All opinions expressed herein are those of the authors and should not be reproduced, quoted in publications, or
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Oak Ridge National Laboratory is managed by UT-Battelle, LLC, for the U.S. Department of Energy.
Published by Oak Ridge National Laboratory
Building CR 5600 P.O. Box 2008 Oak Ridge, TN 37831-6169, USA
Editor: James A. Rome Issue 147 April 2015
E-Mail: jamesrome@gmail.com Phone (865) 482-5643
On the Web at http://www.ornl.gov/sci/fed/stelnews
Recent milestones in commissioning
Wendelstein 7-X
As briefly described in Stellarator News 145, the commissioning
of Wendelstein 7-X (W7-X) in Greifswald, Germany,
started in the summer of 2014 with the cryostat
vacuum system.
A major milestone in July was the first evacuation of the
cryostat vessel. In the following weeks, 29 leaks were
detected and repaired. Simultaneously, mechanical loads
and stresses on the shell structure, resulting in deformations
of the vessels and ports, and relative movement
between the cryostat and the plasma vacuum vessel were
measured during pumpdown at predefined values of 700,
300, and 50 mbar. Of special interest were the port bellows,
specifically those bellows at large oval and rectangular
ports where significant deformation is expected due
to the vacuum load. The measured stresses and deformations
were in rather good agreement with the results of the
finite element (FE) modeling and confirmed the structural
stability of the cryostat.
In fall 2014 the local commissioning of the cryo supply
was started with the recommissioning of the cryoplant, the
cleaning of all helium circuits, and the commissioning of
the machine- and cryo-instrumentation. In parallel to this,
the trim coils were commissioned (November/December
2014), the neutron counter system was calibrated (January
2015), and various water cooling circuits were commissioned
(October 2014–March 2015). In December 2014 a
Paschen test of the complete superconducting (SC) coil
system confirmed the quality of the insulation of the magnet
system.
In mid-March 2015, the cooldown of the magnet system
(50 non-planar and 20 planar SC coils with the bus bar
system and the central support ring) was started (Fig. 1).
Fig. 1. Cool-own progress of the complete superconducting
system.
In this issue . . .
Recent milestones in commissioning Wendelstein
7-X
The W7-X superconducting coils have been cooled
down to 4 K. Evacuation of the vacuum vessel has
started and the leak search is ongoing.................... 1
International Workshop on the Strategy
for Stellarator/Heliotron Research
This workshop was held 4–6 March, 2015, in Nagoya.
The purpose of the workshop was to discuss plans for
stellarator/heliotron research with a view toward a
DEMO device. Emphasis was placed on free discussion
based on the present status of theoretical and
experimental research and reactor-relevant design
studies, rather than on gaining consensus among participants.
.................................................................. 2
Announcement: 14th Coordinated Working
Group Meeting
The 14th meeting will be held in Warsaw, Poland,17–
19 June 2015. ......................................................... 6
Stellarator News -2- April 2015
The cooldown of the coils and support structure to 18 K
succeeded in just 3 weeks without showing any leaks (let
alone cold leaks). In the following week we continued to 4
K to test for thermoacoustic oscillations. In fact, no oscillations
were found. Presently, the system runs at a component
temperature of about 10 K, in the so-called shortstandby
mode (SSM), which will be used during weekends
and overnight. In this mode the required operation tests are
conducted, e.g., improving the settings of the control system,
adjusting all cooling circuits, and performing energy
balances. After that, the standard mode (SM) at about 4 K,
which will be used for magnet operation, will be commissioned.
Recently, evacuation of the plasma vessel started and the
leak search is ongoing.
Dr. Hans-Stephan Bosch
Leiter Wendelstein 7-X Betrieb
Max-Planck-Institut für Plasmaphysik
Wendelsteinstr. 1
17491 Greifswald, Germany
Tel. (03834) 88 - 2212/2000, Fax: -2709, Handy 0173 89 79 128
E-mail bosch@ipp.mpg.de
International Workshop on the
Strategy for Stellarator/Heliotron
Research
This workshop was held 4–6 March, 2015, in Nagoya. It
was organized by O. Kaneko, head of the Research
Enhancement Strategy Office, National Institute for
Fusion Science. The purpose of the workshop was to discuss
plans for stellarator/heliotron research toward a
DEMO device. This workshop may have been the first
such meeting, and emphasis was placed on free discussion
based on the present status of theoretical and experimental
research and reactor-relevant design studies, rather than on
gaining consensus among participants. Consensus will be
gradually formed in continuing workshops. Most of the
workshop participants are identified in Fig. 1. (Those who
did not attend the last day of the workshop are not in the
photograph.) The framework of the workshop was composed
of the following seven discussion points.
1. To review the fusion research program and the roadmap
to DEMO in the world, and to know the position of helical
research in this roadmap
Discussion points focused on the critical paths in the roadmap
to a tokamak DEMO, the R&D schedule and milestones,
and the risks, their mitigation, and the need for
alternatives. To address these assignments, roadmaps
toward DEMO primarily in Europe, the United States, and
Japan were briefly introduced. Presenters were C.
Hidalgo, T. S. Pedersen, H. Yamada, M. Zarnstorff, B.
Kuteev, B. Blackwell, and V. Moiseenko.
Construction of a tokamak DEMO is expected to start
around 2030 in the European Union and in Japan. Successful
ITER operation demonstrating Q = 10 is the prerequisite
for this time schedule. W7-X has an essential place in
the future of helical system research in Europe and probably
in the United States. In Japan the Committee for
Fusion Science & Technology, Council for Science Technology
and Innovation, MEXT has organized a Joint-Core
Team headed by H. Yamada to establish technological
bases required for the development of the tokamak
DEMO, maintaining a good balance between the tokamak,
and the helical system and laser fusion activities. The stellarator/
heliotron is widely considered to be an alternative
to the tokamak because of its long pulse capability and
freedom from major disruptions, for example. Research on
a pilot hybrid plant (PHP) is conducted primarily in Russia
on the basis of tokamak and molten salt technologies.
Among numerous issues, divertor geometry for heat and
ash exhausts is identified as being of particular importance.
Energetic discussion was held throughout the workshop,
such as in debating whether an ITER-like helical
device for burning plasma physics is necessary for building
the helical DEMO. Physics and technology databases
from ITER and the tokamak DEMO should be fully utilized
for the helical DEMO from the viewpoint of risk mitigation.
Lessons learned from ITER construction are
important. Perhaps we should ask specialists who have
been engaged in the ITER construction to participate in the
workshop in the future? The necessity of a fusion neutron
source using, for example, compact toroidal devices was
noted for testing materials under fluences up to 15–20 dpa.
An encouraging experimental result from JET, which
showed weak confinement degradation with power at high
beta plasmas, was reported.
2. To review the status and near-term future plan of helical
research in the world
A summary of LHD experimental results was presented by
Y. Takeiri. Steady progress is seen in LHD experiments,
which have attained simultaneously an ion temperature of
6 keV and electron temperature of 7–8 keV at a density of
1.4 × 1019 m3. The beta of 4.1% was obtained at 1 T.
Long-pulse operation of 48 minutes was sustained with 2
keV temperature and 1.2 × 1019 m3 density. The research
plan of the deuterium experiment was described and is
expected to clarify the isotope effect in a toroidal plasma.
The Numerical Simulation Reactor Research Project,
which was reported by R. Horiuchi, consists of eight task
groups that cover a wide range of fusion plasma simulations.
Achievements from each task will be integrated
around 2020 toward construction of the Numerical Test
Reactor. Research results of the Fusion Engineering
Stellarator News -3- April 2015
Research Project, reported by A. Sagara, were summarized
into the design of several options for what was originally
called a force-free helical reactor, but now known as
just FFHR. The magnetic energy reaches approximately
150 GJ. Five fundamental R&D projects are ongoing:
superconducting (SC) magnet, blanket, low-activation
materials, plasma-facing materials, and tritium handling.
T. S. Pedersen reported that the first plasma of W7-X is
scheduled for this fall using a limiter configuration and
2 MW of electron cyclotron heating (ECH) with a duration
of < 1 second. Commissioning has started successfully;
the cryostat has been pumped down and initial cooldown
was completed. The research plan of W7-X Operation
Phase 1.1 (OP1.1) includes correction of error fields down
to the range of 105, that will be made possible by using
flux surface mapping and the trim coil system.
In Heliotron J, strong gas fuelling using supersonic molecular
beam injection (SMBI) or short high-intensity gas
puff (HIGP) increased the electron density to more than
1020 m3 with electron and ion temperatures remaining at
200 eV, resulting in an increase in the stored energy. This
was reported by T. Mizuuchi. High-density operation is a
unique capability widely observed in helical systems.
C. Hidalgo reported that flux surface asymmetries in TJ-II
plasma potential were observed using the dual HIBP system,
and the role of ion mass on zonal flow-like structures
was investigated.
Fusion plasma research on stellarators in the United State
is regarded as part of the three-dimensional (3D) plasma
physics program, which includes tokamaks. Regarding
collaboration with W7-X, the trim coil system was made
by the United States, and scraper elements were designed
so as to mitigate the risk to steady-state water-cooled target
structures when island structures are shifted from the
nominal configuration. These were reported by H. Neilson.
In L-2M experiments, which were reported by B. Kuteev,
heating power density up to 3 MW/m3 was attained and
heat loading exceeding 0.5 MW/m2 was expected near
separatrix corners. A DEMO-FNS (fusion neutron source)
is being designed using compact tori at the Kurchatov
Institute.
Effects of H-mode transition on divertor flow characteristics
were investigated in U-3M RF plasmas. A conceptual
design study on a fusion-fission reactor based on a stellarator-
mirror device is being carried out at the Kharkov
Institute, as reported by V. Moiseenko.
B. Blackwell reported that Australian activities were conducted
following the mission of detailed understanding of
the basic physics of magnetically confined hot plasma in
the H-1 Heliac.
These ongoing activities continue and near-future plans
are being made while being conscious of issues which
ITER will face. It was reported that financial support for
helical system research may not be adequate for execution
of the planned research, unfortunately.
3. To review design studies on helical reactors
Ongoing reactor design studies were reported from NIFS,
MPI, and PPPL, by A. Sagara, D. Hartmann, and M. Zarnstorff,
respectively. Two comments were made: from the
viewpoints of the utility grid, initial tritium loading and
other issues by K. Okano, and of aspect ratio by S. Okamura.
Advantages and disadvantages of the helical system compared
with the tokamak are rather well known. Advantages
are as follows: steady-state operation, high-density
operation, no major disruption, low recirculating power.
Disadvantages are a complicated 3D structure and usually
a larger size.
The FFHR design is based on the LHD magnetic field
configuration, a coil major radius of 15.6 m, and a plasma
minor radius of 2.54 m. The plasma performance of the
FFHR is deduced through direct profile extrapolation from
LHD experimental data. The design showed the following
attractive features: divertors are placed behind the blankets,
neutron wall loading can be reduced by making the
device size large, and we can obtain a large open space
inside the torus. It was pointed out that the construction
cost does not depend so much on the device size because
the total weight of the coil support scales as R0.4 due to the
decrease in the magnetic field necessary to keep the confinement
time constant. Design window analysis shows
that reductions in stored magnetic energy (160 GJ) and
neutron wall load (1.5 MW/m2) are achieved by reducing
the blanket space (approximately 0.8 m) on the inboard
side. To make the SC helical coils, in addition to continuous
winding using cable-in-conduit conductor (LTS),
winding of segmented helical coils using gas helium
cooled conductor (HTS) has been investigated and an
experimental result of 100 kA for a reasonable cross-section
has been obtained. By using Flibe+Be/ferritic steel for
the blanket it is possible to get the tritium breeding ratio
(TBR) larger than unity and also radiation shielding for the
SC magnet. Removal of the decay heat in the blanket
when it was under maintenance was discussed. Large port
openings are needed for divertor pumping and remote handling.
Reduction in toroidal non-uniformity of divertor
heat flux is important because the heat flux is found to be
at a tolerable level if the toroidal distribution is uniform.
Discussion on the maintenance method and the construction
process for FFHR has been started.
The magnetic configuration of the HeliAS power plant is
based on W7-X, in which the configuration is optimized
Stellarator News -4- April 2015
on plasma performance from various physical viewpoints.
Basic parameters of HeliAS 5-B are as follows: major
radius of 22 m, minor radius of 1.8 m, magnetic field on
coils of 10.5 T, and fusion power of 3 GW. Modular coils
of HeliAS 5-B are of the same size and of performance as
those of ITER, thus ITER coil technology is applicable to
HeliAS. A double-hull structure is employed for the
plasma vessel, similar to the ITER vacuum vessel. Blanket
structure and remote maintenance are the key factors to
determine the size of HeliAS 5-B. The thickness of the
blanket is 80 cm while the space between coil and plasma
is 1.3 m. As an intermediate step, an ITER-like HeliAS is
supposed to reduce uncertainties in extrapolation to the
DEMO reactor. System studies are very sensitive to confinement
enhancement/degradation. For the near future,
IPP activities will be focused mainly on scientific exploitation
and on confirmation of design criteria rather than on
power plant studies.
As for the compact stellarator strategy, ARIES-CS based
on NCSX was primarily discussed, but there are also
ARIES-ACT activities in the United States. The average
major radius is 7.75 m, and the average minor radius is
1.72 m. Compactness has been pursued by stressing the
low aspect ratio, which is in contrast to FFHR or HeliAS.
The low aspect ratio gives minimization of surface area
normalized by volume and higher ion temperature under
ion temperature gradient (ITG) mechanism for ion transport.
A tapered blanket structure similar to that of FFHR is
adopted to minimize coil-plasma stand-off. Because the
most important issue for the stellarator is complexity, a
simplified modular coil system is being pursued. It was
shown that not only neoclassical aspects but also anomalous
transport can be optimized by small adjustments in
plasma shape. Complete stabilization of ITG/ETG is
found in some configurations. A pilot plant with a major
radius of 4.75 m as a step to power plant has been
designed to integrate power plant science and technology
and to generate net electricity, QENG > 1. It is noted that
producing electricity is found to be easier in a stellarator
compared with pilot plants based on an advanced tokamak
or spherical torus. If NSCX-scale experiments start soon, a
decision on the pilot-plant can be made around 2030.
Electricity produced by a fusion plant should match the
demand within the error bar of +/- zero percent. In other
words, strict stability is required or the system must have
an energy reservoir. Strict stability means that even a
minor disruption is unacceptable. Unscheduled outage is
at most 0.5 events or less in a year. Also, the system
should have load-following capability. We must recognize
these points when talking about a fusion power plant.
Comments on the aspect ratio were made from the viewpoint
of the initial cost. The initial investment cost should
be large when the aspect ratio becomes large, although it
was pointed out that the total weight of coil support is proportional
approximately to R0.4. To reduce the initial cost
we need to take account of compactness more intensively.
4. To discuss the strategy of helical system research in the
ITER era
Free discussion, chaired by O. Kaneko, took place on the
following subjects:
• What are the common issues for tokamaks and helical
systems?
• How can helical system research contribute to ITER?
• What we should learn from ITER?
First of all, we are indebted to stakeholders, i.e., government,
for the fusion budget. This means outreach activities
are important, in which almost all institutes engage. Japanese
researchers are fortunate because budgets for helical
research are separated from those for tokamaks, and the
budget situation is rather stable. In Europe, the budget situation
for W7-X also seems to be stable although re-evaluation
is scheduled in 2015. It is expected that W7-X will
receive substantial support for next 10 years. In the United
States the priority of research is placed on basic science
rather than on DEMO-like projects. Stellarators are
regarded as a backup to tokamaks, but if arguments for
stellarators are pushed too hard, counter-arguments will
occur, because stellarators are thought to be behind tokamaks
by 1 or 2 generations.
Collaboration with tokamak researchers is important; as an
example, through ITPA the helical community is participating
in discussion with the tokamak community. The
mission of NIFS as an Inter-University Research Institute
includes not only helical system research but also fusion
science as a whole. The contribution to ITER from helical
research might simply be said to be 3D physics; however,
we should look at all aspects of engineering for improving
ITER, because common issues in tokamaks and helical
systems might constitute most of engineering issues, with
the exception of SC coils. One example of contributions to
ITER from MPI and NIFS is an RF-based negative ion
beam source in collaboration with the Padua group.
The following points were also discussed:
• Are burning physics results from the ITER experiment
sufficient or not?
• How can we manage engineering or manufacturing
development for SC, divertor, and blanket?
• How should we manufacture a big machine in general.
Stellarator News -5- April 2015
5. To discuss the way to helical DEMO
Free discussion was chaired by K. Matsuoka on the following
subjects: how should we appeal to the stakeholders
in order to survive, what and how can we contribute to
ITER, and what lessons could we learn from ITER?
Consensus was obtained about the definition of a helical
DEMO, which meets the following requirements: gross
fusion output larger than several hundreds of megawatts,
viable as a pathway to commercialization, and tritium
breeding ratio more than unity. This definition is the same
as that for a tokamak DEMO.
Regarding FFHR, it was pointed out that the LHD experiment
should be promoted toward satisfying DEMO physics
requirements. Among the goals of the LHD experiment
steady-state operation at 3 megawatts for 1 hour duration
should be pursued more intensively because steady-state
operation is the best sales point for superiority over a tokamak.
We would be happy if a discharge lasting for hours,
days, and weeks is realized. An encouraging result of
4.1%  at 1 T was obtained. In LHD the Shafranov shift is
not small at that time, so some method for reducing the
shift might be required. The deuterium experiment in LHD
takes a step toward an integrated research of science and
engineering. The effect of isotopes on confinement
improvement is expected to be clarified.
W7-X has been optimized for core plasma physics. We are
looking forward to the start of experiments. Concern
would be focused on divertor operation because the divertor
area might be smaller than that of LHD. Island divertors
are inherent to modular stellarators. This is due to an
anti-helical winding component that does not allow a helical
divertor structure along helical windings such as those
of LHD.
Energetic discussions were held on the need for of a helical
ITER-level device. The majority of participants agreed
it is a] necessity. Burning physics in a helical system might
be different from that in a tokamak and the gap from present-
day experiments to the helical DEMO is too large.
In any case, the community is looking forward to experimental
results from LHD and W7-X. Plan for the future
should be drawn up on the basis of experimental results.
6. To discuss possible international collaborations that
assist domestic helical system research activities in each
country
The status of international collaboration was reported by
all participating institutes. The reporters were B. Blackwell,
K. Nagasaki, V. Moiseenko, T. S. Pedersen, C.
Hidalgo, S. Masuzaki, B. Kuteev, and H. Neilson. International
collaboration is taking place in almost all areas of
fusion research including tokamaks.
H. Neilson announced that a stellarator session would be
held at the 26th Symposium on Fusion Energy (SOFE), 31
May–4 June 2015, in Austin, Texas, in the United States.
In session SO-22: Design and Analysis Tools for Stellarator
DEMO Devices, several invited talks on HeliAS,
FFHR, and other topics will be given. Another piece of
useful information was the Announcement of Third IAEA
DEMO Programme Workshop to be held at Hefei, China,
11–16 May 2016. International programme committee
members and a tentative program were introduced. T.
Muroga is the chairperson of the international programme
committee, and H. Neilson is a member of the committee.
In the report by M. Yokoyama on the IEA Implementing
Agreement for Cooperation in Development of the Stellarator-
Heliotron Concept it was announced that the chairmanship
will change to R. Wolf from A. Komori at the
next executive committee meeting in October 2015. The
20th International Stellarator-Heliotron Workshop will be
held at Greifswald 5–9 October 2015. The next Coordi-
Fig. 1. Meeting participants. First row (from the left): A.Sagara, V.Moiseenko, B.Kuteev, B.Blackwell, O.Kaneko, M.Osakabe, T.Mizuuchi,
Y.Takeiri, M.Zarnstorff, T.Mito. Second row (from the left): K.Nagasaki, K.Matsuoka, R.Horiuchi, K.Okano, H.Yamada, T.Goto, S.Imagawa,
S.Okamura, C. Hidalgo, T.Muroga. Third row (from the left): S.Masuzaki, H.Sugama, M.Yokoyama, D.Hartmann, J.Miyazawa, N.Yanagi,
H.Neilson, T.Pedersen, S.Ishiguro.
Stellarator News -6- April 2015
nated Working Group Meeting (CWGM) will be held in
Warsaw 17–19 June 2015.
7. Summary
A summary was presented by O. Kaneko. He made the following
points: Why don’t we make our research more
DEMO-oriented? We don’t have much time left to make a
helical system roadmap to the first candidate of a helical
DEMO that is competitive with a tokamak. In order to
accelerate the research, we should pick specific critical
issues for DEMO, and various groups should proceed to
research the subjects and discuss the results, as is done by
the ITPA for ITER. This may be possible for scientific
issues. How about for engineering issues? This was the
original function of the CWGM, and why don’t we restart
the engine? We hope to discuss this subject at the next
CWGM (June) or Stellarator/Heliotron Workshop (October).
Dr. Keisuke Matsuoka
National Institute for Fusion Science
The Research Enhancement Strategy Office
Toki 509-5292 JAPAN
TEL: +81-572-58-2344
FAX: +81-572-58-2309
E-mail: matsuoka@nifs.ac.jp
Announcement: 14th Coordinated
Working Group Meeting
We cordially invite you to the 14th Coordinated Working
Group Meeting (CWGM) of the International Stellarator-
Heliotron research community. This year it will take place
in Warsaw, Poland 17–19 June 2015. The objective of the
CWGM is to gather numerous experts in the field of magnetic
confinement fusion and to debate the hottest results
and the future of stellarator-heliotron research.
The conference announcement is appended. For further
information please consult the CWGM website http://
cwgm2015.ipplm.pl, which will be regularly updated with
information about the venue, program, and hotel and travel
tips. You may also send your questions directly to
cwgm2015@ipplm.pl; we will be happy to help you.
We are looking forward to seeing you in Warsaw!
Krzysztof Gałązka, on behalf of the Local Organizing Committee
Institute of Plasma Physics and Laser Microfusion, Warsaw
Stellarator News -7- April 2015
THE COORDINATED WORKING GROUP MEETING
of the International Stellarator-Heliotron research
Warsaw, 17-19 June, 2015
Dear Colleagues,
The Institute of Plasma Physics and Laser Microfusion have the honor to invite to:
14th COORDINATED WORKING GROUP MEETING ‘ 2015
which will be held in Warsaw, June 17-19, 2015.
It is worth to mention that the Coordinated Working Group Meeting (CWGM) implements and coordinates international collaborations in
[S]tellarator-[H]elietron research. The work is intended to contribute to the International Stellarator-Heliotron Confinement and Profile
Database (ISH-C(P)DB).
The annual Meeting allows to gather numerous experts in the magnetic confinement fusion field. The main objective of the Meeting is to discuss
the recent results and the way laying ahead of plasma research.
The general character of the programme provides attendees with sessions coordinated by a topic leader. Session leaders will be announced in
due course.
The possible list of topics is as follows:
 SH Confinement and Profile Database
 Highlights in experiment, invitation to joint experiment
 Framework for collaborations
 Diagnostics collaborations
 3D Transport in divertors
 Impurity transport
 Flows and Viscosity, Transport
 Energetic particles, Alfven modes
 3D Equilibrium
 Reactor/Systems code
 Plasma Startup
With this announcement you are invited to propose a talk in the frame of topics listed above. The deadline for abstract submission is the 26th
April, 2015.
More information can be found on the website: http://cwgm2015.ipplm.pl/
Correspondence or questions should be sent to: cwgm2015_ipplm@ipplm.pl
We are looking forward to seeing you in Warsaw!
Local Organizing Committee