A. Sagara

J. Miyazawa

N. Yanagi

T. Goto

T. Tanaka

National Institute for Fusion Science (NIFS-Repository)

    In accordance with a steady progress in the LHD ex-perimental research, a lot of achievements have been made in terms of a refinement of the database, physics analyses, and engineering R&D for helical systems. This study ad-vances a conceptual design of the helical DEMO reactor FFHR-d1 by utilizing these achievements and by cooperat-ing with wide-ranged researches of the core plasma and the reactor technology through cooperative researches in NIFS. This study also aims at establishing an engineering basis that enables an engineering demonstration for a DEMO and contributing to a progress in the nuclear fusion research by clarifying issues and prospects of each research field.  This study has been advanced under a new research project, fusion engineering research project, launched at the beginning of this fiscal year. A conceptual design activity for FFHR-d1 and related engineering R&D have started by 3 research groups (superconducting magnets, in-vessel components, reactor system design) that consist of 13 task groups.  The design integration task group has conducted de-sign window analyses for FFHR-d1 using a system design code.1) A design with a smaller size and higher magnetic field strength compared with a reference design (FFHR-2m22)), which aims at a commercially-competitive power plant, is being considered (see Fig. 1).  The core plasma task group has proposed a new method for extrapolating the plasma parameters obtained in the LHD experiments to a reactor condition, direct profile extrapolation (DPE) method, to enhance a reliability of the core plasma design.3)  The superconducting magnet group has proceeded an upgrade design for the existing high-field testing device, low-temperature structural analysis, and so on. To reduce the size of a reactor, the blanket task group has carried out an optimization study for a thickness of the shielding layer under neutronics calculations by selecting advanced shielding materials (e.g., WC) and an adequate placement of materials (see Fig. 2). This group also has carried out an optimization study for the internal structure of the blanket system by coupling neutronics calculations with thermo-fluid and structural analyses.  The tritium task group has established particle bal-ance model that can deal with an entire plant system4) (see Fig. 3). This model revealed that fueling efficiency has a great influence on particle balance. In relation to this, the fueling task group has started a development of a calcula-tion model for a steady-state fuel supply and an evaluation using the model.   Other task groups (heating system, diagnostics, pow-er supply system, etc.) have started a consideration for de-velopment issues.  Some new ideas that utilize advantages of helical systems (e.g., a consideration of heating scenario for a high-density operation5), a new divertor sweeping scheme using HTS coils6) to reduce particle and heat loads on di-vertor plates) have also been proposed.                         Fig. 1.  Design window analysis for FFHR-d1.                      Fig. 2. Result of the neutronics calculation .                 Fig. 3. Schematic view of the tritium balance model.  1)  Goto, T. et al.: Proc. of ITC20, Toki, Japan, Dec. 7-10, 2010, O-23, to be published in PFR (2011). 2)  Sagara, A. et al.: Fusion Eng. Des. 83 (2008) 1690. 3)  Miyazawa, J. et al.: Proc. of ITC20, Toki, Japan, Dec. 7-10, 2010, P2-63, to be published in Fusion Eng. Des. 4)  Sagara, A. et al: Plenary Talk at 19th TOFE, Las Ve-gas, USA, 2010, Fusion Sci. Technol. 60 (2011) 3. 5)  Mitarai, O. et al: Proc. of 23rd IAEA Fusion Energy Conference, Daejeon, Korea, Oct. 11-16, 2010, FTP/P6-19. 6)  Yanagi, N. et al: Proc. of 23rd IAEA Fusion Energy Conference, Daejeon, Korea, Oct. 11-16, 2010, FTP/P6-21, submitted to Nucl. Fusion. i) Development of system design code for LHD-type heliotron reactorsA helical system with a net-current-free plasma hassuitable properties for a DEMO and a commercial fusionreactor. Especially, heliotron system with two continu-ous helical coils has achieved several remarkable achieve-ment in the experiments in the Large Helical Device(LHD)1). Therefore, a design feasibility study on anLHD-type heliotron DEMO reactor is quite meaningful.For the quantitative understanding of the relationbetween design parameters, a system design code forLHD-type heliotron reactors, HELIOSCOPE (Heliotronsystem design code for reactor performance evaluation),has been developed2). The most important and diﬃ-cult issue is an evaluation of the shape of the magneticsurfaces. The shape of magnetic surfaces of heliotronreactors is strongly coupled to the geometry of external(helical) coils. Therefore, the parameters related to thegeometric conﬁguration of the magnetic surfaces, whichare needed to evaluate plasma performance, cannot begiven as input parameters but need to be obtained froman equilibrium calculation. Such an equilibrium calcula-tion, however, is time-consuming and cannot achieve thecomputational speed fast enough for an application onparametric scans (less than 1 sec for one parameter set).For this reason, database of the magnetic surface con-ﬁgurations for various shapes of the helical coils with aﬁxed major radius Rc has been established separately us-ing a ﬁeld line tracing code and the 3-D equilibrium codeVMEC. HELIOSCOPE refers this database and appliesit to diﬀerent values of Rc by a similar extension.ii) Design window analysis for heliotron DEMOreactors Design windows of a fusion reactor is de-termined by balancing various design factors. In thedesign study of FFHR-2m23), which is aiming at acommercially-competitive helical fusion reactor, the fol-lowing engineering constraints has been assumed: storedmagnetic energy Wmag ≤ 160 GJ to wind helical coilsbased on the ITER-relevant technology, an average neu-tron wall load ⟨Γnw⟩ ≤ 1.5 MW/m2 to suppress neutrondamage on blanket structural materials < 100 dpa af-ter 30 full power years (FPYs) operation with the long-life blanket concept4), and the inboard minimum blanketthickness ∆in ≥ 1.0 m to enable a simultaneous achieve-ment of suﬃcient shielding of the coils from fast neutronsand a net tritium breeding ratio ≥ 1.05. Thus, Rc = 17m and Bt,c = 4.7 T (an average toroidal ﬁeld strengthon the helical coil winding center) has been selected as acandidate design point for FFHR-2m2. This design pointcan be realized when a conﬁnement improvement rela-tive to the present LHD experiments of HLHD = 1.3 isachieved with an assumption on physics conditions: den-sity and temperature properties as ne = ne0(1− ρ2)αn ,Te = Te0(1− ρ2)αT with αn = 0.25 and αT = 1.0, al-pha heating eﬃciency ηα = 0.9, and helium ash fractionfα = 0.05. However, HLHD depends strongly on thesephysics conditions and the dependence does not changeso much by Rc, Bt,c, and a fusion power. Therefore, itis important to ﬁnd a design window with a lower HLHDto secure the design robustness.Figure 1 shows a design window of an LHD-type he-liotron reactor with the same assumption in the physicsconditions as those of FFHR-2m2. It can be seen thata reduction of the blanket space by ∼ 20 cm can reduceHLHD from 1.3 to ≤ 1.2 with keepingWmag. This reduc-tion can be achieved by the use of an advanced shieldingmaterial (e.g., tungsten carbide). Since the ﬁrst prior-ity of a DEMO reactor is a demonstration of a steady-state, self-ignition plasma operation, a long-term (∼ 30FPYs) operation and a fusion output of 3 GW are notnecessary conditions. Thus, the design point plotted asa circle in Fig. 1 is a possible selection as a DEMO reac-tor. If a higher ⟨Γnw⟩ is accepted, HLHD becomes moresmaller. On the other hand, Wmag can be reduced ifHLHD ≥ 1.3 is achieved. The reduction of the blanketspace also enables a reduction of helical coil current den-sity or a ﬂexible selection of a magnetic conﬁguration(e.g., low plasma aspect ratio, etc.). Consequently, itenhances design robustness of a DEMO reactor.Fig. 1: The design window of an LHD-type heliotronreactor with the same physics constraints as those ofFFHR-2m2 (plotted as a star symbol). The circle is apossible design point for a DEMO reactor if the inboardminimum blanket space of 80cm is acheived.1) A. Komori et al., Fusion Sci. Technol. (2010) vol. 58pp.1-11.2) T. Goto et al., to be published in Nucl. Fusion.3) A. Sagara et al., Fusion Eng. Des. (2006) vol. 81pp.2703-2721.4) A. Sagara et al., Nucl. Fusion (2005) vol. 45 pp.258-263.229§2. System Design of an LHD-type Heliotron DEMO ReactorGoto, T., Miyazawa, J., Tanaka, T., Yanagi, N., Sagara, A.

System Design of an LHD-type Heliotron DEMO Reactor

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System Design of an LHD-type Heliotron DEMO Reactor

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