Studies of Impurity Seeding and Divertor Power Handling in Fusion Reactor

Transcription

1 1 FIP/P8-11 Studies of Impurity Seeding and Divertor Power Handling in Fusion Reactor K. Hoshino 1, N. Asakura 1, K. Shimizu 2 and S. Tokunaga 1 1 Japan Atomic Energy Agency, Rokkasho, Aomori Japan 2 Japan Atomic Energy Agency, Naka, Ibaraki Japan Corresponding Author: hoshino.kazuo@jaea.go.jp Abstract: The power handling in the divertor is one of the most crucial issues for a fusion reactor design. In the previous study of development of the power handling scenario for a compact DEMO reactor, SlimCS, further reduction of the target heat load was required even in the case where more than 90 % of the exhausted power from the core plasma was radiated by the argon impurity gas seeding. In this paper, the impact of the fusion power and the impurity seeding on the power handling in the fusion reactor divertor has been investigated by using an integrated divertor code SONIC. With decreasing the fusion power, the divertor plasma detachment is extended and the target heat load decreases. The SONIC simulation showed the operational regime, i.e., the target heat load less than 6 MW/m 2 for a tungsten mono-block divertor with a ferritic steel water-cooling pipe, at the fusion power less than 2 GW. It is also shown that the impurity radiation fraction on the exhausted power can be reduced to 80 % at the fusion power of 2 GW in the case of a copper-alloy water-cooling tube. 1 Introduction Huge power handling in the SOL/divertor region is one of the crucial issues for a tokamak fusion DEMO reactor, such as SSTR[1], Demo-CREST[2], ARIES-AT[3], PPCS[4, 5], SlimCS[6, 7] and so on. In a DEMO reactor, the desirable heat load on the divertor target is less than or comparable to the ITER design of 10 MW/m 2 because the available material is restricted by the strong neutron irradiation environment. The power removal capability is evaluated to be 5-7 MW/m 2 for a mono-block target with tungsten and reduced activation ferritic martensitic steel as the structure and the water-cooling pipe materials, [6]. The primary technique for reduction of the divertor heat load to such a desirable level is enhancement of the radiation loss by impurity gas seeding. Power handling scenario for a compact DEMO reactor (SlimCS: major radius R of 5.5 m, a fusion power P fus of 3 GW) has been investigated [8]. The target heat load q target was 16 MW/m 2 even in the case where more than 90% of the exhausted power from the core plasma (P out ) was radiated by the argon (Ar) impurity gas seeding. The effects of the seeded impurity species (neon and krypton) and the divertor geometry on the divertor

2 FIP/P performance have been also investigated. Due to such effects, q target could be decreased but it was still larger than the heat removal capability. One of the possible solutions for further reduction of q target is an advanced divertor concept, such as short super-x divertor [9]. Another solution is change of machine specifications, such as fusion power, machine size, etc. In this study, the impacts of fusion power and the impurity seeding on the divertor power handling are investigated. 2 Numerical model A suite of integrated divertor codes SONIC [10] is applied to the study of the power handling in the DEMO divertor. The SONIC suite consists of the 2-D plasma fluid code (SOLDOR), the neutral Monte- Carlo code (NEUT2D) and the impurity Monte-Carlo code (IMPMC). The transport of bulk ion is described in the fluid approximation by the SOLDOR code, while neutral dynamics of atom and molecule are treated by the NEUT2D code. Figure 1 shows the divertor geometry and the numerical grid. The input parameters are as follows: the fuel ion density of m 3 is fixed at the core interface boundary (CIB) of r/a = The power flux across the core side boundary P CIB is assumed to be 500 MW for P fus = 3 GW and 320 MW for P fus = 2 GW. The fuel gas puff of Fpuff mid = s 1 and Fpuff OD = s 1 are injected from the mid-plane and near the outer divertor target, respectively, as shown in Fig. 1. The pumping speed is assumed to be 200 m 3 /s. The spatially uniform diffusion coefficients for the particle and heat are assumed to be Z (m) backflow exhaust slot #MSH0306B#mesh fuel&imp. puff sub-divertor chamber pumping R (m) FIG. 1: grid. fuel puff divertor geometry and numerical D = 0.3 m 2 /s and χ = 1 m 2 /s, respectively, as same as the ITER divertor simulation [11]. The Ar impurity gas is injected from the outer divertor to reduce q target and the seeding rate is adjusted by feedback to keep f rad = 92 % 80 %, where f rad is a fraction of the impurity radiation power P rad on P CIB. In order to save a calculation time, the impurity back flow from the sub-divertor region to the divertor region as shown in Fig. 1 is assumed [12].

3 3 FIP/P8-11 FIG. 2: The radial profile of n e, T i and T e at (a) the inner target and (b) the outer target. Bottom figures show the radial profile of the heat load including the plasma heat flux, surface recombination, radiation power, and neutral flux. 3 Impact of the fusion power on the divertor power handling In the case of P fus = 3 GW, large f rad of 92 % (P rad = 460 MW) is achieved by the Ar impurity gas puff of 14.5 Pa m 3 /s. The radiation power is 186 MW (40.5 % of P rad ) in the outer divertor, 76 MW (16.6 % of P rad ) in the inner divertor and 197 MW (42.9 % of P rad ) in the SOL/edge region. Figure 2 shows the radial profile of the electron density n e, the electron temperature T e, the ion temperature T i and q target. Due to the large impurity radiation, the inner divertor plasma is fully detached (T e, T i < 2 ev). On the other hand, the outer divertor is the partial detachment, i.e., T e and T i are less than 2 ev near the strike point (distance from the strike point r sep < 7 cm), while they increase at the outer region (r sep > 7 cm) due to the low n e and low collisionality. The plasma heat load along the field line is decreased to less than 4 MW/m 2 at the both target by z (m) r (m) FIG. 3: The distribution of the Ar impurity radiation power near the outer divertor target.

4 FIP/P FIG. 4: The radial profile on the outer target for P fus = 2 GW: (a) n e, T i and T e, and (b) the heat load. the full/partial detachment. The heat load due to the surface recombination of the ion flux is 4 MW/m 2 near both inner and outer strike point because n e is high. At the attached region of the outer target (r > 7 cm), the impurity radiation load is high. As shown in Fig. 3, the large impurity radiation power is seen close to the target and this radiated power becomes the large target heat load. The neutral heat load of less than 3 MW/m 2 is widely distributed on the target. Consequently, the peak of the total heat load is about 10MW/m 2 on both targets, which is larger than the heat removal capability of the target. In order to solve the power handling issue, impact of the fusion power of on the power handling is investigated. The heat flux across CIB is changed from P CIB = 500 MW to P CIB =320 MW, which corresponds to P fus = 3 GW and 2 GW, respectively. Total impurity radiation power of 320 MW (f rad = 0.92 )is distributed to 43.3 % in the outer divertor, 16.3 % in the inner divertor and 40.3 % in the SOL/edge region. The distribution of the impurity radiation power is similar to the case of P fus = 3 GW. Figure 4 shows the radial profile of the plasma and the heat load on the outer target. Due to the low P CIB with large Ar impurity radiation fraction of f rad of 92 %, the temperature, especially T i decreases and the detached region (T i, T e < 2 ev) is extended to 12 cm. Sum of the plasma heat load and the surface recombination of the ion flux is less than 4 MW/m 2 due to the partial detachment. In addition, the impurity radiation load decreases to less than 2 MW/m 2 because the impurity radiation moves upstream due to extension of the detachment. Consequently, peak of the total heat load becomes 6 MW/m 2, which can be handled by the present divertor target concept. 4 impact of the impurity seeding In above case, f rad = 92 % was assumed but it is difficult to extrapolate such large f rad from the present experiments accompanied by the energy confinement degradation and fuel dilution appropriate for the fusion reactor plasma. When the copper-alloy cooling tube is utilizable for the divertor target, the heat removal capability increases to 10

5 5 FIP/P8-11 MW/m 2. In the case of P fus = 2 GW with f rad = 92 %, there is some margin to reduce f rad. Therefore, the impact of the reduced f rad on the divertor performance is investigated at P fus = 2 GW. With decreasing f rad from 92 % to 80 %, the Ar impurity radiation power in the upstream, i.e., the SOL and edge region (r/a > 0.95) decreases from 115 MW to 81 MW. The electron heat flux through the divertor throat is increased and then the area of the low T e less than 2 ev becomes narrow as shown in Fig. 3(a), i.e., the detachment becomes weak. The heat load due to plasma heat transport and the surface recombination increases to 6 MW/m 2, as shown in Fig. 5. In addition, the impurity radiation region moves to the target and radially expands due to increase in T e near the strike point. Due to increase in FIG. 5: The radial profile of the heat load on the outer target for P fus = 2 GW with f rad = 80 %. the radiated power near the target, the radiation load on the target is increased. Consequently, peak of the total target heat load increases to 10 MW/m 2, but it can be still handled by the tungsten mono-block divertor with a copper-alloy cooling tube. 5 Summary The impacts of fusion power and the impurity seeding on the divertor power handling were investigated using a integrated divertor code SONIC. In the case of the fusion power of 3 GW with f rad = 0.92, the fully and partial detached divertor plasma were obtained in the inner and outer divertor, respectively. The target heat load due to the plasma transport was less than 4 MW/m 2, while the heat load due to the surface recombination and the impurity radiation is relatively large even in the detached region. As a result, the peak of the total heat load exceeds the heat removal capability of 5-7 MW/m 2 for a mono-block target with tungsten and reduced activation ferritic martensitic steel as the structure and the water-cooling pipe materials. With decreasing the fusion power from 3 GW to 2 GW, the divertor plasma detachment was extended and the total target heat load decreased to less than 6 MW/m 2, which is within the heat removal capability of the target. It is also shown that the impurity radiation fraction on the exhausted power can be reduced to 80 % at the fusion power of 2 GW If a copper-alloy is utilized as a cooling tube. In this study, some operational points in the viewpoint of the divertor power handling were shown by the SONIC stimulation. However, f rad is still high compared with the present experiments with high performance core plasma. In addition, assumption of high ion density at the core interface boundary may limit the operational window. To explore the operational windows consistent with the burning core performance, impact of other key parameters, such as edge density profile, a major radius, etc., on power handling will be investigated.

6 FIP/P Acknowledgement This work is mainly carried out within the framework of the DEMO Design Activity under the the Broader Approach collaboration between Euratom and Japan, implemented by Fusion for Energy and Japan Atomic Energy Agency, and partially supported by Grantsin-Aid for Scientific Research of Japan Society for the Promotion of Science. SONIC simulations were carried out using the HELIOS super-computer system at International Fusion Energy Research Centre, Aomori, Japan, under the Broader Approach. References [1] Kikuchi, M., et al, Recent directions in plasma physics and its impact on tokamak magnetic fusion design, Fusion Eng. Des. 16 (1991) 253. [2] Hiwatar, R., et al., Demonstration tokamak fusion power plant for early realization of net electric power generation, Nucl. Fusion 45 (2005) 96. [3] Najimabadi, F., et al., The ARIES-AT advanced tokamak, Advanced technology fusion power plant, Fusion Eng. Des. 80 (2006) 3. [4] Lackner, K., et al., Long-term fusion strategy in Europe, J. Nucl. Mater. 307 (2002) 10. [5] Maisonnier, D., et al., Power plant conceptual studies in Europe, Nucl. Fusion 47 (2007) [6] Tobita, K., et al., SlimCS compact low aspect ratio DEMO reactor with reducedsize central solenoid, Nucl. Fusion 47 (2007) 892. [7] Tobita, K., et al., Compact DEMO, SlimCS: design progress and issues, Nucl. Fusion 49 (2009) [8] Asakura, N., et al., A simulation study of large power handling in the divertor for a Demo reactor, Nucl. Fusion 53 (2013) [9] Asakura, N., et al., Investigation of Advanced Divertor Magnetic Configuration for DEMO Tokamak Reactor, Trans. Fus. Sci. Tech. 63 (2013) 70. [10] Kawashima, H., et al., Development of integrated SOL/divertor code and simulation study of the JT-60U/JT-60SA tokamaks, Plasma Phys. Control. Fusion 49 (2007) S77. [11] Kukushkin, A. S., et al., ITER divertor performance in conditions of carbon reerosion, J. Nucl. Mater (2005) 50. [12] Hoshino K., et al., Development of the Backflow Model for Simplified Impurity Exhaust in Monte-Carlo Calculation, Contrib. Plasma Phys. 54 (2014) 404.