S14-4
Inevitable Measurement Parameters for the Control of Fusion DEMO Reactors
原型炉の制御と必須な計測諸量
Akira Ejiri, Tsuyoshi Akiyama1
江尻晶, 秋山毅志1
Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa 277-8561, Japan東京大学大学院新領域創成科学研究科 〒277-8561千葉県柏市柏の葉5-1-51National Institute for Fusion Science, Toki 509-5292, Japan1
自然科学研究機構 核融合科学研究所 〒509-5292岐阜県土岐市下石町322-6
The diagnostics in fusion reactors are quite important to control burning plasmas. However, due to theradiation environment and due to the limited access area, we should select the minimum and inevitablediagnostics. In this report, a list of such diagnostics is presented assuming a tokamak type and a helicaltype reactors. The number of diagnostics is reduced to a reasonable level (10 sets) considering the available
access area in DEMO reactors.

1. Introduction
In order to operate fusion reactors safely andefficiently, the plasma should be controlled
precisely. For that purpose, various diagnostics
have been developed. However, due to technicaldifficulties, such as the radiation environment, wemust select suitable diagnostics or we must
develop them. In addition, not to degrade thetritium breeding performance,
the available
access area for diagnostics is very limited.
Therefore, we should find out what are theinevitable diagnostics for fusion power plants,
such as DEMO devices. It is expected that theoperation scenario is fixed in DEMO devices andthe behavior of the plasma is known. Therefore,
the diagnostics will be used to identify andcontrol the state of the plasma.
In this report, we list up inevitable diagnostics,
assuming a tokamak type and a helical type
DEMOs considering those situations mentioned
above.
2. List of Diagnostics
Table I shows the list of inevitable diagnostics,
which based on the discussion in Ref. [1] and thediagnostics in ITER. Some additional modifications
have been made. In the table, parameters to bemeasured, purposes (i.e. control targets), methods,
numbers of measurement points, required
accuracies and time resolutions are shown.
Hereafter, we describe each item briefly.
Position and shape control of magnetic flux
surface is a principal control for tokamaks, while
they are determined mainly by external coils inhelicalsystems.
Thus,variousmagnetic
measurements are inevitable for tokamaks.
Although loops and coils are simple instruments forthese purposes, some problems must be solved. One
such problem is the drift of time-integrated signals
due to radiation induced offsets on the raw signals.
The effect becomes unavoidable in the longdischarge durations. Therefore, some DC fieldmeasurement method (e.g. Hall sensor) should beequipped and used for the calibration of the loops.
On the other hand, the role of magnetic
measurements in helical devices is not sosignificant as in tokamaks.
Once the position and the shape are controlled,
the plasma should be fuelled and heated to sustain aburning state. The state can be represented bytemperature and density, while their responses tothe heating source (and the power loss) and theparticle source (and the particle exhaust) can berepresented by various transport rules such as 0-d
energy confinement scalings. The resultant burningstate must be monitored and controlled bymeasuring neutrons. We should also pay attention
to various instabilities, such as beta limits, densitylimits.
In order to minimize the access area of thediagnostics, not only the number of the types ofdiagnostics, but also the numbers of channels
should be minimized. If one can predict the profile,
one can reduce the number of channels. In the caseof tokamak, where H-mode and/or internal transportbarriers are considered to be the inevitable core
structures, at least a few spatial measurement points
are required to identify the profile. On the otherhand, in the helical reactor design FFHR-d1,profiles are a function of a base profile and fewer
parameters are required to identify the profile. The
resultant numbers of spatial points become much
lower than those in ITER, assuming suchpredictability.

Since tokamak reactors rely on bootstrap current,pressure profile should be speculated from limited
information. In helical systems, edge density isimportant to avoid radiation collapse, while intokamaks, H-mode pedestal pressure is important tocontrol ELMs and fast time-response diagnostics
are require to monitor the ELMs.
Divertor plasmas should be measured from thefollowing two points of view. The divertor plasma
is a particle (D, T, He and impurities) exhaust
and/or influx route to the core plasma. In addition,
detachment and radiation should be controlled toreduce the heat flux to the divertor plates.
3. Discussion and Conclusion
Besides the radiation environment, available
access area for diagnostics at the plasma facing
surface is very limited to preserve the tritium
breeding area. The available area is thought to beless than 10 m2, while the area in ITER is about 50
m2. Excluding the magnetic measurements, thenumber of diagnostics in Tab. I is 10 sets (markedwith * in the table) for tokamak, while that in ITER
is about 40. Therefore, the number in Tab. I isreasonable from the view point of the area.
However, the number of spatial points for eachdiagnostics is reduced by one-two orders ofmagnitude. Thus, it is quite challenging to identifyand control the plasma with such little information,
and we need experiences and a predictable model
for burning plasmas.
Acknowledgments
We would like to thank the participants of theworkshops on the Diagnostics for Control of the Fusion
DEMO Reactors for valuable information anddiscussions.
References[1] Program Committee of Technical Study on theDiagnostics for Control of the Fusion DEMOReactors: NIFS-Memo-68 (2014).

Table I. List of diagnostics
Measured parameters
Plasma current
Plasma current
Poloidal flux
Local BZaxbp
DC magnetic field

Purpose(Conrol/Measurement Target)
Equilibrium
VDE detection
Position, Shape, Equilibrium,
Instability dection (e.g. RWM)
VDE detection
Equilibrium
Calibration of magnetic
measurements

RogowskiRogowski
Flux loop, Saddle loop
Pickup coil
Flux loop
Diagmag. Loop

No. (Units, Lines, Points)/Accuracy/Dt
Tokamak
Helical
1/1%/1ms
1/10%/1s
1/30%/0.1ms
100P/ 0.7%/10ms20/?/0.1ms
20/?/0.001-1ms
20/?/0.001-1ms
6P/30%/0.1ms
1P/0.3%/10ms-Hall sensor

200P/0.5%/100s

Diagnostics-Tangential or radial
Interferometer (and/or2L/1%/10ms
1L/0.1%/10ms
polarimeter)
LIDAR Thomson
Electron density
Radiation collapse, Detachment
10P/1%/100ms
and/or Reflectometer
Pressure profile（Bootstrap
Edge:Thomson, Center:
Electron temperature*
2P/5%/100ms
current）
VUV
Radiation collapse, Detachment,
10P+10P/10eV/100
Electron temperature
LIDAR Thomson
Mag. Axis
msPressure profile（Bootstrap
Edge:CXRS,Ion temperature profile*
2P/5%/1s
current）, Aux. heating
Center:Crystal
Edge and divortor line
ELM・H-mode trans., Edge D/TSpectrometer
10L/10%/0.1 ms
5L/10%/10msemiss. (D, T, W, He, Ar)* ratio, Detachment
Divertor Radiation*
Divertor, Detachment, Impurity 2D-Imag, bolometer
1U/?/1s
Divertor particle/heatElectrostatic probe?
Divertor, Detachment
20/10%/10ms
20/10%/10ms
flux*
IR-camera?
Line av. Zeff*
Core impurity, Core He
Vis. Spectrometer
１L/20%/1s
１L/20%/1s
EUV radiation/Bremss. Radiation Collapse
EUV spectrometer?
1/10%/100ms
DT neutron*
Burning state
Neutron (camera)
20L/5-10%/100ms 20L/5-10%/100ms
Neutron spectometer
DD neutron*
Core nD/nT
1L/10%/1s
1L/10%/1s
or g spectrometer?
Micro-fission
Total neutron yield*
Plant, Calibration
5P/10%
5P/10%/100ms?
Chamb./Activation Foil

Line averaged density*

Fuelling, Pressure profile（Bootstrap current）