III. FUSION REACTOR DESIGN STUDY
1. Conceptual Design of DEMO Reactor

Fusion DEMO plant is requested to demonstrate 1) an electric power generation of 1GW level, 2) self-sufficiency of T fuel, 3) year-long continuous operation. From the economical aspect, the reactor size should be as compact as ITER. To meet these requirements, a DEMO reactor concept named SlimCS was proposed in 2005 [1-1].

SlimCS produces a fusion output of 2.95GW with a major radius of 5.5m, aspect ratio (A) of 2.6, normalized beta (βN) of 4.3 and maximum field of 16.4T. The conceptual view is depicted in Fig.III.1-1. It is expected that the zero output at the sending end is obtained at βN = 2, n/nGW = 0.4 and fBS = 0.35 and that a step-by-step power-up eventually attains 1GWe output at βN = 4.3, n/nGW = 1.1 and fBS = 0.77, where n/nGW and fBS are the line-averaged electron density against the Greenwald density and the bootstrap fraction, respectively. SlimCS uses technologies foreseeable in 2020's such as Nb3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel F82H as the blanket structural material, and tungsten monoblock divertor plate. Neutron wall load is designed at 3MW/m2. Divertor heat flux, which can be a critical issue for such a compact reactor, is mitigated to 10MW/m2 at the peak by small inclination (15) of divertor plates and flux expansion in the divertor reg ion.


SlimCS can be as compact as advanced commercial reactor designs such as ARIES-RS and CREST (Fig.III.1-2), eIIIen with the assumption of relatively conservative plasma parameters. This is because such a low-A plasma, being stable for higher elongation (κ), can have higher nGW and βN limits. Another merit of low-A is that the first wall area on the low field side, where smaller electromagnetic (EM) force acts on disruptions, is wide compared with that of conventional-A. This means that tritium can be efficiently breeded with large blanket modules on the low field side. As a result, the demand for tritium breeding on the high field side is comparatively reduced so that small blanket modules, being robust to stronger EM force but less efficient for tritium breeding, can be arranged on the side.

References
1-1 Tobita, K., et al., Fusion Eng. Des., 81, 1151 (2006).

2. Non-Inductive Current Ramp Simulation

From the practical control aspect of a compact, CS-free tokamak reactor concept "VECTOR", a fully non-inductive, very slow current buildup scenario were investigated via a consistent simulation using Tokamak Simulation Code [2-1, 2-2, 2-3]. The L-mode based, improved core confinement transport model, e.g. current diffusive ballooning mode (CDBM), has clarified detailed dynamics of the stable formation of the internal transport barrier (ITB) by non-inductive means of off-axis current sources. First, in accordance with the strong ITB formation, the bootstrap (BS) current was confirmed to substantially increase by more than fbs > 50% and to enhance the current buildup efficiency, saving a great deal of the driving power of the non-inductive current sources. Second, the integrated, non-inductive scenario was show n to meet the following control and physics requirements set by (a) plasma shaping compatible with recharging of the coil currents, (b) available NB-heating power, (c) avoidance of Current Hole formation under over driving, non-inductive current sources, (d) reasonable HH factor = tE/tE,y2 < 1.3 and (e) allowable Greenwald density limit of n < nGW. Third, a safe plasma takeoff from limiter to diverter configuration, as well as a safe landing to limiter structures at discharge termination, was also demonstrated. Furthermore, a new operation scenario was computationally examined to control the ITB structure by means of small, but long-duration perturbation (~ 80sec in reactor plasmas) of negative or positive inductive current sources. Thus, the q-profile was first shown to undergo a drastic change over a wide range from positive to negative magnetic shear configuration, and vice versa.

References
2-1 Nakamura, Y., et al., "Simulation Modeling of Fully Non-Inductive Buildup Scenario in High Bootstrap Current Tokamaks without Center Solenoids," Proc. 32nd EPS Conf. on Plasma Phys., P2-051 (2005).
2-2 Nakamura, Y., et al., "Non-Inductive Operation Scenario of Plasma Current Ramp-Down in CS-Less, Advanced Tokamak Reactor," Proc. 15th International Toki Conf. on "Fusion & Adv. Technol.," PS2-22 (2005).
2-3 Nakamura, Y., et al., "Computational Study of Non-Inductive Current Buildup in Compact DEMO Plant with Slim Center Solenoid," 1st IAEA TM on First Generation of Fusion Power Plants -Design and Technology-, PPCA1-V (2005).

3. Study of Advanced Shield Materials

In general, a hydrogen-rich material has the potential to be an effective neutron shield because the contained hydrogen nuclei work as a moderator of fast neutrons, reducing the fast neutron flux. It is nota ble that some hydrides have a considerably higher hydrogen content than polyethylene, water and solid hydrogen. The material that we have focused attention on is borohydrides which has been developed for a fuel cell [3-1]. The anticipated hydrogen concentration of Mg(BH4)2, which will probably be a new candidate shielding material, is as high as 1.32x1023 H-atoms/cm3, surpassing those of already known VH2 (1.05x1023 H-atoms/cm3) and TiH2 (9.1x1022 H-atoms/cm3).

In order to assess capability of such hydrides as a advanced shield material, neutronics calculation was carried out for the SlimCS design [3-2]. In the design, the shields are located on the inboard and outboard sides, and originally they were designed to be 30 and 70cm in thickness, respectively, using steel-and-water. When the steel-andwater is replaced with steel-and-hydride, it was found that Mg(BH4)2, TiH2 and ZrH2 could reduce the thickness of the outboard shield by 23, 20 and 19%, respectively. When Mg(BH4)2 is mixed with ferritc steel (F82H) at the ratio of 1:1, the gamma-ray flux is reduced to 1/300 compared with that for pure Mg(BH4)2. These results indicates that borohydrides in conbination with steel can work as an attractive shield material for fusion.

References
3-1 Orimo, S., et al., Mater. Sci. Eng., B 108, 51 (2006).
3-2 Hayashi, T., et al., Fusion Eng. Des., 81, 1285 (2006).