Proceedings of ITC/ISHW2007

Design Integration toward Optimization of
LHD-type Fusion Reactor FFHR
A. Sagara, O. Mitaraia, S. Iamagawa, Y. Kozaki, T. Watanabe, T. Tanaka, N. Yanagi,
K. Takahata, H. Tamura, H. Chikaraishi, S. Yamada, T. Mito, A. Shishkinb,
Y. Igitkhanovc, H. Igami, T. Morisaki, M. Kobayashi, R. Sakamoto,O. Motojima and FFHR design group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japana

Kyushyu Tokai University, 9-1-1 Toroku, Kumamoto 862-8652, Japanb

Kharkov Institute of Physics and Technology, Kharkov-108, Ukrainec

Max-Planck-Institut fur Plasmaphysik, IPP-EURATOM Ass., Greifswald, Germany(Received/ Accepted)

Recent activities on optimizing the base design of LHD-type helical reactor FFHR2m1 is presented. Three
candidates to secure the blanket space are proposed in the direction of reactor size optimization without
deteriorating α-heating efficiency and with taking cost analyses into account. For this direction the key engineering
aspects are investigated; on 3D blanket designs, it is shown that the peaking factor of neutron wall loading is 1.2 to
1.3 and the blanket cover rate over 90% is possible by proposing Discrete Pumping with Semi-closed Shield(DPSS) concept. Helical blanket shaping along divertor field lines is a next big issue. On large superconducting
magnet system under the maximum nuclear heating of 200W/m3, CICC and alternative conductor designs areproposed with a robust design of cryogenic support posts. On access to ignited plasmas, new methods are proposed,in which a long rise-up time over 300 s reduces the heating power to 30 MW and a newproportional-integration-derivative (PID) control of the fueling can handle the thermally unstable plasma at highdensity operations.
Keywords: helical reactor, blanket, COE, nuclear heating, superconducting magnet,

ignition

1. Introduction
On the basis of physics and engineering results
established in the LHD project [1], conceptual designs of
the LHD-type helical reactor FFHR have made continuous
progress from 1991 [2-5], aiming at making clear the key
issues required for the core plasma physics and the power
plant engineering, by introducing innovative concepts
expected to be available in this coming decades. Those
design activities have led many R&D works withinternational collaborations in broad research areas [6].
Due to inherent current-less plasma and intrinsic
diverter configuration, helical reactors have attractive
advantages, such as steady operation and no dangerous
current disruption. In particular, in the LHD-type reactor
design, the coil pitch parameter γ of continuous helicalwinding can be adjusted beneficially to reduce themagnetic hoop force (Force Free Helical Reactor: FFHR)
sagara.akio@LHD.nifs.ac.jp

while expanding the blanket space, where γ=(mac)/(lRc)
with a coil major radius Rc, a coil minor radius ac, a pole
number l, and a pitch number m.
As a key feature of helical reactors, the blanket space
directly couples to the helical coils configuration as well as
the core plasma performances under physics andengineering key constrains. Therefore, as the second stepafter concept definition of the initial FFHR1 (l=3) design[2], optimization studies have begun on the reactor size,
based on the LHD-type (l=2, m=10) compact design
FFHR2 (γ=1.15, Rc=10 m) [3]and modified FFHR2m1(γ=1.15 and outer shifted plasma axis, Rc=14 m) andFFHR2m2 (inward shifted plasma axis, Rc=17 m) [4].
This paper presents recent activities on optimizing
FFHR2m1 as a base design to make clear key issues,
mainly focusing on blanket space, neutronics performance,large superconducting magnet system, and plasma
operation.

Proceedings of ITC/ISHW2007
Table 1. Design parameters of helical reactor
Design parameters
LHD
Polarityl2
Field periodsm
10
Coil pitch parameter
γ
1.25
Rc
Coil major Radius
m 3.9
ac
m 0.98
Coil minor radiusRp
m 3.75
Plasma major radius
ap
Plasma radius
m 0.6130
Plasma volume
Vpm3
Blanket space∆
m 0.12
B0
Magnetic field
T
4
B max
T 9.2
Max. field on coils
53
Coil current density
j MA/m 2
Magnetic energy
GJ 1.64
PF
Fusion power
GW
Γn MW/m 2
Neutron wall load
External heating power PextMWηαα heating efficiency
Density lim.improvement
H factor of ISS95
Zeff
Effective ion chargen e(0) 10^19 m -3
Electron density
Ti(0)
keV
Temperature
Plasma beta<β>%MW
Plasma conduction loss PLΓdiv MW/m 2
Diverter heat load
Total capital cost
G$(2003)COE
mill/kWh

FFHR2

2
10
1.15
10
2.3
10
1.24
303
0.7
10
14.8
25
147
1
1.5
70
0.7
1
2.40
1.40
27.4
21
1.6

4.6
155

FFHR2m1 FFHR2m2

2
10
1.15
14.0
3.22
14.0
1.73
827
1.1
6.18
13.3
26.6
133
1.9
1.5
80
0.9
1.5
1.92
1.34
26.7
15.8
3.0
290
1.6
5.6
106

2
10
1.25
17.3
4.33
16.0
2.80
2471
1.15
4.43
13.0
32.8
118
3
1.3
100
0.9
1.5
1.76
1.35
19.0
16.1
4.1
463
2.3
6.9
87

Fig.1. The 3D illustration of the FFHR2m1.

2. Candidates to secure the blanket space

Fig.2 The loss rate of alpha particle in the
FFHR2m1, where the two cases on loss
boundary are shown as a function of theposition of the magnetic axis Rax.

1016
10

JLF-1
JLF-1 (70vol.%)+B4C (30vol.%)
B4C
WC

15

1014
1012
1011
1010

(Magnet)1013
PlasmaFast neutron flux （>0.1MeV)(n/cm2/s)

The design parameters of FFHR2 are listed in Table 1,
which newly includes the recent results of cost evaluation
based on the ITER (2003) design [7]. Figure 1 shows the3D view of the FFHR2m1. In this base design, one of themain issues is the structural compatibility between blanket
and divertor configurations. In particular, the blanket spaceat the inboard side is still insufficient due to theinterference between the first walls and the ergodic layers
surrounding the last closed flux surface. To overcome thisproblem, helical x-point divertor (HXD) has been proposed
to remove the interference [8]. In this concept, very
effective screening of recycling neutrals with 99%
ionization is expected according to 3D simulations [9].
From the point of view of a-heating efficiency over
0.95, the importance of the ergodic layers has been foundby collisionless orbits simulation of 3.52MeV alphaparticles as shown in Fig.2 [10]. Therefore, threealternatives without adopting HXD are considered. One isto reduce the shielding thickness only at the inboard side.
Fig. 3 shows that the WC can reduce about 0.2 m in theshield thickness in comparison to the standard FFHR
design with B4C and JLF-1 [11]. The second is toimprove the symmetry of magnetic surfaces around themagnetic axis, without shifting the magnetic axis inward,
by increasing the current density at the inboard side of thehelical coils while decreasing at the outboard side.
Modulation of the current density can be practicallyobtained by splitting the helical coils [12, 13]. The third is
to increase the reactor size. In this case, as shown in Fig.4,
it is expected that there is an optimum size around Rc of

109

Breeding
layer

108

[Flibe+Be]

107(32cm)

Radiation
shield(60 cm)

180 200 220 240 260 280 300
Radial position（cm)Fig.3 Distributions of first neutron flux (> 0.1 MeV)in the standard FFHR blanket, depending onmaterials composition of the radiation shield.

Proceedings of ITC/ISHW2007

200

15

0.14

Wg (magnetic energy)

150

10
COE

B0

100

5
50

TCC (total capital cost)

FFHR2

5

10

0.10

Ion density: n∝(1-r2)
Temperature: T∝(1-r2)
Max. Temp.: 16 keV

0.08
0.06

p（ｒ） ∝ n2・<σv>(T)

0.04
0.02
0.00

0

50

100

150

200

250

Position (cm)FFHR2m1

15

0
20

Coil major radius Rc (m)Fig. 4

Helical source
Torus uniform source

0.12

FFHR2m2

Pf (fusion output)0

Source distribution (a.u.)

blanket space > 1 m

Wg (GJ), COE (mil/kWh)Pf (GW), TCC (G$), Bo (T)J = 25 ~ 33 A/mm2
Γn = 1.3 ~ 1.5 MW/m2

Fig. 5 Neutron source distributions used forneutronics calculations in the FFHR2m1.

Rc dependences of the fusion output Pf, thetotal capital cost (TCC), magnetic field B0 at
the plasma center, cost of electricity (COE)and magnetic energy Wg under almost same
conditions on neutron wall loading
andcurrent density J on helical coils.

First wall of blanket
on helical coilMax. 1.8
for helicalsourceMax. 2.0
for uniform
sourceAverage 1.5

15m by taking into account the cost of electricity (COE),
the total capital cost, and engineering feasibility on large
scaled magnets. More detailed and integrated optimization,
by selecting or mixing those three candidates, is one of key
next issues.

First wall of blanket
at divertor area

0
(unit : MW/m2)

Fig.6 The 3D top view of calculated neutron wallloading distributions. (72o of the torus)

3. Progress and issues on 3D blanket designs
In the direction of optimizing neutronics
performances, the 3D distribution of neutron wall loading
is basically important. Using the recently developed 3D
neutronics calculation system for non-axisymmetric helicalsystems as LHD [14], two cases of neutron sources havebeen compared as shown in Fig.5: one is a centralized
torus uniform source which represents a peaked plasma
profile, the other is a helical source which comes from atypical parabolic distributions of plasma density andtemperature in the elliptic cross section with the long andshort radii of 2.4 m and 1.8 m, respectively. Under the
averaged neutron wall loading of 1.5 MW/m2, themaximum loading for the uniform and helical source to be
2 MW/m2 and 1.8 MW/m2, respectively, at the first wall of
blankets on the helical coils as shown in Fig.6 [14].
Therefore the peaking factor is estimated to be 1.2 to 1.3.
The FFHR blanket designs have been improved toobtain the total TBR over 1.05 for the standard design ofFlibe+Be/JLF-1 and long-life design of Spectral-shifter andTritium Breeding (STB) blanket [4, 14] by enhancement of
the blanket cover rate to 80%. More increase of the cover

3D design withDPSS

Fig.7 Discrete Pumping with Semi-closed Shield(DPSS) concept, where the helical divertor
duct is almost closed with partly opened atonly the discrete pumping ports.

rate over 90% will be effectively possible by a newproposal of Discrete Pumping with Semi-closed Shield(DPSS) concept as shown in Fig.7, where the helicaldivertor duct is almost closed with partly opened at only

Proceedings of ITC/ISHW2007

Fig.9 The tentatively modified design to avoid
intersection of divertor null points withblankets, where an inward shifted plasma
configuration at γ=1.15 is selected as areference case for optimization.

Fig.8 The first neutron fluxes at the poloidal andhelical coils (a) without and (b) with the DPSS,
where the flux at the rear side of helical coils ishigh in (a) and one order reduced in (b).

4. Base design of large superconducting magnet
systemThe base design for the FFHR2m1 superconducting

Fig. 10 Concept of helical winding with CIC
conductors of current 90 kA with Nb3Al
strands, where the maximum length of acooling path is about 500 m.

101

3

Nuclear heating (mW/cm )the discrete pumping ports. This DPSS is very important
not only to increase the total TBR over 1.2 but also to
reduce the radiation effects on magnets. In fact, as shownin Fig.8, the first neutron fluxes at the poloidal coils justout side the divertor duct and at the side of the helical coils
are successfully reduced to the acceptable level lower than
1E22 n/m2 in 40 years. The total nuclear heating is also
reduced from 250kW to 40kW, which means thecryogenics power to be about 12MW and acceptable level
below 1% of the fusion output.
When the HXD is not adopted as mentioned in theprevious section, the blanket design in divertor area should
be largely modified, because the intrinsic divertor nullpoint deeply intersects the blanket [8]. Figure 9 shows thetentatively modified design, where the radial position of
the divertor area blanket is moved outward. In this case thelocation of poloidal coils should be also moved in theminor radial direction, resulting in about 11 % increase of
the magnetic energy [15]. It is noted in Fig.9 that
re-adjustment of helical blanket shaping should be a next
big issue by taking into account the blanket space,
distribution of wall loading and divertor field lines.

100

Neutron+Gamma
Gamma

10-1
10-2
10-3
10-4
-20

0 20 40 60 80 100
Radial position (cm)Fig.11 The nuclear heating distribution calculatedon the uniformed torus model of FFHR
helical coils shown in Fig.10.

Proceedings of ITC/ISHW2007

Fig.12 The present design of cryogenic support
posts for the FFHR2m1, adopting the same
type of the LHD support post.
magnet system has been preliminary proposed [16] on theengineering base of ITER-TF coils as a conventional
option. Figure 10 shows the cross sectional structure ofcontinuous helical coil, where the magnet-motive force ofhelical coils is about 50 MA and the cable-in conduit
conductors (CICC) of current 90 kA with Nb3Al strands
are wound in the grooves of the internal plates. In thisconcept, react and wind method is preferred to use
conventional insulator and to prevent huge thermal stress.
The maximum length of a cooling path is about 500 m thatis determined by the pressure drop for the required mass
flow against the nuclear heat of 1000 W/m3. This value has
a 5 times margin of the maximum nuclear heating
calculated on the FFHR helical coils as shown in Fig.11, in
which the gamma-ray heating is dominant and the
maximum is about 200 W/m3.
Advanced concepts for the FFHR magnet system is ofimportance as alternative candidates. "Indirect cooling" ispromising, because it solves the issue of the pressure drop.
The preliminary design using Nb3Sn has been proposed for
the FFHR helical coils, where a conventional quench
protection circuit using an external resistor is employed by
dividing the coil into several subdivisions [17].
The total weight of the coils and the supporting
structure exceeds 16,000 tons. This weight is supported bycryogenic support posts which are set on a base plate of acryostat vessel. Fig.12 shows the present design of the post[18] adopting the same type of the LHD support post,
which is a folded multi plates consisted of Carbon Fiber
Reinforcement Plastic (CFRP) and stainless steel plates.
The FEM analyses indicate that the LHD-type support postis also valid for the large sized device such as FFHR inmechanical and thermal points of view. The modal anddynamic response analysis using typical earthquake
vibrations are the next issue for design optimization.

Fig.13 The maximum feedback controlled heating
powers to reach self-ignition for various
fusion power rise-up times in FFHR2m1.

5. New proposals on access to ignited plasmas
Minimization of the external heating power to accessself-ignition is advantageous to increase the reactor design
flexibility and to reduce the capital and operating costs ofthe plasma heating device in a helical reactor. While thefusion power rise-up time in a tokamak depends on the OH
transformer flux or the current drive capability, any fusion
power rise-up time can be employed in a helical reactor,
because the confinement field is generated by the external
helical coils. It has been recently found that a lower density
limit margin reduces the external heating power, and over
300 s of the fusion power rise-up time can reduce theheating power from such as 100 MW to minimized 30 MW
in FFHR2m1 as shown in Fig.13 [19].
A new and simple control method of the unstable
operating point in FFHR2m1 is proposed for the ignited
operation
withhigh-density
plasma[20].
Proportional-integration-derivative (PID) control of the
fueling has been used to obtain the desired fusion powerwith the fusion power error of e(Pf)=(Pfo-Pf) in the stable
operating point. It has been discovered that in the unstableregime the error of the fusion power with an opposite signof e(Pf)= - (Pfo-Pf) can stabilize the unstable operating point.
Around the unstable operating point, excess fusion power
supplies fueling and then increases the density anddecreases the temperature. Less fusion power in the
sub-ignited regime reduces the fueling, decreases thedensity, and increases the temperature. The operating point
approaches the final unstable operating point as oscillationis damped away.

6. Summary
Recent activities on optimizing the base design of
LHD-type helical reactor FFHR2m1 is presented. ThreeProceedings of ITC/ISHW2007
candidates to secure the blanket space are proposed in the
direction of reactor size optimization without deteriorating
a-heating efficiency and with taking cost analyses into
account. For this direction the key engineering aspects areinvestigated.
On 3D blanket designs, it is shown that the peaking
factor of neutron wall loading is 1.2 to 1.3 and the blanket
cover rate over 90% is possible by proposing Discrete
Pumping with Semi-closed Shield (DPSS) concept. Helicalblanket shaping along divertor field lines is a next big
issue.
On large superconducting magnet system under the
maximum nuclear heating of 200W/m3, CICC designs of
500 m cooling path and 90 kA with Nb3Al strands andalternative Indirect cooling Nb3Sn conductor designs areproposed with the LHD-type robust design of cryogenic
support posts. The modal and dynamic response analyses
are the next issue for design optimization.
On access to ignited plasmas, using the advantage ofcurrent-less plasma, new methods are proposed, which area long rise-up time over 300 s to reduce the heating power
to 30 MW and a new proportional-integration-derivative(PID) control of the fueling to handle the thermallyunstable plasma at high density operations.

Acknowledgments
This work has been performed under the research
programs of NIFS (code ULAA116).

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