lOE/ET-0116/1 	BISON.
)C-20, A thru G . 
Fusion Technology Development 
August 1979 
• 	REACTOR STUDIES PLASMA RADIUS 
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• 	EXPERIMENTAL DEVICE PLASMA RADIUS 
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'YEAR
SUNY at Buffalo The Libraries 
THE PROJECTED IMPROVEMENT IN PLASMA 
1 3 24 1 PERFORMANCE IS MOST DRAMATICALLY ILLUSTRATED BY 
THE REDUCTION IN REACTOR SIZE 

f)epository Copy • 
RADIUS • 1.5 m 
POWER DENSITY • 5 MW/m3 
POWER OUTPUT • 750 MWe 
f--13m---l 
1977 
U.S. Department of Energy 
Office of Fusion Energy Division of Development and Technology 

DOE/ET-0116/1 
·Fusion Technology Development 
August 1979 
U.S. Department of Energy Office of Fusion Energy 
Division of Development and Technology Washington, DC 20545 
Available from: 
National Technical Infonaation Service (NTIS) 
U.S. Department of Commerce 5285 Port Royal RoadSpringfield, Virginia 22161 
Price: 	Printed Copy: $ 8.00 Microfiche: $ 3.00 
AN OVERVIEW OF INTERNATIONAL FUSION TECHNOLOGY PROGRAMS 

F. E. Coffman, J. E. Baublitz, D. S. Beard, M. M. Cohen, E. N. C. Dalder, 
c. 
R. Finfgeld, G. M. Haas, c. R. Head, M. R. Murphy, G. R. Nardella, 

R. 
N. Ng, T. C. Reuther, H. s. Staten, K. M. Zwilsky, Office of Fusion Energy, u.s. Department of Energy, Washington, DC 20545-This paper


discusses current progress and future plans for the u.s. and world fusion technology programs. In the area of conceptual design, the 
international INTOR tokamak study, the Garching Ignition Test Reactor Study, the u.s. Engineering Test Facility conceptual design, the ArgonneNational Laboratory Commercial Tokamak Study, mirror conceptual designs,and alternate concepts and applications studies are summarized. In the area of magnetics, progress to date in the large coil program and the pulsed coil program is summa_rized. A new Nb Sn high field test facility at Lawrence Livermore Laboratory, with associated test coils, is described. In the area of plasma heating, fueling and exhaust, emphasis is on a new advanced positive ion source R&D program at Lawrence Berkeley Laborator-Y and Oak Ridge National Laboratory. Negative ion work is also described. 
Tradeoff considerations for RF heating alternatives are made and a new 60-100 GHz ECH R&D program is outlined. Progress and future plans for solid nydrogen pellet injector development is discussed, and future plansfor a divertor technology initiative are delineated. The U.S. and world materials R&D program is reviewed with emphasis placed on new directions and initiatives in the areas of materials radiation test facilities and ferritic steels. A brief review of the U.S. alternate applications and environment and safety program is also provided. 
F. E. Coffman, Director Development and Technology Division Office of Fusion Energy 
U.S. Department of EnergyMS G-234 Washington, DC 20545 

Table of Contents 
I. INTRODUCTION ••••••••••••••••••••••••••••••••••••••••••••••••••• 
II. CONCEPTUAL DESIGN STUDIES ...................................... 

II I. MAGNETICS •••••••••••••••••••••••••••••••••••••••••••••••••••••• 
A. Overview •..................•..••••.••.••..•.••••••....••••• 

B. Large Coil Program ••••••.••••••••••.••••••••••••••••••••••• 
C. Pulsed Coil Efforts •••••••••••••••••••••••••••••••••••••••• 
D. The 	12 Tesla Coil Program •••••••••••••••••••••••••••••••••• 
IV. PLASMA HEATING, FUELING AND EXHAUST •••••••••••••••••••••••••••• 
A. Neutra 1 Beams •••••••••••••••••••••••••••••••••••••••••••••• 
1. Neutral Beam Systems for Current Projects •••••••••••• 
2. 	Neutral Beam Requirements for Upgrades to Existing Facilities and for Future Projects ••••••••••••••••• 
3. Neutral Beam Development Elsewhere in the World •••••• 
B. RF Heating ..••..•......••••.•••••••••.•••.••.•••••••••••••• 
1. Introduction ••••••••••••••••••••••••••••••••••••••••• 
2. Requirements ••••••••.•.•.••.•••••••••.••••.•••••••••• 
3. Technical Approach •.•....•..•••••..••••••..••.••••••• 
c. Fueling •••••••••••••••••••••••••••••••••••••••••••••••••••• 
1. Introduction ......................................... 

2. Requ i remen ts ••••••••••••••••••••••••••••••••••••••••• 
3. Technical Approach ••••••••••••••.••..•••••••••••••••• 
D. Divertor Technology ••••••••••••••••..•••••.•••••••••••••••• 
v. MATERIALS FOR FUSION REACTORS •••••••••••••••••••••••••••••••••• 
A. Introduction ••••••••••••••••••••••••••••••••••••••••••••••• 
B. Radiation Test Faci 1ities •••••••••••••••••••••••••••••••••• 
c. Fission Reactors ••••••••••••••••••••••••••••••••••••••••••• 
D. High Energy Neutron Sources •••••••••••••••••••••••••••••••• 
E. Broad Spectrum, High Flux, High Energy Neutron Sources ••••• 
F. Fusion Reactor Test Environment •••••••••••••••••••••••••••• 
G. Progress in. Plasma Materials Interaction ••••••••••••••••••• 
H. Progress in Alloy Development .............................. 

VI. ALTERNATE APPLICATIONS ......................................... 

VII. ENVIRONMENT AND SAFETY 
VI II. APPENDIX -CAPSULE SUMMARY OF U.S. DEVELOPMENT AND TECHNOLOGY PROGRAM ••••••••••••••••••••••••••••••••••••••••••• 
Page 
1-1 
11-1 
II 1-1 II1-1 111-2 III-3 111-6 
IV-1 
IV-1 
IV-2 
IV-9 IV-11 IV-16 IV-16 IV-17 IV-18 IV-24 IV-24 IV-25 IV-26 IV-29 
V-1 V-1 V-4 V-6 V-7 V-8 V-9 V-9 V-14 
VI-1 
VII-1 
VI II-1 
i 

INTERNATIONAL FUSION TECHNOLOGY PROGRAMS: 
CURRENT STATUS, FUTURE DIRECTIONS 

I. INTRODUCTION With the continued depletion of fossil and uranium resources in the coming decades, the world will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. The recent energy crisis has remi~ded us of the finiteness of our energy resources. We have in past years been accustomed to the idea that we have boundless resources in such fossil fuels as coal, gas, and oil. However, it is now apparent that our oil and gas resources are adequate for only a few more decades. There are extensive coal reserves that would provide a substantial part of our energy needs for several decades to come, but at a serious environmental cost. The more accessible fission fuels (uranium at $10 to $40 per pound of u3o8) also have limited reserves to the extent that utilities, in planning for new light-water reactors, cannot be assured of reasonably priced fuel for the 30-yr. operating lifetime of the reactor. Water, wind, tidal, geothermal, and solar power will provide a continual increment to our power needs in future years. However, these options are limited to certain geographic areas and, in any event, cannot be made adequate to meet future energy needs. 
The primary hope for extending fission fuel resources lies with utilizing breeder reactors to convert the extensive reserves of u238 and Th 232 into 
I-1 

233
the fissile fuels Pu 239 and u, respectively. Successful utilization of 
fast breeder technology could supply a major portion of our power require
ments needs for many centuries. 
A very attractive long-range option, of course, is to successfully develop 
commercial fusion power. The world-wide fusion research and development 
program has now progressed to a point which is very exciting; namely, it is 
no longer a question of whether scientific feasibility can be achieved. It 
is a question of whether fusion power can be made a technically and econom
ically viable energy option in the coming decades. 
The world program is engaged in a four-pronged effort during the next few years. The first part of this effort is the physics. Within the next two years, in the U.S. alone, seven major experiments will operate at full 
parameters; namely, PLT, PDX, Doublet-III, Alcator C, ISX-B, reversed field pinch, and TMX. This development implies that within the next two to three years. with the exception of alpha particle physics. we can anticipate having the physics information to confidently design a tokamak device such as ETF or 
INTOR to operate in the late 1980's. This physics information will be augmented by operation of the TFTR in the u.s., JET at Culham, JT-60 in Japan, and T-11 in the USSR. 
The second prong of the world effort consists of the conceptual design programs in various countries to provide a commercial tokamak which has 11 desirable technical and economic characteristics ... These designs are 
I-2 

pointing out that, if the physics and technology programs continue to be successful, we will have a commercial plant design which could be competitive in the world energy market place in the year 2000 timeframe. 
The third part of the fusion effort is the ETF, INTOR, and ignition test reactor design studies being conducted in the U.S., USSR, Japan, and the Commission of European Communities (CEC) member nations. These studies are of paramount importance in that they are defining what device the world should construct as the next step in the fusion energy program. They are also defining the physics and technology efforts which must be put into place during the next few years to make experimental power reactors and commercial fusion a practical reality. 
The last prong of this effort, which is the subject of this document, is the world fusion technology program. As discussed in following sections, substantial progress has been made and no unsurmountable obstacles have been identified. Fusion technology is truly at a provocative point in that 
within the next two or three years, major, convincing demonstrations will be completed. Superconducting mirror and tokamak confinement prototype coils will be constructed and tested; full scale tritium processing and cleanup 
systems wi 11 be tested, reactor seale neutral beam systems will be developed; ETF/INTOR pellet fueling injectors will be tested, and many other items which support the decisions to proceed with the next step in the fusion program will come to fruition. It is our firm conviction that within three years we will have the demonstrated technology base to allow us to proceed witp the next generation device beyond TFTR, JET, and JT-60. 
I-3 


II. CONCEPTUAL DESIGN STUDIES A broad range of conceptual design studies are conducted by the u.s•• USSR, Japanese, and European fusion programs as a means of assessing the feasibility and desirability of various options for future fusion research and development. A second reason for conducting these studies is to identify areas of fusion physics and technology which require additional development in order to bring the level of understanding up to that needed to proceed beyond the current generation of experimental devices such as JET, T-11, JT-60, and TFTR. Commercial power reactor designs define the, ultimate goal toward which we are working, while concurrent studies of tte next step in the program are conducted to assess the current status of our understanding and define the near term direction of program activities. 
The commercial power reactor stud;es cover all potentially viable concepts, with primary emphasis being placed on the concepts which have the greatest potential and less but still significant emphasis being placed on exploring 
alternate concepts. The world fusion community currently considers the 
tokamak to be the confinement concept which is most likely to be used in the first fusion power reactor, based on its substantial lead in demonstrated plasma physics performance. Several design studies of commercial tokamak power reactors have been conducted in the U.S., starting with the UWMAK-I stu~ 
in 1973 and running up through the recently released NUWMAK design. Outside the U.S., the Culham Laboratory in Great Britain has produced the Mark I and 
II 
Mark II tokamak designs, the KFA/Julich in the Federal Republic of Germany has done a similar study, and the Japanese have developed a design called JAERI-II. 
11-1 

The early (and now obsolete) UWMAK era. studies were a first exiPloratory look 
at what fusion reactors might look 1ike and they served a very useful purpose as problem identifiers. These designs, with their nominal 13 1meter major rad;; and 5 meter minor radii, were properly dubbed "superconducting cathedrals." They showed the need for reductions in plant size, increases in power density, improvements in maintainability and provided a reference against which to check ~he adequacy of supporting technology. Subsequently, a number of additional 
commercial fusion power reactor studies have been performed to explore various technical alternatives for improving on the earlier designs. The major characteristics of these designs are given in Table II-1. 1-11 The trend, as shown 
in Figures Il-l and II-2, is toward much more believable designs. 
Currently, in the U.S.A., the Argonne National Laboratory is conducting a 
two-year, approximately $2 Million design study which is intended to pull together the best ideas from the commercial tokamak power reactor design This study wi 11 use the! recently
studies which have been done to date. 
demonstrated favorable plasma physics results from PLT, Alcator C, and ISX-B, 
and add new innovations, especially in the areas of maintainability and depth 
of design, to provide an up to date, detailed conceptual desi~Jn of the most 
realistic, desirable tokamak power reactor which we can currently envision. 
The report of this design study (called "Starfire") is schedulled to be issued 
before the end of 1980. The current Starfire characteristics are given in 
Table II-2. 12 
II-2 
6 
• 	.REACTOR STUDIESPLASMA RADIUS5 UWMAK.I • UWMAKII 
• 	EXPERIMENTAL DEVICEPLASMA RADIUS 
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--ISXA. ALCATOR C
•
74 76 78 	80 
YEAR 
FIGURE 11~1 
Figure ll-2 
THE PROJECTED IMPROVEMENT IN PLASMA 
PERFORMANCE IS MOST DRAMATICALLY ILLUSTRATED BY 
THE REDUCTION IN REACTOR SIZE 

II 
· 4 I"I(LIIJN COOLANT • --SUPPLY LIN( 
~ ~ ' 
(li.T(N0£0 "o'' 
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MAGNET 
RADIUS = 1.5 m 
POWER DENSITY = 5 MW/m3 
POWER OUTPUT = 750 MWe 
..... 
COMPRESSION HUB'\ 
I POWER DENSITY = 1 MW/m3 Of;l COILS. 
BLANKET MODULE
~ POWER OUTPUT = 1700 MWe ICASSE T TE OPT ION I, VACUUM 
REGION
.:nr ,. 1 
. 
TOROIDAL COILS 
SHIELD~· 
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13m ,., f.-sm~
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19771975 
Design studies of commercial mirror fusion reactors hav·e also been done, primarily by the Lawrence Livermore Laboratory. The .i ni ti al results, 13 as with the tokamak studies, portrayed undesirable designs. The primary problem areas uncovered were unacceptably low power gain (Q), high recirculating power fractions, reliance on very high .thermal power conversion efficiencies, high field magnets, large steady state neutral beams, and cumbersome direct energy conversion systems. As with tokamaks, identification of the problem areas created pressure for innovation to improve the mirror concept. This led to a shift from standard mirror configurations to tandem mirror14 and field 
. 15 f" t•
reversed m1rror con 1gura 1ons. Even with the tandem mirror concept, the original iteration retained most of the problems of the standard mirror, and 
added a requirement for very high energy neutral beams (1.2 MeV). However, continued study has led to further innovation so that the tandem mirror reactor designs currently being developed16 are beginning to show promise. Plasma power gains of from 10 to 20 are being projected, with end plug magnetic field strengths of 12T at the coil and 200-400 keV neutral beam injectors. Nevertheless, much work remains to be done to reduce the size of the end plugs and simplify the technology of the newly introduced end plug thermal barriers. 
Other concepts are also being considered, such as the Elmo Bumpy Torus (EBT), Reverse Field Pinch {RFP), and Torsatron. Table 11-3 gives a more extensive tabulation of the full range of concepts under consideration. These concepts are generally handled by doing a new conceptual design study every few years to assess new ideas and new physics results which have become available. For 
11-5 

the newer and/or less well understood concepts, the u.s. has established a screening process of graded studies in successively greab~r degrees of detail to separate concepts worthy of additional development fran those which have little or no potential. 
For major near term devices, most emphasis is focused on tokamaks, with a smaller back-up effort on mirrors. The U.S. effort is primarily being applied to design of the ETF (Engineering Test Facility). ETF will be the next step after energy breakeven is demonstrated in TFTR (the Tokamak Fusion Test Reactor). It is envisioned as an ignited, D-T burning device which will be used to demonstrate the technological feasibility of fusion power and then serve as a test facility in which to qualify technology and components for the first generation of power reactors, whether they are tokamaks or some other confinement concept. 
The ETF design is being carried out as a cooperative effort among the U.S. 
fusion 1aboratories and industry. The effort is oei ng mcmaged by the ETF 
Design Center which is hosted by Oak Ridge National LabOl~atory. This design 
project was started in early 1979. The current schedule calls for precon
ceptual design and scoping studies through FY 1981, conceptual design in 
FY 1982 and FY 1983 followed by project approval ( 1 i ne item funding by 
Congress) in FY 1984. Initial operation is targeted for 1990. At the 
present time, the ETF Design Center is using the parameters given in 
Table II-417 as the reference set of ETF'parameters. 
·1 
II-6 
A similar study is being pursued on an international basis through the auspices of the IAEA (International Atomic Energy Agency). This study was started in response to a suggestion made by the USSR representative on the IAEA International Fusion Research Council and is called the INTOR 
(International Tokamak Reactor) study. The INTOR effort started in 1979 with a broader charter than ETF in that the IAEA simply requested that a study be conducted to define the characteristics of the next logical step in the fusion program. It is interesting to note that the design which is evolving is almost identical to ETF, as shown in Table 11-5. 18 The INTOR study is intended to be the precursor to design and actual construction of the device as an international project. 
References 
1. 	
R. Hancox' 11 Cul ham I Reactor Design, II International School of Fusion Reactor Technology, Erice (1972), EUR-4999e, p. 135. 

2. 	
University of Wisconsin Fusion Technology Program Document UWFDM-68, 11 A Wisconsin Toroidal Fusion Reactor Desi gn11 by R. W. Conn et al., March 1974. · 

3. 	
Princeton Plasma Physics Laboratory Document MATT-1050, 11 A Fusion 
Power Pl ant11 by R. G. Mi 11 s et al., August 1974. 


4. 	
University of Wisconsin Fusion Technology Program Document UWFDM-112, 11 A Conceptual Tokamak Power Reactor Design11 by R. W. Conn et al., October 1975. 

5. 	
University of Wisconsin Fusion Technology Program Document UWFDM-150, 11 A Nonci rcul ar Tokamak Power Reactor Design.. by R. W. Conn et al., July 1976. 

6. 	
University of Wisconsin Fusion Technology Program Document UWFDM-175, .. What Is Past Is Prologue: Future Directions In Tokamak Power Reactor Design Research.. by R. W. Conn, 1976. 


11-7 

7. 	
Oak Ridge National Laboratory Document ORNL/TM-5813, 11 0RNL Fusion Power Demonstration Study: Interim Report11 by Don Steiner et al., March 1977. 

8. 	
General Atomic Co. Document GA-A13992, 11 Conceptual Design Study of a 
Nonci rcul ar Tokamak Demonstration Fusion Power Reactor11 b:Y C. C. Baker, 


D. W. 	Kearney et al., November 1976. 

9. 	
International Atomic Energy Agency Document STI/PUB/487, Vienna, 1978, 

p. 185, 11 Conceptual Design of a Compact Tokamak-Reactor.. (IAEA-TC-145/13) by S. Forster, P. Kupschus, F. K. Altenhein et al. 

10. 	
Massachusetts Institute of Technology Plasma Fusion Cente,r Research 

Report RR-79-2, 11 High Field Compact Tokamak Reactor (HFCTR) Conceptual Desi gn 11 by D. R. Cohn et al., January 1979. 

11. 	
University of Wisconsin Fusion Technology Program Document UWFDM-330, 11 NUWMAK -A Tokamak Reactor Design Study .. by R. w. Conn E!t al., March 1979. 

12. 	
Minutes of a briefing to the U.S. Office of Fusion Energy on July 12, 

1979 by c. C. Baker, D. DeFreece et al. of Argonne National Laboratory and the McDonnald-Douglas Corporation. 

13. 	
Lawrence Livermore Laboratory Document UC ID-17644, 11 Stanelard Mirror Fusion Reactor Design Study 11 by R. W. Moir et al., January 1978. 

14. 	
Lawrence Livermore Laboratory Document UCRL-52302, 11 Prel iminary Design Study of the Tandem Mirror Reactor (TMR) 11 by R. W. Moir E~t al., July 1977. 

15. 	
Lawrence Livermore Laboratory Document UCRL-52467, 11 ConcE~ptual Design of the Field Reversed Mirror Reactor11 by G. A. Carl son et al., May 1978. 

16. 	
Lawrence Livermore Laboratory Document UCID-18158, 11 Parametric Studies of Tandem Mirror Reactors .. by G. A. Carl son et al., April 1979. 

17. 	
Engineering Test Facility Design Center Newsletter, D. Steiner Editor, Oak Ridge National Laboratory, July 1979. 

18. 	
Minutes of a briefing to the u.s. Office of Fusion Energy on July 19, 1979 by W. M. Stacey of the Georgi a Institute of Technology. 


II-8 

Table Il-l  
COMMERCIAL  FUSION  POWER  REACTOR  PARAMETERS  
TITLE  CULHAM MK-I  UWMAK-I  PPPL MATT-1050  UWMAK-II  UWMAK-III  JAERI-II  ORNL  DEMO  CULHAM MK-II  GA  DEMO  JlJLICH CTR  HFCTR  NUWMAK  
Design  Year  1972  1973  1974  1975  1976  1976  1976  1976  1976  1977  1978  1979  
Therma 1 Power (NW)(average continuous)  5830  4665  5146  4712  4735  2000  2150  5830  1676  5000  2470  2097  
Total  Electric Power  (MWe)  1889  2405  1807  2050  750  754  870  725  
.......  Net Electric Power  (tviWe)  250U  1473  2030  1709  1985  Not Given  2500  611  775  660  
.......  
I \0  Auxiliary Heating Power (MW)  15  None  200  200  100  100  100  80  
Duty Cycle  (%)  93.3  97  94.2  94.7  95  94.7  96  92  
Burn Time  (Seconds)  5400  6000  5400  1800  1200  800  500  225  
Major Radius  (M)  12.5  13  10.5  13  8.1  10.5  6  7.4  7  6.93  6  5.13  
Minor Radius  (M)  2.5  5  3.25  5  2.7  2.7  1.5  2.1  1.8  1.82  1.2  1.13  
Plasma Beta  (%)  1.5  5.2  4  6.4  8.3  3.3  15  9.3  10  15  4  6.5  
Toroidal Field (T)(on plasma axis)  9.5  3.8  6  3.6  4  6  3.6  4.1  3.9  3.7  7.4  6  
Plasma Current (MAl  9.7  25.7  14.6  14.9  ·15.8  10.4  4  11.6  18.6  8.8  6.7  7.2  
Reference  1  2  3  4  5  6  7  6  8  9  10  11  

Table II-2 Preliminary Starfire Parameters 12 
Average Thermal Power 
Duty Cycle 
Net Electrical Power 
Gross Thermoqynamic Efficiency 
Burn Time 
Recycle Time 
Major Radius 
Minor Radius 
Auxiliary Heating Power 
Toroi da 1 Beta 
Toroidal Field (on plasma axis) 
Plasma Current 
3600 MW 

94% 
1250 MWe 
36% 

455 sec. 
45 sec. 
6.92 m 
1. 73 m 
80 MW 6% 
5.3 T 
13.4 MA 
II-10 

Table II-3 Summary of Non-Tokamak Magnetic Confinement Systems Being
Proposed or Cons1dered as Reactors Magnetic Confinement Concept 
o 	Toroidal, steady-state 
Torsatron (T-1)


Elmo Bumpy Torus (EBT) 
Surface Magnetic Confinement (Surmac) 


o 	Toroidal, long-pulsed 
Reversed-Field Pinch (RFPR) 


o 	Linear, steaqy-state 
Tandem Mirror (TMR)


Field-Reversed Mirror (FRMR) Ion-ring compressor 

o 	Linear, pulsed 
Slowly imploding liners (LINUS) 


o 	Very dense, pulsed (mainly linear) systems 
Fast-Imploding Liners (FLR)



Dense Z-Pinch 
II-11 

Table II-4 
Preliminary ETF Parameters17 

Geometry Plasma radius, a 1.2 m Plasma elongation, o 1.6 Aspect ratio, A 4.17 Major radius, R 5.0 m Plasma edge-.to-wi ndi ng distance, t. 1.28 m 
Plasma Beta, a 7.0% Plasma temperature, T 12 keV 
3
Ion density, rr 2 x 1020 m-
Impurity indicator, Zeff <1.5 
Safety factor, q 3.8 
Plasma current, Ip 5.0 MA 
Burntime (steady-state), TB 500 sec 
Duty cycle, de 89% 
Cycle duration 560 sec 
Energy confinement time, TE 1. 2 sec 
Pellet fueling rate 4.6 X 10:~2 

part i c 'I es/ sec 
Beam power into plasma, Pinj 50 MW 
Beam energy, Einj 150 keV 
Microwave power, Prf 1 MW 
Microwave frequency, f 120 GHz 

3
Plasma volume, Vp 227 m
Fusion power density, Pd 5.0 MW/m:3 
Power (fusion), Pf 1140 MW 

a
Total power, Ptot 1450 MW 
Electromagnetic Toroidal field on axis, BT 5.3 T Maximum toroidal field at coil, Bm 10.9 T Number of TF coils 12 Magnetic field ripple at plasma edge, 6a 1. 5% peak to average TF coil type Superconducti ng
(Nb3Sn) 
OH solenoid type Superconducti ng
(NbTi) 

PF coil type 
Inside TF bore Copper 
Outside TF bore Supercon11ucti ng 

Total volt-seconds fvs 83 
Magnetic field in OH solenoid, B0H 8.0 T 

Thermal First wall coolant Water Shield coolant Borated 1~ater Refrigeration loaa (TF coils), Pref 48 kW at SK Power to first wall, Pu/2 113 MW Power to divertor, Pu/2 113 MW Power to shield, Ps 1224 MW 
3 X 1023
Plasma edge particle loss rate partie'! es/sec Nuclear Neutron wall loading, Lw 2.4 MW/nf
a22.4 MeV/event assumed. I I-12 
Table II-5 
ETF/INTOR PARAMETER COMPARISON 

Preliminary Parameter Major Radius (m) Minor Radius (m) Plasma Elongation Toroidal Fi~ld (T)
(At the magnet) Beta (%) Burn Time (sec) Auxiliary Heating
Power (mw) Auxiliary HeatingEnergy (keV) 
ETF INTOR18 
5 4.8 
1.2 1.2 
1.6 1.6 
10.9 10.5 
7 6 > 100 > 100 50 75 
150 200 
11-13 


I I I. IV1AliNET ICS 
A. Overview High magnetic fields are needed to confine and control the ultrahot plasmas in fusion reactors. lienerating these fields requires large amounts of electrical energy when conventional conductor materials such as copper are used in the · electromagnet windings. In fact, the energy needed is so great that net power production would not be possible. Although it has been necessary and economical to use pulsed electromagnets with conventional conductors to develop plasma physics information on small experimental devices, stea<t)' state superconducting magnets that consume much smaller amounts of energy will be necessary for future fusion power facilities. 
To provide the various superconducting magnets that will be needed for fusion, various technology development and demonstration efforts are underway. In the U.S., the technology will be demonstrated by integrally considering the results from the large 7.5 Tesla steaqy state magnet being fabricated for the Mirror Fusion Test Facility at Lawrence Livermore Laboratory, the 8 Tesla steaqy state magnets being fabricated for the Tokamak related Large Coil Program, the 12 Tesla superconductor development for both Mirror and Tokamak applications, the pulsed superconducting coil program, and a supporting low temperature structural materials program. 
The approach is to demonstrate that large steaqy state and pulsed 7 to 8 Tesla magnets based on NbTi superconductors can be manufactured and operated successfully at design conditions. These magnets will confirm the design 
codes, the superconductor manufacturing capability, the structural materials 
I II-1 
and fabrication techniques, the cooling option-s, and the supporting electrical 
and cryogenic equipment. In parallel; a highe-r field supE!rconductor based on 
Nb3Sn is being developed. It is felt that the higher field superconductor 
can be directly combined with the experience gained from the large 7 to 8 
Tesla magnets into a 10-12 Tesla design that can be fabricated with a high 
assurance of success. 
B. Large Coil Program To develop the steaqy state, 8 Tesla magnets for Tokamaks, an international 
cooperative program, the Large Coil Program, is underwczy. The Large Coil 
Program'consists of six large magnets and a test facility in which to test 
them at the Oak Ridge National Laboratory, U.S.A. Three coils are being 
fabricated in the U.S. and three are. being supplied for testing from Euratom, Japan, and Switzerland under an International Energy Agency agreement. 
The six magnets are being constructed to common geometric dimensions and will weigh about 40,000 kilograms each. Each magnet is being independently designed and constructed to maximize design and construction innovations. Table III-1 compares the features of each of the coils. 
The Large Coil Test Facility under construction at ORNL will have the 
11 011
capaoil i ty of testing the shaped 2.5 X 3.5 meter bore superconducti ng coils singularly and in two, three and six coil toroidal arrays at peak 
magnetic fields of 8 Tesla. With sligh-t modifications, the facility will have the capability of testing at 12 Tesla fields and will accommodate one or two large superconducting coils of the size anticipated for the next large tokamak 
IIl-2 
fusion device, the Engineering.T~st Facility (ETF), a U.S. design or the 
INTOR, an_international design ..Csee section II). Figure III-1 presents the 
features of the test facility •. Testing of the magnets is scheduled to begin 
in 1981 and will run through 1984 .as shown in Figure III-2. 
C. Pulsed Coil Efforts In addition to steaqy state magnetic fields, there are other components for fusion devices that provide pulsed magnetic and electric fields which are needed to heat and control the plasma. Tokamaks in particular will need a pulsed superconducting magnet to induce a current of several million amperes to ohmically heat the plasma as well as to stabilize and position it. 
Development efforts are underway to provide the technology needed for pulsed superconducting magnets. The primary need is for a high current superconducting cable that can be charged and discharged rapidly (in about 1-5 seconds) and that creates minimal heat due to mechanical movement and electrical losses. The losses should be kept at 0.3% or less. Low losses are extremely important because these losses which appear as heat have to be removed at liquid Helium temperature (4.2°K) which will directly effect the amount of helium cryogenic equipment needed (1 watt of heat generation in the superconductor results in about 350 watts of refrigeration power requirements). 
Substantial progress has been made with pulsed superconducting coils. Model tokamak ohmic heating systems have demonstrated that pulsed superconducting coils can be swung in a bipolar fashion at rates of 12 to 20 Tesla per second without quenching. A effort is underway at Los Alamos Scientific 
III-3 

FIGURE 111-1 LARGE COIL TEST FACILITY 
DESIGNED FOR 8-12 TESLA ACCEPT ONE TSA SIZE COIL 5%M x 7%M 
CRYOGENIC SYSTEM LIQUID HELIUM LIQUID NITROGEN 
........ 

........ 

........ 

•I ORMAK TANK 
~ 
SEGMENT TESTS 
S,S VACUUM VESSEL 11 M DIA. x 13M HIGH LIQUID NITROGEN 
COOLED WALLS 
6-COIL TORUS ARRAY PULSE MAGNETIC FIELD SIMULATED NEUTRON 
HEATING NbT1 & Nb3 Sn COILS POOL BOILING & FORCED FLOW POWER SUPPLIES & INSTRUMENTATION 
LARGE COIL 
2.5 M x 3.5 M BORE -8 TESLA; 40,000 Kg 
U.S. & FOREIGN COl LS 
FIGURE 111-2 MILESTONES AND SCHEDU.LE FOR THE'LCP 
RECEIVE
u.s.
START RECEIVE Nb3Sn 
BEGIN
COIL START START COIL u.s. COIL FULLDESIGN LCTF FABRICATION NbTi AND TORUS(GO, GE, W) CONST. (GO, GE, W) COILS lEA COILS TESTS 
SIGN
LCP lEA
PLAN TASK 

...... ! ! ~
.......

.......

I
(.}'J 
U.S. COILS 
3-COIL
LCTF PULSE 
8-COIL

1
FY 1977 1978 . 1979 1980 1981 1982 1983 
Laboratory to design and fabricate a 50 KA, 20 MJ coil which can be swung! 8 Tesla in one second. A 50 KA cable has been tested but its losses would be too high at 1 second pulse times. A 50 KA cable incorporating a copper-nickel resistive barrier which reduces coupling losses at short pulse times is now being fabricated. If the cable tests are satisfactory, a 20 MJ coil ,will be fabricated and tested in 1984 to demonstrate that this component is sufficiently developed for utilization in ETF or INTOR. 
The pulsed coil is one component of the tokamak ohmic heating system, which also includes high power switches and a rotating inertial energy storage device, to momentarily accept the energy from the pulsed coil and theri efficiently feed the energy back to the coil in a reversed polar1t:Y. In this way, 
the pulsed superconducting coil provides a maximized change of magnetic flux. 
D.C. homopolar motor generators and A.C. flywheel generators "'ith solid state rectifiers have been considered as the pulsed inertial energy storage component. At pulse times of one second or less,· ·the homopolar machines appear to be best; at times greater than a second, the A.C. flywheel mac:h1nes with 
rectifiers appear to be the better choice. Longer start-up times for Tokamaks substantially ease the ohmic heating system development requir·ements. 
D. The 12 Tesla Coil Program The need for the development and qualification of large superc:onducting magnets for tokamak toroidal field coil service, capable of reliable J>erfonnance to fields of 11-12 Tesla, has emerged as a result of ongoing studies by various 
Tokamak reactor conceptual-design teams. The results of a sbJQy by the ORNL 
LCP program manager indicated that redirection of one or more of the present 
III-6 

~ Tesla coils to produce a like number of 12 Tesla coils would incur a 1-2 year delay in delivery of these coils, relative to projected delivery dates 
for the~ Tesla coils, and additional costs of "'$10 million (for one 12 Tesla 
coil) to "'$26 million (for three 12 Tesla coils). Hence, it is imperative and cost-effective to complete the present Lar~e Coil Project on schedule, as this will provide answers to many of the questions regarding the feasibility of 
designing, faoricating, and operating large, high-field, superconducting magnets. Just as clearly identified, however, is a need for the design, 
development, construction, and operation of some type of test coil, capable of supplying test results that demonstrate the feasibility of various combinations of superconducting material, superconductor design and fabrication method, and coil cooling method, that functions reliably in fields to 12 Tesla. To meet this challenge is· the objective of the program described below. 
The objective of the 12 Tesl a Coil Program is to provide test-demonstrated superconductor:cooling schemes, in th~ form of a 1 m o.d. x 0.4 m i.d. x 0.2 m thick superconducting magnet that s~ccessf~lly operated at 12 Tesla, to the ETF Project Manager for his consideration and possible incorporation into the ETF conceptual design. This objective is being achieved through the design of superconductors and test coils, development of necessary superconductor processing, and incorporation of these accomplishments into four coils. The coils will be instrumented and-tested in fields to at least 12 Tesla in the Livermore High Field Test Facility (HFTF). 
III-7" 

The 12 Tesla coil/coolant/superconductor parameters were derived from requirements for the ETF toroidal field coils1 , as shown in Table III-2. 
The overall organization of the program and each industry and national laboratory•s responsibilities are summarized in Table III-3. There are four coil project teams; each consisting of a national laboratory or general engineering contractor with overall project-management responsibilities, an industrial concern with experience in coil and conductor-design and fabrication, and a superconductor-supplier. Depending on the detailed organizational capabilities within each team, the responsibility for coil and conductor design and testing 
is shared in various proportions between the national laboratory and the 
industrial members of the team. A description of each team and their activi
ties follows: 
Team One is designing and building a coil using alloyed NbTi conductor and pool boiling sub-cooled He to produce fields in the 11-12 Tesla range. The team is managed by General Atomic Company who is participating in the design 
of the conductor and test coil with Magnetics Corporation of America (MCA), a 
manufacturer of superconductors and magnets with extensive experience in 
processing unalloyed and alloyed NbTi. MCA, in addition to producing the 
conductor and building the test coil, is also doing scale!-Up studies on 
alloyed Noli compositions developed both by themselves and by researchers at 
the University of Wisconsin. 
Team Two is designing and building a coil using rrultifilalmentary (MF) No 3sn 
conductor and forced-flow cooling for operation at 12 Tesla. The team is led 
oy the Francis Bitter National l~agnet Laboratory (FBNMLL. which is sharing 
III-8 

design and fabrication of the conductor and coil with the Large Rotating 
Apparatus Division of Westinghouse Corporation. Conductor process development 
is being done by both Airco Superconductors and Supercon, as both firms have 
experience with fabrication of force-cooled conductors. Selection of the 
conductor supplier will be made in 1980. 
Team Three is designing and building a coil using a MF Nb 3Sn conductor based 
on the General Dynamics-Convair LCP conductor concept2 and pool-boiling at 
4.2K for operation at 12 Tesla. The team is headed by Oak Ridge National 
Laboratory (ORNL), and consists of General Dynamics-Convair who provides coil 
and conductor design and fabrication services in conjunction with Airco 
Superconductors, who is also responsible for conductor process developmen~. 
Team Four is designing and building a coil using a MF NB 3Sn conductor based 
on the General Electric Company LCP conductor2 with the cooling concept to 
be chosen early in the first year's design efforts. The team is led by ORNL, 
and the Large Superconducting Projects Department of General Electric Company 
is providing coil design and fabrication services. Intermagnetics General Corporation (IGC) is providing conductor design, process development, and fabrication services. Should evaluation of Nb 3Sn conductor produced by the 
.. external diffusion process.. 3 prove unsuccessful, use of material produced 114
by the 11 bronze process wi 11 provide an acceptable alternative. 
The Test Facility is the Livermore Laboratory High Field Test Facility (HFTF), 5 which is being upgraded to 12 Tesla capabilities. Development and evaluation 
of the MF Nb3Sn conductor for the HFTF is a joint responsibility of Livermore and Airco Superconductors. Detailed test facility: test coil interface 
III-9 

requirements are coordinated by Lawrence Livermore Laboratory. Test plans for each coil are the responsibility of each coil project manager and are being developed in conjunction with Livermore and will be reviewed by a program advisory group whose composition is shown in Table III-4. 
Much of the first year's efforts have been spent in definition,, negotiation, and placement of the many sub-contracts by the National Laboratory project manager with eight industrial concerns who are performing the bulk of the work. Specific technical accomplishments are described below: 
Team One have concentrated their efforts in three main areas: 
o 	Preparation of a reference conceptual design of a ETF-cornpatible toroidal field coil to the design requirements of Table III-2, with particular emphasis being placed on identification of alternative con
ductor designs capable of using alloyed Nb-Ti cooled by pool boiling He in the 2-3°K range. These efforts are underw~ and are scheduled to be completed by March 1980. 
o 	Investigation of potential ductile Nb-Ti-X (X =Ta or Ta plus Hf) alloy candidates, using enhancement of the upper critical field at 2°K as the primary screening too1. 6 Most promising results have been . seen.for Ta added at the expense of Nb in the range 10-30 atom percent (A/o) with a Nb-65 A/o Ti -10 A/o Ta alloy representing a promising candidate for scale-up studies (Figure III-3). Note that the effect of upper critical
field-enhancement of Ta becomes more pronounced as the temperature decreases (Figure I II-4). 
III-10 

FIGURE 111-3 
UPPER CRITICAL FIELD AS A FUNCTION OF 
COMPOSITION FOR Nb-Ti-Ta ALLOYS AT 2°K 

Ti 
ATOMIC/ \ATOMIC o/o Ti/ 

\o/o Nb 
•11.9 
•12.9 
15.2 15.3 
•15.3 e15.4 •15.2 
15.1••15.45 15.35 

14.75 •15.1 •15.25 •15.2 • 

•14.9 •14.7 
•111 •114 
50 40 30 20 10 
ATOMIC o/o Ta 
-He = 14.2 TESLA IE He OF Nb 46.5 W/0 Ti 
2 2 
He ~ 15.3 TESLA 

2 

III-11 


FIGURE 111-4 
EFFECT OF TEMPERATURE C•N UPPER 
CRITICAL FIELD OF COMMERCIJ~L Nb-Ti AND 
A Nb-Ti-Ta ALLOY 

16 -Nb-43 W/0 Ti-25 W/0 Ta 
<t 
...J 
en 
w 
-1-15 c 
...J 
w
-
LL 
...J 14
<t 
(.)
-
t
-
a: 
(.) 
a: 13 
Nb-46.5 W/0 Ti
w 
Q. 
Q. ::J 
12 
11----~--~--~--~----~· 
0 1 2 3 4 
TEMPERATURE, °K 

III-12 

o 	Investigation of the effects of small amount of Ta, up to 2 w/o, on the critical current performance of the commercial Nb-46.5% Ti alloy at 
I 
both 4.2 K and 1.9 K. Results are summarized in Figure III-5 which also includes some data from previously published studies. 7' 8 Note that these small Ta additions increase the critical current-density of the basic alloy at all fields and temperatures investigated, with the greatest increase occuring in the 0.5% Ta alloy. 
Team Two has concentrated its efforts in two main areas: 
o 	Preparation of a preliminary design of a 1m diameter test coil, using a force-cooled multi-filamentary Nb3Sn superconductor. The preliminary design has recently been completed9 and is presently under review by the proposed fabricator, the Large Rotating Apparatus Division of the Westinghouse Corporation. 

o 	Superconductor development is underway by ooth Airco Superconductors and Supercon. Emphasis of the former company is being placed on modification and extending the range of service of their Large Coil Project internal oronze process conductor10 from ~T to 12T by optimizing Nb 3Sn activation heat treatments for 12T performance. The latter company is undertaking an extension of their modification of the internal bronze 


11
process , wherein Cu-Sn rods are wrapped in several thicknesses of Nb foil and the resulting sub-assemblies placed in a pre-drilled Cu billet, extruded and drawn to size, and reacted to form Nb 3Sn on the inside of the Nb foil. The main advantages of this process are {1) the Nb foil 
III-13 

FIGURE 111-5! 
1cf 
~,N CRITICAL CURRENT-DENSITY:
~~
<(0 EXTERNAL FIELD PERFORMANCE 
~ ~ 
FOR Nb-46.5o/o Ti AND Nb-46.5o/o Ti 
enzw ALLOYS WITH SMALL Ta
ADDITIONS AT 4.2 °K 

0
t-I 
z 104 
w 	4 8 12
a:
a:
:J CRITICAL CURRENT-DENSITY:
0 EXTERNAL FIELD PERFORMANCE

....1
<(
0 FOR Nb-46.5o/o Ti AND Nb-46.5o/o Ti 
!:: ALLOYS WITH SMALL TA

a: ADDITIONS AT 1.9 °K 	~ ~ 
0 	Ci)z
w .,:.0 
z
w
a:
a:
:J0 
....1
SYMBOL COMPOSITION 	<(
0
Nb-46.5% Ti
• 	E
0 Nb-46.5% Ti-0.5% Ta 
a:
Nb-46.5% Ti-1.0% Ta 	0•6 Nb-46.5% Ti-2.0% Ta 
------------~----~~----~----~----~--~~1~
0 	4 8 12 
EXTERNAL FIELD, TESLA 
III-14 
tubes serve as diffusion barriers so that Ta barriers are not needed, and (2) each Nb3Sn element is surrounded by a high conductivity Cu matrix, hence transfer lengths are shorter than in multi-filamentary Nb3Sn superconductors made by other means. 
Preliminary results on both Airco•s and Supercon•s efforts are presented 
in Figure III-6, wherein ranges of critical current densities at 12 
Tesla for both efforts are compared with published values for commercial 
Nb	Sn superconductors12 , 13 and to the calculated design point for
3
the M.I.T. lm coil. 8oth sets of results are encouraging. 

Team Three has concentrated the efforts in two areas: 
o 	Airco Superconductors has undertaken a potentially vital cost reduction modification to the manufacturing cycle for their internal bronze process Nb3Sn superconductors, which involves replacing the usual pre-drilled bronze matrix filled with Ta-wrapped Nb rods with an assembly 
of bronze rods stacked in a copper can, each rod having a hexagonal 
exterior and a cylindrical hole containing a Nb rod. Commercial mills 
are being used to evaluate fabrication methods for the bronze tubes, and 
no problems have been encountered throUgh final tube-reduction operations. Upon successful production of the bronze tubes on existing commercial equipment, an extrusion billet will be assembled, extruded and drawn to wire, with Nb3Sn heat treatment cycles optimized for 12T service and subsequent product evaluation. 
III-15 

FIGU~E IH-6 
' CRITICAL CURREr\IT-DENSITY ', VERSUS EXTERN~~L FIELD AT ' 4.2 °K FOR COMMIERCIAL AND ', EXPERIMENTAL MULTI', FILAMENTARY Nb3Sn ' SUPERCONDUCTORS 
\ \ \ \ ' ' ' ' ' ' ' '\ 
\ 
\ ' ' ' ' \ 
en D.. N 
::E ::E <( 0 
' 
~ 
\ 
en 
\
z 
w \ 
c ' ' ' ' ' ' ' ' ' ' ' ' ' ' \ 
.... \
z I 1o5 
w \ 
a: 

a: 


;::) 
0 
...I 
<( 

0 
!::: ' ' ' ' \ 
I 
0 
M.I.T. COIL DESIGN POINT 
a: ' ' @ 
A RANGE OF AIRCO VALUES 
\\ Ts 
s RANGE OF SUPERCON VALUES 
I RANGE OF I.G.C. VALUES 
'\I
---RANGE OF VALUES FOR DOMESTICALLY-PRODUCED COMMERCIAL Nb3Sn CONDUCTORS 
,~~----------------~~----~-----------~-----
0 4 8 12 
EXTERNAL FIELD, TESLA 
111-16 
o 	ueneral Dynamics-Convair have completed the first cycle through a conceptual design of a ETF toroidal field coil based upon the ETF T.F. coil design specification (Table III-2) using a pool-boiling cooled Nb3sr~ superconductor. Preliminary results indicate that a potentially serious proolem may be the low working stress of the strain-hardened Cu stabilizer at the low strains of U.l% to 0.3% necessary to insure integrity of the brittle Nb3Sn conductor. Both ueneral Dynamics and Airco are collaborating in the design of the conductor, using the ETF T.F. coil design specification as a starting point. Emphasis is being placed on concepts similar to the General Dynamics LCP conductor2• 
Team Four has concentrated its efforts in two areas. 
o 	Intermagnetics General Corporation (IGC) is concentrating their efforts on extending the maximum size Nb 3Sn superconductor that can De manufactured by the 11 external bronze .. process from about 0.00611 diameter3 to o.u1211 diameter wire with an overall critical current-density of 5X104 Amp per sq. em. at 12 Tesla. In the past, attempts to produce conductors with diameters above about u.U06 11 were unsuccessful due to separation of the plated Sn layer through lines of voids formed during the reaction heat treatment cycle. Through work done both on IGC funding and at 
M.I.T. by R. Rose and co-workers14 , it has been found that these voids are formed by heterogeneous nucleation of vacancies from a local supersaturation, such as exists at inclusion/matrix interfaces. The voids nucleate preferentially at cold-welded interfaces, such as interfaces in the original extrusion billet between the Nb rods and Cu stabilizer and 
III-17 

between the Cu stabi 1i zer and Cu extrusion can, and are probably found at impurity particles in these interfaces. While application of these concepts to successful production of large diamete1~ (> 0.01011 ) Nb3Sn conductors by the 11external bronze process.. is stilll in the 11 proof-ofprinciple11 stage, the preliminary results obtained by IGC on this program and shown in Figure III-6 are quite encourclging. 
General Electric is performing a scoping study of a 12 TE~sla ETF T.F. Coil, using the ETF coil design values as a reference. They ar·e also doing a trade-off study between cooling by pool-boiling and forcE~d flow. Both of these efforts are being carried out to. enable selection of the 11 best11 Nb3Sn superconductor for use in the '1m diameter coil to be made. Work accomplished to date includes definition of E.T.F. T.F. c:oil-geometry and oegintiing the analyses for magnetic field and force, conductor stress and strain, structure stress, and thermal cooldawn stress. 
HFTF Status Actual design of the facility is complete and fabricatiol'll of the coils and structure is underway. Two backing coils using NbTi conductor (Coils #1 and #3 in Figure III-7) are complete and have been 11 proof-tested11 during verification tsting of the MFTF test coil 15• Two other NbTi beitcking coils (#2 and #4 in Figure III-7) are presently being wound. Some difficulties thdt arose with fabrication of prototype lengths of the NbSn conduc:tor for the inner
3coils (#5 and #6 in Figure III-7) have been overcome by ct change in conductor design, and fabrication of the conductor for these two coils is undeno~ay. Expected completion time for the facility is early 1981. 
III-18 

FIGURE 111-7 
12T HIGH FIELD TEST FACILITY 
IN 2 METER DIA. CRYOSTAT 

~ 
I 
L 
tJ.....f! ~ ·-IJ . 
()' .t> 
.,' (J • 
. 
. 

...
u 
/0\ 
~· I 
q 
.. 
. e 
q ·. 
0· 
·0
} 0. ' 
I I II 
' 
I 
: 
.. 
0
0 
() . ; . ·t:~· 'I? 
" ()'·(/ 
I I  A  
w  
I  o·., . ~ . ·a . tJ . • :  
I  ...  
D·. ,.  
o:  
.. ...  
I  
I  r: :  
,.u,  
/  ~  I  ·. o· .. 0 ' t)..  
0  
0·  
~  o·.  
I I  ~  5 CM.  GAP  
I  I  I~  + 1  
I I  [>'\ ~ 11 CM [?(~ I o'  
0  
0  0 ...  
~  .  
·o ·-=· &>  o ••  

III-19 

References 
1. 	o. Steiner (Editor), .. Engineering Test Facility Design Center Newsletter.. ,July 1979, Pg. 2. 
l. 	P. N. Haubenreich, J. N. Luton, and P. B.Th·ompson, 11 The Role of the 
Large Coil Program in the Development of Superconducting Magnets for 
Fusion Reactors .. , I.E.E.E. Transactions in Magnetics, MAG 15(1),
Jan. 197i, Pg. 520-524. 

3. 	
w. D. Fietz, C. D. Henning, and R. M. Scanlan, ..Multi-Filamentary Nb3Sn 

Conductor For Fusion Research Magnets.. ~ I.E.E.E. Transactions in Magnetics, ivtAG 11(2), March 1975, Pg. 299-302. 

4. 	
E. Gregory, w. Marancik, and F. Ormand, 1'Co111>osite Conductors Containing1VIany Filaments of Nb3Sn .. ,· I.E.E.E. Transactions on Magnetics, MAG 11(2),March 1975, Pg. 295-z9~. 

5. 	
D. w. Deis, D. N. Cornish, J. P. Zbasnik, R. L. Nelson, S. J. Sackett, 
and C. E. Taylor, .. Superconducting Magnet Deveiopment Program Annual 
Report11 , 7/22/77, UCRL-50031-76, Pg. 56-57. 


6. 	
D. G. Hawksworth and D. C. Larbalestier~ 11 Enhanced Values of Hc 2 in 
Nb-Ti Ternary and Quarterary Alloys'!, Paper BA-5, ICMC Conference,


August 1979. 	· 

7. 	
C. N. Curtis and w. K. MacDonald, .. Suitability of NbTi Containing up 
to 2% Ta for Use in Fabrication of Superconductors.. , I.E.E.E. Trans
actions on Magnetics, MAG 15(1), Jan. 1979, Pg. 768-771. 



H. R. Segal, K. Hemachalam, ~T. de Winter, and D. A. Colling, ..Develop
~
ment of Nt>Ti Conductors for lOT-14T Operations .. , final report on D .O.E. 
Contract Eu-77-C-02-4180, 9/30/78. 
9. 	M. o. Hoenig, 11 M.I.T. 12T Test Coil Preliminary Design.. , Specification 
~IT-790308, 7/9/79. 

lU. 	P. A. Sanger, E. Adam, E. Gregory, N. Marancik, E. Mayer, G. Rothschild, and M. Young, 11 Developments in Nb Sn Forced-flow Conductors for Large Magnets 11 , I.E.E.E. Transact1ons in Magnetics, MAG 15(1) Jan. 1978,Pg. 789-9. . 
11. 	
D. Cornish, R. Hoard, R. Randall, R. Scanlan, J. Wong, and J. Zbasnik, 

11 Fabrication and Properties of Multiflamentary Nb3Sn Prepared by the Internal Bronze Techni que11 , Paper CA-9, ICMC, Augast 19'79. 

12. 	
c. R. Spencer, P. A. Sanger, and M. Young, 11 The Temperature and Magnetic Field Dependence of Superconducting Critical Current Densities of i'vtultifilamentary Nb3 Sn and NbTi Composite Wires .. , I.E.E.E. Transactions in Magnetics, ~AG 15(1), Jan. 1979, Pg. 76-79. . 


111-20 

13. 	
~. A. Fietz, "Nb3Sn in 1978: State of the Art", I.E.E.E. Transactions in Magnetics, MAG 15(1), Jan. 1979, Pg. 67~73. 

14. 	
S. Cogan, D. s. Holmes, and R. M. Rose, "On the Elimination of Kirkendall Voids in Superconducting Composites", Accepted by the Journal of Applied 

Physics, Letters, July 1979. 	· 

15. 	
o. P. Cornish, Ji P. Zbasnik, R. L. Leber, D. G. Hazel, J. E. Johnston, and A. R. Rosdahl, 11 M.F. T .F. Test Coi 1 Construction and Performance .. , 


I.E.E.E. Transactions on Magnetics, MAG 15{1), Jan. 1979. Pg. 530-533. 
JII-21 

Core Bore (specified) 
Peak Field (design, specified) 
Ampere-Turns (design) 
Conductor Current (design) 
Conductor Material 
....... 

....... 

....... Conductor Configuration 

I N N 
Helium Conditions 
winding Configuration 
Strt:ctural !'1ateria1 
Structure Configuration 
GO/CONVAIR 

2.5 x 3.5 m 
8.0 T 
6 
6.65 X 10 10,200 
NbTi 
Cable in extendedsurface copper strip 
Pool Boi11ng
(4.2 K, 1 atm) 
Edge wound in layers (14) 
304L SS 
Fully welded case 
Table lll-1 LCT TEST COIL FEATURES 
GENERAL 
ELECTRIC 	WESTINGHOUSE 

2.5 x 3.5 m 	2.5 x 3.5 m 
H.O T 	~.0 T 
6 	6 
6.98 X 10 	7.36 X 10 
10,450 	16,000 
NbTi 	Nb3Sn 
16 sui>elements, Cable (insulated spiraled aroun~ strands) in square copper core conduit 
Pool Boiling 	Supercritical, 
(4.2 K, 1 atm) 	Forced-Fl(JIII 
Flat wound in Laid in spiral pancakes (14) grooves in 26 pancake plates 
316LN SS 	2219-T87 plates 
A286 bolts 

Welded case 	Grooved pancake 
with bolted 	plates, bolted 
closure 
EURAT(J4 

2.5 x 3.5 m 
8.0 T 
6 
6.62 X 10 
11,000 
NbTi 
22-subelements spiraled around CrNi core, inside rectangular conduit 
Supercritical, 
Forced-Fl(JIII 
14 pancakes 
Stainless steel 
similar to 316LN 
Welded case, 
bolted or welded 
closure 
JAPAN 

2.5 x 3.5 m 
8.0 T 6 
6.76 X 10 
10,300 
Nbli 
Cable in roughened copper strip 
Pool Boiling 
Edge wound in pancakes (20) 
Stainless steel similar to 304L 
Welded case, 
bolted side 
closure 
SWITZERLAND 

2.5 x 3.5 m 
8.0 T 
6 
6.6 X 10 15,000 
Nbli 
Solder-filled cabl~ in square conduit 
Supercritical, 
F orced-F 1ow 
24 pancakes 
Stainless steel 
simi iar to 316Lr.i 
Bolted case 

TABLE III-2 
ENGINEERING TEST FACILITY TOROIDAL FIELD COIL PARAMETERS 

ADAP1ED FOR USE BV THE 12 TESLA COIL PROGRAM PARTICIPANTS 
1. 
Plasma Major Radius 	~ 5 m 

2. 
Coil Major Radius ~ 	5.8 m 

3. 
Coil Bore Size -	6 m horizontal X 10 m vertical 

4. 
Coil Shape -Modified D-Shape 

5. 
Number of Coils -12 

6. 
Field on Plasma Axis~ 5.8 

7. 
Peak Field at the Winding -121 

8. 
Conductor -Nb 3Sn or Alloyed Nb-Ti 

9. 
Field Profile -Maximum on the center line 

10. 
Amp Turns/Coil ~ 12 MA 

11. 	
Current Density over Winding Pack ~ 1700 A/em~ 
over Conductor ~ 2200 A/em 


12. 
Operating Current -10 to 15 kA 

13. 
Stored Energy /Coi 1 ~ 1500 MJ 

14. 	
Stability Margin -1/2 turn length at ~he high field region can withstand 100 mJ/cm without undergoing a quench 

15. 
Tolerance for Pulsed Fields -


A. Normal operation: 
1. B1: 0.15 T/sec for 1 second up, 1 second down 
2. Bu: 
II II II II II II II II 
II 
3. Repetition rate: 	1 pulse every 5 minutes 
B. Upset condition: 
Simulation of a plasma disru~tion by a downpulse in Bv of O.ST in O.OlU second with a repetition rate of once a week. 
16. 
Tolerance for radiation: 1 X 109 	rads 

17. 	
Vacuum Topology: Bell jar with re-entrant holes II I -:~3 


TABLE III-3 
PROJECTSa IN 12 TESLA COIL PROGRAM 
Participants 
Project Management and General Atomic FBNML ORNL ORNL Technical Direction 
Conductor Design Wisconsin/MCA Supercon IGC AIRCO and Production 
Co i1 Design and 	GA/MCA Westinghouse General Electric General Dynamics
Fabrication 
• 

Concepts 
...... 
...... 
...... 
Conductor Material 	NbTi Nb 3Sn· Nb 3Sn Nb3Sn
I 
1"-'
-s::. 
Conductor Configuration 	b Internally cooled b Externally cooled 
cable in conduit 	cable or braid joined to stabilizer 
Coolant Conditions 	Low-temperature Forced-flow b Pool-boilingpool-boiling supercritical (possiblysupercri tical) 
aDeve 1opme,rlt of 12 Tesl a conductor for the Mirror Program by LLL AIRCO is di sti net from tokamak\ boriented work, but the two are being closely coordinated. To be detenmined during conceptual design phase of project. 
TABLE II I-4 

MEMBERSHIP OF 12 TESLA COIL PROGRAM REVIEW COMMITTEE 

1{. ~oom U. of Wisconsin 
u. 
N. Cornish LLL 

J. 
File PPPL 

P. 
N. Haubenreich ORNL 

M. 
0. Hoenig FBNML 

M. 
s. Lubell ORNL 

J. 
R. Purcell GA 

E. 
N. C. Dal der DOE, ex officio 

L. 
Dresner ORNL 

w. 
Fietz ORNL 

J. 
Alcorn GAC 


I II-25 


IV. 
PLASMA HEATING FUELING AND EXHAUST 

A. 
Neutral Beams Neutral beam heating systems are a basic component of the current tokamak and mirror experimental confinement devices as well as the next generation of tokamak and mirror systems. Since toroidal and minimum-S mirror magnetic field geometries permit many Coulomb collisions during a plasma confinement time, it is possible to heat the plasma using neutral beam injection with beams of kinetic energy many times greater than the desired plasma temperature. The injected neutrals interact with the plasma via ionization or charge exchange, and the resultant trapped energetic ions slow down gradually by collisions with the plasma electrons and ions. 


Recent results from PLT 1 show that.reactor-like temperatures are achievable in tokamaks and that the potential deleterious effects of various instabilities that were predicted to exist at low collisionalities do not seriously degrade plasma confinement at these temperatures. These results coupled with recent analysis indicate that it is reasonable to expect to be able to heat a large tokamak to ignition with positive ion neutral beam systems of energy in the range of 120-200KeV. As will be discussed below, neutral beam systems are beginning operation at 120KeV for TFTR and work has been initiated in the USA 
to increase the energy and to extend the pulse length capability of the systems to several seconds. 2 
The confinement of a simple mirror is also directly related to neutral beam energy. The injection current requirements of a tandem mirror are considerably reduced over those of a single mirror cell, although the 
IV-1 

required beam energy remains high {"' 1MeV). Recent new concepts to improve tandem mirror performance by use of RF energy have resulted in further reductions in neutral beam requirements in that neutral beam en1ergies of a few hundred KeV appear reaso nab 1e. While this sti 11 requires development of negative ion neutral beam systems, it is a far less ambitic>us requirement. The tandem central solenoid neutral beam energy is chosen in much the same· way as for a tokamak of comparable n,T and -r to provide ad,equate heating or two component operation as desired and to replenish particle losses. 
1. Neutral Beam Systems for Current Projects The neutral beam needs for current generation confinement devices are 1isted in Table III-A-1. The neutral beam system for PLT {40 KeV, 60 A, 0.3 sec) was developed by Oak Ridge National Laboratory {ORNL). 3 Using this system on PLT, record high temperatures for tokamak plasmas have been reached. 1 The neutral beam injection systems for ISX and PDX are based on the PLT Neutral Beam System. 3 ISX is using the PLT source initially and then an upgraded, 100 A s.ource will be installed shortly. The PDX system will also utilize a 100 A source, upgraded to 50 kV operation at a longer pulse length. The POX beam 1 i ne is essentially the same as used for ISX, except that some changes are required to accommodate the longer PDX pulse 1 ength ( 1/2 s~econd versus 1/10 second on ISX). The PDX Neutral Beam System is shown in Figure IV-A-1 and the source in Figure IV-A-2. 
The neutral beam injectors for TFTR, MFTF, and Doublet-III are based upon the Lawrence Berkeley Laboratory {LBL) 10 em X 40 em source. 'By using essentially the same source, development costs can be reduced while reliability and 
IV-2 

ORNL/DWG/FED-77858 
COOLING LINES FOR DEFINING PLATES 
CRYOGENIC 
PUMPING 

ADJ. BEAM 
STEERING DEFINING PLATESf\ MECHANISM 
YOKE SPACER 
...... ' 
< 
I 
w 
100omp ION NEUTRALIZER
" 
SOURCE " GAS CELL 
l
:: II 
COIL Ill ~ 
13.3 1 keV 
H+ 
CRYOGENIC 
PUMPING 

I 
I I 
-------~----1-+

,1\-..:-_:-lf-lL---------r--,f--+,"1 
:  I  I  
I  I  I  I  
1 ----------L  JJ.  

'Ji. , "'--_ _ I 1l II I ---t ! -----I II II 
1 1

I ----------~--r)..,IJ 
I I I ----4-~ --r ~ 
I I I -----T --~-.-' 

I II ------1
1 --+~--------I 
-~ J J.--I 
I

(y--
I I 
L-------------1-
FIGURE IV-A-1 POX BEAM LINE CONCEPT 


.::,:,,_,,~tt',•;>.v····~~J. l _ .1 n SOURCE COIL  f· I'I~·'•'I.;' 1-II  20 COLUMNS 17.8cm LONG CUSP FIELD s  
ANODE 1  
....... < I ~  N -N  
SECTION A-A  
·il LJ  
NEUTRAL  BEAM  

FIGURE IV-A-2 PLT/ISA/PDX STYLE SOURCE 
2.. 
-i 
maintainability can be enhanced. The sources progress from a 1/4 scale (10 01 X 10 em) full-power density, full-pulse length proof test on LBL test stand 
III-A, through full-size short-pulse testing on LBL test stand 111-B, to full-power, full-pulse length shakedown on the Lawrence Livermore Laboratory High Voltage Test Stand (HVTS) prior to delivery of a prototype source to the 
projects. The LBL 1/4 scale source is shown in Figure IV-A-3 and the TFTR 
beam line in Figure IV-A-4. 
The neutral beam source development has gone in parallel with development of supporting electrical systems. Good source operation is highly dependent upon and closely integrated with matching electrical systems. There have been 
significant problems to overcome. The accelerator power system not only can 
easily destroy the neutral beam source if not carefully controlled, but just the energy stored in the stray capacitance is sufficient to prevent conditioning up to rated voltage of the higher energy sources if additional measures 
are not taken to absorb this energy elsewhere upon the occurrence of a source 
arc. Not only have these problems been solved, but significant reductions in 
power system complexity and cost have been effected. The complete neutral 
beam system has also shown significant improvements in cost and perfonmance. 
This is shown in Figure IV-A-5 where neutral beam costs are given on a per 
unit injected power basis. Note the dramatic reductions in costs that are 
attributable to economics of scale and more efficient utilization of systems. 
Energy recovery systems have been tested at the parameters listed in 
Table IV-A-2 at the ORNL and at the LLL/LBL. The ORNL system uses a PLT 
style source operating nearly at ground with magnetic blocking to suppress 
IV-5 

Pulsed gas valve Cathode cover plate Gas inlet with diffuser 
0.005 inch thick Kapton 
-Filament plate 
+ Filament plate 
Filament chuck 
......."F Vacuum en wall 
insulator 
111-Tubular
Gradient grid 
insulator
(+100 kV) Suppressor grid (-2 kV) 
Exit grid (0 kV) 
0 5 10 
CENTIMETERS 
F.IGURE IV-A-13 CROSS SECTION OF THE 1/4 SCALE 120 kV~ 0.5 SEC. SOURCE 
LIQUID HELIUM DEWAR 
189.00 
MAGNETIC SHIELDINGDUMPR LN DISTRI6UTION MANIFOLD
(I\LORIV1E.TER ReTRACTOR 2 
125.00 
CLAMP SEAL 
Et:AM SOURCE SCR·\F'ERS 
217.11 
165.31
..... 
< 
I 
....... 

jr 
1o:.so 
• 72.J0 
·---_ ___!._ 
LCHEVRONS 
---~---------··218.00 .FIGURE IV-A-4 SIDE ELEVAriON OF A TFTR NEUP~~AL BEAM LINE 
FIGURE IV-A-5 
4 
~ ® 
~ 
0 3 UJ 
t-
UJ 
0 
(.) @@)X
~ 
ct 
2
w 
m 
...I 
ct 
a: @D
t
::;:) 
w 
2 1 
@ 
1978 1980 1982 1984 
APPROXIMATE OPERATIONAL ClATE 
TV-3 

electrons. Electron losses have not caused great difficulty. The LLL/LBL system is optimized for electrostatic electron suppression but has also been operated with magnetic blocking. Initially, considerable electron suppression problems were encountered in the LLL/LBL tests, but these seem to be overcome by use of redesigned electrodes and careful electrode conditioning. 
2. 	Neutral Beam Requirements for Upgrades to Existing Facilities and for 
Future Projects The neutral beam requirements for upgrades of existing facilities and for future projects are listed in Table IV-A-3. 
In planning development activities for these devices, the following factors were considered: 
o 	Positive ion neutral beam sources have been demonstrated up to 120 KeV. Currents up to about 100 A per source are reasonable. Extensions to ~ 200 KeV with energy recovery appear feasible. 3' 4' 5 

o 	Negative ion neutral beam systems will be more efficient and costeffective than positive ion systems with energy recovery above some energy; probably 150 to 200 KeV. 6 

o 	Negative ion sources of 1 A have been demonstrated although not as an engineered device. In the near to intermediate term, currents of 5-10 A per source appear feasible. 7' 8' 9 


Given this background, neutral beam needs are broken down into two energy categories; namely, <200 KeV -medium energy, and >200 KeV -high energy. The 
IV-9 

·· near-tenn and potentially 1 ong-term needs in the medi urn energy range can 
be met with advanced positive ion systems. The high ener!lY needs motivate 
further development of negative ion systems. Si nee posi t'ive ion systems are 
the most advanced today, they will be called upon to car~t out the near-tenn 
high power experiments. 
The positive ion needs for near term fusion facilities ar«~ sufficiently similar that initial work can be accomplished using a common source and test facilities. Specific sources for each application may differ in detail such as grid curvature and spacing, but this does not negate the advantages of common test vehicles for development. As such, an Advanct~d Positive Ion Source (APIS) has been defined. This source includes act·ive water cooled ~lectrodes, magnetic cusp field plasma chamber, long lived cathodes, and energy recovery of the unneutral ized fraction of the beam.. The Oak Ridge National Laboratory and the Lawrence Berkeley Laboratories are competing in development of this source. Basic parameters are listed ·in Table IV-A-4. ~nitial partial size versions are being fabricated and the R&D test facilities are being upgraded to test them. Full scale source tests should begin in October 1980. This activity is receiving high priority in the U.S. program. 
The negative ion needs are based on tandem mirror needs. However, it may be desirable to provide some fraction of the total injectt~d power to a tokamak at higher energies as well. Initial tandem mirro1r reactor calculations indicated high ("' 1 MeV) beam energies would be required. More recent analysis has indicated that beams of lower energy may be adequate. While some of these analyses can be int~rpreted to indicate positive ion neutral beams may suffice, they also indicate a rapid increase in tandem mirror reactor 
IV-10 

power gain and neutral beam utilization efficiency as beam energy increases 
to the range where negative ion beams will be required. 
Negative ion source development has proceeded at a more modest pace. Use on a 
plasma device is yet to occur. Negative ion currents of 1-2 amperes have been 
obtained at Brookhaven National Laboratory with 1 ampere being accelerated to 
120 KeV. The LLL/LBL program has extracted 300 rna of H-and D-current and 
accelerated 100 rna to 60 KeV by use of cesium double charge exchange. LBL 
has also obtained ~ 400 rna of negative ions using a self extraction bucket 
source. The ORNL Program is developing a modified duopigatron negative ion 
source, a small version of which has yielded 1 ampere of negative ions for 200 msec. ORNL is also developing a penning source that is capable of 60-100 rna 
and has been operated for 20 second pulses. 
Successful operation of the Tandem Mirror Experiment (TMX} at the Lawrence Livermore Laboratory (LLL} is expected to provide added emphasis for the need to develop negative ion sources. In response to this anticipated emphasis, a proof of principle goal has been established with the belief that considerable work can be done at modest cost if acceleration to higher energies with the needed large test facilities is deferred until a source with acceptable properties is demonstrated. The proof of principle source parameters are listed in Table IV-A-5. It is planned to choose the best source for further development on the LLL High Voltage Test Stand and subsequent application to MFTF-B. 
3. Neutral Beam Development Elsewhere in the World Neutral beam development programs are in place in the USSR, Germany, France, England, and Japan. In the USSR the T-11 experiment is utilizing 25 A, 35 KeV 
IV-11 

injectors and a more advanced 120 amp 80 KeY system is bE!ing developed for 
T-10M. This latter system is intended for 1 to 2 second pulse length and utilizes a LaB 6 cathode. Negative ion sources are also being developed ··utilizing double charge exchange and direct extraction. The sodiiD double 
charge exchange system has produced .1. 4 ampere of H-at ~~0 KeY. The goal is 
_for 10 amps at 100 KeV and 1 second pulse length. At No,,osibirsk work is proceeding on surface negative ion sources with 0. 9 amps having been obtai ned at 30 KeV for 1 msec. Original plans for T-20 included 60-100 MW of 400 KeY injection. 
At Culham, England, the DITE experi~ent is operating with 1.2 MW of 30 KeY 
beams and work is underway to double this power, They at·e also developing a 1 MW 30 KeV system for Wendelstein Vll ~' .a stellerator at Garching. Work on 80 KeV neutral beams are under development for JET. This work includes direct 
energy recovery of individual beaml ets•. 
At Fontenay-Aux-Roses, France, the TFR tokamak is being heated by 35 KeV, 
60 kw sources. Plans are to inject 4 MW by use of the Fontenay-Aux-Roses 
-peripl asmatron. They are also developing _a periplasmatron injection system 
for ASDEX at Garching. Successful direct recovery experiments have been 
conducted using a wire grid electrostatic electron suppression system. 
Fontenay is developing 160 KeV o-systems utilizing direct recovery for 
later application to JET. 

At Karlsruhe, Germany, an interesting system of cluster acceleration is being pursued. Singly charged clusters of atoms are accelerated to yield .5 to 1 KeV/AMU energy for cesium double charge exchange to negative ions. If 
IV-12 

successful, this cluster system should circumvent the space charge blow up problems of low energy high current sources presently being used for cesium 
double charge exchange experimentation. 
In Japan, the Gamma-6 Tandem Mirror utilizes a single 20 KeV, 10 A source per end cell. JAERI plans to use 20 MW of hydrogen injection at 75 KeV on JT -60. Over the years they have developed several duopigatron style sources up to 30 
KeV, 	30 Ampere, and 1 second pulse length. 
References 
1. 	
H. Eubank, et al, 11 PLT Neutral Beam Heating Results, .. IAEA-CN-37-C-3. 

2. 	
Draft INTOR Internal Report on Heating. D. L. Jassby, et al, 
May 18, 1979. 


3. 	
H. H. Haselton, 11 Positive Ion Systems -State of the Art and Ultimate 


11
Potential , Proceedings o·f the Heating Developments Requirements 
Workshop, December 1977. DOE Report CONF-771241. 
4. 	K. H. Berkner, et al, 11 The LBL/LLL Positive Ion Based Nuetral Beam 
Development Program Status, .. Proceedings of the Heating Developments 
Requirements Workshop, UCID 3987, Dec. 1977. 
5. 	
R. W. Moir, et al, 11 Neutra1 Beam Based on Positive Ions with Direct Energy Conversion, 11 Proceedings of the Heating Developments Requirements Workshop, UCRL 80418, Dec. 1977. 

6. 	
L. D. Stewart, Discussion Summary, Proceedings of the Heating Developments Requirements Workshop, December 1977. 

7. 	
K. Prel ec, 11 Negative Ion Systems Based on Direct Extraction Sources, 11 

Proceedings of the Heating Developments Requirements Workshop, BNL 23583, December 1977. 

8. 	
Th. Sluyters, 11 BNL High Energy Neutral Beam Development Plans, .. 

Proceedings of the Heating Developments Requirements Workshop, BNL 23582, December 1977. 

9. 	
W. L. Stirling, 11 0RNL Negative Ion Program, .. Proceedings of the Heating DevPlopments Requirements Workshop, Dec. 1977. 


IV-13 

TABLE IV-A-1. Current Neutral Beam Source Development Requirements 
Energy Current Pulse Length Species Operating Prototype (keV) (Amperes) (Seconds) ($Full Energyt Gas Need Date 
ISX-B 40 100 1/10 > 90% H,D 12/78 PDX 60 100 1/2 > 90% H,D 2/79 TFTR 120 65 1/2 -1 1/2 > 85% H,D 10/78 Daub 1et-I II 80 80 1/2 60% H only Mid-'79 MFTF 80 80 1/2 > 85% H,D Mid '79 
TABLE IV-A-2. Energy Recovery Parameters 
ORNL LLL/LBL 
Beam Energy 35 keV 100 keV Total Ion Current 2.2 A 1.6A Size at Converter 15 em di am. 8.6 em X 25 em Size at Source 17 em diam. 9cmX9cm Pulse Length 100 msec 600 msec Net Recovered Power 55 + 15 kW 95 kW 
(returned-to power supply) Power Recovery Efficiency of Total Charged Components 80 + 20% 65% 
(independent of time and pressure) 
TABLE IV-A-3. Neutral Beam Requirements for IF aci 1i t,l 
Upgrades and Future Projects 

Energy Pulse Length Need Project (KeV) Power (MW) (SEC) Date 
TFTR/U 120 (D) 50 ~ 1 1/2 1983 MFTF -B 150 (D) 30 1983 
> 200 (H) cw 1986 ETF/INTOR 120 -200 (D) 50 5s-cw 1991 Tandem Mirror 300 (H) 
ETF Daub 1et-I II 80 (H) 40 5 1985 
IV-14 

TABLE IV-A-4. Advanced Positive Ion Source Parameters 
Energy:  150-200 KeV  
Current:  100 A  
Pulse Length:  cw  
Optics:  < 1/2° HWHM  
Gas Efficiency:  > 50%  
Atomic Yield:  > 90%  
Energy Recovery Included  
High Reliability Required  

-TABLE IV-A-5. Negative Ion Source Proof of Principle Goals 
Energy: "' 20 KeV 
Current: 10 A ions (SA neutrals) Pulse Length: cw Optics: Adequate to allow 1/2° HWHM operation at 200 KeV Gas Efficiency: > 10% 
IV-15 

B. 	RF Heating 
1. 	Introduction The extrapolation of neutral beam heating systems to appl·ication on ETF /INTOR devices may not be straightforward. In addition to extrapolation of system size, energy and power; reliability and maintenance in a radio
active environment are important concerns. Radi a-frequency heating of 
magnetically confined plasmas, although not as well under~;tood from a 
plasma physics viewpoint, offers a potentially attractive alternative 
heating system because, in theory, the rf source can be lj:>cated relatively 
far 	away from the reactor and the rf can be transmitted tt:> the plasma 
through coaxial pipes or waveguides. The antenna systems used to couple 
the 	rf to the plasma and the physical processes which abs,orb the power 
will vary depending on the frequency of the rf. Thus, a 1l'lide variety of 
power generation, transmission, and coupling systems might be employed 
over the range 1u MHz to 1ti0 GHz {at up to 5 tesla magnetic field). These 
are 	summarized in Table IV-B-1. The technical areas of each heating 
system are divided roughly into the power source, the transmission system 
and the coupling system. For lower hybrid heating schemes, the klystron power sources are essentially 11 0ff the shelf11 technology. On the other hand, the power source for electron cyclotron heating, the gyrotron, is highly developmental and, in fact, most of the U.S. funds for rf technology development are for gyrotron development. Many of the components of the transmission systems are straightforward in,that coax and waveguide design are well understood. However, the design of coupling systems {antennas) 
is often highly dependent on a particular application and is quite often 
IV-16 

beyond the state-of-the-art. For example, ion cyclotron experiments typically use partial loop antennas inside the experimental device vacuum vessel. These loops are driven by very high voltages and require vacuum tight, insulating feedthroughs. In typical lower nYbrid or ion ~clotron heating experiments, the rf heating frequency will be equal to the electron cyclotron resonance frequency at some point between the toroidal field coils and the surface of the vacuum vessel. Therefore, it is considered necessary to place an rf window in the waveguide near the vacuum vessel surface in order to keep any low level plasma away from the resonance region. (Electron cyclotron heating experiments with ports on the low field side do not have this problem and may not need these windows.) At present, the design and fabrication of coupling systems and the R&D required for insulators and windows are carried out by the major experimental groups that conduct the heating experiments. 
Longer range development of higher power, radiation resistant windows and 
insulators for ETF/INTOR is recognized as a need and some work is being 
initiated in the u.s. as part of the special purpose materials development 
program. The design of coupling systems for ETF/~NTOR devices is contin-I 
gent on the results of physics experiments in current/future devices. 
2. 	Requirements So far rf has been discussed as an alternative to neutral beam heating for tokamaks. However, certain types of rf heating, particularly electron cyclotron heating, may be absolutely essential for the operatjon of certain devices. The best example is the EBT concept which employs 
IV-17 

high-beta relativistic electron rings for plasma stabili~~ation. Another potential application is in the Tandem Mirror concept wh•~re ECH heating in the end of a mirror would provide a thermal barrier against plasma energy loss. A possible unique application of high power ECH in tokamaks is for preheating the plasma and for profile control. Thus, some experiments are aimed at exploring alternative heating schemes whereas other experiments such as EBT are aimed at testing mainline and possibly the only approach to accomp1i shi ng certain plasma experimental goals. Tablie IV-B-2 summarizes current, planned, and proposed experiments. 
3. 	Technical Approach The priorities for development of rf heating components/systems can be stated simply. First, up to lower hybrid frequencies, sources of power are readily available and no significant development of these sources is necessary at this time. Second, the design and fabricatiion of transmission and coupling systems (rf windows, antennas) will be pursued by individual experimental groups for the near term. As clear candidates for very high power (> 2U MW) systems are identified, longer range devE~lopment will be pursued. Third, the success of electron cyclotron heating, which is essential for operation of the EBT, is most highly dependent on the development of the electron-cyclotron maser (gyrotron, gyroklystron) as a source of continuous (> ls-cw) high power. See Figure IV-B-1. 
Given the above, all funds for source development are alllocated to the 
gyrotron development program. The most immediate need is for a 28GHz, 200 
kw, 	cw gyrotron for the EBT-S at Oak Ridge National Labor·atory. Although 
IV-18 

FIGURE IV-B-1 
IV-19 

initial results were quite favorable (> 250 kw, pulsed) mar~ difficulties including microwave heating of insulators and the output wiindow have been encountered in the design of a cw tube. So far tubes have been operated at 100 kw, cw for 5 minutes before output window failure and <tt 50 kw, cw for "' 7 hours before insulator failure. It is hoped that the design goals wi 11 be achieved soon but due to the nature of the problem,, no date can be ni ven. 
Initial preparations for llOGHz, 200 kw, cw gyrotron development for EBT-P are also underway. The progress of this effort will, of cuurse, depend on successful completion of the 28GHz tube. At this point, it is assumed these efforts will require 3-31/2 years to complete at a cost of around 
$1M per year per cont~actor for two separate contractors. 
The substantial progress of the Soviet Union in gyrotron dE~velopment should be noted. _A rather extensive program has been under'Way for several years and the Soviet group has announced 1 MW, lOOpS pulses from a lOU GHz 
tube and 3UU kw, "'20ms pulses from a 85GHz tube. Also, Japanese industry 
has initiated a gyrotron development effort aimed at producting a 35 GHz, 
cw tube for the Japanese bumpy torus experiments. 
In summary, the development of sources of rf power is devoted solely to 
the gyrotron with other sources being considered off the shelf. Development of transmission/coupling systems is being carried out by individual 
users at present. As clear candidates for very high power (> 20 MW) 
IV-20 

systems are identified, longer range development will be pursued. Preliminary work on radiation resistant windows and insulators is being conducted in preparation for use on ETF/INTOR types of devices. 
IV-21 

Technique 
Alfven 
Ion Cyclotron 
....... 

Lower Hybrid
< 
I N N 
Electron Cyclotron 
TABLE IV-B-1. Summary of RF Heating Techniques 
Frequency 
~ 10 MHz (determined by cavity size and plasma density) 
5/7.5/15 1v1Hz pertesla of magenticfield (first harmonic for tritium/deuterium/ hydrogen atoms) 
f. <ff.=6g3 MHz p~~ t~sla of magnetic field 
28GHz per tesla of magnetic field for 
first harmonic 
Power Klystron 
Klystron 
Klystron Gyrotron 
f i = ion cyclotron frequency
f~i = ion plasma frequency (a function of plasma density) fee= e~ectron cyclotron frequency 
Transmission Coax 
Coax, 1oaded waveguide, slotted waveguide 
Waveguide 
Waveguide 
Coupling 
Slotted Cavity, Antenna Loops 
Same as for Alfven plus open, multi-portedwaveguide. 
Open, multi-portedwaveguide 
Open waveguide, microwave horn 
TABLE IV-B-2 
Summary of Current and Planned Major RF Heating Experiments 

Technique 
Alfven 
Ion Cyclotron 
Lower Hybrid 
Electron Cyclotron 
Experiment 
Petul a (France) 
PLT 
TFR Alcator-C 
Alcator-C 
PLT 
JFT-2 (Japan) 
EBT-S 
ISX-B 
EIH-P 
T-lll (USSR)T-15 (USSR) 
Frequency, Power, Pulse Length 
UNK, 10Mw, UNK 
25MHz, 2Mw, 150ms 4U-55MHz, 2.5-5Mw, 33ms UNK, 1Mw, UNK 183MHz, 1Mw, > 200ms 
4.6GHz, 4Mw, 250ms 
BUUMHz, 1Mw, 4UOms 
2BGHz, 200 kw, 40ms 
28GHz, 2UO kw, cw 
28GHz, ~ 150 kw, lOOms 
35GHz, ~ 150 kw, 10ms 
110liHz , 2Mw, cw 
80-90liHz, 2UO kw, cw 
85liHz, 150 kw, 25ms 
1UOliHz, 5Mw, 1.5 s 
Status 
In agreement with theory. 
August 1979 October 1979-80 August 1980 January 1980 
August 1980 March 1980 
February 1980 1979 September 1979 August 1979 1983 1983 August 1979 1983 
UNK = Unknown at time of publication. 
IV-23 

C. 	Fueling 
1. 	Introduction All plasma confinement devices which operate for manY particle containment and fuel burning times will require fueling. In existing devices, where no plasma burning is taking place, particle containment losses are the single factor which determines the nature and magnitude of various fueling techniques. In open mirror systems, such as the Tandem Mirror Experiment at Lawrence Livermore Laboratory, refueling is accomplished by gas puffing and by the neutral beam systems which are used to heat the mirror plasmas. In present tokamaks, initial plasma is supplied by electrical breakdown of the fill gas and density is increased by puffing in more gas at the plasma edge. As particles escape and are neutralized, other neutral particles reenter the plasma, are reionized, and maintain the plasma density. In larger tokamaks and in tokamaks with divertors, this edge refueling process may be inadequate and active means of fueling may be required. High energy neutral beams can play a partial role in fueling; however, in order to avoid tnermal runaway, some other means of fueling, with lower energy per particle, will be needed to maintain steaqy state fueling during the burn phase of ETF/INTOR-like devices, especially if divertors are used. Low particle energy methods which have been considered for reactors include plasma guns, cluster acceleration~ and solid tzydrogen pellet acceleration. 
Plasma guns and cluster acceleration are two techniques for the electrical acceleration of fuel particles with low atomic mass number (M=l-500) at 
IV-24 

very high velocities. These techniques have been discussed in detail in 
other reports 1 and are being examined in Europe and Japan. Due to both.· 
phYsics and technology concerns about fuel penetration, there are no plans 
for aggressively pursuing major development of either of these techniques 
in the u.s. 
On the other hand, acceleration of solid hydrogen pellets {M > 1019) appears to be a very promising technique for fueling. At one point there was serious concern that a point source of cold fuel in a plasma such as a solid hydrogen pellet, would cause major plasma disruptions. However, recent experiments on the ISX-B Tokamak have shown these fears to be largely unfounded. 2 Plasma densities have been increased by factors of 2-3 oy a single pellet with no damaging effects. In addition, the lifetime of the pellet in the ISX plasma corresponds very closely to theoretical predictions. Thus, there is great confidence that pellet fueling will be a suitable technique for fueling ETF/INTOR devices. 
l. 	Requirements Technological requirements for pellet injection {size, velocity, rep. rate) will be determined primarily by the required pellet penetration depth. Early calculations of penetration depth were based on parabolic density profiles, which required a fuel source at the center of the plasma. These predicted requirements {> lOkm/s) are now considered overly conservative. Self-consistent transport calculations3 indicate that ETF sized plasmas may operate in an optimum manner with rather flat density 
IV-25 

profiles and in this case only shallow fueling, with greatly reduced technological requirements, will be required. These requirements are listed in the following section. 
3. 	Technical Approach ivtany techniques including electrical (charged pellet), magnetic (with a metal driver), mechanical (centrifugal), pneumatic (gas), and laser ablation have been proposed for accelerating these pellets to high velocities. In addition, the high pressure injection of liquid jets has been proposed as an alternate means of fueling with 11 solid11 particles. The pneumatic and mechanical techniques appear to be the most promising approaches for producing 3-5 mm diameter pellets with velocities of 2-4 km/s and with rep rates of up to 50 pellets/s per accelerator. Based upon certain assumptions concerning plasma profiles in reactor devices, the above specifications have been calculated to be adequate to fuel ETF/INTOR devices; therefore, all effort in the U.S. and much effort in Europe/Japan are focused on developing these approaches. Table IV-C-1 summarizes the status and future plans for developing these two approaches. Second generation pellet accelerators are being designed and should be available for use on experimental devices within 1-1 1/2 years. It is hoped that experiments on PDX, D-Ill, and TFTR will verify the current theoretical predictions of the physics/technological requirements for fueling. In that case the advanced development as shown in Table IV-C-2 should be adequate for fueling ETF/INTOR sized devices. 
IV-26 

Other techniques such as electrical 4 or laser ablation 5 have been proposed for attaining higher pellet velocities(> 10 km/s); however, these techniques are not well understood and will require significant development. Given that 10 km/s is not recognized as a definite need, we do not consider it prudent to embark on such efforts at the present time. As discussed, the development of the pellet fueling option has achieved significant positive progress Doth from a plasma physics and acceleration technology viewpoint. The development goals, which are based on reasonable assumptions about plasma behavior, appear to be within reach. A program is in place to develop two separate approaches, mechanical and pneumatic, to achieve these goals and both approaches are strong contenders for use on current devices as well as ETF/INTOR sized devices. 
References 
1. 	
Proceedings of the Fusion Fueling Workshop, Held at Princeton University,November 1-3, 1977, DOE Report #CONF-771129. 

2. 	
ISX-B Status Report Presented to Office of Fusion Energy, June 29, 1979. 

3. 	
Fusion Fueling Workshop, Reference 1. 

4. 	
Fusion Fueling workshop, Reference 1. 

5. 	
Model of Ablative Acceleration at Moderate Laser Intensities, F. s. FelDer General Atomic Report #GA-Al4944. 

6. 	
Conversation with Stan Milora, Oak Ridge National Laboratory, July, 1979. 


IV-27 

TABLE IV-C-1 
Summarl of Current and Projected· Pellet InJection Develo2ment 
(a~~roximate ~arameters)6 
Current 	Near Term Goals Long Range Goals 
Status 	(1980-81) (ETF/INTOR)Program 
U.S. 	ORNL Pneumatic Pneumatic Pneumatic andMechanical 
(ISX, ORNL) 1 mm diameter 1-2 rnm dhmeter 3-5 mm diameter (llDX, PPPL) 1 km/s 1-2 km/s 2-4 km/s 
(IJ-I II, (;jA) 	single shot lO pellets/s so pellets/s 
(TFTR, PPPL) Mechankal Mechanical 
1 mm diameter i-2 mm diameter .30 km/s 1-2 km/s 150 pellets/s 150 pellets/s 
(;jermanyGarching Pneumatic Pneumatic 
(ASIJEX) 	1 mm diameter 1 mm diameter UNK 
1 km/s 1 km/s
single shot 2 shots 

Mechanical Mechanical 
.4 mm diameter 1 mm UNK .33 km/s .75 km/s single shot 100 pellets/s 
England UNK Pneumatic 
Culham

(IJITE) · .4 mm diameter UNK 
.40 km/s
5-10 shots 

Denmark
Riso Pneumatic 

(UArHE) 	.5 mm UNK UNK 
.33 km/s
single shot 

UNK =Unknown at time of publication. 
IV-28 
o. Divertor Technology Impurity control in plasmas is receiving more attention lately as the fusion program moves closer to the construction of an ignition device. This is because impurity control is necessary both to reach ignition (keep high Z impurities out of the pfasma), and to sustain ignition (limit the helium accumulation). ·operation of a tokamak ignition device without helium removal from the _plasll)a and pumping the gas means that the burn time will be limited to 20 to 30 seconds, which is too short to be considered reactor-relevant. In_addi.tion, by guiding the particles along magnetic field lines to specific p~rticle collectors, one hopes to reduce first wall erosion due to ene_rgetic particle-bombardment. The divertor, by controlling the plasma boundary condition, can also produce a flatter plasma temperature profile, result1ng in a higher average power densi~. In short, divertors of different types can be used t( both reduce plasma contamination and guide particles to a region where pumping is more feasible. 
Various types of magnetic field divertors have been devised for toroidal devices. All are based upon the idea of generating a null at some point or points in a component of the magnetic field. This generates a separatrix and· field lines on the outside of this separatrix are carried away from the plasma surface as they pass the neutral point (null point). Magnetic flux surfaces on the inside remain closed within the plasma volume. The separatrix then becomes the boundary of the toroidally confined plasma. The separatrix is sometimes referred to as a magnetic limiter. The basic idea of the divertor 
IV-29 

is that as particles diffuse across the separatrix, they follow along the field lines and are carried away from the plasma to special particle collectors. The implied assumption is that particle mobility along the field is greater than cross-field diffusion. 
Three different types of· divertors are shown in Figure IV-0-1. The toroidal divertor has been used on stellerators at Princeton. Th'is type of divertor generates a null in the toroidal magentic field. A disadvantage is its substantial perturbation on the symmetry of the toroidal field {field ripple). In the Model C stellarator, the toroidal magnetic field perturbation was about 60% at the center of the plasma. 
A variation on the toroidal divertor is the so-called bundle divertor recently proposed by a Culham group. Two opposing current loops adjacent to each other divert a bundle of magnetic flux. The main advantage of this approach is that it produced only a minor perturbation {~ 1% -2%) in the magentic field at the center of the plasma. 
The poloidal divertor diverts the poloidal magnetic field of the plasma using currents located outside the blanket and shield and perhaps even outside the main non magnets producing the toroidal field~ This concept has been used 
in the FM-1 divertor and is basis for the POX experiment at Princeton Plasma 
Physics Laboratory. There is some freedom in the location of the null 
p9ints and thus in the field lines that guide the particles away from the 
plasma. One can have a single null point.on the midplane.at the inside of 
the torus or two null points symmetrically above and below the midplane 
IV-30 

,figure IV-D-1 :Different Types of· Divertors 

,Particle  
coli  
\Diverted  
Particle  flux bundle  
Vcollector  
Separatrix  

TOROIDAL DIVERTOR (Top view of torus) BUNDLE DIVERTOR 
Current 
Projection of a diverted field line 
'Projection of a 
diverted field line 

To collection region 
SINGLE NEUTRAL POINT 
DOUBLE NEUTRAL POINT 
(Currents not shown) (Not all currents shown) 

(CROSS-SECTION OF TORUS) 

IV-31 


towards the inside of the torus. The double-null divertor configuration requires less hardware in the central core of the torus, where space and access are at a premium, compared to the single null divertor. From an optimization viewpoint and the desire for a small aspect ratio, it would be most desirable to place the null points on the outside of the torus. However, this does not appear to be feasible because the vertical field required for radial equilibrium has the opposite direction to that required for a null point on the outside. This in turn would cause an even more grossly non-uniform field variation in comparison with that producing a null point towards the inside of the torus, i.e., more ripple in axial field. 
Poloidal divertors preserve the axisymmetry of the tokamak. As a consequepce the banana orbits of particles are not affected and superbanana orbits, with their much larger radial excursions, are not produced. The toroidal and bundle divertors destroy the axisymmetry and lead to superbanana effects, especially near the separatrix. Even if the corresponding enhancement of diffusion is tolerable, a significant number of alpha particles born on superbanana orbits near the plasma edge may hit the first wall at high energy and produce erosion by blistering. This suggests that the poloidal divertor may better protect the first wall when the plasma is thermonuclear. 
Recent results from the ST Tokamak indicate that a substantial amount of wall-originated impurities are released into the plasma during the early stages of the discharge. In order to intercept these impuri ties, the di vertor must operate during the initial stages of the current-rise phase as well as during the burn. With proper programming of the external currents, the 
IV-32 

discharge could be initiated at a multipole null in the center of the vacuum chamber and a poloidal divertor configuration established with a small plasma at the center of the chamber. This plasma could then be allowed to expand in a controlled manner during the current rise in such a w~ as to avoid skin currents, maintain reasonable safety factor in the plasma, and maintain the poloidal divertor geometry. In this chain of events, the external currents both establish a divertor and an expanding magnetic limiter. Some of these possibilities are being considered for PLT and PDX. 
A comparison of the two basic divertor types is shown in Table IV-D-1. 
The poloidal divertor is considered better from a physics standpoint because it does not cause field-ripple; however, from a maintenance standpoint in a reactor, it does not appear to be practicable in its present configuration. PPPL has recently suggested a segmented poloidal divertor to allow maintenance, but the effect of the cross-over current leads on field ripple has yet to be evaluated. 
The toroidal or bundle divertor has the advantage of maintainability, but 
the effect of field ripple on particle loss rates has not yet been explored experimentally. Again, recent ideas put forth by ORNL/Westinghouse may 
provide designs with ripple low enough to be accepable. 
The current state of magnetic divertor technology is given in Table IV-D-2. The results to date have been very limited, ~s shown i'1 Table IV-D-3. Experiments in the next two years on PDX at PPPL and ISX-B at ORNL will yield 
poloidal and bundle divertor pnysics data with plasmas at higher temperature 
IV-33 

and density than previous experiments. Near-tenn experiments are also planned' using hot, high beta, collisionless plasmas (ISX-B, POX, PLT) to determine just how critical ripple effects and helium transport are in reactor-relevant plasma regimes. 
In order for a divertor to be successful in controlling plasma impurities, the following issues must also be addressed: 
1. 	Heat flux at divertor collector plates The heat flux at the divertor collector plates can be intense and sputtering can cause rapid material erosion. 
There are no applicable techniques for handling steaqy state heat flux 
levels above lu kw/cm2 for a large number of repeated pulses. Even if 
the 	head load could be distributed over a larger surface area, the pulsed 
nature of tokamak operation could create fatigue problems and the particlE 
flux 	would cause sputtering for a conventional metal-clad heat-exchanger. 
~ack streaming from the divertor could thus put sputtered material back 
into 	the plasma. 
l. 	Pumping in the divertor chamber 
Since the primary aim of the divertor is to rid the plasma of impurities, in particular helium, the first consideration of the divertor pumping should be the pumping speed for helium. Table IV-0-4 shows the various options for pumping along with estimates of pumping speeds and limitations. 
IV-34 

In addition to pumping helium, the divertor will also pump the tritlum from the plasma; which in turn must be regenerated and ultimately returned to the plasma as fuel. 
Conclusion The state-of-the-art of divertor technology today is best characterized as primitive, mostly because of the lack of a sufficient physics data-base upon which selection and ranking of critical technology issues may be based. The 
u.s. does not have a divertor technology program in place at present, but preparation is oeing made to establish a program in FY 1980. This program will be very closely coupled to the emerging pnysics results from divertoroperation in PDX and ISX-B. 
The divertor system technology is complicated by the need to satisfy many technology and physics requirements simultaneously, such as: 
o Helium pumping 

o Tritium pumping and regeneration 

o Impurity pumping without back streaming 

o High thermal heat loads 

o No plasma unipolar arcing 

o Accessibility and maintainability 

o Accommodation of high magnetic stresses 


IV-35 

Because the state-of-the-art is primitive, there is much room for innovation. Within just the last year new ideas have evolved which may reduce the maintenance problem with the poloidal divertor and which have taken the bundle divertor from being unacceptable to the point where there exists some optimism that technically and economically viable options for ETF/INTOR and power reactors can be defined. 
IV-36 

TABLE IV-D-1. State of Divertor Technology 
Physics 
Engineering 
....... 

< 
I 
w 
....... 

Poloidal Advantage Symmetric -no Field Ripple 
Disadvantage Probably must be located inside toroidal field coils -very difficult to maintain. 
Experiments POX ASDEX DIVA Toroidal (Bundle} Disadvantage Asymmetric -field Ripple 1-2'.t (New designs may be able to reduce 
ripple to < 0.5'.t} 
Advantage Can be more easily maintained because divertor can be located between toroidal coils. 
DITE ISX-B 
TABLE IV-D-2. Current Magentic Divertor Technology Status 
o 	Dite bundle divertor continues to show favorable physics results -Culham is confident that engineering improvements can be made. 

o 	POX will shortly test poloidal divertor action with a relatively low temperature plasma. 

o 	Design activities are underway for bundle divertors for ISX-B. 

o 	The ORNL/(~) TNS Program included innovative designs for a compact poloidal divertor and an improved bundle divertor. 

o 	The ORNL/GAC TNS Program in FY 1 78 proposed a bundle divertor with a liquid lithium collector. 

o 	Bundle divertor designs for more stringent reactor conditions were completed for the (W) DTHR and CTHR reactors and development issues identified. 

o 	Innovative advanced concept development, especially those efforts led by Dr. H. Furth (PPPL) and Dr. T. Yang (~), is progressing and suggests considerable design and performance improvement is possible. 


IV-38 

TABLE IV-0-3. General Divertor Performance Conclusions 
from Recent Experiments 
o 	Z effective (resistivity) reduced by a factor of 3. 

o 	Plasma flows in the scrape-off l~er u11 ~ u.2 -0.5 Cs (speed sound)._ 

o 	Impurities flow at the same speed as the plasma. 

o 	Cross field diffusion in the scrape-off l~er ~ BOHM. 

o 	Impurity sweeping effect has been proven. 

o 	Impurity shielding effect has been proven but shown dependent on ionization state and species. 

o 	Radiative power loss reduced by as much as a factor of 5. 

o 	Operating regime of current experiments: B=1.0 -2.0 Tesla


0 
R = 0.5 -1.2 m

0 
A = 0.1 -0.2 m 

Ip = 14 kA -40 kA 
N = 1u12 1013 3

-cm-
TE = 200 -700 ev 
q = 3, 9 

IV-39 

TABLE IV-D-4. Helium Pumpin~ Options 
Helium Hydrogen Limitations Turbo Molecular Pumps 
-1 -2
(or diffusion pumps) 5R.s em Large wall required Cryopumps and Cryosorpti on 
-1 -2
Pumps 2R.s em 	Need regularregeneration 
Gettering 1-2R.s em 	No He pumping 
Liquid Lithium Not Known 	Direct ion Impurity Source trapping 
IV-40 

V. MATERIALS FOR FUSION REACTORS 
A. Introduction Understanding of mate~ials requirements for fusion reactors has changed rapidly over the past five to seven years. The design concepts which began to evolve in the early 1970 1 s specified relatively high heat loads and thermal efficiencies. 1-3 First wall temperatures were projected in the 1000° to 1100°C temperature range. At these temperatures, refractory metals-notably niobium and its alloys or ceramics and metal-ceramic combinations-were usually ~efined for the 'first wall-structural materials since high temperature properties and, high thermal conductivity were the principal identified requirements. 
In the 1973-75 time period, significant changes in design philosophy based on emp~asizing .. existing technology .. resulted in designs featuring lower power densities and with it lower wall temperatures. Materials selection shifted to austenitic stainless steel operating at 500°c4 , nickel base alloys such as PE-16 operating at 660°c5 , and Incoloy 800 operating at 625°c6• Of perhaps greater importance was the fact that during this time period designs began to include consideration of heat transport and secondary system factors including temperature and type of working fluid. For instance, the 500°C temperature in UWMAK-1 was adopted because of compatibility considerations between lithium and stainless stee1 4• 
In the period from 1975 to 1977, still different ideas were explored in reactor design. These took advantage of the realization that with the large mean free path of high energy neutrons, the first wall could be operated at 
V-1 

substantially lower temperature while maximum temperatures could occur in the 
blanket region away from areas of intense radiation. Specific designs, for 
instance, were made based on using austenitic stainless steel at 300°C, a 
7
temperature regime where helium embrittlement is not a problem , and the use of aluminum-base alloys at 250°-300°c. 8 
During the past two years the situation has changed once again to higher wall 
loadings and wall temperatures. The driving force has been a new realism in 
reactor designs which for reasons of cost and size, require more compact 
designs and with it higher power densities. This trend has been supported by 
recent physics results which indicate a good possibility for operation at 
higher plasma power densities. One recent design is based on titanium as 'the 
first wall material because its• much higher thermal conductivity results in 
improved thermal stress limitations9• 
Based on the changing materials requirements occasioned by changes in design philosophy, the question has been for some time how best to plan a materials 
program. In the U.S. this problem was approached by outlining a multi-path alloy development program based on a number of goals listed in Table V-1 10 •11 • 
It should be recognized that there is a distinct difference between goals and 
requirements. Recent TNS studies in the U.S. have been based on wall lifetime~ of 5 to 10 MW-years/m2 which does not alter the fact that a long term goal for commercial power reactors is a lifetime of at least 20 and possibly 40 MWyears/m2. With the recognition that it was not possible at this point in time to single out one ideal structural material, the U.S. Alloy Development Program is based on Austenitic Stainless Steels (Path A), Iron-Chrome-Nickel 
V-2 

Alloys (Path B), Reactive/Refractory Metals (Path C), and Innovative Concepts (Path D). A brief listing of the advantages and limitations of each alloy system is given in Table V-2. Of special note is a new development in Innovative Concepts including recognition of the advantages of ferritic (martensitic) steels for fusion reactor applications. 
Development of materials for fusion reactors requires facilities for testing in the fusion environment. Specifically, the effects of high energy neutrons resulting from the DT reaction on materials have to be studied. This assessment has resulted in the construction of two neutron sources, the Rotating Target Neutron Source at the Lawrence Livermore Laboratory (RTNS) and the Fusion Materials Irradiation Test Facility (FMIT) at the Hanford Engineering Development Laboratory. These facilities are described later on. 
The required research and development in the area of plasma-wall interaction 
has been more clear-cut than in structural materials. Charged particle 
loadings of first walls in Tokamak confinement will be on the order of 20% of total wall loads. This particle flux must be considered both from the point of view of damage to the wall and introduction of impurities into the plasma. The latter of these has to date proven to be of far greater importance to attainment of fusion burn. Phenomena including sputtering, blistering, absorption and reemission of gases, hydrogen reflection and backscattering, desorption and arcing have been addressed by the international community in many forums including three international conterences12 -14• In the U.S. such problems have been assessed and solutions outlined in a Plasma-Materials Interaction Program Plan15• Since impurity problems relate closely to 
V-3 

plasma physics, it is not surprising that efforts in plasma-materials 
interaction exist at almost all major plasma physics laboratories in the 
world. 
The strategy for materials testing and some recent results from the structural development and plasma-materials interaction work are given in the ·following paragraphs. 
B. Radiation Test Facilities In the area of radiation facilities for fusion materials testing, the most visible and perhaps greatest strides have been made in the recognition of testing and facilities needs and in the extension of current neutron production technologies. 
Although it was known in 1974 that fusion structural materials would be 
subject to bombardment by high energy, pri.nari ly 14 MeV neutrons as well as the blanket-moderated lower energy neutron spectrum, very little data existed regarding the effects of high energy 'neutrons on fusion materials. High energy neutron test facilities with s'ufficiently high flux capability and large enough volumes to be useful in accumulating the extensive data base required for fusion reactor design were unavailable. 
In addition to the use of systems studies and estimates of reactor 1i fetime needs, it is, of course, necessary to analyze available and anticipated radiation test environments and estimate the time required to test promising materials for reactor applications. The complexity of the fusion reactor 
environment, however, leads to the inevitable conclusion that fully prototypic 
V-4 

engineering testing can be achieved only in an actual fusion reactor. Such a facility will not be available for the next 10-15years and, even when available, may provide real time testing rather than the accelerated testing conditions required for long range materials development. In the absence of a fusion test reactor for materials studies, it is necessary to implement a strategy which takes full advantage of all presently available testing facilities and provides for the construction of new high energy neutron sources to approximate materials damage. The strategy that has been developed includes the use of fission reactors, D-T neutron sources, broad spectrum high energy neutron sources, ion simulation, and, of necessity, a strong understanding of the physical processes that occur during irradiation. The goal must be to predict materials performance in the fusion environment prior to the completion of the first fusion reactor. To accompl :sh this, it is imperative that extrapolation techniques be established to extrapolate from fission reactor data to the fusion environment. 
In overview, the strategy is: (1) To utilize effectively those characteristics of various available radiation sources that in some respect either mimic the fusion reactor environment or permit the controlled study of a significant variable or damage mechanism, (2) .To develop the correlation procedures between these radiation sources and the fusion environment using the FMIT for confirmation prior to the existence -of a fusion reactor. 
V-5 

C. Fission Reactors Alloy development and evaluation programs aimed at radiation resistance require large (multi-liter) radiation test volumes. At present, such facilities exist only as fission reactors whose energies are too low for present theories to adequately extrapolate fission-produced damage to the high energies of a blanket moderated fusion spectrum and to relate high gas generation rates to mechanical property changes. 
Although thermal and mixed spectrum research reactors with fluxes in t~e 2-4 X 1014 n/cm2-sec range can match displacement rates equivalent to wall loadings of 1-2 MW/m2 , the approximation of helium generation rates in fission reactors can be made only for nickel bearing alloys. This scheme depends on the two-stage transmutation reaction vi a the sequence 
Cross sections for these reactions are high enough that helium production rates on the order of those occuring in fusion reactors can be produced in materials containing nickel. Calculations made for the Oak Ridge Research Reactor and several other thermal reactors show that helium/DPA ratios at wall loadings up to 2 MW/m2 can be simulated in real time while mixed spectra of reactors such as the High Flux Isotope Reactor are suitable for accelerated helium production. It is possible to span the entire range of 
V-6 

first wall conditions in terms of helium production and displacement damage 
in these reactors.* 
Fast reactors do not have the correct spectra for generating the large 
quantities of helium equivalent to that of a fusion environment in any 
material. However, they can produce displacement rates equivalent to 
2-4MW/m2 and for that reason can be used for screening of materials to 
high DPA's. 
D. High Energy Neutron Sources 
At present one neutron source, the Rotating Target Neutron Source (RTNS-11), is in operation at the Lawrence Livermore Laboratory. In this facility, 14 MeV neutrons are produced by the acceleration of deuterium onto a solid rotating target with a coating of titanium-tritide on the surface. The operating parameters for this low flux, monoenergetic 14 MeV source are shown in Table V-3. 
The damage information which is already being generated in RTNS-11 is an 
essential element to relate data from the fission reactor environment to the fusion environment at low and medium fluences. In addition, the RTNS-II can 
be used for a) synergistic surface studies on materials (interaction of neutron damage with other plasma radiation), b) neutron and helium crosssection measurements, c) comparison with irradiation data from charged 
*Two relevant materials exposure parameters for radiation damage assessment 
have been adopted. The first is displacement per atom (DPA), the second 
helium, expressed as atom fraction. The rate and ratio of these exposure
terms are considered the most appropriate basis on which to characterize 
irradiation environments and provide a basis to account for the energy,
flux, and fluence dependence of neutron exposure to specific materials. 

V-7 

particles, and d) for direct end of 1 ife testing of 1 ow fl uence coin"pon'e'nts .. ' such as superconductors and magnet insulators. The data from the monoenergetic RTNS-II also will be used for damage comparison with other high energy spectra such as those arising from the Be {D,n) and Li {D,n) stripping reactions, the latter of which forms the basic reaction for the Fusion Materials Irradiation Test Facility {FMIT) under construction at the Hanford Engineering Development Laboratory. 
E. Broad Spectrum, High Flux, High Energy Neutron Sources The most important criterion in evaluating potential irradiation test environments is that the test environments be capable of irradiating materials to the projected end-of-life levels of a fusion reactor first wall. To date, production of the most intense neutron beams has been by stripping reactions 
which result in a broad spectrum of neutrons. 
Early in 1984, the FMIT should become operational. On the basis of displacement production and helium generation, this Li {D,n) stripping source appears to match more closely the damage production in a fusion reactor first wall than any other avail able· or projected test environnent. FMIT has an experi
3
mental volume of 5-10 cmat a maximum flux of 1015 n/cm2-sec and approximatel~ 500 cm3 at a flux of 1014 n/cm2-sec. The operating parameters for FMIT are shown in Table V-3. 
FMIT will be utilized for high fluence end-of-life testing of materials, development of a correlation methodology for-the utilization of fission 
reactor observations and those from other irradiation environments, and alloy 
V-8 

development--especially for non-nickel containing materials. The facility will form the basis for evaluation of data from mixed spectrum fission reactors in materials containing nickel, and for the evaluation of fast reactor irradiations of reactive and refractory metal alloys. 
F. Fusion Reactor Test Environment In the early 1990's the ETF will become available for materials testing. Thus a reactor will have the proper fusion spectrum and a large volume for testing. Accumulation of neutron damage, however, will still be at a rate of "real time" at best, and sources such as FMIT are still needed for accelerated testing and materials development. 
G. Progress in Plasma Materials Interaction With the recognition that thorough characterization of surface conditions in confinement experiments is essential to manage impurity release, heavy emphasis has been placed on in situ diagnostics. Depth profiling of surface composition using sputter Auger has been applied to stainless samples exposed in PLT, revealing carbon and oxygen as well as iron, chromium and nicke1. 16 Deuterium concentrations in silicon and stainless steel exposed in PLT have been obtained using nuclear reaction profiling16 , with significant implications both for hydrogen neutral flux to the PLT wall and eventually for tritium inventory. Techniques to observe impurity release and transport in the near-surface region are now under development. Laser fluorescence holds out the prospect of giving both impurity concentrations, and velocity distributions through Doppler shifts, while providing fast data acquisition, high space and time resolution, and minimal experimental perturbation. 
V-9 

Fluxes from plasmas to first walls are also critical determinants of surface processes. The primary components are hydrogen isotopes lost through charge exchange neutralization. This flux has been calculated for TFTR by Cohen 17 
and for UWMAK II by Conn and Kesner18• The energy distributions are generally Maxwellian, and the integrated fluxes approach 1015 H/cm2-s in both cases. 
The study of individual surface processes has progressed in support of the in ~itu work. Most of the relevant yields for physical sputtering by normally incident hydrogen at 500 ev -10 keV are now in hand19 • They are in the 10-2 -10-4 atom/ion range depending on target element, with maxima at 5-10 keV, falling slowly at higher energies and sharply at lower energies. Some information has been obtained on yields at grazing incidence, angular distributions, and energy distributions of sputtered particles. The charged fraction of sputtered particles, which is usually small, can be increased greatly by promoting oxides on some targets, such as titanium20 • Sputtering at higher incident energies, of relevance in neutral beams and in mirror 
machines, and surface chemical processes leading to impurity release, remain subjects for further study. Sputtering by energetic neutrons was briefly feared to be a significant mechanism of impurity release and wall erosion. A cooperative experiment utilizing energetic neutrons from several sources on two types of cold-rolled niobium targets led to the conclusion that the total 
. 1 1 -4 I 21
probable y1e ds were no arger than 10 atoms neutron • 
V-10 

Combining hydrogen sputtering data with accumulating information on helium 
blistering, Bauer et. a1. 23 gave an impressive example of synergistic 
effects, and of the emerging predictive capability of fusion surface science. 
In early D-T burners, there is expected to be a fraction (~ 10%) of the 3.5 
MeV alphas which will lie within the loss cone and reach the first wall 
unmoderated. It would have a range of several microns and could cause serious 
exfoliation, although it will have an angular distribution tending to broaden 
the range. Hydrogen sputtering leads simultaneously to surface recession and resultant range broadening, and the occurrence of blistering is thus determined by competition between surface erosion by D-T sputtering and the accumulation of helium at the end-of-range. Figure V-1 summarizes the likelihood of blistering in stainless steel as a function of homologous wall temperature, which determines the helium release rate and the surface velocity of recession. The surface velocity is determined by the plasma edge temperature which establishes the energy distribution of hydrogen neutrals and consequent sputtering yield. 
The attractiveness of first wall coatings, which enable the designer to select one material with optimum properties under plasma exposure and another with optimum structural/neutronic properties, was recognized several years ago. Coating development has become a major element within the U.S. Plasma/Materials 
Interaction Program. Work has been primarily on coated limiters and other 
high heat-load components, providing engineering solutions to problems in confinement systems. Coatings of interest include TiB 2 and B4c, which have been deposited on graphite. Coated limiters for ISX-B and components for POX have been completed and have successfully passed all of the qualification tests 
V-11 

STAINLESS STEEL 
.2 s·'
14
(D+T)FLUX=7.6XI0 em 
3.5MevHe FLUX= lxl012 cm·2 s·' 

1000 eV edge temp 
NO BLISTERS 
-• 
200 eV edge temp 
0 
0
-
~ BLISTERS>~ 
'-'
C) 80 eV edge temp
_, 
loU
> 
loU -11
'-'I0
cI.L. 
~
=
en 
300K BOOK BOOK 
0.5
0.1 0.3 
T(K) . 
HOMOLOGOUS WALL OPERATING TEMPERATURE 
Tm (K) 
Figure V-1 Prediction of plasma and wall conditions under which blister formation is likely to occur due to injected helium1231• 
related to erosion and adherence under thermal cycling. These are to be 
installed into the respective tokamaks in the fall. 
International cooperation in plasma-materials interaction has increased as the problems have become better defined. A formal cooperation exists on TEXTOR (Torus Experiment for Technology Oriented Research) which is under construction at the KFA Julich, with participation by Canada, Euratom, Japan, Switzerland, Turkey and the U.S.A. It is intended to help provide the data which will be required to build an ignited, long pulse tokamak in the near future. To make this possible, the objectives of TEXTOR are: 
o 	to evaluate the relative importance of the various processes leading to accumulation of impurities in tokamaks and to damage of the first wall under different operating conditions 

o 	to search for appropriate first wall materials, structures and 
temperatures that are optimized with respect to particle release 
and wall material behavior 


o 	to develop and test methods to control the plasma boundary. 


With these objectives in mind, the machine was designed with special features, which include: 
o 	accessibility for diagnostics and plasma edge layer examination 

o 	easy change-out of wall components 

o 	controllable inner wall temperature up to 600°C 


V-13 

o 	long pulse duration (3 sec) 

o 	portholes for injectors, bundle divertors, scrape-off limiters, beam dumps and plasma edge diagnostics. 


A recognized forum for the publication of results in the plasma-materials interaction area has evolved in the series of biennial conferences on Surface Interactions begun at Argonne in 1974, San Francisco in 1976, and Culham in 197812-14• The next conference will be held in April 1980 at Garmisch 
" 
Partenkirchen, FRG, and will be organized by the Max-Planck-Institut fur Plasmaphysik Garching. 
H. Progress in Alloy Development As was discussed above, materials research and development for fusion reactor concepts has experienced a metamorphosis in the short time period of the early to late 197o•s. This metamorphosis includes the creation of a materials community that is focused on fusion, the interaction of that community with the reactor design community to establish commonly understood goals for both materials development and for reactor design problem analysis, program planning relative to those goals, and the output of the first generation of experimental studies which were tailored specifically to fusion reactor concepts. 
Within the U.S. programs, the materials research and development on the· general subject of irradiation damage falls within two task areas: Alloy Development for Irradiation Performance (ADIP) and Damage Analysis and Fundamental Studies (OAFS). ADIP has as a primary Qbjective the research and development that will lead to the selection and application of the principal 
V-14 

structural materials for fusion reactor blanket and first wall applications while the OAFS activity is to establish the basis for projecting from the presently available test environments to that of the fusion reactor. These 
26 
program activities were described at the 1978 Santa Fe meet1ng24-as well as 
in the comprehensive materials Program Plans published by the Departm~nt of 15 27 28
Energy. 11 , , , This metamorphosis also may be seen by examining the 
principal international conferences on the subject in 1975 at Gatlinburg, Tennessee29 and in 1979 at Miami, Florida. 30 
The keynote papers in the 1975 conference29 made it very clear that materials in general and the influence of the fusion reactor environment on the reactor materials in particular may be expected to limit the performance and lifetime of fusion reactor systems. A technical assessment "Alloys for the Fusion 
Reactor Environment" 31 was performed within the U.S. program for the purpose of examining the technological base for structural materials that would be subject to intense neutron irradiation in fusion reactors, dominantly the blanket and first wall application. Potential candidate alloys were considered relative to properties that are important in design and critical to performance 
of reactor systems. Both technical and economic analyses were included. 
It was determined that technologically the pacing factor to alloy development 
and establishment of an engineering data base is the problem of irradiation 
testing in the absence of a fusion reactor test environment. The programmatic 
approach to this problem is discussed above. The scope of the required 
testing exceeds even that of the LMFBR program because of the number of 
materials to be considered and because cyclic phenomena--fatigue and 
V-15 

fatigue crack propagation--are domi nent in determining tokamak reactor 
structural performance limits and lifetimes. These factors must be 
considered in addition to the conventional design properties. 
The alloys that were identified for long term development in the U.S. fusion 
program are listed in Table V-4. The rationale behind those selections are 
reviewed in Ref. 25. As program results become available, this list will be 
shortened. Performance factors that are easier to establish than irradiation 
damage, e.g., coolant or breeding compatability will be important to such an 
evaluation. For the near term focus on ETF, an even more immediate set of 
criteria, including special emphasis on current industrial capability, must be 
applied. For ETF, austenitic stainless steel is the leading contender; 
however, the limits imposed by its relatively poor thermal properties and the 
recent ev1"dence32-35 f d" . . .
o outstan 1ng res1stance to 1rradiat1on damage by the family of 9-12% chromium ferritic (martensitic) steels has led the U.S. program to the conclusion that these steels may offer clear advantages over the austenitics for ETF as well as for long term fusion applications. Tradeoff and design studies36 indicate the advantage could be a~ much as a factor of five in lifetime, if concerns regarding fracture behavior and welding of complex structures are favorably resolved. 
Returning to the subject of testing, major advances have been made within the fusion materials program including the evolution of miniature test specimens for an array of properties, special tailoring of both materials, and test environments to better approximate the fusion reactor environment, in test hardware, and in the interpretation of test environments to 
V-16 
more completely account for differences in spectrum. These are illustrated in Figures V-2, and V-3. 
V-17 

Figure V-2 Superimposed test results for fatigue crack growth and comparative illustrations of conventional COD type of specimens and miniaturized center-notched test specimens. 
Figure V-3 
Irradiation capsule for RTNS-11 testing and loading assembly used to insert test specimens into the assembly. 

References 
1. 
A. P. Fraas, "Conceptual Design of the Blanket and Shield Region and 

Related Systems of a full Seale Toroidal Fusion Reactor, .. ORNL-TN-3096,( 1973}. 

2. 	
R. A. Krakowski, F. L. Ribe, T. A. Coultas, A. J. Hatch; 11 An EngineeringDesign Study of a Reference Theta Pinch Reactor (RTPR}," LASL 5336 and ANL 8019 (1974}. 

3. 	
G. A. Carlson and R. W. Moir, 11 Mirror Fusion Reactor Study, .. LLL ReportUCRL-76985 (1975}. 

4. 	
B. Badger et. al., 11 UWMAK-I, A Wisconsin Toroidal Fusion Reactor Design, 11 Volumes I and II, University of Wisconsin, UWFDM-68, (1973, 1974}. 

5. 	
R. G. Mills, 11 A Fusion Power Plant," Princeton Plasma Physics Laboratory,


MATT-1050 (1974}. 
6. 	
K. Sako et. al., 11 Conceptual Design of a Gas-Cooled Tokamak Reactor, .. 
Nuclear Fusion Special Supplement, p. 27, IAEA (1974}. 


7. 	
B. Badger et. al., "TETR, A Tokamak Engineering Test Reactor to QualifyMaterials and Blanket Components for Early DT Fusion Power Reactors,"University of Wisconsin UWFDM-191 (1977}. 

8. 	
J. R. Powell, "Preliminary Reference Design of a Fusion Reactor Blanket Exhibiting Very Low Residual Radioactivity, BNL 19565 (1974}. 

9. 	
. B. Badger et al 11 NUWMAK, A Tokamak Reactor Design Study," University of Wisconsin UWFDM-330 (1979}. 


10. 	
Fusion Materials Program Bulletin 2, "First Wall Structural Goals for Economic Fusion Power, u.s. Energy Research and Development Administration (1976}. 

11. 	
The Fusion Materials Program Pl.an, Section I: 11 Alloy Development for Irradiation Performance, .. DOE/ET-0032/1 (1978}. 

12. 	
H. Wiedersich, M. S. Kaminsky and K. M. Zwilsky 11 Surface Effects in Controlled Fusion," North Holland Publishing Company (1974}. 

13. 	
w. Bauer, C. R. Finfgeld and M. S. Kaminsky, 11 Surface Effects in Controlled Fusion Devices," North Holland Publishing Company {1976}. 

14. 	
G. M. McCracken, P. E. Stott, and M. W. Thompson, 11 Plasma Surface Interactions in Controlled Fusion Devices, .. North Holland PublishingCompany (1978}. 


V-20 

15. 	
The Fusion Materials Program Plan Section I II: 11 Pl asma Materials Inter
action," DOE/ET-0032/3 (1978). 


16. 	
s. A. Cohen, H. F. Dylla, S. M. Rossnagel, S. T. Picraux, J. A. Borders 
and C. W. Magee, J. Nucl. Mat. 76 &77 (1978) 459. 


17. 	
S. A. Cohen, J. Vac Sci Technol. 13 (1976) 449. 

18. 	
R. W. Conn and·J. Kesner, J. Nucl. Mat. 63 (1976) 1. 

19. 	
J. Roth, J. Bohdansky and W. Ohenberger, to be published. 

20. 	
D. M. Gruen, A. R. Krauss, R. B. Wright and M. B. Liu, to be published

in Proceedings of the First Topical Meeting on Fusion Reactor Materials·,Bal Harbour, Florida, January 1973. 

21. 	
R. Behrisch, 0. K. Harling, M. T. Thomas, R. L. Brodzinski, L. H. Jenkins,

G. J. 	Smith, J. R. Wendelken, M. J. Saltmarsh, M. Kaminsky, S. K. Das,
C. M. 	Logan, R. Meisenheimer, J. E. Robinson, M. Shimatomai and 
D. A. 	Thompson, J. Appl. Phys. 48,_ 394 (1977). 

22. 	
G. L. Kulcinski, University of Wisconsin UWFDM-228, 6 (1977). 

23. 	
W. Bauer, K. L. Wilson, C. L. Bisson, L. G. Haggmark, and R. J. Goldston,

J. Nucl. Mat. 76 &77 (1978) 396• 

24. 	
K. M. Zwol sky et al, "Materials Program for Magnetic Fusion Energy, Proc. Third Topical Meeting on the Technology of Controlled Nuclear Fusion, 11 Santa Fe, N.M., May 9-10, 1978, p. 549-553 CONF 780508, (1978). 

25. 	
E. E. Bloom, et al, "Alloy Development for Irradiation Performance: Program Strategy~ Ibid p. 554-564. _ 

26. 	
D. G. Doran et al, 11 The Damage Analysis and Fundamental Studies Program, 11 Ibid p. 575-587--. 

27. 	
The Fusion Materials Program· Plan, Section II: Damage Analysis and Fundamental Studies DOE/ET-0032/2· (1978). 

28. 	
The Fusion Materials Program Plan, Section IV: Special Purpose Materials,


DOE/ET-0032/4 ( 1978). . . . . 
29. 	Radiation Effects and Tritium Technology for Fusion Reactors, Producing··of the International Conference,· Gatlin burg, TN, October 1-3, 1975,CONF-750989 Vol. 1 (March 1976) •. 
30. 	
First Topical Meeting rin F~sib~ Reattor Materials, Bal Harbour, Florida,January 29-31, 1979, Americas Nuclear Society. 

31. 	
D. L. Kummer, Alloys for the Fusion Reactor Environment: A Technical Assessment, U.S. Department of Energy, DOE/ET-0007, January 1978. 


V-21 
32. 	J. J. Laidler, J. J. Holmes and J. W. Bennett, 11 U.S. Programs 
on
Reference and Advanced Cladding/Duct Development, 11 p. 41-61, RadiationEffects in Breeder Reactor Structural Materials, M. L. Bleiberg and
J. W. 	Bennett (Ed), The Metallurgical Society, AIME (1977). 
33. 	
M. M. Porton, B. A. Chen, E. R. Gilbert and R. E. Nygren, 11 Comparison ofthe Inreactor Components of Selected Ferritic Solid Solutions Strengthened

and Participation Hardened Alloys, .. J. Nuc. Matl. 80 p. 144-151 (197Q). 

34. 	
J. Erler, A. Maillard, G. Brun, J. Lehmann, J. M. Dupouy, 11 Le Comportement d1 Aeiers Ferri tiques Sous Irradiation en Neutrons Rapid, 11 ProceedingsConference on Irradiation Behavior of Metallic Materials for FastReactor Core Components, Corsica, France, June 4-8, 1979. 

35. 	
A. Little, D. R. Harries, D. R. Arkel, G. w. Lewtliwaite and 


T. M. 	Williams, .. Development of Ferritic-Martensitie Steels for Fast
Reactor Applications, .. Ibid p. 31. 
· 
36. 	S. N. Rosenworsser, W. E. Toffolo, W. Chen, J. A. Delesandro, J. M. Rawlsand P. Miller, 11 The Application of Martensitie Stainless Steels in LongLifetime Fusion First Wall/Blankets, .. First Topical Meeting on Fusion
Reactor Materials, Bal Harbour, Florida, January 29-31, 1979. 
V-22 

• 
TABLE V-1 
GOALS OF ALLOY DEVELOPMENT PROGRAM 

Time Integrated Exposure: 20-40 MW-years/m2 
Maximum Temperature: Based on Heat Conversion System 
550°C for Na:H 2o 
730°C for He:H 20 
900°C for He Turbine 
Compati bi 1ity: Compatible With At Least One: 
Li 
He 
Molten Salt 
Qualitative Factors: Fabricable, Joinable, Available, Reasonable Cost, Low Activation, Maintainable, Electric and Magnetic Compatibility 
V-23 
TABLE V-2 
ALLOY DEVELOPMENT PROGRAM 
POSITIVE FACTORS 
PATH A: AUSTENITIC ALLOYS 
1. 
LARGE DATA BASE 

2. 
GOOD FABRICABILITY 

3. 
MODEL MATERIAL 


PATH B: Fe-Cr-Ni ALLOYS 
1. GREATER STRENGTH THAN PATH A 
< 2. HIGHER TEMPERATURE THAN PATH A
•
N 
3. LESS VOID SWELLING THAN PATH A 
~ 
PATH C: REACTIVE/REFRACTORY METALS 
1. 
GOOD RADIATION RESISTANCE 

2. 
HIGH THERMAL CONDUCTIVITY 

3. 
LOW THERMAL EXPANSION 

4. 
GOOD Li CORROSION RESISTANCE 

5. 
HIGH STRENGTH/HIGH TEMPERATURE CAPABILITY 


PATH D: INNOVATIVE CONCEPTS 
LIMITING FACTORS 
1. 
STRENGTH AT TEMPERATURE 

2. 
CORROSION IN LITHIUM 

3. 
FATIGUE RESISTANCE 

4. 
DUCTILITY AT ELEVATED TEMPERATURES 

5. 
LOW THERMAL CONDUCTIVITY 


1. 
FABRICABILITY/JOINABILITY

2. 
CORROSION IN LITHIUM 

3. 
DUCTILITY 


1. 
FABRICABILITY 

2. 
JO INABILITY 

3. 
DUCTILE/BRITTLE TRANSITION 

4. 
INTERSTITIAL PICKUP 


FERRITIC (MARTENSITIC) STEELS, LONG RANGE ORDERED ALLOYS, RAPIDLY COOLED MATERIALS, COMPOSITES, OTHERS 

TABLE V-3 
OPERATING PARAMETERS FOR RTNS-11 AND FMIT 

RTNS-.11 FMIT 
4 X 1013 2 X 1016
Source Strenth (n/sec) 2 X 1013 1 X 1015
Maximum Flux (n/cm2-sec) 
Experimental Volum! atMaximum Flux (em ) 1.0 7 -10 
Experimental Volu~~ at F~uxGreater than 10 n/cm -sec (cm3) 0 500 
Operation Date Operating 1984 
v..25 
TABLE v..4 
ALLOYS PRESENTLY UNDER INVESTIGATION IN THE FUSION REACTOR ALLOY DEVELOPMENT RPOGRAM 
Path A: Reference Alloy_ 20% cw 316 Prime Candidate Alloy_ Fe-16 Ni-14 Cr-2 Mo-(Mn, Ti, Si, C) Path B: Base Research Alloy_s Fe-25 Ni-10 Cr-1 Mo Plus Fe-40 Ni-12 Cr-3 Mo Ti, Al Fe~30 Ni-12 Cr-2 Nb C, Si Fe-40 Ni-12 Cr-3 Nb Mn, B, Zr Fe-75 Ni-15 Cr-1 Nb Path C: Scopi ng A11 oy_s Ti-6 Al-4 V Ti-6 Al-2 Sn-4 Zr-2 Mo-0.25 Si Ti: Ti-5 Al-6 Sn-2 Zr-1 Mo-0.25 Si Ti-3 Al-8 V-6 Cr-4 Mo-4 Zr V-20 Ti 
V: 	V-15 Cr-5 Ti Vanstar-7 V~9 Cr-3.3 Fe-1.3 Zr-0.054C Nb: Nb-1 Zr Nb-6 Mo-l Zr Path D: Scoping Alloy_s 2 LRO Alloys 
Ferritic 	(Martensitic) Steels 9 Cr-1 Mo 
V-26 

VI. ALTERNATE APPLICATIONS Eighty percent of the deuteri um-triti urn fusion reaction energy is carried away by high energy neutrons. The high energy and penetrating power of these neutrons, and their abundance on a thermal power basis, make possible consideration of fusion for production of portable (synthetic) and fissile fuels as 
. 
alternate applications to direct production of electricity. Programs both in the u.s. and abroad have been exploring these options for several years. 
The use of fusion to produce portable fuels would address an energy demand which currently exceeds that for electricity by about 3 to 1. This is illustrated in Figure VI-1 which shows the flow of energy utilization in the u.s. for 1975. 1 If fusion can be developed to help meet this demand for liquid and gaseous fuels, it could contribute significantly to meeting 
the acute shortages of petroleum and natural gas expected early in the next century without requiring expensive conversion of end use technology. 
With this in mind, in October 1977 the Office of Fusion Energy of the U.S. Department of Energy convened an ad hoc adivsory panel to evaluate the 
potential for portable fuel production from fusion and to identify the most promising processes. The panel report2 concluded that the potential 
for portable fuel production from fusion was favorable, and listed high 
temperature electrolysis (HTE), thermochemical cycles, and radiolysis as 
the suggested processes for study, in decreasing order of promise. 
Brookhaven National Laboratory (BNL) began work on fusion HTE under DOE sponsorship in 1976. Figure VI-2 shows the potential attractiveness of 
VI-1 

ENERGY SOURCES ·ENERGY CONVERSION FACILITIES USEFUL ENERGY APPLICATIONS 
falling hydroelectric 
water -----------, eneration 
illumination !light uranium --------• 
industrial drive refrigeration coal 
other electric (power 
transportation ·and farm work 
cooking and oil water heating 
< I space heat 
-
N 
lheat 
industrial process natural heat 
·.': ..:,·
gas l 
,,·. 
conversion losses stack lossesFigure Vl-1 Flow of energy through the United 
States economy in the mid-1970s. from major UNAVAILABLE ENERGY 
energy sources (left) through conversion facilities 
(middle) to useful applications (right), and 

unavailable energy at the bottom. Width of a chanReference Vl-1nel is proportional to the amount of energy; 
Figure Vl-2 	HYDROGEN PRODUCTION· FROM FUSION REACTORS WITH HIGH TEMP. ELECTROLYSIS (HTE) . 
80 
> 
(!) 
a: 
w 
z 
w 
70 
z 
0 
UJ 
::J 
u.. 
1
< 
>::J
...... 	60
00..
I 
w z~
w.._ 
-> 
~(!)
u..a: 
u..w 
wz w 501 HHV 

...J 
w 
::J 
u.. 
N 
:I: 
-40% EFFICIENT ELECTRIC-POWER CYCLE
\a
::J 
HHV -HIGHER 	HT VALUE
D.. 
1-

LHV -LOWER HT VALUE 
1-	••
0 ::J @r .>LHV 	AGE SYSTEM ::J? 
CONV ELECTROLYSIS 301-• LURGISYSTEM 
20 I  !  !  !  !  -' .':..  -'·  -'~  .':..  L  _!.,.  -··  -~ .L  -~- I_  J- I  -'  I  
TEMPERATURE-100's °K  
Reference Vl-3  

HTE as process temperatures are increased. 3 Compared to conventional low 
temperature electrolysis, significant improvement in the useful hydrogen 
energy content produced compared to input energy invested can be attained 
(46% compared to 32-37%). Part of Brookhaven's work has focused on identi
fication of candidate blanket materials, recognizing this issue as the key 
to producing high temperature blankets for fusion portable fuel production. 
Preliminary results from weight loss tests of materials exposed to high temperature flowing steam have been encouraging, with Zirconia (Zr02) and Alumina (Al 2o3) the most promising. The HTE process has been selected by the U.S. program as having the highest potential for hydrogen production 
from fusion, and future efforts will concentrate in this area. A small bench-scale demonstration unit is planned for FY 1980 to confirm the basic process, with the subsequent possibility of building pilot scale modules for hydrogen generation in the Fusion Materials Irradiation Test Facility 
(FMIT) or Engineering Test Facility (ETF). A schematic of a proposed demonstration system using FMIT neutrons is shown in Figure VI-3. 3 
Work on thermochemical decomposition of water to produce hydrogen has been underway in the u.s. and abroad, independent of fusion development, for many years. Thermochemical cycles currently under active research and development internationally are shown in Figure Vl-4. 4 The U.S. Magnetic Fusion Program has taken advantage of the ongoing work at Los Alamos Scientific Laboratory 
(LASL) on the Bismuth Sulfate-Sulfuric Acid Cycle to evaluate coupling a fusion reactor to this cycle. The basic three-step cycle is shown in Figure VI-5. 4 The third step is actually a complex series of chemical steps 
Vl-4 
I Figure Vl-3 HTE PILOT PLANT FOR 
i 

HYDROGEN PRODUCTION 
720 L/min H2 .--~ PURIFICATION BED 
19 Llmin COOLING 
360 Llmin 02 H20 ~t

+ .. 

....------~---~ 
~-----~STEAM H2-STEAM
NEUTRON ELECTROLYZER 
a:

SOURCE / FUSION 800°C 
w 

35 kW Q ?-BLANKET en 
z

~ 
........_ MODULE -w L-r-------.,....,f4-+ c 
..... z POWER <:J 014-11°C At -1 (J (J
SUPPLY 0 

87.8 kW (") 0 

3.8 L/min H20 20VDC
COOLING 
H20 

1.21 Kg/min H20 STEAM 
~ 150°C ' 
5 atm 
STEAM 

I--
H20GENERATOR 
0.5 Kg/Min 1 atm 
POWER 

1.21 
SUPPLY 

L/min 55kW 
5 atm 
110 VAC 
r 

RECYCLE H20 n.& Kg/hr
h SUMP 
\._1 --TANK 

PUMP 
·~~ 
MAKEUP H20 

0.76 L/min 
VI-5 Reference Vl-3 



Figure Vl-4 THERMOCHEMICAL CYCLES UNDER ACTIVERESEARCH AND DEVELOPMENT 
CONTINUOUS-CIRCUIT, BENCH-SCALE TESTS: 
• 	HYBRID SULFURIC ACID CYCLE (WESTINGHOUSE ELECTRICCORPORATION, EURATOM (MARK 11)) 
• 	SULFURIC ACID -HYDROGEN IODIDE CYCLE (GENERAL ATOMICCORPORATION, EURATOM (MARK 16)) 
• 	HYBRID SULFURIC ACID -HYDROGEN BROMIDE CYCLE
(EURATOM (MARK 13)) 

ALTERNATE .CYCLES UND.ER RESEARCH: 
• BISMUTH SULFATE -SULFURIC ACID CYCLE (LOS ALAMOS< s·CIENTIFIC LABORATORY)
........ . 	..

~ • MAGNESIUM-IODINE CYCLE (NATIONAL CHEMICAL LABORATORY FOR. INDUSTRY, JAPAN) · 
. 
• COPPER SULFATE CYCLE (INSTITUTE OF GAS TECHNOLOGY, USA)• POTASSIUM IODIDE -AMMONIA CYCLE (ARGONNE NATIONAL •LABORATORY) 
• 	BARIUM HYDROXIDE -COPPER CYCLE (OAK RIDGE NATIONALLABORATORY) 
• 	ZINC OXIDE CYCLE (LAWRENCE LIVERMORE LABORATORY)
• 	CESIUM OXIDE -HYDROCHLORIC ACID CYCLE (LOS ALAMOSSCIENTIFIC LABORATORY) 
• 	CALCIUM BROMIDE -IRON OXIDE CYCLES (UNIVERSITY OF TOKYO,JAPAN) Reference Vl-4 
Figure Vl-5 HYBRID BISMUTH SULFATE THERMOCHEMICAL CYCLE (LOS ALAMOS SCIENTIFIC LABORATORY) 
• A THREE STEP CYCLE: 
S02(g) + 2H20U)-+ H2S04 (sol) + H2(g) H2~04(sol) + %Bi203(s) -+ %Bi203 • 3SOa + H20U)
< 
..... 
I 
%Bi203 • ~S03h;) -+ %Bi203(s) + S02(g) + Yz 02(g)
...... 
Bi203 • 3S0a(s) -+ Bi203 • 2S0a(s) + SOa(g) Bi203 • 2S0a(s) -+ Bi203 • SOa(s) + S03(g) Bi203 • SOa(s,l) -+ Bi203(s,l) + S03(g) so3(g) ... so2 + Yz02(g) 350K (ELECTROLYSIS) 350K 900-1100K 
Reference Vl-4 
at varying temperatures to decompose the Bi 203"3S03• The key to the process and its relationship to the fusion driver is the potential for increasing the overall process efficiency by conducting this series of decomposition steps using high temperature heat from a fusion blanket. This is shown schematically in Figure VI-6. 4 The USSR is actively investigating a two step iron oxide cycle for fusion fuel production. The close cooperation between the Institute of.High Temperature and the Kurchatov Fusion Institute has benefited this work and demonstrates a high degree of interest in this area by the Soviet fusion program. Future work in the U.S. in this general area is expected to focus on identification of other thermochemical cycles which can be simpler and make more direct use of the fusion high temperature heat. 
As mentioned previously, the key to successful development of a fuel-producin~ fusion reactor is development of a high temperature blanket system. This has been the focus of DOE-sponsored work at the Argonne National Laboratory (ANL). A conceptual design has been developed for a falling bed blanket using ceramic spheres as the heat transfer medium. 5 The falling bed design, shown schematically in Figures VI-7 and Vl-8, 5 removes any structural integrity requirements from the high temperature IDaterial (ceramic spheres) and permits the structural components to operate at substantially lower temperatures. Future work in this area will concentrate on experimental verification of ~hermal-hydraulic characteristics and radiolytic effects for this design concept. 
VI-8 

Figure Vl-6 	SCHEMATIC DIAGRAM OF BISMUTH-SULFATE/ SULFURIC-ACID THERMOCHEMICAL HYDROGEN CYCLE DRIVEN BY A FUSION REACTOR 
-r--H20  de ELECTRICAL ENERGY H 2  
ELECTROLYZER 2H20 +S02+-H2S04 + H2  IH2S04 (A~  BISMUTH SULFATE REACTION Bi203 • S03 + H2S04 ... Bi203 • 2S03 + H20  
H20  
< ....... I \.0  so2 02 SULFUR DIOXIDEOXYGEN SEPARATION  so2. 02  Bi203 • S03 BISMUTH SULFATE DECOMPOSITION Bi203 • 2S03-Bi203 • S03 +S03  Bi203 • 2S03  
so3  
PLANT AUXILIARIES t I IL __  ELECTRICITY GENERATION  
FIRST-WALL COOLANT LOOP  FUSION REACTOR HIGH TEMPERATURE BLANKET (LITHIUM. BOILERI  

Reference Vl-4 
Figure Vl-7 	SCHEMATIC DIAGRAM OF FUSION REACTOR HIGH TEMPERATURE BLANKET CONCEPT USING FALLING BED OF CERAMIC PELLETS 
HEAT EXCHANGER 
I 
AIR OR STEAM 
Reference Vl-5 
VI-10 

-

( -...
~ ' 
Figure Vl-8 DETAIL OF Li BREEDING ZONE AND FALLING BED ZONE FOR FUSION REACTOR BLANKET USING FALLING BED OF CERAMIC PELLETS 
PLASMA 
REGION 

PLASMA 
SLAB CONFIGURATION 
PIN CONFIGURATION 
Li BREEDI ZONE 15cm 
PANEL WALL-H20 COOLED 
INSULATION-AI302 
(2.5 em) 
FALLING 
BED 

Hit--+-----20-60 em ----n--~'H 
ZONE 
40cm 

0 0 0 0 0 0 0 
SHIELD 
ZONE 
90cm 

111111 0 5 10 15 20
~ 
SCALE. em Reference Vl-5 
VI-11 

Radiolysis schemes for portable fuel production have been judged less favorable due to the low thermal efficiencies (~1~) which have been identified. u.s. work in this area has been limited largely to examination of methane production by KMS Fusion, Inc. Although favorable results have been reported, confirmation has been difficult due to the proprietary nature of the work. Some preliminary work has also been done in the USSR using pulsed, inertial confinement (Relativistic Electron Beam) driver concepts to deliver neutron and gamma energy directly to a co2 energy capsule for high temperature (4100°K) dissociation. The potential attractiveness of this system has not been fully evaluated. 
The production of fissile fuel from fusion has also been a subject of great interest, internationally, in recent years. Figure VI-9, taken from a report prepared jointly by the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency and the International Atomic Energy Agency (IAEA), 6 shows that world-wide uranium resources are likely to be exceeded by demand early in the next century, even if only modest nuclear growth and the use of fissile fuel recycle are assumed. On the other hand, the expansion of nuclear power in the U.S. is currently uncertain and the need for fissile fuel breeders is not assumed by current policy. The U.S. fusion program is evaluating the role of hybrid breeders within the context of DOE fusion policy to develop fusion•s potential as a virtually inexhaustible resource as fully as possible. The u.s. program began looking at hybrids in detail about 3 years ago, and has examined a classical mirror hybrid reactor, the possibility of a hybrid blanket for a TFTR-scale tokamak, a near-term, major hybrid demonstration tokamak and various hybrid blanket 
VI-12 
Figure Vl-9 WORLD CUMULATIVE 
URANIUM REQUIREMENTS (2000-2025) 

Cumulative Uranium Requirements 1106 tonnes Ul 
20  (No recycle)  
15  I 1  High Power Growth (Recycle)+ Committed**  
10  
5  

19n-19BO 1990 2000 2010 2020 2030 Year Reference Vl-6 
VI-13 

designs. The program is now focusing a major effort on producing two commercial scale nybrid reactor designs, one a tokamak and one a tandem mirror. These design studies will provide, in 1980, a broad basis for evaluation of hybrid energy systems considering economic, environment/safety and nonproliferation issues. ·comparisons also will be possible of alternate confinement concepts, blanket designs and fuel cycles for hybrid applications. The results from these major studies will determine the future directions of the U.S. hybrid program. International interest in the hybrid has been particularly high in the USSR where this application comprises the main thrust of their program. The Soviets are currently working on conceptual designs of tokamaks, tandem mirror, and inertial confinement hybrid reactors, as well as molten salt blanket concepts and a renewable fission fuel cycle similar to the 
7U.S. "refresh" concept. 
For fusion as a whole, pursuit of the hybrid option could enable an acceleration of fusion•s energy impact. If fission reactors are built in substantial numbers over the remainder of this century, fusion hybrids could supply fuel to an existing industry immediately upon their initial deployment. This could possibly produce important energy impacts from fusion earlier than would be required to assimilate fusion reactors into existing utility networks for electricity production. This is illustrated in Figure VI-10, where the 
shaded area represents the number of fission convertor reactors which could be supplied with hybrid fuel beginning with deployment of the first hybrid. 8 
VI-14 

Figure Vl-10 HYBRID DEVELOPMENT AND ENERGY CONTRIBUTION TIME SCALE 
~r-----~--------------------------------------------~~----~ 

en
... 
0 
t) 200 
C'CI 
Cl) 
CE: 
0 
-
... 
Cl) 
.Q 

E 100 
::::J 
z
< 
...... 
I 
...... 
(.J"' 
0 
..,.___ ..,. 
Engineering decision 
en 
Cl) 
g) 
C'CI

... 
en 
c 
~ 
CE: 
.. _... 

Test-facility operation 
Fission converter reactors supplied with hybrid 
produced fuel 
... 

Prototype reactor operation 
•

Demo reactor operation 
Fusion-tlsslon hybrid reactors 
Hybrid deployment 
1990 2000 2010 2020 
Reference Vl-8 
In summary, the investigation of alternate applications for fusion is 
intended to assure its development for the widest feasible range of end 
uses. If successful, this approach can both broaden and accelerate the 
useful exploitation of fusion for solving world-wide energy problems. 
References 
1. 	
11 Encyclopedia of Energy, .. D. N. Lapedes, Editor-in-chief, McGraw-Hill, New York, 1976. 

2. 	
CONF -770593, 11 Fusi on Energy Applied to Synthetic Fuel Production, 11 

U.S. DOE, October 1977. 

3. 	
BNL-26334, 11 Status of Brookhaven National Laboratory Fusion SynfuelProgram, 11 BNL, May 29, 1979. 

4. 	
LA-UR-79-1115, 11 Synfuel (Hydrogen) Production from Fusion Power, 11 

R. A. Krakowski, et al, Los Alamos Scientific Laboratory, presented at 14th Intersociety Energy Conversion Engineering Conference,· August 5-10, 1979, Boston, Massachusetts. 

5. 	
Undistributed Report, 11 Fusion Energy for Alternative Applications:The Generation of Synthetic Gaseous Fuels, .. ANL Fusion Power Program,May 30, 1979. 


6.. 	11 Urani urn Resources, Production and Demand, 11 joint report by the OECD Nuclear Energy Agency and the International Atomic Energy Agency,
Organization for Economic Cooperation and Development (OECD), Paris 1977. 
7. 	
..Proceedings of the Third US/USSR Symposium on Fusion-Fission Reactors, January 22-26, 1979, 11 in press. 

8. 	
11 The Promise and Timetable for Fusion-Fission Hybrid Systems, .. 


R. E. Balzhiser (EPRI) and E. E. Kintner (DOE), presented at the Second International Forum on Fundamental World Energy Problems, UNESCO, June 11-15, 1979, Santiego de Compostela, Spain. 
VI-16 


VII. ENVIRONMENT AND SAFETY The ultimate acceptability of fusion power reactors as a major energy source will depend heavily on the safety and environmental impact of their operation. 
~ 
The objective is to ensure that the development and operation of fusion reactors occurs with minimum adverse effect on the environment and maximum safety .for both the public and reactor operating personnel. 
Eventually, a fusion reactor may be able to use only deuterium or other alternate fuels that are essentially free of any radioactive products and has a truly minimal environmental and safety impact. However, first generation plants will use deuterium and tritium as fuel due to limitations in plasma technology. The important factors from an environment and safety viewpoint of deuterium-tritium fusi~n power are: (1) it produces substantial numbers of neutrons that result in induced radioactivity within the reactor structure, and (2) it requires the breeding and handling of the radioisotope tritium. 
Even though tritium has been handled and used safely for many years in other program areas, much still needs to be done regarding its application in fusion technology. With this purpose in mind, the Tritium Systems Test Assembly (TSTA) is being built at the Los Alamos Scientific Laboratory (LASL) at a cost of $13.5 million. 1 Figure VII-1 is an artists• conception of the completed facility. TSTA, scheduled to operate in 1981, is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle as well as to demonstrate the ability to handle large tritium inventories in fusion reactor-like systems with minimal environmental and 
~ 
safety effects. Figure VII-2 is a schematic flowsheet of the TSTA process 
VII-1 

...... 
c.f) UJ 
...... 
.... 
I 
c.f) > 
~ Q) 
~ 
::I 
en
UJ 
...... u::: 
c.f) 
>
c.f) 
VII-2 

TSTA 
MAIN PROCESS LOOP AND AUXILIARY SYSTEMS 
Figure Vll-2 

TRITIUMEXPERIMENTAL WASTEIMPURITIES CONTAMINATION .... 
~ TREATMENT """"'':"STUDIES 10 gm 
.. 
' 
NEUTRAL VACUUM FUEL
TORUS ....
.....
BEAM ,. . SYSTEM --CLEAN-UP -
MOCK-UP
INTERFACE 6gm 20gm < 
....... 

....... 

'~ 
w I ~ ,w
;'kg/da) 
ISOTOPESTORAGE .... SEPARATION10 gm -104 gm --
MASTER DATA ACQUISITION & CONTROL EMERGENCY TRITIUM CLEAN-UP GAS ANALYSIS WASTE DISPOSAL TRITIUM MONITORING INVENTORY CONTROL 

systems. The Japanese have proposed building a facility similar to TSTA in 
the 1983-1984 timeframe, which would also demonstrate the total fuel cycle for fusion. 2 
The effort at TSTA is supplemented by work in other areas such as tritium 
permeation, extraction, and monitoring. There is also a substantial effort 
by the Biomedical and Environmental Research Divisions within the U.S. DOE 
to identify and develop a better understanding of the radiological effects, 
transport pathways, etc., of tritium and other important radionuclides upon 
biological and environmental systems. These efforts also include an under
standing of the biological effects of magnetic fields, the development of 
personnel dosimeters to assess occupational exposure to magnetic fields, the 
development of instrumentation and techniques to monitor tritium releases 
and transport, and the development of methods for fire protection. 
The major objective and future direction of all of this work is to develop a solid base of understanding from which it can be demonstrated that tritium, 
the most severe potential hazard with magnetic fusion power, can be controlled and contained. 
Another important and potentially hazardous material that will be used in a 
D-T fusion reactor is lithium. Lithium spill experiments have been conducted including lithium reactions with a variety of materials (concrete, air, insulation), fire extinguishment, and aerosal air cleaning. In conjunction with this experimental work, there is a continuing development of analytical 
tools such as lithium fire models to aid in the understanding and control of this material. As a result of this program and other considerations, the use 
VII-4 
of liquid lithium as a coolant is a design option which appears, as the program has progressed, to be less 1 ikely. Other more enviro1111entally acceptable coolants such as helium are now being considered more seriously than before. 
Of course, the overall obJective of the safety program is to develop and 
demonstrate magnetic fusion energy as an acceptable reactor concept. 3 In order to help reach that goal, the Department of Energy has chosen Edgerton, Genneshausen and Grier (EG&G), Idaho Inc. as having the 11 lead lab11 role in Fusion Safety Research. They are developing a program plan which will identify th€ safety information needed for magnetic fusion power. In addition, EG&G is conducting programs in .other areas of safety research such as activation products, superconducting magnet safety, and safety design criteria. 
In order to establish the environmental acceptability of this energy source, the U.S. DOE's Office of Fusion Energy has initiated studies leading to the prepartt.ti on of an Environmental Impact Statement for the entire magnetic fusion program. Completion of a draft statement is expected in 1981. Additional studies are being initiated to_ consider the commercialization and public perception issues associated with fusion reactor systems. 
Initiatives are also developing internationally. The IAEA is planning to 
institute international cooperative programs in fusion environment and safety. With this in mind, a Consultants• Meeting on the Environmental Aspects of Fusion is being planned by the IAEA for later this year to lay the groundwork for this cooperation. From recent experiences with other energy sources, it 
VII-S 

cannot be overemphasized that it is extremely important to demonstrate the environmental acceptability of fusion, otherwise this viable energy resource could be inhibited from deployment. 
In conclusion, a vigorous effort is needed in order to demonstrate the environmental and safety aspects of fusion power. For this to be accomplished, the environmental and safety programs must keep pace with the technological development. 
References 
1. 	
Tritium Systems Test Assembly: Design for Major Device Fabrication 

Review, LA-6855-P Los Alamos Scientific Laboratory (LASL) June,1977. 

2. 	
IAEA, International Tokamak Reactor (INTOR) Workshop, JapaneseReport of Group 13: Tritium, June 1979. 

3. 	
U.S. Department of Energy Environmental Development Plan for the Magnetic Fusion Program, March 1978. 


VII-6 

VIII. CAPSULE SUMMARY OF U.S. DEVELOPMENT AND TECHNOLOGY PROGRAM 

VIII-1 

< 
....... 

....... 

....... 

I 
N 

U.S. DEVELOPMENT AND TECHNOLOGY 
PROGRAM OBJECTIVES 

• 	
PROVIDE SYSTEMS ENGINEERING SUPPORT AND COMPONENT DEVELOPMENT FOR MAJOR PHYSICS EXPERIMENTS 

• 	
DEVELOP TECHNOLOGY BASE FOR FUSION TO BECOME COMMERCIAL REALITY 

• 	
DESIGN AND OPERATE FUSION ENGINEERING TEST FACILITIES 


U.S. DEVEVELOPMENT AND 
TECHNOLOGY MAJOR PROGRAM 

ELEMENTS 
• REACTOR STUDIES 
• ENGINEERING TEST FACILITY DESIGN < • PLASMA FUELING, HEATING AND EXHAUST 
...... 
...... 
...... 
w 
I • MAGNETIC SYSTEMS 
• 
MATERIALS 

• 
ENVIRONMENT AND SAFETY 

• 
TRITIUM TECHNOLOGY 

• 
ALTERNATE APPLICATIONS 


REACTOR  
CONCEPT  
TOKAMAKS  
· 
<  
......  
......  
.......,  MIRRORS  
""'  
EBT  

PROGRESS IN MAGNETIC CONFINEMENT REACTOR STUDIES 

1974 
• 	
INITIAL DESIGNS PRO-JECTED LARGE, COM-PLICATED REACTORS 

• 	
SAME AS ABOVE 

• 	
LOW POWER AMPLIFICATION (0<1.5) 

• 	
NO DESIGN 


1979 	1984 
• 	
MAJOR SIZE REDUCTIONS • ETR AND INTOR DESIGNS COMPLETED 

• 	
DEVELOPAaLETECHNOLOGY • DESIGN FOR MAINTAINABILITY INCORPORATED 

• 	
GOOD EXPERIMENTAL BASIS • SAFETY AND ENVIRONMENTAL PROTECTION INCLUDED IN DESIGN 

• 	
0 = 5 REACTOR DESIGNED 


e 	0>10 REACTOR CONCEPT IDENTIFIED 
e 	INITIAL DESIGN PROJECTED LARGE, NONCOMPETITIVE REACTOR 
e 	LIMITED EXPERIMENTAL BASIS 
PROGRESS IN MAGNETIC FUSION
PLASMA HEATING 

1974 	1979 
1984 NEUTRAL BEAMS • 20keV, 50ms OPERATING • ISX-B OPERATING AT 40keV, • 150keV, 30 SEC. WITH ENERGY150ms RECOVERY 
• 	DESIGNS FOR PLT, • SOURCES FOR DOUBLET-Ill ANDDOUBLET-Ill STARTED MFTF OPERATED A.T 80keV,
..
. . 
500mSEC.
...... 
j.' •
......
......
I • TFTR SOURCE 01-'ERATED AT

U"l -' 	'. 
1201('eV 1.5SECS. 
ECH 	• NO PROGRAM • 
28 GHz 	• 110GHz
250kW 	200kW, cW
50kW cW 
PROGRESS IN MAGENTIC FUSION FUELING AND IMPURITY REMOVAL 
1984
PROBLEM 1974 	1979 
., 
• 	PELLET FUELING ON ISX-B • TRITIUM QUALIFIED, RELIABLE
FUELING • NO PROGRAM 'INCREASES DENSITY BY ETF/REACTOR PELLET INJECTOR 
• 	CONSIDERED POTENTIALLY FACTOR OF 5 AVAILABLE UNSOLVABLE
<
.......

.......

.......

I
C"l DIVERTORS • NO PROGRAM • 	PRACTICAL CONCEPTS BE lNG • ETF DESIGN CONCEPTS TESTED STUDIED ON TEST STAND 
• 	IMPRACTICAL CONCEPTS • POX OPERATIONAL • STRUCTURAL, HEAT TRANSFER, AND PUMPING PROBLEMS RESOLVED 
• 	ETF DIVERTOR PROTOTYPE CONSTRUCTION 
PROBLEM 
SUPERCONDUCTING MAGNETS 
< 
~ ~ ~ 
I 
........ 

PROGRESS IN MAGNETIC FUSION 
SUPERCONDUCTING MAGNETS 

1974 
• 	BASIC SUPERCONDUCTIVITY PROGRAM IN INITIAL PHASE 
1979 
• 	
LARGE COl L TEST fACILITY 20% COMPLETE 

• 	
THREE LARGE COl LS IN CONSTRUCTION 

• 	
NBTI CONDUCTOR A COM-MERCIAL PRODUCT 

• 	
MIRROR FUSION TEST FACILITY MAGNET WOUND SUCCESSFULLY 


1984 
• 	
ALL ETF SUB-.SI ZE COl LS TESTED 

• 	
ETF PROTOTYPE COl L NEAR COMPLETION 

• 	
ETF COIL TEST FACILITY ON LINE 

• 	
NB SN CONDUCTOR A COMMER-CIA PRODUCT 


PROBLEM 
MATERIALS DEVELOPMENT 
< 
...... 
...... 
...... 
1 
co 
PROGRESS IN MAGNETIC FUSION MATERIALS DEVELOPMENT 

1974 
• 	
NO HIGH ENERGY NEUTRON SOURCES AVAILABLE 

• 	
PROGRAM PLANNING STARTED 

• 	
SMALL-SCALE FUNDAMENTAL STUDIES ($1.5M) 


1979 
• 	
ROTATING TARGET NEUTRON SOURCE-II OPERATING 

• 	
FMIT APPROVED IN TITLE-I 

• 	
DETAILED PROGRAM PLANS AVAILABLE 

• 	
BALANCED MATERIALS PRO-GRAMS IN PLACE ($9M), ALLOY DEVELOPMENT, DAMAGE ANALYSIS, PLASMA-MATERIALS INTERACTION 

• 	
LOW Z COATINGS PROGRAM IN PLACE, LIMITERS COATED FOR'ISX-B, POX 

• 	
WORK ON AUSTENITICS, NICKEL-BASED ALLOYS AND REFRACTORY METALS INITIATED, FERRITICS TEST PLAN DEVELOPED 


1984 
• 	
FUSION MATERIALS IRRADIATION TEST FACILITY OPERATING 

• 	
DAMAGE ANALYSIS AND DEVELOP-' MENT EXPERIMENTS FOR FMIT IN PLACE 

• 	
FERRITIC AND AUSTENITIC SS QUALIFIED FOR ETF 

• 	
LOW Z, ACTIVELY COOLED LIMITERS FOR TFTR & ETF FABRICATED AND TESTED 

• 	
UNDERSTANDING OF HIGH ENERGY NEUTRON EFFECTS AT LOW AND . INTERMEDIATE EXPOSURES ESTABLISHED I 


' 
J 
PROGRESS IN MAGNETIC FUSION -TRITIUM HANDLING 
PROBLEM 1974 1979 1984 
TRITIUM • SMALL STUDIES • TSTA 50% COMPLETED, WILL • FULL-8CALE ETF FUEL CYCLE ASSOCIATED WITH OPERATE IN 1982 AND CONTAINMENT/CLEANUP < REACTOR SCOPING SYSTEM OPERATED & OPTIMIZED
...... 
...... 
DESIGNS • TFTR TRITIUM SYSTEM IN 
...... 
I FABRICATION • FUELERS, NEUTRAL BEAM
\.0 
INJECTORS, VACUUM PUMPS 
TRITIUM QUALIFIED AND 
TESTED IN TSTA 
J 
PROGRESS IN MAGNETIC FUSION-ENVIRONMENTAL STUDIES 

1974 	1979 
ENVIRONMENT/ • NO PROGRAM • GENERIC PROGRAM EIS SAFETY INITIATED 
• 	INEL SELECTED AS LEAD SAFETY RESEArK:H LAB 
• LITHIUM FIRE SAFETY TESTS < IN PROGRESS 
...... 
...... 
...... 
I 
_. • LOW ENVIRONMENTAL 0 HAZARD REACTOR DESIGN STARTED 
• 	PROGRAMS UNDERWAY IN 
1984 
• 	
GENERIC PROGRAM EIS ISSUED 

• 	
DESIRABLE ENVIRONMENTAL AND SAFETY FEATURES FOR ETF, FUSION REACTORS IDENTIFIED AND INCORPORATED INTO DESIGN 

• 	
FUSION REACTOR SAFETY RESEARCH PLAN ISSUED 


TRITIUM PERMEATION, SAFETY 
ANALYSIS METHODOLOGY, 
ACTIVATION PRODUCT 
ANALYSIS 

• 	EIS AND SAR COMPLETED FOR INS,FMIT,TFTR,TSTA 
PROGRESS IN MAGNETIC FUSIONALTERNATE USES FOR 
FUSION ENERGY 

1974 1979 	1984 
PORTABLE • NO PROGRAM PROGRAMS IN PLACE INVESTIGATING: DEMONSTRATION OF HYDROGEN 
< 
....... 

....... (SYNTHETIC) 	PRODUCTION USING PROCESS 

....... 

• 	HIGH TEMPER'ftTURE ELECTROLYSIS 
I FUEL 	I 
....... 	EXTRAPOLABLE TO FUSION 

....... 

PRODUCTION 	• THERMOCHEMICAL CYCLES APPLICATION 
• 	HIGH TEMPERATURE FUSION i BLANKET DESIGN 
I 
ENGINEERING TEST FACILITY (ETF) 
• 	
THE DEVICE IN WHICH THE TECHNOLOGICAL FEASIBILITY OF FUSION ENERGY TO PRODUCE USEFUL POWER WILL BE DEMONSTRATED. 

• 	
A TEST BED TO DEVELOP AND DEMONSTRATE THE TECHNOLOGICAL REQUIREMENTS FOR A USEFUL FUSION POWER REACTOR. 


< 
...... 
...... 
...... • A PROJECT TO PROVIDE PACE AND RELEVANCE TO 
....... 

I 	ENTIRE PROGRAM . 
N 
A 
• 	
WILL BE THE FIRST DEVICE TO LIGHT A LIGHT BULB USING FUSION POWER. 

• 	
LIKELY TO BE A TOKAMAK. STUDIES INDICATE TOKAMAK TECHNOLOGY ADEQUATELY DEMONSTRATES TECHNOLOGY FOR ALTERNATE CONCEPTS. 

• 	
FOLLOWS DEMONSTRATION OF SCIENTIFIC FEASIBILITY IN TFTR. 



·~ ETF PROGRAM PLAN 

• 	
COMBINE THE CAPABILITIES OF THE NATIONAL LABORATORIES AND INDUSTRY INTO AN ETF DESIGN CENTER 

• 	
DEVELOP A DETAILED DEFINITION OF THE ETF MISSION 


< 
....... 

....... 

....... 

f:; • ASSESS THE R&D NEEDS TO SUPPORT ETF DESIGN AND CONSTRUCTION 
• 	
COMPLETE ETF CONCEPTUAL DESIGN TO BEGIN TITLE I IN FY 1984 

• 	
COMPLETE CONSTRUCTION IN FY 1992 



ETF STATUS 
• 
DESIGN CENTER, MANAGER SELECTED 

• 
INITIAL STAFF ESSENTI.ALLY COMPLETE 


I 
< 
...... • RFP FOR 50% IND.USTRY PARTICIPATION ISSUED 
...... 
...... 
I
__. 
+::-• FINAL MISSION STATEMENT PREPARED 
• 
R&D NEEDS ASSESSMENT DUE PRIOR TO END OF FY 1979 

• 
FY 1979 BUDGET~$2.0M 



ETF CHA.RACTERISTICS 
• 	IGNITED D-T PLASMA 
• 	LONG BURN PULSE (TBURN~ 100 SEC.} 
• 	HIGH DUTY CYCLE (5 50%) 
<
..... 

!: • BETA t3 4%
<..n 
• 	R ~5M, a= 1.2m/1.9m 
• 	MODULAR BLANKET CAPABILITY AND MODULE INTERCHANGEABILITY 
• 	A TEST BED FOR COMPONENTS, MAT-ERIALS AND OPERATION 
• 	1MW(e) TEST LOOP 
INTOR ORGANIZATION 

IAEA 
IFRC 

I 
I J 1 
JAPAN 
CEC U.S.S.R. S.MORI G. GRIEGER B. KADOMTSEV 
<
-

-

... 

-

..;.., 
cri 
U.S.A. W.M.STACEY 
I 
"PL,ASMA PHYSICS 
I 
MATERIALS 
I 
STARTUP, BURN AND SHUTDOWN 
-
I 
FIRST WALL 
BLANKET AND 
SHIELD 

.. ' 
I'., ' 
I 
SYSTEMS INTEGRATION 
I 
TRITIUM 
-
I 
FUELING, EXHAUST 
AND 
IMPURITY CONTROL 

I 
POWER SUPPLY 
AND TRANSFER 
1 
VACUUM 
I 
HEATING 
1 

ASSEMBLY 
AND 
MAINTENANCE 

I 

ENVIRONMENT 
AND 
SAFETY 

-
I 
MAGNETICS i 
I
AND CRYOGENICS 
I 
I 
STABILITY 
CONTROL 

I 
DIAGNOSTICS 
INTOR CHARACTERISTICS 

!= 
I ,.. ~ I 
a • D-T PLASMA, Q >5 WITH IGNITION AS A GOAL. 
~ 
~ 

... 
i I 
t: 
;:: • BURN PULSE= 60 SEC. 
• DUTY CYCLE = 75% 
< 
...... 
...... 
...... 
I 
..... 
...... 
• (3=6% 
• 
PHYSICAL Sl ZE: R = 4.8m, a = 1.2m/1.9m 

• 
REACTOR RELEVANT TECHNOLOGIES' 

• 
PROVIDES FACILITY FOR TESTING BLANKET MODULES 

• 
DEMONSTRATES ELECTRICITY GENERATION 


\ '' .. ' •• l •• \' \ 
500 MWo TANDEM MIRROR REACTOR 
BLANKET/SHIELD MODULES WITH PLUG NEUTRAL BEA. SOLENOID COILS 
BARRIER PUMP NEUTRAL BEAMS 
BARRIER COIL TRANSITION COIL PLUG YIN-YANG COILS CIRCULARIZING COIL DIRECT CONVERTER