J. Plasma Fusion Res. SERIES, Vol.1 (1998) 53-56

The WENDELSTEIN 7-X Project
GRIEGER Guenter and the W7-X TeamMax- P lanck- Institut fdr plasmaphysik,

EURATOM Ass., Greifsward and Garching, p.o.B. 1s33, g5740 Garching, Germany(Received: 30 September L997/Accepted:22 October 1997)

Abstract
Before fueezing all essential parameters for the construction of W7-X, their compatibility has
been
re-checked with the evolution of knowledge on HELIAS-type stellarators. It was found thatno modifi-cation of earlier conclusions became necessary. Thus, W7-X is being realized according to the originalplan except for somewhat more space which could be provided for the divertor arrangement.
TheGreifswald branch of IPP is under construction. The buildings will be ready for occupancy aroundtheyear 2000, when also the first parts will be delivered for the construction of the experimentalfacility.

Keywords:
Wendelstein 7-X, HELIAS, stellarator, Ipp Greifswald

1. Introduction

Once both are operating, it will be interesting to com-

As an essential milestone in the course of realization of WENDELSTEIN 7-X (W7-X)[1], the cornerstone was layed for the new Greifswald branch of theMax-Planck-Institut fiir Plasmaphysik on 19 June 1997.

pare their performance.

During this conference, there are several overview
papers on W7-X: on the W7-AS experiment by F.
Wagner, on stellarator theory by J. Niihrenberg, on theW7-X divertor concept by H.Renner, on the W7-X device construction by M. Wanner, on HELIAS-type
power stations by H. Wobig, and on special details bysome others, too. Apart from a short description of thenew institute to house W7-X and its infrastructure, thispaper will recapitulate the general philosophy underlying the concept of W7-X. This is justified because atthe transition from the W7-X design to the construction
phase it is useful to re-iterate the main points of theW7-X is appearing on the scene when the world fusion
programme has embarked on ITER to demonstrateburning plasma operation. Since for the time being thedata base and thus the predictive capacity is largest for
tokamaks, ITER has to be a tokamak. But since a fu-I

sion power station to become attractive has to prove
economic competitiveness and general public acceptance, it is generally agreed that concept improvements
are still necessary. W7-X is right in time to aim at demonstrating the particular ability of HELIAS-type stellarators [2] to provide these needed concept improvements. W7-X will be ready for operation several years
after LHD [3,4] which aims at the same goal. But thesefew years of delay with respect to each other haveallowed the introduction of several new results of stellarator theory in the concept of W7-X, and in thisrespect the two devices are designed using different

basic philosophy, and it is the last chance for re-check-ing their continued validity before everything is frozen.

2. Stellarators for Concept lmprovement
HELIAS-type stellarators were designed by starting from reactor needs. The strong net toroidal currents
inherent to the tokamak concept and responsible for
their potential disruption could be avoided in HELIAS-

criteria, and in this sense they are complementary.
Corresponding author's e-mail: guenter.grieger@ ipp.mpg.de53@1998 by The Japan Society of Plasma
Science and Nuclear Fusion ResearchGrieger G. et at., 'fhe WENDELSTEIN 7-X Projecttype stellarators. This allowed full exploitation of thepossibilities offered by the HELIAS concept and optimization of the magnetic configuration practically independent of the plasma pressure B for < p ) -values ofup to 57o, or even higher. This is fully sufficient for theoperation of a fusion power station. These systems

oEo(!

show the following desired properties:r inherent capability of stationary operation'eo absence of current disruptionso sufficiently small loss cone for particles
o sufficient energy confinement.=3. negligible bootstrap currentC)

oo no need for current drive.9

r no need for feed-back position control
r no density limit except for a soft one arising from im-oxuJpurity radiation.
Thus, compared to the tokamak concepts of today'
operation of power stations based on the HELIAS concept are expected to be much more benign and circulat-Fig. 1 Strategy for reactor development.

ing power saving.

injecting high energy neutral particles into H plasmas,
and by performing some experiments with DD plasmas.
This way of procedure reduces the cost of construction
and operation by order of magnitude and allows much
faster experimentation. The radioactive burden remains
small and under all circumstances access to the experiment remains possible after one or two days at the latest. In addition, no problem will arise with the later decommissioning of the device.

3. Strategy
Within the limits of present-day devices (W7-AS)'

the HELIAS concept has been successfully checked.
But before relying on the reactor properties, an integral
concept test is necessary with parameters close enoughto the reactor ones. This is the task of W7-X. For satisfying these aims, W7-X has. to possess the final reactor configuration, ando to be large enough to yield conclusive predictions onDT plasmas so that extrapolations to reactor conditions are in an acceptable range.
The use of DT plasmas is not necessary in W7-X.
As indicated in Fig. 1., it is rather possible to exploit theexperience on DT plasmas gained with ITER, and tocombine it with the results of W7-X. But this is onlypossible if both experiments, on burning plasmas withITER, and on concept improvement with W7-X, aredone in parallel.

5. Demonstration of Stationary Operation
Due to the absence of the need for strong toroidally flowing plasma currents, stellarators have the inherent property of stationary operation. This property
was considered so important by the European Commission's Ad Hoc Group examining the proposal for
W7-X that it requested W7-X to be made capable ofyielding its demonstration above any doubt. This meansthat W7-X itself must be made capable of truly stationary operation which leads to the following require-

4. Dispense with DT Operation
If W7-X had to be operated with DT to yield con-ments:o The magnet system must be superconducting to ar-clusive results, cost and complications had to be expected close to those envisaged for the construction andoperation of ITER. It has to be considered, however,that the physics of the fusion process itself is well

rive at manageable power consumption'. A 140 GHz, 10 MW, cw heating source must beavailable. This can be achieved by using gyrotrons of
1 MW unit power.

known, and that, practically, it is only the interaction offusion-generated particles with the bulk plasma which is
not covered by the exclusive use of non-DT plasmas'o Continuous handling of 10 MW plasma throughput
power must be possible. This is a design task.
r The divertor system must be capable of handling thisthroughput power and particle flux in stationary

But information on this point will be available fromITER DT operation which can be checked on W7-X by54I

Grieger G. et al., The WENDELSTEIN 7-X proiect

fashion, and of limiting the impurity fraction to toler_able levels. This is a development task for W7_X but
promising concepts exist.

Within the following limits, a coil system superim_
posed to the basic coil configuration shown in Fig.1
allows ample flexibility in the variation of the configu_

Additional heating schemes are necessary for flex_ible experimentation but pulse durations of 10 s or so
are sufficient. 30-110 MHz, 4 MW ICRH, and 55
keV, 3 MW NI(H), or 65 keV,4.5 MW NI (D) will beration :o Twist on axis
0.75-1.01
o Twist on boundary..... 0.g3-1,25
o Shear..... 0.10-0.20
o Mirror field on axis .............. 0.00-0.10
o Magnetic well depth
0.004-0.015
This flexibility inherent to the machine can beavailable. During stage 2 of the experiment, the heating
power will be increased to 20 MW for exploring theB_limit but also for this purpose pulses of 10 s are sufficient.

utilized to search for the optimum parameters to be expected for the three essential cases: long mean freepath, high B, and high performance. These are listed in
the following table:

6. Characteristic Data of W7-X
Evolution of knowledge in physics and related
technology have not yielded any request for changes ofthe basic machine parameters of W7-X. Only the shapeof the inner coil surface and the corresponding shape ofthe vacuum vessel could be somewhat modified to provide more space for the divertor equipment and thus tomake the divertor situation a bit more comfortable. Alllong mean high B

treepathPn(0)[10'z0

the performance relevant parameters were confirmedby the ongoing studies. The final characteristic data of
W7-X are collected in the following table:o Average major radiusr(0) [kev]B(0)......... 5.5 m
0.53 mo Pfirsch-Schl0ter/diamagnetic currents ......... 0.7
o Magnetic well depth.......... L0 %2.s
10m-3] 0.5

T"(0) [kev]. Average plasma radius ..........
r Twist, on axis/boundury ................... 0.84/0.99. MHDstabilitylimit <B. Equivalent ripple ...........

[r]
[Mw]

B

[%]

6.o-e.4
2.6-4.0
2.7-4.3<0> [o/"]0.7-13z.

84-160[ms]

highperformance

r.25
3.0
20
24
1.5
3.0
L.4-2.5 2.2-3.5
r.4-2.3 2.2-3.4
rr-tg
6.0-9.3
3.6-6.9 2.0-3.5
57-109 l5E-291

7. Extrapolation to Reactor Conditions
With respect to the extrapolation to reactor conditions and the predictive power of the expected W7-X
results, the following statements can be made:o The configurations are identical, only lBl is doubled.
r The same normalized mean free path, A*:)"t/zR: 100, can be reached by ECRH for n:0.6X 1020..... L5%o Reduction factor of bootstrap current ...... <0.1
o Magnetic field on axis,/at coils .......... 3.0/6.17
o Total magnetic energy..... 600 MJ
The basic configuration of W7-X is displayed inm-3,the following Fig. 2:

f:3 keV.

o The same ratio between plasma to ion gyroradius will
occur for Qr:a/pt:350 for 4:1 keV (at A*:1.3for n:0.5x1020 m-3 for ECRH).

o The same relative plasma pressure <B>-5o/o will
be produced for full heating power at half B.

o The ignition parameter, 4 X Ti X r=3O[1020 m-3keV sl - P x Qot X B can be approximated only
within one order of magnitude because in W7-X not
all the dimensionless parameters can be made equal
to reactor values at the same time.
c particle physics can be simulated by the injectionof high-energy particles.

The extrapolation is largest in magnetic energy(X100) but this is acceptable since no change intechnology is required.

Fig.2 W7-X, basic configuration.

55

Grieger G. et a/., The WENDELSTEIN 7-X Projecl

Garching. Only Niihrenberg's gtoup on Stellarator Theory has already moved to Greifswald and shares rented
space with the team of engineers responsible for the8. Operation Scenarios
W7-X will be capable of truly stationary operationand the power consumption for operating the confinement device with its superconducting magnet will betolerable. But the power for providing 10 MW of stationary ECR heating may require up to 50 MW to bedrawn from the grid which will yield an electricity billof about 100.000 DM per operating day of 8 hours
duration. It will thus be more economic to consider
quasi-stationary operation for a few minutes with verification of the results by longer operation from time totime. For the exploration of the B{imit, 10 s pulses will
be sufficient in any case.

construction of the institute buildings and the infrastructure.

10. Summary
By also exploiting the results coming from DTusing devices like ITER, W7-X is expected to answer
conclusively all questions about the reactor propertiesof advanced HELIAS-type stellarators. The institutebuildings and the infrastructure are under construction
at the new site at Greifswald.

References
Fusion Technology 21, 1767a/.,e/t1l G. Grieger

9. The Greifswald Branch of IPP
For housing the W7-X activities, a new branch of(Ieez).

IPP has been founded at Greifswald which is located at
the north-east corner of Germany. Construction of theinstitute and all the infrastructure needed for the operation of W7-X has begun. It is expected that the buildings will be ready for occupancy in the year 2000 or2001. Up to this time, the team of engineers responsible for the construction of the W7-X facility as well as
the team operating W7-AS will continue to reside at

t2l G. Grieger et a/., Physics Fluids B 4,2081 (1992).
t3l A. Iiyoshi, M. Fujiwara, O. Motojima, N. Ohyabu

and K. Yamazaki, Fusion Technology 17, 169(1ee0).

[4] M.Fujiwara et al., J. of Fusion Energy 15, 7(1ee6).

56