TEST RESULTS OF 60-cm BORE Nb3Sn TEST MODULE COIL (TMC-I) IN THE CLUSTER TEST FACILITY

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TEST RESULTS OF 60-cm BORE Nb3Sn TEST MODULE COIL (TMC-I) IN THE CLUSTER TEST

FACILITY

T. Ando, S. Shimamoto, T. Hiyama, H. Tsuji, Y. Takahashi, M. Nishi, E.

Tada, K. Yoshida, K. Okuno, K. Koizurmi, et al.

To cite this version:

T. Ando, S. Shimamoto, T. Hiyama, H. Tsuji, Y. Takahashi, et al.. TEST RESULTS OF 60-cm

BORE Nb3Sn TEST MODULE COIL (TMC-I) IN THE CLUSTER TEST FACILITY. Journal de

Physique Colloques, 1984, 45 (C1), pp.C1-101-C1-104. �10.1051/jphyscol:1984123�. �jpa-00223678�

JOURNAL DE PHYSIQUE

Colloque Cl, supplbment au no 1, Tome 45, janvier 1984 page CI-107

SYSTEMS ENGINEERING APPROACH T O THE DESIGN

MAGNETS

FUSION DEV

CES

K.E.

Wakefield

Princeton Plasma Physics Laboratory, Princeton University, Prineeton, New Jersey, U.S.A.

Rgsume - Revue des facteurs critiques

considcrer pour.la definition des objectifs, de la forme et des dimensions du dispositif de confinement torofdal devant faire suite aux projets TFTR et JET.

Abstract -

review of the critical factors considered in defining the objectives, form and size of the toroidal confinement device to follow TFTR and JET.

Introduction

Initial operation of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory and the Joint European Torus

(JET) at the culham Laboratory having begun within the past few months and full operation of both these machines under reactor-like plasma conditions anticipated within the next two or three years, ever-increasing attention is being given the problem of determining the objectives, form and size of the toroidal confinement device best suited to continued advancement toward a full scale fusion power re- actor. This problem has, of course, been the subject of considerable debate during recent years and its solution the object of a great deal of scientific and engineering effort. The fact that the ques- tion has not yet been resolved is, paradoxically, not due to a lack of technically feasible embodiments, since the numerous proposals to date could, for the most part, be so described. The difficulties have and continue to stem from considerations of cost and perceived risk, and are compounded by the nascent shift in program emphasis away from the need for further plasma physics advances, and toward the need for demonstrated technological capability in areas such as large scale superconducting fusion magnets and first wall materials capable of extended life in the fusion environment, to name only two.

The studies usually adopt the strategy of defining at the outset a

"point design" based on a given set of project objectives and physics/

technology parameters.

the details take shape, the design is re- fined to better meet the objectives. The process yields a great deal of valuable information and data and often spurs important development programs along lines that might otherwise have been overlooked or postponed. With some exceptions, however, a stage is reached where one or more critical factors make it prudent or necessary to modify the project objectives or physics/technology parameters, and the pro- cess is repeated.

This is a natural process characteristic of most creative undertakings but it needs shortening. How this is being achieved and some prelim- inary results of recent attempts to determine the objectives, form and size of the optimum device are the subjects of this paper.

Article published online by EDP Sciences and available at http://dx.doi.org/10.1051/jphyscol:1984123

J O U R N A L D E PHYSIQUE

Turn-around Time

Years ago, when the Computer and Fusion Ages were in their infancy, the organization that is now the Plasma Physics Laboratory faced the problem of designing an axisymmetric divertor. The primary design tool was a computer program that traced the lines of magnetic force for a defined coil geometry. The state of the art was such that we were able to run one or two test cases a week. In desperation and a flash of inspiration the author devised a simple network of resistors that enabled one to map the critical flux surfaces within minutes, leading to a workable divertor design concept in a relatively short time.

An

"point design" is being achieved through the use and extensions of the Systems Codes that are being developed at the Fusion Engineering Design Center (FEDC), the Massachusetts Institute of Technology (MIT) and elsewhere. The motivation for their development, which began several years ago, was to quickly calculate the incremental costs of small variations in geometry, size or other machine parameters. The cost derivatives (with respect to each variable) thus obtained served to point the user toward lower-cost designs. Costs are calculated using algorithms based on data derived from actual costs on other machines, hand cost analyses or quotations.

The development of the Toroidal Fusion Core (TFC) concept has been sped up considerably through use of the FEDC and other codes. Since algorithms for mechanical stresses, temperatures, voltage stresses and other quantities critical in a design can be easily constructed, the codes are used to optimize a configuration from many engineering or design points of view, including choice of the configuration it- self. For example, one might use a complex finite element (FEA) stress code to determine the variations in mechanical. stress with re- spect to several parameters and use the results to construct an algo- rithm to be used in the Systems Code.

Application to TFC

Previous studies, such as the International Tokamak Reactor (INTOR) and the Fusion Engineering Device (FED) envisioned modestly large advances in physics and technology with the result that the objectives of proposals based on the concepts characteristic of those studies were too ambitious for the investment in resources that potential sponsors were willing to commit.

There is, however, a strong feeling in the fusion community, espe- cially among those involved in technology, that the time is over-ripe for the development and construction of a superconducting TF coil sys- tem as one feature of a D-T burning, long-pulse tokamak. On the other hand, a number of serious proposals have been made for water or liquid nitrogen cooled copper TF and PF coil systems for machines with simi- lar operational goals.

In the absence of a consensus on that major question, other consider- ations tend to be given increased weight, among which are: physics topics (ignition, long-pulse operation, alpha behavior, impurity con- trol, helium transport, lower hybrid current drive (LHCD), ion cyclo- tron resonance frequency (ICRF) heating and need for OH current drive): cost (superconducting coils are perceived, perhaps unjustly, to be more expensive); degree of risk (the step size for supercon- ducting TF coils is thought by some to be too large); potential sched ule delays; reliability; capability for significant neutron fluences;

ALL

SUPERCONDUC

FIRST W A L L

S.C. TF COlL

R 3.75m

S,C, PF

COIL

HYBRID

~s.c.

ALL COPPER

COPPER TF COlL

RO 3.0a

Figure

- Cross-section views of initial versions of three TF coil concepts for the TFC.

blanket studies; and possible use of existing facilities.

Figure 1 illustrates the essential features of the TF coils for three of the copfigurational concepts considered during the TFC studies.

The common elements of these are: ignited, long-pulse operation; elim-

ination of neutron fluence objectives; high wall-loading; include re-

cent tokamak physics results in the optimization process; and some

development of tokamak reactor engineering concepts. Cost and sched-

ule will be strong drivers in the selection of the configuration(s)

for more detailed study.

JOURNAL

DE

Table 1

Parameters of three configurational concepts.

Item Variation 1

TF Coil type

-

Major Radius, m 3.75

Minor Radius, m 1.07

Aspect Ratio 3.5

Magnetic field, T 4.3 Plasma current,

MA

TF Power, MW

-

Beta, % 5.9

Ignition Safety Factor 0.7

Variation 2 Hybrid

3.6 0.97 3.7 4.8 7.2 50 5.6 0.7

Variation 3 Copper

3.0 1.20 2.5 3.8 11.0

350 8.3 1.25 Plasma elongation (1.6)

,

,

In performing these studies, every attempt was made to ensure that each variation resulted in the same "physics" machine. Reference to Table 1 will show that this was fairly successful. The lower aspect ratio for the copper version, however, couldn't be helped. This ver- sion also allows higher plasma current and beta. It, therefore, has a larger margin of safety for reaching ignition.

A significant factor that permits a smaller major radius in the copper version is the abandonment of high neutron fluence objectives, which allows the copper coil to act as a partial neutron shield for the superconducting PF coils. The nuclear heating in the superconducting coils is the same for all versions.

Any of the versions could utilize copper PF coils. Doing so would add about 350 MW to the required magnet power. This consequence must be weighed against potential risks in developing and constructing the 11 meter diameter superconducting PF coils.

The work upon which this paper draws was performed by PPPL and FEDC staff members under the direction of George V. Sheffield of the Plasma Physics Laboratory.

This work was supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073. Neither the United States nor the United States Department of Energy, nor any of their em- ployees, nor any of their contractors, subcontractors, or their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights.