III. Fusion Reactor Design Study
1. Progress in Compact DEMO Reactor Study

The design study of fusion DEMO reactor based on a slim CS (center solenoid coil) concept has been continued and still now be in progressing way. The slim CS concept leads to low aspect ratio plasma. The lower aspect ratio plasma is considered to have higher plasma performance (high elongation plasma and high beta) and then is considered to be preferable from viewpoint of economical aspect. The size of CS coil (radius of 75cm) is decided from the minimum magnetic flux requirement of 20 volt-second.

The SlimCS reactor 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 toroidal field of 16.4T.

Among the design progress issues, two issues concerning a plasma profile consistency and a tritium breeding are given in the followings.

Since plasma parameters are determined by a systems code based on a point model, the parameters should be checked for correctness using a one-dimensional (1-D) code. For this purpose, an ACCOME code [1-1] was used to investigate the consistency between the assumed plasma profiles and key parameters such as Pfus, βN, n/nGW, HHy2 and fBS. In the 1-D analysis, we attempted to find a solution with q-profile other than strongly reversed shear (RS). This is because strong RS does not seem to be appropriate as the standard operation mode of SlimCS from the points of view of disruptivity and the controllability of q in the central region. Figure III.1-1 shows a solution for weakly RS [1-2]. Although the profile is not reasonably optimized because of a single ECRF beamline, the following information was obtained from the analysis:

  1. Most of the design parameters by the point model are consistent with the 1-D calculation;
  2. The location of NBCD is restricted to the peripheral region because of beam attenuation even for 1.5 MeV NB;
  3. Use of ECRF as the main CD tool requires a high CD power due to its lower current drive efficiency than NBCD.
In addition, the calculation indicates that q-profile control in the central and peripheral regions is important to maintain the bootstrap fraction around the design value for various density and temperature profiles. For example, suppose that the BS current around an internal transport barrier is dominant. Then fBS is strongly dependent on q-value around the ITB, i.e., the total current driven inside the ITB. For this sense, q-profile control can be key technology to maintaining fBS at a design value in fusion plasma especially with high fBS. In connection to this, the interplay between q-profile and pressure profile (including ITB structure) will be an essential issue governing the controllability on fBS. A concern about the analysis is consistency between the obtained q-profile and the given density/temperature profiles. This is an open question to be resolved with further understanding on plasma transport. The result that even 1.5 MeV NB deposits in the peripheral region suggests the necessity to reconsider the role and beam energy of NBI. A possible idea for it is to use lower energy (e.g. ~0.5 MeV) NBI for driving plasma rotation as well as peripheral beam current.


Another issue is a systematic analysis of tritium breeding ratio (TBR) based on a one-dimensional neutronics code THIDA with FENDL2.0 library. Findings of the analysis are:

When the actual tritium production is larger than expected, for example by 5%, surplus production of tritium amounts to about 25 g/day, which will be extracted from the fuel cycle system and stored in the on-site fuel storage. Since the surplus production of tritium reaches 9 kg for one-year operation, an in-situ TBR control method will be required to avoid excessive production. Borated-water is promising for the purpose in that water borated with 0.7 wt% of H3BO3 reduces the local TBR by 0.07 as shown in Fig.III.1-2, corresponding to a reduction of 0.05 in the net TBR. This indicates that borated-water injection into coolant is useful as in-situ TBR control.

References
1-1 Tani K., et al., J. Comput. Phys. 98, 332(1990).
1-2 Tobita K., et al., Nucl. Fusion 47, 892(2007).

2. Numerical Study on Beta Limit in Low Aspect Ratio Tokamak

The critical beta of low aspect ratio tokamak for toroidal mode number n=1 is analyzed by using MARG2D code developed in JAEA, which solves 2-dimensional Newcomb equation as an eigen-value problem [2-1].

The maximum stable beta value is considered to be mainly limited by RWM mode that is induced by ideal kink ballooning mode, and we analyze the ideal kink ballooning mode as a first step. As for the equilibria, typical up-down symmetric configuration of SlimCS (elongation : κ=2.0, triangularity : δ=0.36) is used, and the profiles of plasma current and pressure are obtained from the correlation function using the experimental profile data of JT-60U plasma with ITB (Internal Transport Barrier) [2-2], where the minimum values of safety factor are set to be greater than 3. Critical normalized beta 2.7 is obtained in the case of no ideal wall. To obtain stable βN=4.3 SlimCS plasma, the conformal ideal wall must be placed at the position of 1.2 times plasma minor radius (Fig.III.2-1). Profile optimization is necessary to obtain higher critical normalized beta.

References
2-1 Aiba N., et al., Comp. Phys. Commun. 175 269 (2006).
2-2 Kurita G., Nagashima K., et al 1998 15th Annual Meeting of Japan Society of Plasma Science and Nuclear Fusion Research.

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