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Physics design study of the divertor power handling in 8 m class DEMO reactor
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Fusion Engineering and Design xxx (2017) xxx–xxx
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Physics design study of the divertor power handling in 8 m class DEMO reactor Kazuo Hoshino a,∗ , Nobuyuki Asakura b , Shinsuke Tokunaga b , Yuki Homma b , Katsuhiro Shimizu a , Yoshiteru Sakamoto b , Kenji Tobita b , Joint Special Design Team for Fusion DEMO a b
National Institutes for Quantum and Radiological Science and Technology, Naka Fusion Institute, Ibaraki, Japan National Institutes for Quantum and Radiological Science and Technology, Rokkasho Fusion Institute, Aomori, Japan
h i g h l i g h t s • • • •
The power handling for DEMO with 1.5 GW fusion power was studied. The detached divertor plasma was obtained by large impurity radiation. The SONIC simulation showed the target heat load less than 8 MW/m2 . Dependence of the impurity concentration on the fuel gas puff was also studied.
a r t i c l e
i n f o
Article history: Received 3 October 2016 Received in revised form 13 March 2017 Accepted 14 March 2017 Available online xxx Keywords: DEMO reactor Divertor plasma design Power handling Radiative cooling SONIC
a b s t r a c t The divertor plasma performance and the power handling are studied for 8 m class DEMO reactor with the fusion power of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target are obtained for a relatively low electron density of 1.8 × 1019 m−3 at the outer mid-plane separatrix. The SONIC simulation shows the target heat load less than 8 MW/m2 , which can be handled by the ITER-like divertor target, for both target. However, at the outer target, the ion temperature is still high which may cause the target erosion. For the divertor power handling and suppression of the target erosion, the divertor design study have to be further proceeded as well as the core plasma design. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff is also studied. With increasing fuel gas puff rate, the mid-plane separatrix density increases and the Ar impurity concentration at the outer mid-plane decreases to 0.5%. © 2017 Elsevier B.V. All rights reserved.
1. Introduction Handling of the huge power exhausted from the core region to the SOL/divertor region is one of the crucial issues for a DEMO reactor design. The power handling scenario has been investigated for a 3 GW class fusion reaction with an ITER-size plasma, SlimCS [1]. The numerical simulation showed that the target heat load qtarget was 16 MW/m2 even in the case where more than 90% of the exhausted power from the core plasma (Pout ) was radiated by the argon (Ar) impurity gas seeding. In order to investigate design window of the divertor, dependence of qtarget on the fusion power and the impurity
∗ Corresponding author. E-mail address: [email protected] (K. Hoshino).
radiation has been studied [2]. In the simulation, qtarget < 10 MW/m2 was found for the fusion power of ∼1.5 GW and the large impurity radiation power (>0.8Pout ) by the Ar gas seeding. Recently, JA-Model 2014 DEMO concept has been proposed [3]. From the previous concept, SlimCS, a major radius was increased to 8.5 m in order to install a center solenoid coil with enough size for full inductive current ramp up, and a fusion power was decreased to ∼1.5 GW for the divertor and blanket design. These changes are preferable for the divertor power handling. However, the operational density became low (n¯e ∼6.6 × 1019 m−3 , where n¯e is the line-averaged density.) compared with SlimCS (n¯e ∼1.2 × 1020 m−3 ), due to the low plasma current, the high aspect ratio, and the resultant low Greenwald density. Generally, high operational density (high SOL density nsep ) is preferable for formation of the divertor plasma detachment and the power handling.
http://dx.doi.org/10.1016/j.fusengdes.2017.03.068 0920-3796/© 2017 Elsevier B.V. All rights reserved.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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In this paper, the divertor plasma performance and the power handling scenario for the 8 m class DEMO with low operational density is investigated by using a suite of integrated divertor codes, SONIC. 2. Numerical model A suite of integrated divertor codes SONIC [4,5] is applied to the physics design study of the DEMO divertor plasma. The fuel plasma transport in the plasma edge (0.95 < r/a < 1), SOL and divertor regions is described by a fluid model. The dynamics of fuel neutral and impurity is solved by a Monte-Carlo particle model in whole region including the sub-divertor region. Although the drift and current model are not included in the present version of the SONIC code, the code can qualitatively reproduce the detached divertor plasma [6]. In this analysis, only deuterium species are considered as fuel species for simplification. The divertor geometry and the numerical grid are shown in Fig. 1. In order to enhance the recycling and formation of detachment, reflector is installed at the inner divertor in addition to the outer divertor. The pumping slot is located at the bottom of the divertor cassette with the pumping speed of 65 m3 /s. The numerical grid for a plasma transport covers the flux surfaces of r < 3.2 cm at the mid-plane. Although the armor material of the divertor target and the blanket is tungsten, its sputtering is not considered in the present analysis. The recycling coefficient of unity is assumed due to the steady-state operation. At the core-edge boundary (r/a = 0.95), exhaust power Pout and particle flux out are given. For a 1.5 GW fusion power DEMO, the alpha heating, the additional heating and the radiation in the core are assumed to 300 MW, 80 MW and 130 MW, respectively, in the analysis. Therefore, Pout is set to 250 MW. From the preliminary core plasma transport analysis [7], out = 1 ×1022 s−1 is assumed. Fuel gas to enhance the divertor recycling is injected from the outer mid-plane with 1.2 × 1022 D/s, which corresponds to D2 gas puff of 25 Pa m3 /s. Diffusion coefficients for thermal and particle are assumed to be i = e = 1 m2 s−1 and D = 0.3 m2 s−1 , respectively, which are the same in the ITER divertor plasma simulation [8]. In order to enhance the impurity radiation in the plasma edge, SOL and divertor regions, Ar impurity gas is puffed into the outer divertor. The Ar seeding rate is controlled by feedback to achieve an assigned frad =
edge
edge,SOL,div
SOL + P div /P Prad + Prad out , where Prad rad
Fig. 2. The radial profile of ne , Ti and Te at the outer mid-plane.
3. Divertor plasma performance and heat load on the wall The radial profiles of the electron density ne , the ion temperature Ti and the electron temperature Te at the outer mid-plane are plotted in Fig. 2. The separatrix density is nsep ∼ 1.8 × 1019 m−3 , which
is
the impurity radiation power in the edge (0.95 < r/a < 1), SOL and divertor region, respectively. Based on the previous study [2], frad = 0.8 is given. For reduction of the impurity Monte-Carlo calculation time, the impurity back flow from the sub-divertor region to the divertor region has been modeled [9].
Fig. 1. The divertor geometry and the numerical grid.
Fig. 3. Spatial distribution of the Ar impurity radiation power Prad (a) and Te (b).
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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Fig. 4. Radial profiles of ne , Ti and Te (a, b) and qtarget (c, d). Left and right figures correspond to the inner and the outer target, respectively. qtarget is evaluated by taking in account the surface recombination of the ion particle, the impurity radiation and the neutral flux in addition to the plasma heat transport.
corresponds to 0.27 n¯e . The plasma temperature of Ti ∼ 1010 eV and Te ∼ 480 eV is relatively high due to high exhausted power and low nsep . Large frad of 0.8 (200 MW) is achieved by the Ar gas puff rate of o−div = 1.47 Pa m3 /s. Distribution of the Ar impurity radiation is Prad SOL+edge
i-div = 82 MW and P 80 MW, Prad = 37 MW. The region of the rad large impurity radiation power can be seen in the inner and outer divertor region, as show in Fig. 3(a). Fig. 3(b) shows the spatial distribution of Te . Due to the large impurity radiation and large volume, low Te region (<2 eV) can be seen in the wide region of the inner divertor and around the outer strike point, i.e., the almost complete detachment in the inner divertor and the partial detachment in the outer divertor are obtained even in the low nsep condition. Radial profiles of ne , Ti , Te and qtarget at the divertor target are plotted in Fig. 4. At the inner target, Te,i < 2 eV, i.e., the detachment can be seen on the almost whole target. The detachment region can been seen around the strike point (≤10 cm) at the outer target, while, Ti is ∼250 eV at the outer flux region due to the low nsep and the resultant low recycling. Since such high Ti can lead significant erosion of the target, expansion of the detached region and reduction of Ti are indispensable. Peak of qtarget is ∼7 MW/m2 on the inner target and ∼8 MW/m2 on the outer target, which can be handled by the ITER divertor technology, i.e., the tungsten mono-block target and the water cooling with the Cu-alloy cooling tube. The contribution to the target heat load is almost plasma heat load at the outer target due to the partial detachment. On the other hand, at the inner target, the plasma heat load decreases but the surface recombination increases regardless of the full detachment. Reproduction of the detachment, especially significant decrease in the particle flux is challenging issue even
for state-of-the-art divertor codes [10]. Therefore, heat load due to the plasma transport and the surface recombination (the ion particle flux) is probably overestimated. Further study and modeling of detachment are necessary for precise analysis. The heat load on the first wall (the blanket surface) is also evaluated in Fig. 5. The impurity radiation load includes contribution only from the calculation domain of SONIC, i.e., r/a > 0.95. The radiation power (including the impurity line radiation, cyclotron emission and bremsstrahlung radiation) from the core region is ∼130 MW (see Section 2). The heat load due to the core radiation is estimated
Fig. 5. The heat load along the blanket surface. The poloidal angle of 0◦ corresponds to the outer mid-plane.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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Fig. 6. Variation of nsep and nz /ne on fuel gas puff rate.
to <100 kW/m2 in average but this is not included in Fig. 5. Main component of the heat load on the blanket is the impurity radiation and the neutral particles. Total heat load on the blanket is less than 150 kW/m2 , which is compatible with allowable heat load of blanket design (500–600 kW/m2 [12]). Here, it is noted that heat load of the plasma transport and the surface recombination are evaluated at the grid boundary for the plasma transport (see Section 2). The plasma heat and particle are transported along the flux tube outside of the numerical grid, and concentrate to local area. Therefore, their heat load can become more peak as analyzed in Ref. [11]. 4. Control of the SOL density and the impurity concentration In the view point of consistency between the core plasma design and the divertor power handling, nsep and the impurity concentration nZ /ne are important parameters. Fig. 6 shows nsep and nZ /ne on the different fuel gas puff rate from 1.2 × 1022 D/s to 4.8 × 1022 D/s. The mid-plane separatrix density, nsep , is changed from 1.8 × 1019 m−3 to 2.3 × 1019 m−3 , which correspond to 0.27–0.35 of n¯e = 6.6 × 1019 m−3 . The impurity concentration nZ /ne can be also controlled by the fuel gas puff. Higher gas puff rate enhances the divertor impurity shielding effect by the friction force and dilution of the impurity concentration by higher nsep . Change of the fuel gas puff from 1.2 × 1022 D/s to 4.8 × 1022 D/s decreases nZ /ne from 1.1% to 0.5%. These parameters may be acceptable for a core plasma design but an integrated analysis between core and SOL/divertor plasma is necessary for evaluation of consistency. 5. Summary The divertor plasma performance and the power handling has been studied for 8 m class DEMO reactor with the fusion power
of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target were obtained even for a relatively low nsep ∼ 1.8 × 1019 m−3 . The SONIC simulation showed the target heat load less than 8 MW/m2 , which can be handled by the ITER-like divertor target. However, at the outer target, reduction of the ion temperature is still necessary for suppression of the target erosion. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff rate has been also studied. With increasing fuel gas puff rate to 4.8 × 1022 D/s, nsep increased to 2.3 × 1019 m−3 and the Ar impurity concentration (nz /ne ) decreased to 0.5 %. Consistency of these parameters with the core plasma design will be investigated by using an integrated analysis between the core and SOL/divertor plasma. To suppress the target erosion, i.e., to control the temperature in far SOL at divertor targets, further optimization of divertor geometry and the position of pump slot, and the exploration of more advanced divertor configurations are necessary in future. In addition, development of high density and high radiative core plasma scenario has to be proceeded. One of the key parameter is 95 [13]. If 95 can be increased from 1.65 (JA model 2014) to 1.75, n¯e can be increased from 6.6 × 1019 m−3 to 7.2 × 1019 m−3 and increased fusion power can compensate impurity radiation loss. Acknowledgements SONIC simulations were carried out using the HELIOS supercomputer system at International Fusion Energy Research Centre (IFERC), Aomori, Japan, under the Broader Approach (BA) collaboration between Euratom and Japan, implemented by Fusion for Energy and Japan Atomic Energy Agency. This work is mainly carried out within the framework of the DEMO Design Activity under the BA-IFERC project, and partially supported by Grant-in-Aid for Young Scientists (B) and Grant-in-Aid for Scientific Research (B) of Japan Society for the Promotion of Science. References [1] N. Asakura, et al., Nucl. Fusion 53 (2013) 123013. [2] K. Hoshino, et al., In: 25th IAEA Fusion Energy Conference FIP/P 8-11, in: St. Petersburg, Russian Federation, October 2014. [3] Y. Sakamoto, et al., In: 25th IAEA Fusion Energy Conference FIP/3-4Rb, in: St. Petersburg, Russian Federation, October 2014. [4] H. Kawashima, et al., Plasma Fusion Res. 1 (2006) 031. [5] K. Shimizu, et al., Nucl.Fusion 49 (2009) 065028. [6] K. Hoshino, et al., J. Nucl. Mater. 463 (2015) 573. [7] S. Tokunaga, et al., 2nd IAEA DEMO Programme Workshop, Vienna, Austria, December 2013. [8] A.S. Kukushkin, et al., J. Nucl. Mater. 438 (2013) S203. [9] K. Hoshino, et al., Contrib. Plasma Phys. 54 (2014) 404. [10] M. Wischmeier, et al., J. Nucl. Mater. 390–391 (2009) 250. [11] Y. Miyoshi, et al., submitted for publication. [12] Y. Someya, et al., Fusion Eng. Des. 98–99 (2015) 1872. [13] N. Asakura, et al., 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices O16, Rome, Italy, May 2016.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
ARTICLE IN PRESS
FUSION-9248; No. of Pages 4
Fusion Engineering and Design xxx (2017) xxx–xxx
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Physics design study of the divertor power handling in 8 m class DEMO reactor Kazuo Hoshino a,∗ , Nobuyuki Asakura b , Shinsuke Tokunaga b , Yuki Homma b , Katsuhiro Shimizu a , Yoshiteru Sakamoto b , Kenji Tobita b , Joint Special Design Team for Fusion DEMO a b
National Institutes for Quantum and Radiological Science and Technology, Naka Fusion Institute, Ibaraki, Japan National Institutes for Quantum and Radiological Science and Technology, Rokkasho Fusion Institute, Aomori, Japan
h i g h l i g h t s • • • •
The power handling for DEMO with 1.5 GW fusion power was studied. The detached divertor plasma was obtained by large impurity radiation. The SONIC simulation showed the target heat load less than 8 MW/m2 . Dependence of the impurity concentration on the fuel gas puff was also studied.
a r t i c l e
i n f o
Article history: Received 3 October 2016 Received in revised form 13 March 2017 Accepted 14 March 2017 Available online xxx Keywords: DEMO reactor Divertor plasma design Power handling Radiative cooling SONIC
a b s t r a c t The divertor plasma performance and the power handling are studied for 8 m class DEMO reactor with the fusion power of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target are obtained for a relatively low electron density of 1.8 × 1019 m−3 at the outer mid-plane separatrix. The SONIC simulation shows the target heat load less than 8 MW/m2 , which can be handled by the ITER-like divertor target, for both target. However, at the outer target, the ion temperature is still high which may cause the target erosion. For the divertor power handling and suppression of the target erosion, the divertor design study have to be further proceeded as well as the core plasma design. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff is also studied. With increasing fuel gas puff rate, the mid-plane separatrix density increases and the Ar impurity concentration at the outer mid-plane decreases to 0.5%. © 2017 Elsevier B.V. All rights reserved.
1. Introduction Handling of the huge power exhausted from the core region to the SOL/divertor region is one of the crucial issues for a DEMO reactor design. The power handling scenario has been investigated for a 3 GW class fusion reaction with an ITER-size plasma, SlimCS [1]. The numerical simulation showed that the target heat load qtarget was 16 MW/m2 even in the case where more than 90% of the exhausted power from the core plasma (Pout ) was radiated by the argon (Ar) impurity gas seeding. In order to investigate design window of the divertor, dependence of qtarget on the fusion power and the impurity
∗ Corresponding author. E-mail address: [email protected] (K. Hoshino).
radiation has been studied [2]. In the simulation, qtarget < 10 MW/m2 was found for the fusion power of ∼1.5 GW and the large impurity radiation power (>0.8Pout ) by the Ar gas seeding. Recently, JA-Model 2014 DEMO concept has been proposed [3]. From the previous concept, SlimCS, a major radius was increased to 8.5 m in order to install a center solenoid coil with enough size for full inductive current ramp up, and a fusion power was decreased to ∼1.5 GW for the divertor and blanket design. These changes are preferable for the divertor power handling. However, the operational density became low (n¯e ∼6.6 × 1019 m−3 , where n¯e is the line-averaged density.) compared with SlimCS (n¯e ∼1.2 × 1020 m−3 ), due to the low plasma current, the high aspect ratio, and the resultant low Greenwald density. Generally, high operational density (high SOL density nsep ) is preferable for formation of the divertor plasma detachment and the power handling.
http://dx.doi.org/10.1016/j.fusengdes.2017.03.068 0920-3796/© 2017 Elsevier B.V. All rights reserved.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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ARTICLE IN PRESS K. Hoshino et al. / Fusion Engineering and Design xxx (2017) xxx–xxx
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In this paper, the divertor plasma performance and the power handling scenario for the 8 m class DEMO with low operational density is investigated by using a suite of integrated divertor codes, SONIC. 2. Numerical model A suite of integrated divertor codes SONIC [4,5] is applied to the physics design study of the DEMO divertor plasma. The fuel plasma transport in the plasma edge (0.95 < r/a < 1), SOL and divertor regions is described by a fluid model. The dynamics of fuel neutral and impurity is solved by a Monte-Carlo particle model in whole region including the sub-divertor region. Although the drift and current model are not included in the present version of the SONIC code, the code can qualitatively reproduce the detached divertor plasma [6]. In this analysis, only deuterium species are considered as fuel species for simplification. The divertor geometry and the numerical grid are shown in Fig. 1. In order to enhance the recycling and formation of detachment, reflector is installed at the inner divertor in addition to the outer divertor. The pumping slot is located at the bottom of the divertor cassette with the pumping speed of 65 m3 /s. The numerical grid for a plasma transport covers the flux surfaces of r < 3.2 cm at the mid-plane. Although the armor material of the divertor target and the blanket is tungsten, its sputtering is not considered in the present analysis. The recycling coefficient of unity is assumed due to the steady-state operation. At the core-edge boundary (r/a = 0.95), exhaust power Pout and particle flux out are given. For a 1.5 GW fusion power DEMO, the alpha heating, the additional heating and the radiation in the core are assumed to 300 MW, 80 MW and 130 MW, respectively, in the analysis. Therefore, Pout is set to 250 MW. From the preliminary core plasma transport analysis [7], out = 1 ×1022 s−1 is assumed. Fuel gas to enhance the divertor recycling is injected from the outer mid-plane with 1.2 × 1022 D/s, which corresponds to D2 gas puff of 25 Pa m3 /s. Diffusion coefficients for thermal and particle are assumed to be i = e = 1 m2 s−1 and D = 0.3 m2 s−1 , respectively, which are the same in the ITER divertor plasma simulation [8]. In order to enhance the impurity radiation in the plasma edge, SOL and divertor regions, Ar impurity gas is puffed into the outer divertor. The Ar seeding rate is controlled by feedback to achieve an assigned frad =
edge
edge,SOL,div
SOL + P div /P Prad + Prad out , where Prad rad
Fig. 2. The radial profile of ne , Ti and Te at the outer mid-plane.
3. Divertor plasma performance and heat load on the wall The radial profiles of the electron density ne , the ion temperature Ti and the electron temperature Te at the outer mid-plane are plotted in Fig. 2. The separatrix density is nsep ∼ 1.8 × 1019 m−3 , which
is
the impurity radiation power in the edge (0.95 < r/a < 1), SOL and divertor region, respectively. Based on the previous study [2], frad = 0.8 is given. For reduction of the impurity Monte-Carlo calculation time, the impurity back flow from the sub-divertor region to the divertor region has been modeled [9].
Fig. 1. The divertor geometry and the numerical grid.
Fig. 3. Spatial distribution of the Ar impurity radiation power Prad (a) and Te (b).
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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Fig. 4. Radial profiles of ne , Ti and Te (a, b) and qtarget (c, d). Left and right figures correspond to the inner and the outer target, respectively. qtarget is evaluated by taking in account the surface recombination of the ion particle, the impurity radiation and the neutral flux in addition to the plasma heat transport.
corresponds to 0.27 n¯e . The plasma temperature of Ti ∼ 1010 eV and Te ∼ 480 eV is relatively high due to high exhausted power and low nsep . Large frad of 0.8 (200 MW) is achieved by the Ar gas puff rate of o−div = 1.47 Pa m3 /s. Distribution of the Ar impurity radiation is Prad SOL+edge
i-div = 82 MW and P 80 MW, Prad = 37 MW. The region of the rad large impurity radiation power can be seen in the inner and outer divertor region, as show in Fig. 3(a). Fig. 3(b) shows the spatial distribution of Te . Due to the large impurity radiation and large volume, low Te region (<2 eV) can be seen in the wide region of the inner divertor and around the outer strike point, i.e., the almost complete detachment in the inner divertor and the partial detachment in the outer divertor are obtained even in the low nsep condition. Radial profiles of ne , Ti , Te and qtarget at the divertor target are plotted in Fig. 4. At the inner target, Te,i < 2 eV, i.e., the detachment can be seen on the almost whole target. The detachment region can been seen around the strike point (≤10 cm) at the outer target, while, Ti is ∼250 eV at the outer flux region due to the low nsep and the resultant low recycling. Since such high Ti can lead significant erosion of the target, expansion of the detached region and reduction of Ti are indispensable. Peak of qtarget is ∼7 MW/m2 on the inner target and ∼8 MW/m2 on the outer target, which can be handled by the ITER divertor technology, i.e., the tungsten mono-block target and the water cooling with the Cu-alloy cooling tube. The contribution to the target heat load is almost plasma heat load at the outer target due to the partial detachment. On the other hand, at the inner target, the plasma heat load decreases but the surface recombination increases regardless of the full detachment. Reproduction of the detachment, especially significant decrease in the particle flux is challenging issue even
for state-of-the-art divertor codes [10]. Therefore, heat load due to the plasma transport and the surface recombination (the ion particle flux) is probably overestimated. Further study and modeling of detachment are necessary for precise analysis. The heat load on the first wall (the blanket surface) is also evaluated in Fig. 5. The impurity radiation load includes contribution only from the calculation domain of SONIC, i.e., r/a > 0.95. The radiation power (including the impurity line radiation, cyclotron emission and bremsstrahlung radiation) from the core region is ∼130 MW (see Section 2). The heat load due to the core radiation is estimated
Fig. 5. The heat load along the blanket surface. The poloidal angle of 0◦ corresponds to the outer mid-plane.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068
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Fig. 6. Variation of nsep and nz /ne on fuel gas puff rate.
to <100 kW/m2 in average but this is not included in Fig. 5. Main component of the heat load on the blanket is the impurity radiation and the neutral particles. Total heat load on the blanket is less than 150 kW/m2 , which is compatible with allowable heat load of blanket design (500–600 kW/m2 [12]). Here, it is noted that heat load of the plasma transport and the surface recombination are evaluated at the grid boundary for the plasma transport (see Section 2). The plasma heat and particle are transported along the flux tube outside of the numerical grid, and concentrate to local area. Therefore, their heat load can become more peak as analyzed in Ref. [11]. 4. Control of the SOL density and the impurity concentration In the view point of consistency between the core plasma design and the divertor power handling, nsep and the impurity concentration nZ /ne are important parameters. Fig. 6 shows nsep and nZ /ne on the different fuel gas puff rate from 1.2 × 1022 D/s to 4.8 × 1022 D/s. The mid-plane separatrix density, nsep , is changed from 1.8 × 1019 m−3 to 2.3 × 1019 m−3 , which correspond to 0.27–0.35 of n¯e = 6.6 × 1019 m−3 . The impurity concentration nZ /ne can be also controlled by the fuel gas puff. Higher gas puff rate enhances the divertor impurity shielding effect by the friction force and dilution of the impurity concentration by higher nsep . Change of the fuel gas puff from 1.2 × 1022 D/s to 4.8 × 1022 D/s decreases nZ /ne from 1.1% to 0.5%. These parameters may be acceptable for a core plasma design but an integrated analysis between core and SOL/divertor plasma is necessary for evaluation of consistency. 5. Summary The divertor plasma performance and the power handling has been studied for 8 m class DEMO reactor with the fusion power
of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target were obtained even for a relatively low nsep ∼ 1.8 × 1019 m−3 . The SONIC simulation showed the target heat load less than 8 MW/m2 , which can be handled by the ITER-like divertor target. However, at the outer target, reduction of the ion temperature is still necessary for suppression of the target erosion. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff rate has been also studied. With increasing fuel gas puff rate to 4.8 × 1022 D/s, nsep increased to 2.3 × 1019 m−3 and the Ar impurity concentration (nz /ne ) decreased to 0.5 %. Consistency of these parameters with the core plasma design will be investigated by using an integrated analysis between the core and SOL/divertor plasma. To suppress the target erosion, i.e., to control the temperature in far SOL at divertor targets, further optimization of divertor geometry and the position of pump slot, and the exploration of more advanced divertor configurations are necessary in future. In addition, development of high density and high radiative core plasma scenario has to be proceeded. One of the key parameter is 95 [13]. If 95 can be increased from 1.65 (JA model 2014) to 1.75, n¯e can be increased from 6.6 × 1019 m−3 to 7.2 × 1019 m−3 and increased fusion power can compensate impurity radiation loss. Acknowledgements SONIC simulations were carried out using the HELIOS supercomputer system at International Fusion Energy Research Centre (IFERC), Aomori, Japan, under the Broader Approach (BA) collaboration between Euratom and Japan, implemented by Fusion for Energy and Japan Atomic Energy Agency. This work is mainly carried out within the framework of the DEMO Design Activity under the BA-IFERC project, and partially supported by Grant-in-Aid for Young Scientists (B) and Grant-in-Aid for Scientific Research (B) of Japan Society for the Promotion of Science. References [1] N. Asakura, et al., Nucl. Fusion 53 (2013) 123013. [2] K. Hoshino, et al., In: 25th IAEA Fusion Energy Conference FIP/P 8-11, in: St. Petersburg, Russian Federation, October 2014. [3] Y. Sakamoto, et al., In: 25th IAEA Fusion Energy Conference FIP/3-4Rb, in: St. Petersburg, Russian Federation, October 2014. [4] H. Kawashima, et al., Plasma Fusion Res. 1 (2006) 031. [5] K. Shimizu, et al., Nucl.Fusion 49 (2009) 065028. [6] K. Hoshino, et al., J. Nucl. Mater. 463 (2015) 573. [7] S. Tokunaga, et al., 2nd IAEA DEMO Programme Workshop, Vienna, Austria, December 2013. [8] A.S. Kukushkin, et al., J. Nucl. Mater. 438 (2013) S203. [9] K. Hoshino, et al., Contrib. Plasma Phys. 54 (2014) 404. [10] M. Wischmeier, et al., J. Nucl. Mater. 390–391 (2009) 250. [11] Y. Miyoshi, et al., submitted for publication. [12] Y. Someya, et al., Fusion Eng. Des. 98–99 (2015) 1872. [13] N. Asakura, et al., 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices O16, Rome, Italy, May 2016.
Please cite this article in press as: K. Hoshino, et al., Physics design study of the divertor power handling in 8 m class DEMO reactor, Fusion Eng. Des. (2017), http://dx.doi.org/10.1016/j.fusengdes.2017.03.068