamak fusion reactor, the first wall structure must be strong enough against thermal shock from this radiation. Despite such successful development of disruption mitigation techniques, it is unlikely to increase the current quench time much longer than the expected values of ITER scaling (dIp /dt ∼ S ∼ 27 ms for ITER: S is cross-sectional area of plasma). 2.3. Blanket design withstanding EM force of current quench Most important force on blanket structure due to current quench is an EM force by an induced current as shown in Fig. 3. This force produces a
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Fig. 3. Two types of blanket structures to withstand EM forces in tokamak reactor. Left hand side shows blanket concept of SSTR [16] where FGM insulator is used to eliminate loop current around the blanket box. Right hand side shows recent blanket concept [6] allowing current flowing in the blanket box where induced eddy current loop is shown by “→” and EM force on blanket box is shown by “ ”.
torque to twist the blanket box. The torque M is roughly proportional to Ip ·Bt ·d·t where Ip , Bt , d and t are plasma current, toroidal field at the blanket, toroidal width of blanket, radial thickness of blanket, respectively. There are two ways of blanket design withstanding this EM force. First one is to make a cut for this current loop by a FGM (Functionally Graduated Material: insulator) at the back of blanket box placed vertically adopted for SSTR design [16] as shown in Fig. 3 (left). This design allows easier structural design with keys and separation of tritium breeding blanket to fixed blanket and replaceable blanket (double-layer blanket to reduce radioactive-waste). The feasibility of FGM as a part of blanket structure becomes an important R&D issue for this design. Another approach is to adopt single layer blankets without FGM that are fixed to back plate firmly by using welding and thick keys [6]. It is reported that blanket design withstanding torque of 2 MNm would be possible expected for a current quench time of 30 ms close to a value expected in ITER. For a low aspect ratio DEMO concept [5], difference of toroidal field at low and high field sides allows three times longer toroidal width d for low field side blankets which might be sufficient to achieve overall TBR > 1 since low field side blankets share major part of TBR.
3. Plasma performance enhancement through optimization of blanket–plasma interaction 3.1. High normalized beta plasma sustainment in JT-60U long pulse experiments and needs for high beta steady-state operation above the no-wall limit ITER is an important step towards the technical feasibility of fusion energy. In tokamak confinement geometry, beta limit under no-wall condition is governed by so-called Troyon scaling (βt = βN Ip /aBt : βN ∼ 2.8). Technical objectives of ITER could be achieved at the plasma beta below the Troyon limit. And the long sustainment of ITER relevant high normalized beta was successfully demonstrated in JT-60U by extending the machine capability from 15 to 65 s (high heating power is limited to 30 s). Fig. 4 shows recent typical waveform of long sustainment of high βN discharge much longer than the current diffusion time and the progress of JT-60U experiments in (βN , duration) plane with target regimes of ITER inductive and steady-state operation. Long sustainment of high beta close to ITER advanced operation was successfully demonstrated. However, economically viable fusion reactor needs steady-state operation at higher βN above the ITER advanced operation. The mass power density (MPD)
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Fig. 4. Long sustainment of ITER relevant high βN discharges longer than current diffusion time in JT-60U. High normalized beta βN = 2.3 was sustained for 22.3 s with feedback control of plasma stored energy. Lower figure shows sustained βN as a function of duration with ITER inductive and advanced operation regimes.
defined by MPD = fusion power/total reactor weight is one of major figure of merit of economical potential of fusion power station. The MPD will increase with fusion power density pf = Pf /V for a fixed reactor geometry. Therefore, it is important to increase the fusion power density by raising the plasma beta. To achieve high mass power density (MPD > 100 kW/tonnes), the high normalized beta value βN of 3.5–5.5 is typically required although βN is not a single parameter to determine the MPD. To achieve such high βN , wall stabilization or stabilization of resistive wall mode is required and the optimization of stabilizing shell for DEMO is crucially important. 3.2. Optimization of blanket–plasma interaction for wall stabilization of MHD modes Normalized beta at ideal MHD stability limit is plotted as a function of wall position for two plasma aspect ratios (A = R/a = 2.5 and 3.0) as shown in Fig. 5. Critical normalized beta βN is slightly higher for lower aspect ratio for fixed safety factor, elongation and triangularity. As described in Section 2.3, blanket modules must be segmented toroidally to withstand EM force during the current quench. This segmentation reduces the
stabilizing effect for low n ideal MHD modes. And the wall stabilization against low n ideal modes comes from a shell structure behind the blanket modules. A noble shell structure “Saddle Shell” behind the blanket is proposed [6] using the “Saddle Loop” concept [18] to improve coupling of shell to low n global MHD modes. This shell consists of parallel plates and side plates as shown in Fig. 6 that can produce the eddy current pattern quite effectively close to the no-cut case. It should be noted that stabilizing shell position might be greater than 1.3 and smaller than rw /a = 1.4 if we adopt stabilizing shell behind the blanket module with thickness of 0.6 m and some space between first wall and plasma surface. Critical βN is a strong function of normalized wall radius rw /a as shown in Fig. 5. A big difference in critical βN is expected between rw /a = 1.3 and 1.4. Therefore, it is important to optimize the blanket structure to minimize normalized wall radius rw /a. It may be concluded that fat plasma may be favorable for a fixed blanket thickness and plasma volume to have lower rw /a value. On the other hand, too compact reactor leads to higher rw /a due to smaller plasma minor radius a, and optimization of device size should take this (rw /a varies with device size) into account.
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Fig. 5. Reactor core configuration showing plasma, blanket and shell geometry is shown in left. With segmented blanket, blanket does not have a strong shell effect. And the wall stabilization comes from stabilizing shell behind the blanket. Critical βN for A = 2.5 (triangles) and A = 3.0 (squares) with n = 1 and 2 mode as a function of ideal wall location rw /a in right. Expected shell location is between 1.3 and 1.4 and is shown by 2 lines. Negative shear plasma with qmin = 2.4 in the parabolic pressure profile, and the fixed plasma elongation at κ95 = 1.8 are assumed.
It should be noted also that the stability limit depends on current and pressure profiles. It is difficult to control current and pressure profiles (especially pressure profile) to match the optimized ones. Therefore, an operational beta should have a reasonable margin from the beta limit calculated for “optimum profile” to achieve low probability of off-normal events. To evaluate such reactor-relevant high beta steady-state operation in collision-less regime, National Centralized Tokamak (NCT: JT-60 steady-state high beta device) program is planned and ready for implementation [17].
3.3. Optimal use of ferritic wall for fast particle confinement and MHD stability There are number of candidates for blanket structural materials such as reduced activation ferritic steel (RAF), Vanadium alloy and SiC/SiC composite. Among them, RAF is a most likely candidate of blanket structural material for first generation tokamak reactor with its excellent properties against neutron irradiation and its manufacturability. However, magnetization (µ/µ0 > l) provides two blanket–plasma interactions,
Fig. 6. Basic assembly of “Saddle Shell” is shown in the right. Left figure shows eddy current flow pattern and its assembly proposed in [6]. Side plates are quite effective in compensating “cut” effect.
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Fig. 7. Ripple reduction geometry using ferritic steel in JFT-2M. Ferritic plates are placed outside of vacuum vessel of JFT-2M as shown in left. (a–c) Right three figures show IRTV view without plasma, temperature rise measured with IRTV without ferritic plates, and that with ferritic plates. Most significant change between two cases is seen as a reduction of temperature rise in the ripple trapped loss region. But temperature rise by banana drift ion loss is also reduced.
namely effect of error field on plasma performance and effect of attractive force on MHD stability. Leakage field caused by magnetization can be used to compensate toroidal field ripple as demonstrated in JFT-2M experiments (Fig. 7) [19]. Care must be taken so that low n (for example, n = 1) error field due to miss-alignment of blanket is small enough not to trigger locked modes as well as higher harmonic modes produced by toroidal non-uniformity of plasma-facing blanket modules at port section [20]. High beta experiments have been done in JFT-2M using a full coverage ferritic steel in JFT-2M and have successfully demonstrated DEMO-relevant high βN plasma as shown in Fig. 8 [21]. Effect of the ferromagnetic wall on resistive tearing modes was studied in the JFT-2M tokamak and it was concluded that destabilization effect due to the
attraction of the perturbed magnetic field by non-unity permeability is very small compared with the stabilizing effect of the conducting wall [22]. Effect of the ferromagnetic wall on resistive wall modes was also studied numerically for an application to NCT device as a follow-on device of JT-60U [4]. Left hand side of Fig. 9 shows growth rate normalized by a poloidal Alfven transit time as a function of normalized beta βN for two permeability constants µ = µ0 and 2µ0 . Case with µ = 2µ0 corresponds to saturated magnetization. Normalized growth rate increases with βN and sharp increase happens at some βN where transition of RWM to ideal kink mode occurs. The βN value at this transition region decreases by 10% with the factor 2 increase in permeability constant µ. The right hand side of Fig. 9 shows normalized growth rate and real frequency as a function of normalized wall radius
Fig. 8. Plasma equilibrium geometries in JFT-2M to test effect of full coverage of ferritic steel inside the vacuum vessel in the left figure. Achieved normalized beta as a function of various normalized wall radius rw /a before and after full coverage of ferritic steel. Reactor relevant high βN was achieved.
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Fig. 9. Numerical calculation of normalized growth rate as a function of normalized beta βN (left figure) and normalized growth rate and real frequency showing RWM stability window with plasma rotation with and w/o ferromagnetism (right figure).
rw /a for resistive wall modes under the plasma rotation (V = 0.5Vpa ) [23]. The RWM is unstable (γ > 0) when wall is close to the plasma (smaller rw /a) and becomes stable from some rw /a value until the wall location of ideal kink unstable region. This basic characteristics does not change by a change in the permeability constant µ although real frequency and growth rate are modified by the ferromagnetism as shown in the right hand side of Fig. 9 [4]. In summary, plasma compatibility test of RAF was successfully done in JFT-2M and theoretical studies gives positive prospects for the usage of RAF as a first wall material if careful design were made to minimize the error field.
4. Neutral and ion flux control
tion near the 1st wall. And the control of such 2nd SOL is important for future steady-state tokamak reactors. 4.2. Energy and flux distribution of neutrals under saturated wall in JT-60U Recent long pulse experiments in JT-60U showed a sign of wall saturation that will be common to steady-state tokamak reactors. Evaluation of energy spectrum of neutral flux to the 1st wall is important to evaluate possible 1st wall damage due to energetic neutral. Fig. 11 shows estimated average energy and neutral flux to the 1st wall in JT-60U under the saturated wall condition (E44020) using DEGAS [25]. Average energy of neutrals at the 1st wall is 100 eV and flux of a few ×l019 particles/s/m2 . Fig. 11 also shows the distribution of particle fluxes for energy range of 1–1000 eV. It was found that there
In this section, we address behavior of energetic neutrals/ions incident to first wall. Influx of energetic neutrals/ions to first wall should be sufficiently low to keep sputtering erosion low enough. 4.1. Extension of SOL plasma by 2nd SOL Decay of SOL plasma towards the 1st wall is important to assure the reliability of 1st wall for long pulse operation in a fusion reactor. Second SOL with much longer decay length than 1st SOL was observed in JT60U divertor [24]. Fig. 10 shows typical example of 2nd SOL observed in JT-60U for two densities. Stronger 2nd SOL appears for higher density case. The mechanism for the 2nd SOL is not yet identified but the effective perpendicular particle diffusivity seems to be enhanced. The existence of 2nd SOL would increase the plasma density and hence plasma–blanket interac-
Fig. 10. Electron density as a function of distance from separatrix for two average bulk plasma densities. Vertical dashed line shows location of separatrix. Change of the decay length of density is clearly seen and two SOLs (first and second) are seen.
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Fig. 11. DEGAS2 mesh geometry (left figure), average energy and D neutral flux to wall as a function of poloidal distance L (middle figure), energy distribution of neutral flux to the wall (right figure) calculated with DEGAS2 code for long pulse saturated wall experiment in JT-60U.
is non-negligible population of high-energy (a few 100 eV–1 keV) neutrals. Existence of such energetic neutrals is of concern for future long pulse operation of fusion reactor and left for future studies.
in this field. We also thank to Dr. M. Nagami for valuable comments.
References 5. Summary Progresses of understanding of blanket–plasma interaction from JT-60U, JFT-2M and reactor studies are summarized. Philosophical discussion of disruption was made to treat disruption as off-normal event. And basic design guideline to compromise disruption tolerance and requirement for wall stabilization is described based on status of development of disruption control techniques. It is shown that there seems to be a solution to withstand disruption force for a blanket while keeping the wall stabilization against the kink-ballooning modes with somewhat distant wall location. Compatibility of ferritic first wall to high performance plasma is demonstrated in JFT-2M and also theoretical works show that this ferromagnetism can be manageable in tokamak fusion reactor. Importance of assessment of 2nd SOL and warm neutrals around the main plasma is pointed out for a long-lived first wall.
Acknowledgements Authors appreciate Dr. M. Seki and Dr. H. Ninomiya for giving us an opportunity to summarize our activity
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