DEMO Concept Development and Assessment of Relevant Technologies

Transcription

1 1 FIP/3-4Rb DEMO Concept Development and Assessment of Relevant Technologies Y. Sakamoto, K. Tobita, H. Utoh, N. Asakura, Y. Someya, K. Hoshino, M. Nakamura, S. Tokunaga and the DEMO Design Team Japan Atomic Energy Agency, Rokkasho, Aomori-ken, Japan contact of main author: Abstract. Recent development of a DEMO concept with a medium size (major radius of ~8.5 m) and a lower fusion power (~1.5 GW) is presented together with assessment of relevant technologies. The maximum toroidal field is evaluated at ~13 T, which is nearly independent on strand materials (Nb 3 Sn or Nb 3 Al) unlike a compact DEMO, while the increase of the allowable design stress has a large impact on that. The divertor simulation study indicates that the tolerable level of divertor heat flux (~5 MW/m 2 ) is foreseeable for the medium size DEMO, and the design study of short super-x divertor as an option is progressing to efficiently obtain the fully detached plasma. The assessment of various maintenance schemes indicates that the vertical port maintenance scheme provides advantages in the easy handling, the layout of poloidal coils, the size of toroidal coils, and separate maintenance of the in-vessel components. Finally, the study of the waste management suggests that the ratio of the radioactive waste to be disposed of in shallow land burial can be increased thanks to the lower fusion power. 1. Introduction Previous DEMO reactor study on SlimCS [1,2] focused on a steady state and relatively compact reactor that is capable of producing a fusion power of 3 GW and a net electricity of 1 GWe. A lesson of the design study suggests that such a compact and high power reactor has an intractable problem in huge power exhaust in the divertor and a difficulty in the compatibility between tritium breeding and heat extraction. The subsequent design study also suggests that there are additional advantages of reducing the fusion power in terms of remote maintenance and safety in order to assure the engineering feasibility of DEMO. Based on the results, we have been devoted to concept development of DEMO with the medium size (major radius of ~8.5 m) and the lower fusion power (about 1.5 GW). 2. Outline of the concept The target size is set at major radius of ~8.5 m to mitigate the issues related to heat removal and to enable the pulsed operation using enough flux swing by central solenoid coil during commissioning phase, even for the steady-state DEMO. The system code analysis indicated that the fusion power of ~1.5 GW with large bootstrap current fraction (>60%) would be required to generate a net electricity of GWe using neutral beam energy of 1.5 MeV for non-inductive current drive, and to reduce neutron wall load to ~1 MW/m 2 for increasing tritium breeding ratio. Basic concept of blanket is similar to that for SlimCS, which means that the cooling water is the pressurized water condition ( ºC, 15.5 MPa) taking account of compatibility of RAFM, and that the breeding blanket composed of chemically stable neutron multiplier (Be 12 Ti pebbles) and solid breeder (Li 2 Ti 3 pebbles).

2 2 FIP/3-4Rb 3. Relevant technologies for the DEMO concept The assessment technologies mainly include toroidal field coil, divertor, blanket, remote maintenance, waste management scenario and conducting shell Toroidal field coils TF coil is designed based on ITER technology, such as cable-in-conduit type conductor, radial plate, and wedge support structure. Achievable toroidal magnetic field is assessed by using SCONE code [3]. Previous design study indicated that Nb 3 Al could generate larger maximum toroidal magnetic field (B Tmax ) than Nb 3 Sn for the compact DEMO when thickness of toroidal field (TF) coil and allowable design stress were fixed. For the medium size DEMO, however it is found that there is small difference in B Tmax generated by whether Nb 3 Al or Nb 3 Sn and also that the allowable design stress has a large impact on B Tmax as shown in FIG.1, because the increase in magnetic energy requires the large footprint of stabilizer in the coil case to bear large electromagnetic force. The B Tmax for the medium size DEMO is evaluated about 13 T in the case of allowable design stress of 667 MPa. The choice of the superconductor strand material should be evaluated by considering the strain property of the critical current density of superconducting wires and the ease of large coil fabrication. B Tmax (T) Coil thickness: 1.6m Nb 3 Al Nb 3 Sn S m =667MPa S m =800MPa Major radius (m) FIG. 1. B Tmax generated by Nb 3 Sn (triangles) and Nb 3 Al (circles) for allowable design stress of 667 MPa (closed symbols) and 800 MPa (open symbols), respectively Divertor Basic concept of divertor is single-null conventional divertor configuration with water-cooled tungsten mono-block as plasma facing component, similar to ITER divertor design. However, a huge power exhausted from the main plasma in a DEMO is several times as high as that in ITER. Heat removal capability for DEMO divertor has been assessed by temperature analysis of the component. For the water-cooled tungsten mono-block with ferritic steel pipe (F82H), the heat removal capability is 5 MW/m 2 level which is a half of the ITER divertor with Cu-alloy cooling pipe, while 10 MW/m 2 level for Cu-alloy cooling pipe for DEMO, as shown in FIG. 2. It should be noted that influence of neutron irradiation to the Cu-alloy pipe near the large

3 3 FIP/3-4Rb heat flux region in divertor is evaluated less than 1.5 dpa/year, indicating the Cu-alloy can be partially used as cooling pipe near high heat flux region. These results suggest that reducing the fusion power is required even for the medium size DEMO. In addition, the SONIC simulation showed that the target heat load is reduced to the required level in the case of the fusion power of less than 2 GW and large radiation fraction of 90% for the compact DEMO reactor design with R=5.5m (SlimCS). Simulation study of the conventional divertor for the medium size DEMO reactor started recently, and the results showed that the reduced heat load is obtained with the lower radiation fraction [4]. Engineering and physics design studies of the advanced divertor with comparable size of the divertor length, short super-x divertor, started as an option, and its simulation demonstrates the efficient formation of fully detached divertor plasma [5]. (a) (b) (c) F82H-cooling-pipe CuCrZr-cooling-pipe 5.0-MW/m (W) dpa/year P fus =1.35GW Outer-strikepoint Inner-strikepoint 12.5MW/m (W) 541 -(F82H) PosiFon-of-cooling-pipe-in-divertor-region (CuCrZr) FIG. 2. (a) Result of thermal analysis for tungsten mono-block with F82H cooling pipe, (b) influence of neutron irradiation to Cu-alloy cooling pipe, (c) Result of thermal analysis for CuCrZr cooling pipe 3.3. Blanket Basic concept of blanket is similar to that for ITER-TBM in Japan, which means that the cooling water is the pressurized water condition ( ºC, 15.5 MPa) taking account of compatibility of F82H, and that the breeding blanket composed of chemically stable neutron multiplier (Be 12 Ti pebbles) and solid breeder (Li 2 Ti 3 pebbles). Important roles of blanket is to achieve self-sufficiency in tritium, where tritium-breeding ration TBR should be larger than 1.05, and to extraction heat for electric generation. The thickness of the blanket required for TBR>1.05 by optimizing cooling pipe layout for heat extraction has been assessed by neutronics analysis [6,7]. The result indicates that the TBR increases with decreasing neutron wall load, and the required thickness is 0.6 m for the net TBR>1.05, as shown in FIG. 3.

4 4 FIP/3-4Rb MW/m 2 3MW/m 2 5MW/m 2 Local(TBR Blanket(thickness((cm) FIG. 3. Relation between local TBR and blanket thickness for neutron wall load of 1, 3,5 MW/m Remote maintenance Maintenance scheme significantly affects the configuration of in-vessel components, the arrangement of poloidal field (PF) coils, the size of TF coils, reactor buildings including hot cell, and waste management scenario. The various schemes characterizing maintenance port location, blanket segmentation and divertor segmentation have been studied [8], and the two schemes such as the sector and vertical port maintenance schemes are shown in FIG. 4. It is concluded that the vertical port maintenance scheme provides advantages to the ease of handling, the layout and required currents for PF coils, the size of TF coils, and enables us to replace the blanket and divertor modules according to the their lifetimes. CS#coil PF1 PF2 PF8 PF7 TF'coil PF3 PF4 (a) PF6 PF5 Sector'maintenance'port CS#coil Ver.cal'maintenance'port PF1 PF2 PF8 PF7 PF3 TF'coil PF4 PF5 PF6 (b) Divertor' maintenance'port FIG. 4. Layouts of (a) sector and (b) vertical port maintenance schemes

5 5 FIP/3-4Rb 3.5. Waste management scenario The waste management scenarios are considered. The amount of radioactive waste generated in every replacement (4 years) is estimated to over 9,000 tons (BLK: 5723 ton, DIV: 924 x 4 = 3696 ton). Since the lower fusion power contributes to reduce the activation level, the ratio of the waste to be disposed of in shallow land burial can be increased. Reuse of the neutron shield is a key strategy to reduce the waste furthers [9]. By reusing mixed breeder (881 ton of Li 2 TiO 3 &Be 12 Ti), Conducting shell (372.4 ton of CuCrZr) and Back plate ( ton of F82H and SUS316L), the waste can be reduced by 1879 ton (20%) and disposed in shallow land burial in 16 years (no geological disposal) as shown in FIG. 5. Hot(cell(storage Interim( Shallow(land( storage burial FIG. 5. Time history of contact dose rate of each component 3.6. Conducting shell Plasma elongation is one of the key factors in determining absolute plasma performance, while highly elongated plasma is vertically unstable. Vertical instability can be suppressed by eddy current driven in conducting shell and in-vessel coils, which is considered in ITER as well. However, in-vessel coils are not applicable to DEMO because of large neutron irradiation and maintainability. Furthermore, the position of conducting shell should be far from the plasma surface in DEMO due to the thick breeding blanket. For instance, 0.6 m is required for TBR larger than In this case, wall radius over plasma minor radius become larger than 1.3, where the conducting shell is less effective on vertical stability. In order to assess the available elongation to DEMO, growth rate of vertical instability with conducting shell is investigated by using simple ring coil conducting shell model, as shown in FIG. 6. The results suggest that elongation of 1.65 is sustainable from the viewpoint of vertical stability. It should be noted that higher elongation can be available for lower aspect ratio, but aspect ratio larger than 3 is typically required for inductive I p ramp-up because the CS coil radius decreases with aspect ratio in the case of fixed size of in-vessel components and major radius. On the other hand, the required conducting shell position determines the maximum aspect ratio, though high aspect ratio is desirable to increase n GW ~I p /πa 2 p for compatibility with the divertor detachment.

6 6 FIP/3-4Rb Ac#ve&control&by&PF&coil FIG. 6. Property of vertical stability for different elongations 4. Design Parameters Based on the assessment of relevant technologies described Sec. 3, possible design parameter is shown in TABLE 1, where TF coil is designed with allowable stress of 800 MPa and Nb3Sn conductor. This design parameter makes it possible to operate for steady-state and two hours pulse in the same device configuration. TABLE 1: Possible DEMO design parameters Para. Size;&;ConfiguraOon Absolute;Performance Normalized;Performance S.S.)/) Pulse)(2hr) Ref.) ITER R p ;(m) a p ;(m) A κ δ q V p ;(m 3 ) I p ;(MA) B T ;(T) B T max ;(T) Para. P fus ;(MW) Pulse)) (2hr) S.S. Ref.) ITER P net ;(MWe) W Q P alp ;(MW) P ADD ;(MW) <T e >;(kev) <n e >; (10 19 m W3 ) W th ;(MJ) τ E ;(s) Para. Pulse)) (2hr) S.S. Ref.) ITER HH 98y β N f BS f CD n e /n GW f He Summary The DEMO concept with the medium size (R p ~8.5 m) and the lower fusion power (P fus ~1.5 GW) has been developed together with assessment of relevant technologies. The divertor simulation study indicates that the tolerable level of divertor heat flux (~5 MW/m 2 ) is foreseeable for the medium size DEMO. Lower P fus contributes to enhance the net TBR.

7 7 FIP/3-4Rb Radioactive wastes can be disposed in shallow land burial in 16 years since blanket and divertor replacements. The segmented maintenance scheme provides advantages compared with the sector maintenance scheme for the medium size DEMO. The maximum B T is evaluated at ~12 T with the allowable design stress of 800 MPa. The possible DEMO design parameter indicates that the DEMO can has wide operation range from 2 hours pulse operation to steady state, which would satisfy the DEMO development strategy. References [1] TOBITA, K., et al., SlimCS compact low aspect ration DEMO reactor with reducedsize central solenoid, Nucl. Fusion 47 (2007) [2] TOBITA, K., et al., Compact DEMO, SlimCS: design progress and issues, Nucl. Fusion 49 (2009) [3] UTOH, H., et al., SCONE code: Superconducting TF coils design code for tokamak fusion reactor, J. Plasma Fusion Res. Ser. 9 (2010) [4] HOSHINO, K., et al., Studies of Impurity Seeding and Divertor Power Handling in Fusion Reactor, 25 th Fusion Energy Conference (FEC 2014), FIP/P8-11. [5] ASAKURA, N., et al., Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor, 25 th Fusion Energy Conference (FEC 2014), FIP/3-4Ra. [6] SOMEYA, Y., et al., Simplification of blanket system for SlimCS fusion DEMO reactor, Fusion Eng. Design 86 (2011) [7] SOMEYA, Y., et al., Design study of blanket structure based on a water-cooled solid breeder for DEMO 28 th Symposium on Fusion Technology (2014) P [8] UTOH, H., et al., Comparative evaluation of Remote Maintenance Schemes for Fusion DEMO Reactor, 28 th Symposium on Fusion Technology (2014) P [9] SOMEYA, Y., et al., Waste management scenario in the hot cell and waste storage for DEMO, Fusion Eng. Design 89 (2014)