24D12
Power plant performance optimization of tokamak fusion reactor
with design codes
設計コードによるトカマク型核融合炉の発電プラント性能における最適化
1Akira Furuya, 1Yuya Miyoshi, 2Hiroyasu Utoh, 1Kunihiko Okano and 1Yuichi Ogawa
1
古谷晃，1三善悠矢，2宇藤裕康，1岡野邦彦，1小川雄一

1

The Graduate School of Frontier Science, the University of Tokyo 5-1-5, Kashiwanoha, Kashiwa 277-8568, Japan
2
Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori pref.
1
東大新領域 〒277-8568 柏市柏の葉5-1-52
日本原子力研究開発機構 青森県上北郡六ケ所村
In Tokamak fusion reactor, plasma and coil performance is in a very close relationship. So far, we havedeveloped design code that can analyze the amount of power in self-consistent parameters. The code
calculates plasma physics performance with the toroidal field coils from the radial build on the basis ofengineering constraints. Then, analysis was performed with a particular focus on the net electric power at
the transmission end, and we have considered an optimal design window for DEMO plants.

1. Introduction
As a next step of the experimental reactor(ITER), the DEMO plant is planned to validate
the integrated plant-system to generate net
electric power. In this study, possible design
window for the DEMO plant has been
investigated using a design code based on the
expected performance in plasma physics with amagnetic field that is computed by the conditions
on the coil engineering expected in near future.

parameters of the plasma, where the density,
temperature and current profile are given as theinputs. The coil design is used to calculate various
parameters from plasma physics, and thecalculation of the coil dimensions is achieved
automatically. It is also possible to estimate theVolt-Sec capacity of central solenoid (CS) and thepossible period for inductive (pulsed) operation if
the CS capacity exceeds the minimum requirement
for initial current ramp-up.

2. Research methods
In the plasma physics code developed by
DRIVER of CRIEPI [1], the plasma performance isdetermined self-consistently based on current
distribution, plasma density etc., by using anautomatic external current and power source, suchas NBI current drive. The current distribution isadjusted to the specified current profile. In theDRIVER code, this profile can be determined byMHD equilibrium / stability analysis, but in thisstudy, the profiles have been given as the input.
The shape of the toroidal field coils is determined
under the same condition as the parameter values
used in the calculation. Then the magnetic field at
the major radius is determined under someengineering constraints based on their shapes [2,3].
Self-consistent analysis of power output from thecoil have been done to unify and integrate the code
parameters as described above.
3. Design code
The diagram of design code is shown in Fig.1.
This code is designed to obtain the self-consistentFig.1. diagram of design code

4. Result
The parameters used here for the TF coil design
are equivalent to the SlimCS design by JAEA [4],shown Table I. The size and aspect ratio of a fusion
reactor have been scanned and the relationshipbetween the magnitude of the output are obtained(Fig.2). In this calculation, the plasma densities
are controlled automatically to keep theconditions βN<0.75βNmax and ne/nGW≤1, where amodel of βNmax formulated in [5] has been adopted.
In this formula, higher βNmax can be expected withlower aspect ratio, as well as with higher elongation.
The safety factor q is fixed to the specified values
in each calculation (3.0 ~ 5.0). The relationshipbetween the major radius and center of the magnetic
field strength is shown in Fig.2. With a larger major
radius, the maximum magnetic field on coilbecomes smaller, and the difference between themaximum magnetic field and the center becomes
smaller. Therefore the curves of the magnetic fieldstrength on the plasma center have peaks, as shownin Fig.2.

no doubt more difficult than the inductive ramp-up.
In the right hand side of figure (i.e., larger major
radius), the weight of coils naturally increases.
Considering those restrictions, reasonable design
points will exist in the central area.

Fig.2. major radius and magnetic field strength on theplasma center

Table I. Design parameters

Fig.3. Major radius and net electric output

The relationship between the net electric output
and the major radius is shown in Fig.3, where
Ti=16keV and qΨ=4 have been assumed and aspectratio of Tokamak, A, has been scanned from 2.0 to4.0. With A=2, external power is too much and net
electric output goes negative value. In the lower
yellow area, HH98(y,2) exceeds 1.3. In the upper pinkarea, the wall load exceeds 3MW. In the left blue
area, the capacity of CS coils is less than LpIp, andtherefore non-inductive current ramp-up (NICR)
will be necessary. This NICR will be possible, butReferences[1] K. Okano: J. Society for Simulation Technology 22(2003) 83-88.
[2] H. Utoh, T. Isono, M. Hasegawa, K. Tobita and N.
Asakura: Plasma Fusion Res. SERIES, Vol. 9 (2010)304-309[3] T. Isono, N. Koizumi, K. Okuno, R. Kurihara, S.
Nishino, K. Tobita: Fusion Engineering and Design
81 (2006) 1257-1261.
[4] K. Tobita et al.:Conceptual Design of the SlimCS
Fusion DEMO Reactor (JAEA-Research 2010-019)[5] Y.R. Lin-Liu and R.D. Stambaugh: Nucl. Fusion44(2004) 548-554.