Studies of deuterium-fueled Tokamak reactors

NUCLEAR

INSTRUMENTS

AND

METHODS

144

(1977)

9-16

;

©

NORTH-ttOLLAND

PUBLISHING

CO.

STUDIES OF DEUTERIUM-FUELED TOKAMAK REACTORS GEORGE H. MILEY, FINIS H. S O U T H W O R T H , G L E N N GERDIN and C H A N CHOI

Fusion Studies Laboratoo, Nuclear l:'nt¢ineering Program. Utliversity q/ Illinois, 15rbana, Illinois 61801, U.S.A. The importance of" using non D - T fusion fuels is briefly reviewed. D - D operation where product 3He and T are burned at the same rate as produced (called "'catalyzed-D"t is identified as the only advanced fuel that avoids breeding of fuels and yet potentially offers operation under conditions achievable by present magnetic confinement devices. The conceptual design of a l o k a m a k reactor using catalyzed-D indicates a relatively large reactor / -- 12 G W gross output, 12 m radius torus), but this could still be quite attractive due to potential improvements in efficiency, reduced radioactivity inventory, and increased first-wall lifetime. While a n u m b e r of uncertainties remain, the feasibility of such a reactor does not appear to be too great an extension beyond D - T reactors. Although a fuel such as p-liB could offer further advantages by essentially eliminating neutrons and tritium, suitable methods of confining and burning it require a m u c h larger extrapolation of current experience.

1. Introduction It is generally conceded that first generation fusion reactors will employ deuterium-tritium (D-T) fuel because of the lower plasma temperature (or beam energy) required to burn it compared to other fuels. Actually, if elements through boron are included, some 20-odd reactions have been observed to have non-negligible fusion cross-sections in the energy range q 1 MeV 1,2). Six of the more prominent reactions are listed in table 1, while fusion cross sections are summarized in ref. 3. Reactions involving elements no heavier than helium, along with D-T, are labeled as "prime reactions" since fusion requirements remain relatively mod-

est. Reactions involving heavier elements such as lithium and boron are considerably more difficult to achieve under conditions of interest for practical fusion reactors (i.e. with a net energy release) and hence they are labeled as exotic fuels. All fuels beyond D-T, including the exotic reactions, are generally categorized as "advanced" fuels. The present study was initiated to see if "'conventional" magnetic confinement could be used to burn advanced fuels. The exotic fuels have very attractive features, e.g. p-~B offers virtual elimination of neutron emission. However, the combination of the high temperature required and the large magnetic confining field results in such

TABLE 1

Some important fusion reactions. T h e fusion energy released (Qf), the fraction of the energy carried by charged particles (.lc), and the approximate ignition temperature (Tig) are indicated. Qf(MeV)

,lc

17.6

0.2

-

3.6

0.67

r_ 35

Tig (keV)

A'ear term DT D÷T-n+2

Advanced

4

n ÷ 3He D-D

D+ D "* p+T

Prime

D-3He

D + 3He ~ ~z+ p

18.3

1.0

- 28

(at D

( D - D plus in situ burn of T and 3He) 6D-2~+ 2n+2p

21.6

0.62

-25

p-6Li

p

0:~ 3He

< 5.0

- 0.9

8.7

1.0

none

+ 6Li Exotic

D-6Li p-liB

D p + 7El + ... p + I I B ~ 3~

% 150a

a A s s u m e s that Te - { T i and relatively optimistic cross sections (see p. 50, ref. 2). All other Tis a s s u m e Te - Ti. I. A D V A N C E D

FUEL

FUSION

FEASIBILI]'I'

STUDIES

10

(i

II

MII. I:'~ ct al.

heavy cyclotron radiation losses that it does not seem to be possible to use such fuels with presently conceived magnetic systems ~. Inertial confinement would avoid some of these problems at the expense of incurring other difficulties, but this approach was not considered in the present study. Consequently, we have focused on deuterium fueled reactors, and for reasons described in later sections, have examined a Tokamak design in some detail. 2. Potential advantages and difficulties of advanced-fuel fusion reactors An advanced-fuel reactor offers four potential advantages that are extremely important for future energy sources, namely: l) The larger fraction of fusion energy entering charged particles (see table 1) makes it practical to consider direct energy conversion * to improve the overall efficiency and reduce thermal waste. 2) C o n t a i n m e n t and handling problems associated with radioactive materials are alleviated by the reduction (or in some cases elimination) of trit i u m combined with the reduced neutron flux. Hence, induced radioactivity is lower. 3) The reduced 14 MeV neutron flux (compared to D - T systems) would alleviate radiation damage to the first wall and blanket, extending the lifetime of this structure and reducing maintenance costs . 4) Several of the advanced fuel cycles (e.g. D - D and p-l~B) c o n s u m e fuels that are essentially inexhaustible in supply. In contrast, D - T uses lithium to breed tritium. While lithium resources are adequate to meet the needs of several generations of fusion reactorsT), the supply is quite limited compared to deuterium or boron. While the potential advantages of advanced fuels are large, there are significant obstacles to their * The most promising approach would be a " ' t w o - c o m p o n e n t ' " reactor 2) using proton injection into a relatively cold lIB target plasma, ttowever, preliminary calculations indicate that the energy gain would be too low for a practical reactorS). 1" As discussed in ref. 2, several methods such as direct collection and expansion-cycles appear possible• In contrast, with D - T fuel the large energy associated with neutrons must be processed by a thermal cycle. As pointed out by Steiner and Fraas4), tritium containment in D - T systems is not at all an insurmountable problem. However, the added cost for fool-proof containment and handling m u s t be considered. • This becomes crucial when it is considered that some conceptual D ] reactor designs 5) require s h u t d o w n as frequentb as every !-2 years to replace massi',e blanket parts.

use. For magnetic confinement ~, the oroblems are best summarizea through fig. 1 where the power density achievable with various fuels is plotted as a function of plasma temperature T s). This graph assumes that all systems use the same m a x i m u m magnetic field Bin; thus, pressure balance requires that n T ~s fiB 2, ~ constant, ( I) where n is the plasma density and/3 is the plasma beta-value which defines the trapped magnetic field (assumed constant from case to case here). Then, since the power density P is proportional to n2(av), where
I--

i

,

,

,

D

o_

,

,

,

ITIO

D-D

_ 0001 ~"

,

P 6LI

d o: 0000~

/

/ 2

3

1, , 4 5 ? IO TEMPERATURE,

/ IGNITION L , 20 30 4 0 5 0 70 keY

IOO

Fig. 1. Relative power density vs temperature for various fuels a s s u m i n g a m a x i m u m pressure, i.e., magnet limited system• (From Mills, ref, 8.)

DEUTERIUM

t UELED

provide sufficient cost savings to offset powerdensity considerations; and, the added freedom in site selection due to reduced hazards could translate into significant savings in transmission costs. The effect of these factors will become more apparent in the discussion of the catalyzed-D Tokamak design presented later.

3. Reactor considerations

RI:A(1ORS

TOKAMAK

]1

TABLE 2 Typical values of the radiation parameter qJR. Fuel

Approximate range of ~//R

D-T dithium blanket) Catalyzed D D-3lle

0.08 to 0.16 0.31 to 0.62 0.48 to 0.97

P-°Liexotic fuels

0.50 to 1.0

A basic quantity of interest for reactor operation is the plasma Qp-value defined as fusion energy released Q~ -= energy input via injection, etc."

(3)

It is easily shown 2) that the overall efficiency tl0 for a reactor with thermal converter efficiency t/th, direct conversion efficiency q'DC and injector efficiency r/t is*: ,to = (J -,kR)

I t(~_ '/DC).

,7,~ + ¢. ,l~.: - Q,~

(4)

Here the type of fuel enters through the radiation parameter ~R defined as

~/.', ~ (1-Z.) f~, where./~ is the fraction of the fusion energy going into charged particles and ZR gives the fraction of the latter energy escaping as radiation. Approximate values of ~R are given in table 2 while more complete plots of ~'k are given on pp. 49-50 of ref. 2. To illustrate this result, a plot of q0 vs Qp is shown in fig. 2 for near the maximum that might ultimately be anticipated. For catalyzed-D fuel (see table 1), the maximum value of ~,~ is ~0.62 assuming no radiation losses. A more realistic value is ~0.3, and using this with fig. 2 we conclude the Qp5 1.0 is required to maintain a plant efficiency ~ 50%. However, in this limit, considerable output energy must be recirculated and reinjected into the plasmat. The equipment required for recirculation is expensive. Thus, there is considerable motivation to require that O0~10 to minimize these costs. With this background the key question is What Q~ can be obtained with a given reactor? Since Qp is proportional to the containment parameter nrL (plasma density times energy confinement time), * The prime notation indicates a correction to the unit efficiency for recovery of rejected energy by a thermal bottoming cycle2). t The recirculation fraction is simply (Jl0 Ill Qp+ l) -I.

this question deals with the central question of present experiments; namely, how to scale nz~: to the reactor regime. This remains highly debatable, but some estimates based on simple scaling models developed in ref. 2 are shown in fig. 3 for the three main magnetic confinement devices under study in the U.S.: the Tokamak, thetapinch, and mirror. It must be stressed that these estimates are highly uncertain since various assumptions were made, e.g. the Tokamak results assume relatively low impurity levels, long pulses, and plasma fi-values -0.3. At any rate, based on these resuits, it appears that the Tokamak is the most likely method of obtaining Q~, values near 10 with advanced fuels. This is ironic in that the mirror appears better suited to high energy operation, but the scattering losses are prohibitive, lowering Q~ to a marginal level. This might be overcome with new techniques, e.g. field reversalS), but a ground rule for present purposes is to employ conventional approaches. Similarly, improvements in theta-pinch operation are conceivable, but involve a departure from the current LASL approach. Consequently, the Tokamak was chosen for further study here. i

i

i

,

i1,,

I

r

i

i

,

, ,IT

F

i |

~th--05 D"~ I.O

-'T]Dc

% )

: 09

I0

,z° °.8- ~t ~:r=l.O >-" o.s: 0.82 ~ ' ~ " - - ~

o.6o.~

~

~

~ 0.4-

~

~

-" O02!

i

1

}

% -I%1

/

O0~ 02

05

I0

qp

5

IO

J

" ~50

Fig. 2. Overall efficiency vs the plasma ()p-,,aluc fl~r ',arious levels o f radiation emission indicated by ~//R. ,",1)\ \ N (

I D

I [ I l

t I SI(IN

I t , \ > , I H I I I I '~ :'-;I t D I t /',

12

o,

[ /

D-T / .-- CATD-D/

~

[

,¢,

10O

/ .......

q:2

.

_loLeve /',// / i/•

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y

1 IOI

,

;)¢/*

4/

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/'MIRR

, I/,,

...... !02

I03

ION TEMR OR AVG. ENERGY,keV Fig. 3. Approximate ranges of Qp-values attainable with the m a i n types of magnetically-confined fusion reactions currently u n d e r study. Operation with three different fuels is indicated. Parameters indicated are: Tokamak, q - safety factor, A = a s p e c t ratio; 0-pinch, #v- m a x i m u m c o m p r e s s e d field, r B - b u r n time; mirror, R m = effective mirror ratio.

4. Catalyzed-D Tokamak reactor The principle objective of this study is to carry through a conceptual design of a catalyzed Tokamak in sufficient detail to allow a comparison with equivalent D-T designs. In this way both the feasibility and relative merits of such a reactor can be evaluated. Because the Tokamak is a relatively low-# reactor, large cyclotron losses can result with the hightemperature plasmas required for advanced fuels~°). A non-circular Tokamak was selected in the present study in an effort to maximize /3, hence reduce cyclotron losses< In addition, a catalyzed-D fuel cycle was selected since, as seen from fig. 1, it offers the lowest ignition temperature and the largest power density of the advanced fuels. As indicated in table 1, catalyzed-D operation basically involves D-D fusion where the product 3He and tritium are burned in situ at the same rate they are produced, i.e. the reactions involved are : 3He+n, D+D ~*" T+p,

D D

HD

4DT

?131,e

.

4<~v

nD

"

DD >D3He '

where nj is the density of species j and (av)# is the average fusion reaction parameter for ]-i fusion. These relations vary with plasma temperature T since the <~v) are functions of T. For temperatures of interest, here ( T - 4 0 - 5 0 keV), it is found that nT/nD~O.01 while n3HJno~O.1. The tritium density is small since it burns readily; however 3He densities must be larger to force this slower reaction. The tritium concentration is near that obtained naturally by confinement of the thermalized product. Much of the 3He escapes without burning, however, and it must be separated and reinjected into the plasma to maintain the desired concentration. As discussed in connection with fig. 1, in order for the catalyzed-D to compete with D-T systems from an economics point of view, a maximum wall loading must be achieved. Preliminary estimates ~) of the heat load on the first wall of the T h e s e relations a s s u m e that fusion reaction as the products slow down in the plasma can be neglected compared to subs e q u e n t reactions of the thermalized products. This relation a s s u m e s steady-state operation. A small inventory of tritium and 3He would be required for start-up.

°rtlI IC[ L2

175 25

i

3

R:16

55 15

\\

O~ t 'O4 144 204)~2-8~ ~

,2~~-

8,5

ENCE~ -~-- ~DESIGN

)/// . . . .

~-\~75

I LOAD LINE o

c[ R=I8

-~ n + ~ x .

* To a first approximation, cyclotron losses vary as (1

IIT

io54o4

,

5

4O

3He+D ~ ~+p, T+D

et al.

The net result is the overall reaction given in table 1. The burning of the products is important since these reactions have larger energy yields than the D-D reaction, thus "catalyzing" the reaction in the sense of lowering the ignition temperaturel°). To equalize the production-burn rates of tritium and 3He, it is necessary to maintain their concentration such that t

\ A:5 -
/

O 3He L Economic Level

H, MILE'~

#)/#

d u e to exclusion of the magnetic field from t h e plasma (see p. 386, ref, 2).

ot OI ot Ot ol 16 14 12 I0 8 'ilO 120 pw( M~/m2 ) ]' PTN{GW)

130

140 150 BMAx(KG}

160

170

5

Fig. 4. Relation a m o n g wall loading, reactor power, and magnetic fields. Here B-r is the toroidal field on plasma center-line while BMAx is the field at the surface of the toroidal field coils.

DEUTERIUM-FUELED

catalyzed-D Tokamak indicate that use of a graphite blanket would permit a gross wall loading* of about 2 M W / m 2. This wall loading is chosen, therefore, as the target value. The resulting design should be near optimum in the sense of minimizing the size per unit power t. Based on a Brookhaven studyl2), the graphite blanket approach was selected for the present design in order to capitalize on the potential for a minimum induced radioactivity and achieve high temperature operation. The impact of the wall loading on the size of the catalyzed-D reactor is illustrated in fig. 4. The relations in this figure have been calculated t3'~4) by assuming a plasma-wall separation of 1 m, a blanket and shield thickness of 1.5 m, an aspect ratio of 2.5, a safety factor (q) of 3, a plasma height-towidth ratio (K) of 3, and a plasma temperature of 50 keV. The plasma /3 limit is that prescribed in ref. 15. If conventional NbTi superconducting coils are assumed, the maximum field (Bmax) o n the coil surface will be limited to - 1 1 0 kG by stress considerations. Then, from fig. 4, we see that even for a major radius (R) of 18 m and gross fusion power (PTH) of 14GW, a wall loading of only 0.95 M W / m 2 is obtained. Larger wall loadings require increased R and PTH, but the idea of such a large power plant seems undesirable. The only alternative is to postulate that larger magnetic fields are possible. This may not be too unrealistic since Nb3Sn superconductors could stand up to 170 kG at the coil, and the development of such magnets should not lag far behind the NbTi technology. Then, for example, fig. 4 indicates that a wall loading of 2 M W / m 2 can be obtained at 155 kG with R = 14 m, giving PTH = 17GW. The size is minimized by going to the limit of 170 kG (giving R = 1 2 m , PTH=12GW), and this is selected as the "reference" design. For the blanket-shield dimensions indicated earlier, as seen from fig. 4, this corresponds to a center-line plasma field (BT) of 67 kG. Some feeling for the scale of the reactor can be obtained from the rough cross-section sketch shown in fig. 5. It is envisioned that the elliptical

TOKAMAK

13

REACTORS

plasma will be surrounded by a rectangular graphite blanket, the shape of the blanket being dictated by mechanical considerations and the need for easy removal of modules for maintenanceS2). The large D-shaped superconducting toroidal coils are designed to allow ample space for auxiliary heating and stabilizing coils, diverter coils and maintenance access. The energy confinement time required for ignited operation* (%) of this size is compared to that predicted by turbulent plasma theory ~6) (rTl) in fig. 6. A temperature region below - 50 keV but above the minimum ignition temperature t of 35 keV exists where %
__.8

24

Nb3Sn MAGNET

" GRAPHITE BLANKET * Gross wall loading is defined as the total fusion power div-

ided by the total wall area. Note that the design goal of 2 M W / m 2 is comparable with limits for D - T systems due to radiation damage. If, in that case, heat loadings alone were limiting, values over 20 M W / m 2 might be used, putting catalyzed-D at a large disadvantage.

0

,

L 4

i

i 8

i 12

i

16 . .2.0. . R(m)

24

i 28

I

3'2

i

i

36

I

• 410

Fig. 5. Conceptual cross-section o f the plasma, blanket, and field coil.

1. A I ) \ A N ( E I )

FUEL

FUSION

FEASIBILITY

STUDIES

14

o

tl. MII. t-:'~ et al.

The selection of an o p t i m u m temperature represents a compromise among several factors. In the present case, stress was placed on potential coupling to a direct collector. Since the collection efficiency increases with temperature, higher values are desirable. On the other hand, as seen from table 3, the power retained by the leaking plasma begins to decrease for T ~ 4 5 keV due to increasing radiation losses. Fig. 1 indicates that the m a x i m u m power density occurs at - 3 5 keV, but since the variation is small between 30 and 50 keV, we have opted for T - 4 5 keV in the reference design. Under these conditions, radiation losses reduce the leakage plasma power from the ideal value of 62% (table 1) to about 25%. This is still well above that obtainable with D - T ( - 2 0 ? , , maximum), and is sufficient to make direct conversion potentially attractive. One approach would be to couple an electrostatic direct collector to the plas... ma through a divertor. [ln fact, a detailed study of the possibility of coupling to the direct collection unit by using a bundle divertor 17) is in progress.] At an average charged particle energy of 45 keV, a direct conversion efficiency of about 55% appears to be feasiblelS). The power conversion system might then use a thermal bottoming cycle with an efficiency of about 30%. This would yield an overall conversion efficiency of about 70% for the leaking charged-particle power.

60

T

i

~

i

BT=65KO PTH: I25GW

50

\

~

A=25 /--~-T I

K: 5 0

40

X

q=30 ZI=6

F 30

~

fH:L°°/° tw: 20xlO.9.-rn

2o OPERAT,.G

Io

%0

}"

315

RANGE

I 40

'

t 45

1 50

55

Ti (keY) Fig. 6. Confinement time eslimates for lhe reference catalyzedD reactor.

TABLE 3 Effect of plasma temperature on key parameters a. T(keVt

35

45

55

R(m) Pw (MW/m2) Ymax (KG) PLP (GW) PR (GW)

13.2 1.9 159 2.9 4.5

13.5 1.8 158 3.8 3.3

14.3 1.6 155 3.4 3.2

a Assumes the following parameters are fixed: PTh--12.5GW; A 3;/',:- 3 ; q 3 ; B x 7 5 k G ; f l - 0.12;0.19o of Z 6 i m p u rity; 2>, 10-7#2.m first wall resistivity. Here Pw is the gross wall loading. PLP is leakage plasma power, and PR is the total radiation power escaping the plasma.

Another advantage in efficiency may occur through the removal of tritium breeding requirements. This allows the reactor blanket to be designed purely from heat transfer considerations. With the reference graphite blanket and a helium coolant, it has been estimated that a gas turbine cycle with a 45% efficiency might be feasible~). Combining this efficiency with the previous charged-particle power conversion efficiency, an overall efficiency of about 52% is obtained. Although other parasitic power losses will reduce this somewhat, this still represents a quite attractive efficiency. Another important potential advantage of the catalyzed-D reactor is a longer first wall lifetime, i,e. reduced maintenance costs. The constraint on wall loading appears to be the heat load due to radiation, i.e., bremsstrahlung and cyclotron radiation. The neutron wall loading is substantially smaller than for a D - T reactor at the same gross wall loading; e.g. at P~ = 2 M W / m 2, the neutron wall loading is - I . 6 M W / m 2 of 14.1MeV neutrons for a D - T reactor vs 0 . 6 6 M W / m 2 of 14.1 MeV neutrons and 0.12 M W / m 2 of 2.45 MeV neutrons for the catalyzed-D reactor. Neutron damage effects may, therefore, be substantially alleviated, although a quantitative evaluation requires a better understanding of the relative damage due to a 2 . 4 5 M e V neutron compared to 14.1 MeV neutron. The reactor power of 12.5 G W seems large but it should be noted that 2-unit nuclear power stations are already being constructed with thermal powers of 6-7 GW. Hence, 12.5 G W for a reactor system which will not be built for several decades does not seem unreasonable. In addition, since the

D E U ' I F R I t NI-t-I;ELEI) I O K A M A K

reactor efficiency promises to be substantially higher than for current plants, despite the larger size, the thermal pollution impact should not impose overly restrictive siting limitations. Still, in order to tie in with large-scale power grids, siting would be restricted to appropriate network centers. Some additional flexibility might be gained, however, by extracting some of the energy in nonelectrical form. For example, the shortage of natural gas and other fossil fuels will presumably force the utilization of synthetic fuels such as hydrogen or methane in transportation and other sectors of the energy economy. The possibility of using electricity, radiation, and/or heat from fusion reactors to produce hydrogen or other fuels has been suggested by several authors'). With the relatively large cyclotron output from the present reactor, it is tempting to consider uses for this radiation. Indeed, Eastlund and Levine 2°) have noted the possible use of 120-300 GHz emission to produce methane by irradiation of CO2 and H2 mixtures. This reaction is unique due to the high selectivity for methane production and they cite experiments that suggest yields of product C H 4 and CO exceeding 90%. Use of a significant fraction of the reactor power for synthetic fuel generation would reduce the impact posed by introducing such a large single electrical generating unit on a power grid. The dual output could also serve to maintain the reactor power level by switching off peak load powers for synthetic fuel production:~). Another use of the catalyzed-D reactor might be to breed 3He for use in "'satellite" D-3He reactors. The latter is attractive since it has potential of being substantially free of neutron production ~ and could possibly be developed into small reactors for use at urban sites. Some form of breeding of 3He appears to be necessary, however, due to the small natural isotopic abundance of 3He (the fraction of 3He in natural helium is - 1 0 7). Two breeding processes are possible. One approach would be to run the system somewhat lean in 3He such that some of the product 3He could be collected and shipped to the D-3He reactor. The other approach would involve breeding tritium in the blanket using neutron-lithium reactions as envisioned for D-T reactors. Tritium would then be stored and allowed to decay (12y half-life) to form 3He. While the latter approach would allow maximum Side D-D and D-T reactions could be reduced using a twocomponent type reactor, i.e. injection of high energy deuterium beams into a cold 3He plasma2.5).

REA(_IORS

15

production rates, it would lose several advantages of the non-breeding reactor; namely, the blanket would be restricted to a lithium bearing design and larger tritium inventories would be introduced. In summary, the catalyzed-D Tokamak appears to capitalize on most of the advantages of advanced fuels including elimination of tritium breeding, reduced radioactivity, and increased efficiency. In addition, the technology required only represents a modest extrapolation beyond D-T systems whereas use of other fuels, such as p-~B, would necessitate a radical departure from current approaches. While the conceptual design described here contains a number of uncertainties (e.g. the plasma regime, hence confinement time at the high temperatures required, the details of cyclotron transport when spatial effects are included, the degree of accumulation of impurities, etc.), there appears to be sufficient margin for error to give confidence that ignited operation is feasible. A number of other problems have not been addressed here [e.g. how to reinject the 3He such that it reaches the central core of the plasma; how to start-up without excessive power requirements22); etc.], but none of these problems appear insolvable. Perhaps the most crucial question involves the economics of such large reactors. This question cannot be fully answered without a much broader study that includes consideration of the integration with distribution networks. Again, some flexibility might be afforded through dual purpose plants that offer alternate outputs such as synthetic fuels. The Tokamak study described is part of a joint project with the Brookhaven National Laboratory and the Lawrence Livermore Laboratory funded through the Electric Power Research Institute. References 1) J. R. McNally, Jr., ORNL-TM-3233 (Rev.), Oak Ridge National Laboratory, Oak Ridge, TN (1971). 2) G. H. Miley, Fusion em,;:~ com,crsion (Am. Nucl. Soe., tlinsdale, IL, 1976). 3) G. H. Miley and H. Towner, Proc. APS Conf. on Nuclear cross sections and technology. Washington, D.C., March 1975, CONF-750303 (National Bureau of Standards, Washington, DC, 1975). 4) D. Steiner and A. P. Fraas, Nucl. Safety 13 (1972) 353. 51 C. Bathke, H. Towner and G. H. Miley, Trans. Am. Nucl. Soc. 17 (1973) 41. ~) B. Badger et al., UWFDM-68, vols. 1 and 2, Univ. o1 Wisconsin, Madison, W1 (March 1974 and May 1975). 1. A D V A N C E D FUEL FUSION [EASIBILIT'~

STUDIES

16

•.

H. MILEY et al.

7) j. p. Holdren, UCID-15953, Lawrence Livermore Laboratory, Livermore, CA (1971); also see appendix B, ref. 2. 8) R. G. Mills, TM-259, Princeton Plasma Physics Laboratory, Princeton, NJ (1971). 9) W. C. Condit et al., UCRL-52008, Lawrence Livermore Laboratory, Livermore, CA (1976). to) T. Chu and G. H. Miley, Proc. Syrup. on Technology ol controlled thermonuclear ,lusion experiments and the engineering aspects o].¢ttsion reactors, University of Texas, Austin, TX, November (1972) AEC Syrup. Ser. 31, CONF.7211, National Technical Information Service, Springfield, VA (1974). 11) j. Usher and J. R. Powell, Trans. Am. Nucl. Soc. 23 (1976) 31. t2) j. A. Fillo and J, R. Powell, BNL-21158, Brookhaven National Laboratory, Upton, NY (1976). 13) F. H. Southworth and G. H. Miley, Trans. Am. Nucl. Soc. 23 (1976) 53. 14) F. H. Southworth and G. M. Miley, Proc. 9th Syrup. on

Fusion technology, Garching, W. Germany (1976). 15) j. p. Friedberg and W. Grossman, Phys. Fluids 18 (1975) 1494. t6) S. O. Dean et al., WASH-1295, U.S. Govt. Printing Office, Washington, DC (,1975). 17) G. Miley, F. Southworth, G. Gerdin and C. Choi, Progress Letter for 1 Jan-31 Mar. 1976, EPRI Project //PR645, University of Illinois, Urbana, IL. 18) R. W. Moir, W. L. Barr and R. W. Werner, IEEE Int. Conf. on Plasma science. 76CH1083-5-NPS (1976) p. 55. 19) T. Petrie, A. B. Chilton and G. H. Miley, Trans. Am. Nucl. Soc. 16 (1973) 11. 2o) B. J. Eastlund and L. Levine, Phase 1 Status Report, EPRI Project RP 471-1, Electric Power Research Institute, Palo Alto, CA (1975). 21) j. Powell et al., BNL-18430, Brookhaven National Laboratory, Upton, NY (1973). 22) G. Gerdin et al., IEEE Int. Conf. on Plasma Science, Austin, TX (May 1976) IEEE, Piscataway, N J, p. 53.