Optimization Studies on Conceptual Designs of LHD-type Reactor FFHR

Optimization Studies on Conceptual Designs of LHD-type Reactor FFHR A. Sagara 1, O. Mitarai 2, M. Kobayashi 1, T. Morisaki 1, T. Tanaka 1, S. Imagawa 1, Y. Kozaki 1, S. Masuzaki 1, K. Takahata 1, H. Tamura 1, N. Yanagi 1, K. Nishimura 1, H. Chikaraishi 1, S. Yamada 1, S. Fukada 3, A. Shishkin 4, T. Watanabe 1, Y. Igitkhanov 5, T. Goto 6, Y. Ogawa 6, T. Muroga 1, T. Mito 1, O. Motojima 1 and FFHR design group 1 National Institute for Fusion Science, Japan 2 Kyushyu Tokai University, 3 Kyushu University, 4 Kharkov Institute of Physics and Technology, Ukraine 5 Max-Planck-Institut für r Plasmaphysik,, Germany 6 The University of Tokyo Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies with participation of EU February 5-7, 2007 at Kyoto, JAPAN A.Sagara-1/26

System Integration and Design Activities From 1993,, collaboration activities on LHD-type reactor-system designs have developed with connecting wide research areas on fusion science and engineering (Fusion( Research Network) ) in the universities in Japan. Confinement scaling H.Yamada (NIFS) Helical core plasma K.Yamazaki(NIFS) Magnetic structure T.Morisaki (NIFS) SC magnet & supprt S.Imagawa T.Mito K.Seo (NIFS) Virtual Reality tool N.Mizuguchi (NIFS) LHD Power supply H.Chikaraishi (NIFS) External heating O.Kaneko (NIFS) SC magnet & support Heating Divertor pumping S.Masuzaki, Kobayashi (NIFS) Ignition access & heat flux O.Mitarai (Kyusyu Tokai Univ.) Core Plasma Divertor pumping Fueling (NIFS) surface heat flux Fueling Thermo-mechanical analysis H.Matsui (Tohoku Univ.) Blanket system T.Terai (Univ. of Tokyo) 14MeV neutron In-vessel Components Thermofluid Structural Materials Prot ecti on wall P h = 0.2 M W /m 2 P n = 1.5 M W /m 2 Nd =45 0 d pa/30 y Purifier Pump Thermofluid MHD S.Satake (Tokyo Univ. Sci.) Advanced first wall T.Norimatsu (Osaka Univ.) Se lf-co oled T breeder TBRl ocal > 1.2 Be Tan k Liquid / Gas HX Ra diat ion shie ld redu cti on > 5 ord ers T herm al shield high temp. T-diseng ager Chemistry Blanket shielding materials. Vacu um ve ssel & T bou nda ry SC ma gnet 20 C Pu mp T storage Tur bine Safety & Cost Helical reactor design / Sytem Integration / Replacement A.Sagara (NIFS) Tritium Thermofluid system K.Yuki, H.Hashizume (Tohoku Univ.) Advanced thermofluid T.Kunugi (Kyoto Univ.) Neutronics T.Tanaka, Sagara (NIFS) Alternative blanket T.Muroga (NIFS) T-disengager system S.Fukada, M.Nishikawa (Kyusyu Univ.) Heat exchanger & gas turbine system A.Shimizu (Kyusyu Univ.) March. 200#, A.Sagara A.Sagara-2/26

ASIPP, China Kyoto Univ., Eng., (1,2,3,4) Kyoto Univ., Res. Reactor Inst. (1,3) Kyoto Univ., Inst. Adv. Energy (1,3) Kyoto Univ., Energy Science (1,3) Kyoto Univ. Education (3) Osaka Univ. ILE (3,4) Osaka Univ., Eng. (1,2,3,4) Kinki Univ. (1,4) Osaka Pref. Univ. (3) Osaka City Univ. (2) Osaka Inst. Tech. (1,3) Hyogo Univ. Education (1) Kobe Univ. Marc. Marine (1) Himeji Inst. Tech (1) Okayama Univ. (3) Okayama Univ. Science (1,2) Fukuyama Univ. (3) Hiroshinma Univ., Science (3) Hiroshima Univ., Res. Inst. Nuc. Med. (3) Tottori Univ. (2) Tokushima Univ. (1) Ehime Univ. (1) Kyushu Univ., Eng. (1,2,3,4) Kyushu Univ., Science (1,3) Kyushu Univ., Res. Inst. Appl. Mech. (1) Kyushu Univ., Eng. Sci. (1,3) Kyushu Inst. Tech., Eng. (3) Kyushu Inst. Tech., Comp. System (2) Oita Univ. (2) Kagoshima Univ. (2) Kyushu Tokai Univ. (3) Fusion Engineering Network Niigta Univ. (1,3) Nagaoka Univ. Tech. (4) Toyama Univ. (1) Kanazawa Univ. (3) Nagoya Univ., Eng. (1,3) Nagoya Univ., Integ. Res. Sci. Eng. (1) Shizuoka Univ. (1,3) Toyohashi Univ. Tech. (2,3) FFHR design NIFS (1)Materials and Fuels (2)Magnet, Electromagnetics (3)Reactor System, Safety (4)ICF Technology Hokkaido Univ., Eng. (1,2,4) Hokkaido Univ., CARET (1,3) Muroran Inst. Tech (1,4) Tohoku Univ., Eng. (1,2,3) Tohoku Univ., Science (1) Tohoku Univ., Med. (3) Tohoku Univ., IMR (1) Tohoku Univ., IMR-Oarai (1) Tohoku Univ., Inst. Fluid Sci. (2) Iwate Univ. (2) Akita Univ., Med. (2,3) Gunma Univ. (1,4) Ibaraki Univ., Eng (1) Ibaraki Univ., Science (3) Tsukuba Univ. (1,3) Keio Univ. (3) Hosei Univ. (3) Seikei Univ. (2,3,4) Sophia Univ. (2) Toho Univ. (3) Hihon Univ. (2) Kanagawa Inst. Tech. (2) Tokyo Inst. Tech., Eng. Sci. (3) Tokyo Inst. Tech., Res. Nucl. Reac. (1,2,3) Tokyo Inst. Tech., Eng. (1) Tokai Univ., Eng. (1,3) Tokyo Univ., Eng. (1,2,3,4) Tokyo Univ., Nucl. Eng. Res. Lab. (1,2,3) Tokyo Univ., Res. Artifact (1) Tokyo Univ., Science (2) Tokyo Univ., Isotope (1,3) Max-Planck IPP, Germany Kharkov IPT, Ukraine A.Sagara, ANS-FED Newsletter - June 2004, http://fed.ans.org/ A.Sagara-3/26

Based on LHD, FFHR has 2 Main Features (1) The coil pitch parameter of continuous helical winding has been adjusted to reduce the magnetic hoop force (Force Free Helical Reactor: FFHR), to expand the blanket space By Selecting lower value of =tan (2) Self-cooled liquid Flibe (BeF 2 -LiF) blanket has been proposed, due to low MHD pressure loss, low reactivity with air, low pressure operation, low tritium solubility. A.Sagara-4/26

Main design parameters and recent modification & optimization activities A. Sagara et al. Nucl. Fusion 45(2005) 258-263. Sagara, 17th TOFE(2006) Kobayashi, 17th TOFE(2006) Morisaki,, 15th ITC(2005) Takahata, 24th SOFT(2006) Itoh, 17th TOFE(2006) Yamada, 24th SOFT(2006) Mitarai, 21th IAEA(2006) Sato, 17th TOFE(2006) Tanaka, 21th IAEA(2006) Muroga, 17th TOFE(2006) Nagasaka, 17th TOFE(2006) Kozaki, 17th TOFE(2006) Mizuguchi,, 15th ITC(2005) Design parameters LHD FFHR2 FFHR2m1 FFHR2m2 Polarity l 2 2 2 2 Field periods m 10 10 10 10 Coil pitch parameter 1.25 1.15 1.15 1.25 Coil major Radius Rc m 3.9 10 14.0 17.3 Coil minor radius ac m 0.98 2.3 3.22 4.33 Plasma major radius Rp m 3.75 10 14.0 16.0 Plasma radius ap m 0.61 1.2 1.73 2.80 Blanket space m 0.12 0.7 1.2 1.1 Magnetic field B0 T 4 10 6.18 4.43 Max. field on coils Bmax T 9.2 13 13.3 13 Coil current density j MA/m2 53 25 26.6 32.8 Weight of support ton 400 2880 3020 3210 Magnetic energy GJ 1.64 147 120 142 Fusion power P F GW 1.77 1.9 3.0 Neutron wall load MW/m2 2.8 1.5 1.3 External heating power Pext MW 100 80 100 heating efficiency 0.7 0.9 0.9 Density lim.improvement 1 1.5 1.5 H factor of ISS95 2.53 1.92 1.68 Effective ion charge Zeff 1.32 1.34 1.35 Electron density ne(0) 10^19 m-3 28.0 26.7 19.0 Temperature Ti(0) kev 27 15.8 16.1 <>=2*n*T/(B^2/μ) (parabolic distribution) 1.8 3.0 4.1 COE Yen/kWh 21.00 14.00 9.00 A.Sagara-5/26

Recent Activities in Optimization Phase 1. Heating power 2. Reactor size 3. Nuclear shielding 4. Replacement-free blanket 5. Magnet system 6. Blanket materials 7. Plant system 8. Summary A.Sagara-6/26

Heating Power can be Minimized to < 30MW by longer fusion power rise-up time due to dw/dt effect. O. Mitarai, A. Sagara et al., 21th IAEA (2006) Any fusion power rise-up time can be achievable in a helical reactor. [Example: t rise =300 sec] P EXT (MW) 100 80 60 40 20 0 0 100 200 300 400 500 rise (s) PEXT-Peak PEXT-FB DWDTV-Peak DWDTV-FB AV-DWDTV n(0) (m -3 ) f alpha 4 10 20 3 10 20 2 10 20 1 10 20 n limit (0)/n(0) P EXT (MW) 0 0.04 0.02 0 4 3 2 1 0 150 100 50 NE0 NE0LMT PF PF0 DLM PEXT (b) (c) (a) (d) FALPHA WPV 0 0 0 100 200 300 400 500 Time (s) T DWDTV 40 30 20 10 0 2 1 0 800 400 0 15 10 5 T i (0) (kev) P f (GW) Wp(MW) dw/dt(mw) A.Sagara-7/26

Reactor size Which is strongly coupled with blanket and divertor design LHD-type reactor FFHR2m1 FFHR2m1 P ~ 1GWel W~ 25,000 ton Q > 50 by helical New targets based on LHD: Large scale SC magnet SC magnet of high J candidate Coil design by S.Imagawa constraints: R=14m A.Sagara-8/26

Standard design in FFHR Conventional helical divertor inherently equipped built-in divertor Alternative design in FFHR Helical X-point Divertor (HXD) new concept for the compatibility with blankets T.Morisaki, A.Sagara et al., ITC-15, 2005 / FED 81 (2006) Heat load to target plates --> assuming average width ~ 0.25m (total area: 0.25m x 230m ~ 58m 3 ) --> power loss through divertor ~ 300MW --> ~ 3.2 MW/m 2 Particle recycling ---> 99% ionized ( Kobayashi, 17th TOFE) A.Sagara-9/26

Improvement of 3D neutronics calculation system for helical structure T. Tanaka, A. Sagara, M.Z.Youssef, et al. 21th IAEA (2006) (a) (b) (c) Generation of helical structure according to numerical equations Division of cross-section of helical blanket components into quadrangular meshes. Quick generation of 3-D helical geometry with small size input data. (Coordinates of quadrangular meshes on cross-section drawing) Example of 3-D geometry data with MCNP (~3,000 cells) and helical neutron wall loading A.Sagara-11/26

Shielding issues have been made clear (1) Direct neutrons from core plasma to side surfaces of helical coils (Could be suppressed by enhancement of shielding) (2) Neutron streaming to poroidal coils (3) Neutron reflection on helical coils Geometry for investigation of neutron streaming and reflection Core plasma (1) (2) (3) Horizontal cross-section of geometry Distribution of fast neutron flux on helical coil, troidal coils and support structures A.Sagara-12/26

Discrete Pumping with Semi-closed Shield (DPSS) is proposed Localization of pumping ports Enhancement of pumping efficiency Suppression of nuclear streaming to SC coils Improvement of the total TBR A.Sagara-13/26

Long-life blanket STB for 30 years has been proposed (Spectral-shifter and Tritium Breeder blanket) A. Sagara et al. Nucl. Fusion 45(2005) 258. Neutron damage (dpa/30years) 250 200 150 100 50 0 C Be 2 C STB in FFHR2m1 : after 30 years at 1.5 MW/m 2 JLF-1 JLF-1 0 100 200 300 400 Position (mm) Replacement-free of blanket units < 100dpa in the full-life of 30y Improved nuclear shielding for SC magnets ~ 5x10 22 fast n/m 2 in 30y Issue : Enhancement of cooling capability for one-side heating > 1MW/m 2 is requested. C Flibe40%+Be60% Flibe A.Sagara-15/26

Thermofluid R&D activities Tohoku Univ.: H.Hashizume, K.Yuki et al. for enhancing heat-transfer in such high Prandtl-number fluid as Flibe The performance of SPP flow using alumina spheres (0.5D) is about 4 times higher than that of CP flow. Flow velocity to achieve h=20000w/m 2 K using SPP is just 4 m/sec, which is much lower compared with CP flow(~17m/s). Reduction of MHD effect, erosion, etc. 17th TOFE 20000 4 times To remove 1MW/m 2, h=20000 is necessary. 4 17 Pumping power to achieve h=20000 using SPP is a little higher than that of CP flow. However, in the case of using metal and smaller spheres, the heat transfer performance of SPP can overcome CP flow, which enables lower pumping power conditions. Much lower pumping power in all piping system. 20000 A.Sagara-16/26

Superconducting Magnet System for FFHR (Low-Temperature Superconductor Option) K. Takahata, T. Mito, A. Sagara, et al: presented at 24 th SOFT,11-15 Sept. 2006, Warsaw, Poland, P1-E-285 Poloidal coil Continuous windings with a large diameter, twisted configuration and large magnetic energy Indirect cooling using cooling panels Aluminum-alloy jacketed Nb 3 Sn superconductor 100 ka Superconductor High effective thermal conductivity High mechanical rigidity and strength Quench protection Nb 3 Sn+Cu Ceramic R=14 m B 0 =6.2 T B max =13.3 T E coil =120 GJ SS316 Indirect Cooling Supporting structure Helical coil SS316 Conventional protection circuit using an external resistor =20 s Six subdivisions V max =10 kv R&D plan Hot spot temperature < 150 K Sub-scale 10 ka-class superconductor Model coil test Cross-sectional structure of the helical coil A.Sagara-18/26

Characteristic evaluation of mechanical jointed HTS cable for remountable HTS magnet 17th TOFE SUS304 Rod Tohoku Univ.: H.Hashizume, S.Ito et al. Low Temperature Solder BSCCO 2223 Cable SUS304 Plate BSCCO 2223 Cable Buckling occurs Top loading Dual loading Dual Loading with supporting structure Improving joint performance by optimizing loading system Dual loading with supporting structure can increase the optimum joint stress where joint resistance is the minimum. Dual loading with supporting structure can reduce joint resistance just slightly. Joint surface condition must be improved to reduce joint resistance more. (Target of joint resistance: Order of 100n) A.Sagara-19/26

Tritium Control for V-alloy/Flibe Blanket which can increase operation temperature and T 17th TOFE Application of REDOX control Out of Blanket WF 6 doping by MoF 6 or WF 6 doping Flibe 17th TOFE / NIFS : T.Muroga, T, Tanaka, Z. Li, A. Sagara UCSD : Sze Recovery of TF and W However, conventional REDOX control with Be (TF to T 2 ) results in extremely high T inventory in V-alloy structure Thermodynamics analysis were carried out for another REDOX control (T 2 to TF) by MoF 6 or WF 6 doping This reaction can form corrosion protective Mo or W coating on V-alloy surfaces with self-healing capability Doping of 1ppm MoF 6 or WF 6 can reduce T inventory in blanket below 100 g. Among the issues of the concept are recovery of tritium in the form of TF and Mo or W out of the blanket. Reaction in Flibe T 2 +WF 6 TF + W Reaction at the wall V + WF 6 VF 4or 5 + W V-alloy n 10000 e v n I 1000 m u i t i r T 100 10 In Flibe VF 4 or 5 WF 6 Wcoating (self-healing) Concept of REDOX control by WF 6 doping In V-alloy Structure (Flibe with 1ppm WF 6 ) 1 0.01 0.1 1 10 100 Tririum Molar Fraction as TF in Flibe (ppm) V In V-alloy Structure (Flibe with 1ppm MoF 6 ) T-inventory in blanket as a function of TF level in Flibe, for doping of 1ppm MoF 6 or 1ppm WF 6 A.Sagara-21/26

Tritium cycle of FFHR-2 Flibe self coolant system 17th TOFE: S. Fukada (Kyushu University), A. Sagara (NIFS) Flibe redox control analysis TF concentration in a Flibe blanket of FFHR Flibe-He extraction tower calculation TF (HF) can be reduced to T 2 (H 2 ) in contact with Be in blanket. (JUPITER-II study) The redox control by Be is achieved under the TF concentration expected in the FFHR-2 Flibe blanket.. Flibe-He counter-current extraction system is designed and calculated. A.Sagara-23/26

Dedicated Hydrogen Production by 1 GWe (Optional Design) Breeder Blanket Neutron Shield Flibe loop S. Yamada, A. Sagara et al: presented at 24 th SOFT, 11-15 Sept. 2006, Warsaw, Poland, P3-J-333. Steam production for electrolysis by the waste heat from; divertor plates, neutron shielding parts and so on. Various types of hydrogen production; 24hour full-time hydrogen production, - High pressure gaseous hydrogen (625 t/d @50 MPa) - Liquid hydrogen (574 t/d) Hydrogen production at off-peak time. (L-H2 for 100 t/d & 824 MWe) Diverter Fusion output (3 GW) Electric output (1 GW) Helium loop Pure water Waste Heat Pure Water (221 MW) (6354 tons) Steam Electrolyzer Electrolysis (0.892 GW) Hydrogen Gas Compression (50 MPa) (0.108 GW) Electrolysis (0.82 GW) Liquefaction (0.18GW) Electrcity (1 GW) Electrolysis (150 MW) Liquefaction (26 MW) Electrcity (824MW) Liquid Hydrogen Helium Turbine Generator Electricity Hydrogen Steam Electrolyzer 100 t/d 625 t/d 574 t/d A.Sagara-24/26

System Design Code Design Windows Analysis based on a Cost Model of Helical Reactor 17th TOFE Y. Kozaki, A. Sagara, S. Imagawa Standard Design Case 3GW Standard Plant FFHR2m1 Design Study Cost Comparison to Previous Works Weight-Cost Analysis Magnet Cost Analysis based on ITER Using the ITER magnet cost data we evaluated the cost of helical reactor magnet, which show the similar coil weight and cost between Helical and Tokamak. Magnets Weight and Cost of Tokamak and Helical Reactor Design Windows Analysis PlasmaMagnet Blanket Identifying Critical Parameters Sensitivity Study for taking advantage of Helical Reactor Features Comparative Economics Studies for Fusion Plants Mass-Cost Estimating Model (HeliCos) Mass- Cost Database A.Sagara-25/26

Summary 1. The FFHR design is now on optimization phase. 2. Heating power minimization and a flexible 3D neutronics method have been developed for helical system. 3. Blanket design and divertor design are strongly coupled in helical system. 4. New engineering targets on magnets, blanket concepts and plant system have been proposed. 5. Plasma operation devices on fueling, heating, and He ash removal etc. can be designed based on LHD results. A.Sagara-26/26