Recent directions in plasma physics and its impact on tokamak magnetic fusion design

Fusion Engineering and Design 16 (1991) 253-270 North-Holland

253

Recent directions in plasma physics and its impact on tokamak magnetic fusion design M. K i k u c h i u R . W . C o n n a, F. N a j m a b a d i a a n d Y. Seki b a Institute of Plasma and Fusion Research, University of California, Los Angeles, CA90024, U.S.A. t, Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-01, Japan

Steady progress has been made in recent years towards achieving fusion breakeven and reactor-relevant plasma conditions in the largest tokamak experiments. In particular, the experimental observation of a large self-sustaining bootstrap current in the plasma permits development of steady-state reactor concepts with modest current drive power and relatively high plasma energy gain (Q). The SSTR (Japan) and ARIES-I (USA) designs reported in this paper are first-stability, steady-state tokamak reactors based on "modest" extrapolations from the present tokamak physics database. The plasma is characterized by a high edge safety factor (qa) and a high poloidal beta (tip) in order to increase bootstrap current fraction ( - 70%). This mode of operation is achieved by selecting high values of both aspect ratio (A = R / a = 4-4.5) and toroidal magnetic field on axis (9-11 T) in these reactors. Both reactor studies suggest that the tokamak system can be a steady-state power reactor with net electricity of - 1 GW and with plant efficiency 30-40%. The SSTR is characterized by its technical feasibility in the near future. On the other hand, the ARIES-I focuses on better safety and environmental aspects and a longer time frame. The SSTR and ARIES-I studies show that, with proper R&D programs, tokamak fusion reactors can be developed that will have acceptable cost of electricity.

1. Introduction During the last few years, continuous experimental and theoretical effort has resulted in progress in tokamak plasma physics. The JET tokamak has achieved near breakeven conditions [1]. The T F T R tokamak [2] shows the existence of the bootstrap current and the JT-60 tokamak [3] has attained a bootstrap current fraction up to 80%. Efficient current drive with the lower hybrid wave has also been demonstrated in the JT-60 tokamak [4]. The D I I I - D tokamak [5] has demonstrated the Troyon factor g up to 5 (transiently) at high values of edge safety factor qa. Also, in DIII-D, longpulse discharges are achieved with g = 3.5. In parallel with these achievements in tokamak research, significant efforts are being made on the conceptual design of the International Thermonuclear Experimental Reactor (ITER) as a next step device towards the production of a fusion energy [6]. Recently, two independent studies have pursued conceptual designs for tokamak power reactors based on "modest" extrapolations of tokamak plasma physics from our present data base. The SSTR design [7] has been developed at JAERI in Japan while the ARIES-I study [8] is a US multi-institutional effort led by UCLA.

These reactors have m a n y different aspects compared with those from former comprehensive pulsed or steady-state tokamak reactor design studies, such as U W M A K [9] and S T A R F I R E [10]. These new studies are particularly important in indicating directions for plasma research and for the next step design and engineering R & D efforts. Cross-sectional views of the SSTR and the ARIES-I power reactors are shown in figs. 1 and 2, respectively. In this paper, recent advances in plasma physics are summarized in section 2 and their implications for the ARIES-I and SSTR are discussed. The technical aspects of the two reactors are described in sections 3 and 4 and the implications for fusion research are discussed in section 5. Concluding remarks are given in section 6.

2. Recent implications of the plasma physics 2.1. Steady-state operation and current drive

Steady-state operation is one of the most important issues for the tokamak system. Recent experimental confirmation of a large bootstrap current in the tokamak [3] gives us an opportunity to achieve high-Q

0 9 2 0 - 3 7 9 6 / 9 1 / $ 0 3 . 5 0 © 1991 - Elsevier Science P u b l i s h e r s B.V. A l l rights r e s e r v e d

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M. Kikuchi et a L / Plasma physics and its impact on tokamak design

steady-state operation in a fusion reactor. Operation at high edge safety factor q ( = 4-5) and high poloidal beta fir, ( - 2 ) is required to increase the bootstrap current fraction fb- As a rough estimate, the bootstrap current fraction fb is given by [11], -6- tip) 13, fh = 0"75(~/~-

(1)

where tip is given by 4

Bp - .0

Rp

fP dV.

(2)

High bootstrap fraction is obtained with the increase of total plasma pressure (hence, high fusion output power) and the decrease of the plasma current (A = 4.1, Ip = 12 MA, fb = 0.75 for 3 G W SSTR and A = 4.5, Ip = 10 MA, fb = 0.68 for 1.9 G W ARIES-I). The current profile control is especially important for such high bootstrap current reactors because the bootstrap current is determined by the uncontrollable

Solenoid

pressure profile. The bootstrap current is a hollow current and active current drive, especially at plasma center, is required to obtain an M H D stable current profile. Fast-wave current drive (141 MHz, 100 MW) has been adopted in the ARIES-I due to its low cost. The energy multiplication factor Q is 19 for the total fusion output of 1.9 GW. High-energy NB current drive (2 MeV, 60 MW) has been adopted in the SSTR due to its high current drive efficiency. The energy multiplication factor Q becomes 50 for the total fusion output of 3 GW. 2.2. Plasma beta and MHD stabili(v

The beta limit is an important operational constraint for the tokamak. The scaling law of the beta limit is given by Troyon [12] as follows, Ip(MA) (fit)(%) = g ap(m) Bt(T) .

(3)

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Neutron

~Jectlon Ports =OMW)

)r ence

Ports

inboar¢ Permel ~rts

5(m) =

Cryostat

Fig. 1. Bird's eye view of the SSTR Fusion Power Reactor.

~

5

(m)

M. Kikuchi et aL / Plasma physics and its impact on tokamak design

255

Fig. 2. Bird's view of the ARIES-I Fusion Power Reactor.

3

'

\

'

r'l.8 g=3.5

'

Combining eq. (3) with eq. (2) gives

'

- 4"

ARIES L \ss'r.

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2

t o

ITER

0

I

I

1

2

I

I

3 4
I

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5

6

7

Fig. 3. ((Bt), tip) diagram and the operation points of ITER, SSTR and ARIES-I. Solid line shows g = 3.5 contour for the x = 1.8 case.

(4)

This relation is shown in fig. 3 with the design points for ITER, ARIES-I and SSTR. The steady-state reactors will be operated at high q and high tip in contrast with experimental reactors. The choice of the Troyon factor g - 3.5 is important to attain high bootstrap current fraction. Recent results from D I I I - D are extremely encouraging. Enhancement of the Troyon factor g up to 5 is attributed to current profile control and a steady-state g = 3.5 discharge has also been demonstrated [5]. Stability against high-n ballooning modes and low-n kinkballooning modes has been studied for both ARIES-I

256

M. Kikuchi et aL / Plasma physics and its impact on tokamak design

and SSTR. The ARIES-I adopted low q0 ( = 1.3) to achieve low-n MHD stability without expecting the wall stabilization. The Troyon factor for ARIES-I design is 3.2. The SSTR plasma is stable to both modes when the central q (qo) is larger than 2 due to effective wall stabilization (r w = 1.2a) which enables a larger bootstrap current fraction. A moderate toroidal rotation with the beam injection is expected to be enough for this wall stabilization. The Troyon factor is 3.3 for the SSTR design. Stability against vertical plasma motion is important for non-circular tokamaks. The concept of the twin loop is proposed in ITER (A = 2.9, K95 = 2.0) [13] to stabilize the vertical instability for a highly elongated plasma. The same scheme is adopted in the SSTR with reduced elongation K95 = 1.85 because of the higher aspect ratio. The twin loop is located just behind the replaceable blanket. The growth rate is further reduced with the low-resistance vacuum vessel (one turn resistance of 4 ~tfl) in the SSTR to reduce the forcing voltage of the power supply ( - 2 MVA) [14]. In the ARIES-I design, because of concern related to irradiation damage and nuclear heating in metallic loops very close to the plasma, the stabilizer elements are located behind the blanket resulting in a relatively low plasma elongation of K95 1.6. The active control coils are located behind the TF coils and their power requirements is 2 MVA. =

2.3. Energy confinement

Under the constraints of high bootstrap current fraction (say, fb = 0.75 in eq. (1)) and high Troyon factor (say, g = 3.5 in eq. (4)), the machine parameters for a high-Q steady-state reactor can he determined consistent with geometrical and field constraints [15]. The confinement enhancement factors required for high-Q steady-state operation are then obtained for various scaling laws. The scaling law of the energy confinement proposed by Goldston [16] has accurately predicted the energy confinement characteristics of JET, TFTR and DIII-D. The ITER team assessed the database and proposed two L-mode scalings (power and offset linear scalings) using the confinement database from major tokamak devices [17]. The offset linear scalings give fairly optimistic predictions for power reactors. Increase in the plasma current is effective in improving the energy confinement. However, large plasma current is not desirable for high-Q steady-state operation in power reactors. Goldston and ITER power laws ( z F _ p-0.~) predict that the fusion product n ~ T -

i~,~

~-Equl tE contour

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--

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Ip (MA) Fig. 4. (lp, 14) diagram and the design points for JET, JT-60U, FER, ITER, SSTR and ARIES-I. Solid lines show equi Tt. contours for Goldston and TFTR scalings. The broken line shows 75% equi bootstrap fraction contour for 3 GW fusion output under same wall load.

12p(R/a)" (a >_ 2). These scalings are roughly consistent with JT-60 lower X point experiments (A = 4.5) and the T F T R aspect ratio experiments [18]. So it is possible to reduce the plasma current by increasing the aspect ratio without sacrificing the confinement capability. This is the basic design philosophy of both SSTR and ARIES-I. Operation at low plasma current is also favorable to reduce the electromagnetic force during plasma disruptions. But it should be noted that the plasma elongation must be lowered in the high aspect ratio regime, as discussed in section 2.2. Figure 4 shows the (/p, A) diagram and design points for ITER, ARIES-I and SSTR with the equi-fusion-product contours for Goldston and TFTR scaling laws. We can see that both SSTR and ARIES-I have similar confinement capability to ITER88. But it should be noticed that the required plasma current to achieve high-Q confinement is sensitive to the He concentration (file = n H J n e ) . The He concentration of 5% is assumed in the SSTR resulted in the H factor of 2.0 for ITER power law. The He concentration in ARIES-I is 8% which is estimated based on (%/'r E) ..... = 4 and 50% helium exhaust efficiency. The SSTR and the ARIES-I adopted high toroidal magnetic fields (9 and 11 T at the plasma center, respectively) to increase the bootstrap current fraction. High-field and high-temperature operation results in a large synchrotron radiation from the main plasma. The first wall of the SSTR is metallic with high reflectivity and the synchrotron radiation loss is - 15% of plasma heating power. The ARIES-I first wall is made of SiC composites and, therefore, - 5 0 % of plasma heating power is radiated as synchrotron radiation. The choice of the aspect ratio becomes rather arbitrary if we accept higher fusion output. The bootstrap

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

ii~....

current fraction increases with the inverse aspect ratio for the same value of/3p. This point together with the higher elongation in the low aspect ratio regime gives a possibility to get high Q steady state for higher fusion output regime in the low aspect ratio tokamaks (Rp = 6.25 m, A = 2.9, Ip = 17 MA, ~95 = 2.0, /)fusion= 3.9 GW, Q = 50, B,,~ = 16.5 T) with the confinement enhancement factor less than 2.0 for the ITER power law with the He concentration of 5%.

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2.4. Power and particle control

ne-contours --2.0E.21 -I.OE*21

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Fig. 5. Typical two-dimensional divertor code result showing the equi n e contours in the SSTR. Heat flux (solid line) and radiation heat flux (broken line) to the inboard and outboard divertor plates.

Power and particle control are important issues in magnetic fusion research. Even for experimental reactors (plasma heating power < 200 MW), the heat flux to the divertor plate is much higher in some operating scenarios than that achieved in the present high heat flux components. The total plasma heating power of the SSTR and the ARIES-I are 660 and 485 MW, respectively. High remote radiative cooling in the divertor plasma is proposed in the SSTR by using a strong divertor gas puff with powdery Fe injection. The key consideration in this scheme is the prevention of impurity back-flow into the main plasma. Two-dimensional divertor code analysis has been conducted in the SSTR [19]. It is shown that the impurity does not flow back to the main plasma due to strong friction force from the fuel D / T ions. Equi contour plots of the electron density and the heat flux distribution along the divertor plates are shown in fig. 5. Formation of a cold and dense divertor plasma and a heat flux density less than 10 M W / m 2 are found in this simulation. The ARIES-I design also uses high-recycling divertor. Because of the double-null divertor configuration, high synchrotron radiation losses from the core plasma (52% of total plasma heating power), and high edge density, there is no need for gas puffing or impurity injection. The increase of the confinement enhancement factor due to large synchrotron radiation from the core is modest ( H = 2.56 for ITER power law). Two dimensional scrape-off layer plasma transport indicates that the peak heat flux on the divertor plate is 4.5 M W / m 2 and the peak ion temperature is - 25 eV. The thermostructural design of ARIES-I divertor can handle localized peaking of the heat flux by a factor of two above the nominal design value. Monte Carlo neutral transport calculations indicate that sputtering erosion of the first wall and divertor plates is small and that adequate He ash exhaust can be achieved.

258

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

Table 1 Major parameters of the SSTR

/ j

Items

Value

Major radius Minor radius Elongation Triangularity Aspect ratio Plasma volume Plasma current Toroidal field Maximum field Safety factor (q95) Safety factor (qo) Toroidal beta ((/3t)) Poloidal beta (gp) Troyon factor (g) Average density Average temperature Zeff He concentration Bootstrap current Beam driven current NBI power Beam energy Fusion output Energy multiplication Q Total thermal output Max. neutron wall load Gross electric output Net electric power

7.0 m 1.7 m

3. T h e S S T R

1.85

0.4 4.1 730 m~ 12 MA 9.0 T 16.5 T 5.0 2.0 2.5% 2.0 3.3 1.4×102° m 3 17 keV 1.5 5% 9 MA 3 MA 60 MW 2 MeV 3 GW 5O 3710 MW 5 MW/m 2 1280 MWe 1080 MWe

n

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Fig. 6. Side view of the tokamak and NBI system of SSTR.

(2) low-voltage p l a s m a b r e a k d o w n (l~pp = 4 V) using 5 M W E ~ H pre-ionization; (3) a simple shield integrated v a c u u m vessel having low o n e - t u r n resistance (4 ~t~2); (4) two layers of b l a n k e t structure for periodic rep l a c e m e n t of the p l a s m a facing units; (5) functionally g r a d u a t e d material to reduce electrom a g n e t i c force o n the replaceable blankets:

design

3.1. Major parameters of the S S T R Concept study of the SSTR was m a d e in J A E R I as a D E M O reactor [20,21]. M a j o r p a r a m e t e r s of the SSTR power reactor are s h o w n in table 1. Plan a n d side views of the t o k a m a k a n d the Negative N B I system are s h o w n in figs. 6 a n d 7. T h e 2 MeV, 60 M W N N B I system is used to realize a steady state. F o u r units of steam generators, pellet injector, E C H system a n d two m a i n t e n a n c e systems are located in the reactor hall.

Cockcroft Walton 2MeV accelerator

/ ~ ~ [ I / ~ \ \

/ /

~

/[~]~l~l~J

PlasmaNeutralizingCell

3.2. Key features of the S S T R

"~'Steam Generato~ ..... Key features of the SSTR o n the technical front are: (1) a single-null divertor configuration to reduce magnetomotive force a n d the power supply capacity of the poloidal field coils;

NINI~

Fig. 7. Plan view of the tokamak and NBI system of SSTR.

259

M. Kikuchi et al. / Plasma physics and its impact on tokamak design SSS

(6) reduced activation ferritic steel (F82H) for the blanket structural material; (7) M o - R e power generation divertor with active cooling using the strong divertor gas puff; (8) high current density (Jc = 600 A/ram2), 16.5 T (NbTi)3Sn conductor by a Nb tube method in the JN1 advanced disks having good mechanical properties for the toroidal coil; (9) Cockcroft-Walton accelerator for 2 MeV D o beam injector with high system efficiency ( > 50%); (10) shallow pellet fuelling to reduce hot neutral influx to the blanket first wall and to control edge plasma in the H mode; (11) low tritium inventory ( - 5 kg) by reducing the amount of the isotope separation in the fuel cycle; (12) reliable high-temperature pressurized water cooling system for power generation. 3.3. Superconducting magnet design

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i

6 4

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0

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f

f

I

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15 16 17 18 19 20 21 22 23 24 Peak

Coil

Field,Bmax

(T)

Fig. 8. Non-Cu critical current density as a function of the magnetic field for Ti-doped Nb3Sn conductor manufactured by tube method [22]. The operating non-Cu current density is chosen at 300 A/m_rn2.

,

'=l,=l.!-!,z!~,!

~ ~ l . 6 t

The toroidal magnet consists of 16 toroidal field (TF) coils which produces 9 T at the plasma center. The maximum field becomes 16.5 T at the coil. The conductor, whose operating current is 81 kA, is divided into two gradings which are wound with the modified threein-hand. One is a circular (NbTi)3Sn/Cu cable-in-conduit conductor operated at 16.5 T with the internal reinforcement. Another is a circular N b T i / C u cablein-conduit conductor operated at 5 T. In this design, high critical current density 600 A / m m 2 at 16.5 T is

585

0.5t

(NbTi)3Sn(16.$T)

~~0.5t

1.2t

NbTi(ST)

Fig. 9. Inboard cross-section of the TF coil and the structure of conductors for SSTR.

used for a non-Cu critical current density of (NbTi)3Sn strand. This high critical current density has already been realized using the tube method as is shown in fig. 8 [22]. The conductor is wound in the grooves of JN-1 advanced disks by a semi-wind-and-react method. With this winding concept, an excellent mechanical rigidity of the case and a significant reduction of the bending strain of the conductor become possible. Cross-sections of the inboard toroidal coil and the conductors are shown in fig. 9. Stress analysis of the TF coil is conducted by a three-dimensional finite-element code. The maximum equivalent local stress is 850 MPa which could be attainable with the high-strength cryogenic steel such as JNl(oy > 1200 MPa) [23]. The poloidal magnet consists of 14 blocks which produces a single-null poloidal divertor equilibrium. The maximum field at the conductor is 7 T for the center solenoids and 5.7 T for other equilibrium field coils. Moreover, the changing field is below 1 T / s during plasma disruption due to effective magnetic shielding of the low-resistance vacuum vessel. Thus an inexpensive and simple NbTi conductor is adopted.

260

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

The heat loads for the cryogenic system consists of refrigeration load of 20 kW and liquefaction load of 8630 1/h. '

3.4. Shield integrated vacuum vessel [14]

The vacuum vessel and shield are integrated to form a double-thin-walled structure with 50 mm stainless steel (SS) as is shown in fig. 10. The one-turn resistance of the vacuum vessel is chosen at 4 F~2, considering the recent progresses of the ECH pre-ionization experiments. Reinforcement plates are welded in poloidal and toroidal directions, and form the chambers for the electrically insulated shielding units. Inboard vessel consists of 85% SS and 15% water to improve nuclear shielding. The outboard vessel consists of 50% SS and 50% water. This low-resistance vessel together with a thick twin loop behind the replaceable blanket is effective in reducing the vertical instability growth rate and therefore reduces the necessary power supply capacity of the feedback control system.

I

L

Side Module

Inb~°ard Center Module [Guard LimiterJ

Outboard

Side Module

\\

\.

\\ Inboard Side Module

~

"

Outboard Center Module

l~r ~'~, %, i

3.5. Blanket and divertor [24]

/~ Two layers of blanket structure have been adopted in the SSTR. The 20 cm thick blanket is periodically replaced: every two years for the outboard replaceable blankets and every four years for the inboard replaceable blankets. Such a replacement is required both from the maximum life-time neutron fluence (10 M W a / m 2) of the structural materials (low-activation ferritic steel F82H) and from the 6Li burn up in the front breeding

l

Outboard Side Module

Fig. 11. A set of replaceable blanket modules to be replaced periodically from a vertical port. zone. The replaceable blanket in one toroidal section (22.~5o) is divided into six units as is shown in fig. 11. Periodic replacement would be made from the vertical port during the annual maintenance period ( - 100 days).

~ ~

VacuumVessel

[IOutletHeader 1 ~ ~ [ ~ [t~-.. ~ ~ Header~ ~ ' ~ " "

CoolantFlow ..../-CoolantManifold Cq

__ ~_)_~_l~-~Reinforcing Plat

L

~~ ~ a n t l , ,

"'~

Manifold Zoroldalcoil

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Fig.10.Structureof theshieldintegratedvacuumvesselof SSTRshowingthecoolantflow.

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

A replaceable blanket consists of Li20 solid breeder in pebble bed structure and a beryllium neutron multiplier. The electromagnetic force on this thin blanket is a serious issue and the major eddy-current loop is cut by a functionally graduated material (FGM) made of ceramics sandwiched by the ferritic steel. The permanent blanket is attached to the vessel and is expected to last the reactor lifetime of 30 y. The tritium breeding ratio and the blanket energy multiplication factor of the blanket are 1.2 and 1.35, respectively. The first wall and blanket are cooled by pressurized water of 15 MPa and inlet/outlet temperatures of 285 ° C/325 ° C to generate electricity by a PWR power plant. The divertor plate is designed for a maximum heat flux of 6 M W / m 2. The maximum heat flux to be extracted by the pressurized high-temperature water is around 3 M"W//m2 for a smooth tube. So an improvement of the bum-out heat flux to - 2 0 M W / m 2 is necessary to use the heat to the divertor plate for power generation. A swirl tube is used for the short ( - 50 cm) high heat flux zone in the SSTR. 3.6. N B I system

The 2 MeV, 60 MW negative beam based NBI system is adopted for the central current drive. The system consists of two beam-line modules. Each beam line has eight ion sources which can deliver 5 MW of injection power per ion source. A high system efficiency of 50% is realized by adopting an electrostatic accelerator of Cockcroft-Walton type and a plasma neutralizing cell. Design parameters of the NBI system are shown in table 2. In the beam-line design, the ion beam from each ion source is deflected magnetically. This configuration is particularly important to eliminate neutron damage of the ion sources and makes it possible to install many small ion sources to control the current profile and to compensate for an accidental break-down of an ion source. Small elliptic injection ports are enough for the current profile control because a change of the central current profile affects the bootstrap current profile. This aspect is effective in reducing the neutron streaming loss to the NBI ports. 3. 7. Operation scenario and power supplies

Pure inductive current build-up is adopted in the SSTR. The required volt-seconds are determined by the ITER guide line. The plasma is initiated as a limiter plasma and grows up to form a single-null divertor configuration. Figure 12 shows a series of plasma equi-

261

Table 2 Design specifications of SSTR NBI system Beam energy Beam power Beam species Beamlines Source modules System efficiency Acceleration Stripping loss Geometric loss xxxx Reionization loss Power supply, etc. Ion source type Extraction current Current density Extraction grid Filling pressure Glas flow rate Accelerator Beam current Acceleration stages Beam divergence Beam deflector Neutralizer Gas species Ionization Pressure Beam dump Pumping Power supply Ripple Frequency

2 MeV 60 MW (Max. 80 MW) D 2 8/beamline 50% 90% 10% 5% 5% 80% Volume-Surface 5A 40 mA/cm2 12 × 30 cm 0.3 Pa 0.6 Pa m3/s Electrostatic 3.5 A/module 10 3 mrad 60 o bending magnet + Q magnets Plasma cell Xe or Ar 50% 0.03 Pa Externally finned swirl tube 25 m3/s TMP × 38 × 2 Cockcroft-Walton < + 1% 10 kHz

libria during the current buildup phase and the power supply requirement. The capacities of the poloidal power supply are 972 MW of thyristor convertors and 1140 MVA of transformers to which two 3 G J, 300 MVA motor generators and a 200 MVA line grid are connected. This is actually smaller than that for JT-60. Power loss in the power supply is reduced using by-pass pair operation in a steady-state phase. The power supply for the toroidal magnet is only 3.2 MW, 81 kA. But the capacities of the dump resistor and the circuit breaker are fairly large to absorb toroidal stored energy of 135 GJ. 3.8. Power balance and cost of electricity

An overall power balance of the SSTR and the COE comparison between various energy sources are shown

262

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

in fig. 13. Attainment of high-Q ( = 50) steady state enables a reduction in circulating power to 200 M W e and a net electric power output of 1080 MWe. The circulating power consists of 120 M W e for N B I system, 20 MWe for cryogenic system, 50 MWe for cooling system and 10 M W e for the joule loss of the power supply system. The construction cost is estimated and the cost of electricity is estimated to be 1.5 times that of light-water reactor with a service life of 16 y and an availability of 70%. The cost of electricity cculd approach to that of the steam power plants if we consider effective life of 30 y and an availability of 85% [20].

3. 9. Construction and site plan

The area of the S S T R site is 450 × 400 m. There is no construction room in the SSTR taking into account the reliability of the major components of the tokamak and the cost merit as a D E M O reactor. The major components of the tokamak are constructed with the large side crane during the construction phase of the fusion reactor building. Total construction period is estimated to be 5.5 y after the main contract. In the D E M O phase, we can expect high reliability on the superconductors and there is few possibility for the coil failure. Moreover, the shield integrated vacuum vessel

I2MA 12 /

Initial Magnetization 3.75GJ 81MAT 92vs

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0.5MA,lsec - -

(MVA)

--

(MW)

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t(sec)

100

9.5MA,80sec

5MA,50sec

12MA,100sec

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Fig. 12. Start-up scenario with a series of plasma equilibria and time variation of the power supply output.

M. Kikuchi et aL / Plasma physics and its impact on tokamak design

"--11;7;ooo.w I

60MW

1280MWe 1080MWe ,,i P£ PE(net)

, .3710 w

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120MWe 80MWe

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Fig. 13. Power balance of the SSTR and the COE comparison between various energy sources.

proposed in the SSTR has high reliability and the probability of the severe damage is quite low. Thus, we can adopt simple reactor building structure similar to the commercial light-water reactor. So a significant reduction of the volume of the reactor building becomes possible.

4. The ARIES-I design The ARIES-I study is aimed at developing several visions of tokamak power reactors with enhanced economic, safety, and environmental features. The first design, ARIES-I, is a DT burning, 1000 MWe reactor [25,26]. The physics basis of ARIES-I is as much as possible, consistent with existing tokamak experimental data. From the technological viewpoint, ARIES-I makes choices that in several cases, extend beyond present engineering achievement. However, in all cases, the technology and engineering are supported by laboratory data and by industry trends, and it is expected that with proper R& D, these technologies will be available within the next 20 to 30 y and could be utilized in DEMO and power reactors. In a second study, ARIES-II, the team assumes potential advances in plasma physics such as high plasma beta in the second M H D stable regime, that are predicted by theory but are not well established experimentally. The ARIES team is also exploring the

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potential of advanced fuel cycles, specifically D - 3 H e , in the context of the A R I E S - I l l design. A major goal for all ARIES designs is to maximize the environmental and safety attributes of fusion through innovative design and careful selection of lowactivation materials. As such, the ARIES-I blanket is a low-activation design that has a low tritium inventory and minimum induced radioactivity and afterheat. It meets the US limits for low-level (Class C) waste disposal [27]. Simultaneously, the design has a superior nuclear performance and a high coolant exit temperature (650 ° C), leading to a high-efficiency (49% gross) Rankine-cycle power-conversion system. The major parameters of the ARIES-I reactors are given in table 3. The major radius is 6.75 m, the plasma minor radius is 1.5 m, the average neutron wall loading is 2.5 M W / m 2, and the mass power density is about 100 kWe per ton of fusion power core. The design uses moderately high aspect ratio (A = 1/~ =4.5), low plasma current (Ip = 10 MA), and high magnetic field ( - 11 T at the plasma center). Steady-state operation is presumed, based upon ICRF fast-wave current drive to supplement a large (68%), theoretically predicted bootstrap current. Detailed current drive analysis has been performed to ensure that the sum of both the bootstrap and driven current densities match the equilibrium current density profile, as is shown in fig. 14. Impurity control and particle exhaust are based on high-recycling poloidal divertors in a double-null configuration. Self-consistent core and scrape-off-layer plasma calculations predict that with a ratio of a-particle confinement in the core to plasma energy confinement (~a/~E) of 4, the alpha exhaust efficiency is 50% and the He ash concentration is 8%. 4.1. A R I E S - I fusion power core

The principal structural material used throughout the fusion power core (FPC) is a composite of siliconcarbide (SIC) fibers in an SiC matrix. This composite retains many of the desirable features of bulk SiC ceramic while the addition of SiC fibers greatly reduces the brittle nature of the material and produces a high degree of fracture toughness. The increase in toughness creates more freedom in engineering design and allows both tensile and compressive stress in the composite. Desirable features of the SiC composites include low levels of induced activation and afterheat, high-temperature capability, high strength, and extensive resource availability. It should be noted that the composite industry, while advanced in the aerospace industry and other fields, is far from fully mature. Manufacturing

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M. Kikuchi et a L / Plasma physics and its impact on tokamak design

Table 3 Major Parameters of ARIES-I first-stability tokamak reactor (a) Plasma parameters Plasma aspect ratio Major plasma radius (m) Minor plasma radius (m) Plasma vertical elongation (95% flux) Plasma current (MA) Toroidal field on axis (T) Toroidal beta Poloidal beta Electron density (m 3) Plasma temperature (keV) Current drive method Bootstrap current fraction Current drive power (MW) Impurity control system Fueling method (b) Reactor parameters Neutron wall loading ( M W / m 2) Mass power density (kWe/t of FPC) Unit direct cost ($/kWe) Net electric output (MWe) Fusion power (MW) Thermal power (MW) Thermal cycle efficiency Recirculating power fraction Net plant efficiency (c) Fusion power core parameters Coolant Structural material Breeder/Multiplier Blanket energy multiplication ratio Tritium breeding ratio Coolant inlet temperature Coolant outlet temperature Coolant pressure (MPa) Peak surface heating ( M W / m 2) First wall Divertor Pumping power (MW) First wall and blanket Divertor (d) Reactor costs (Million $) Land and land rights Structure and facilities Reactor plant equipment Turbine plant equipment Electric plant equipment Miscellaneous plant equipment Total direct cost Construction Home office engineering Field office engineering Owner's cost Project contingency

4.5 6.75 1.5 1.6 10.2 l 1.3 0.02 2.1 1.5 × 1020 20 ICRF fast wave 0.68 100 Double-null divertors Pellet injection 2.5 100 2140 1000 1925 2544 49% 20% 39% He SiC composite Li 2ZrO3/Be 1.3 1.2 350 o C 650 o C 10

Table 3 (continued) (d) Reactor costs (Million $) Interest during construction Total cost Cost of electricity (mills/kWh)

technology for h i g h - p e r f o r m a n c e , large-scale SiC-composite c o m p o n e n t s must b e developed a n d the costs must be lowered from present levels. A large effort in this direction is already in place outside the fusion p r o g r a m [28-32]. T h e fusion-specific R & D issues include the d e t e r m i n a t i o n of irradiation effect o n properties of SiC composites, the leak tightness of components, a n d the m a n u f a c t u r i n g of reliable j o i n t s between composites a n d metal c o m p o n e n t s . The t r i t i u m - b r e e d i n g material in the blanket is lithium zirconate (Li2ZrO3), a ceramic solid breeder. Lithium zirconate was selected as the reference solid breeder because the d a t a base d e m o n s t r a t e s long-term a n d h i g h - t e m p e r a t u r e stability. Isotopically tailored zirconium is necessary to reduce the level of decay heat a n d improve the waste disposal rating of the blanket. The tritium inventory in the solid breeder is only l g. O t h e r solid breeders such as lithium oxide ( L i 2 0 ) a n d lithium orthosilicate (Li4SiO4) c a n be used as alternative breeder materials. Lack of data a n d concerns a b o u t the effects of irradiation on chemical stability and tritium r e t e n t i o n at high l i t h i u m b u r n - u p levels have precluded the choice of these low-activation, low-afterheat breeder materials. Since the d o m i n a n t safety h a z a r d in the A R I E S - I b l a n k e t is the L i z Z r O 3 breeder (even after extensive isotope tailoring), there is a strong incentive to develop such alternative solid tritium breeders. 0.2

0.5 4.5 19 35 5 340 1363 245 137 50 2140 214 214 214 107 214

5t 3 3617 66

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0.4

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Fig. 14. Components of the current density for a typical equilibrium examined for ARIES-I. The solid line is the required equilibrium current density profile, G(~)-= ( j H B ) / (B2). Schematics of one of the ARIES-I blanket modules. The chain-dashed curve is the bootstrap current contribution and the other curves are the contributions from seven RF current drive rays. The dotted curve is the sum of the bootstrap and the driven current densities and it matches the required G (~).

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

265

COIL SUPERSTRUCTURE

FPC MODULE

Inboard Blanket /

C~

Inboerd

Fig. 15. Schematics of one of the ARIES-I blanket modules. The ARIES-I FPC comprises 16 identical and selfcontained toroidal modules. Each module consists of one toroidal-field coil, two inboard and two outboard blanket and shield submodules, two upper and lower divertor targets, and a section of the vacuum vessel as shown in fig. 15. Each sub-module has 17 nested Ushaped SiC-composite shells. The sphere-pac solid breeder and beryllium-neutron multiplier mixture is

BLANKET COOLANT SHELL

PLENUM

//POLOIDAL PLENA

BREEDER

FLARED SHELL END

Fig. 16. Internal structure of one of the ARIES-I blanket submodules.

ARIES-1 FUSION POWER CORE

Fig. 17. Schematic of steps in replacing one of ARIES-I blanket modules. located in the space between the shells, as is shown in fig. 16. Tritium is recovered by a low-pressure, helium purge gas, flowing slowly between the shells. Sphere-pac pellets of beryllium metal are used as the neutron multiplier in order to boost the tritium-breeding ratio to 1.18 and achieve a high value ( M = 1.3) of blanket-energy multiplication. The development of the beryllium recycling technology is necessary because of the predicted resource limit of this material. The ARIES-I maintenance philosophy is based on high degrees of standardization and automation. Therefore, the ARIES-I FPC comprises 16 large-scale, identical modules. The FPC modules are self-contained and structurally independent of any neighboring component. Each module is replaced as a single unit during both scheduled and unscheduled events, as is schematically presented in fig. 17. Since the TF coils are removed during module replacement, a removable cryostat seal is used which does not require the cutting and joining of welds to remove cryogenic structures, further reducing the time required for maintenance. All modules are pretested prior to installatio m the reactor vault so that undetected defects become :vident and high reliability can be achieved. As a re., It, new modules are always available for immediate installation in the reactor vault. Minimum repair is performed in the reactor vault. Instead, the module containing the failed component is removed and replaced with a new module. After the reactor is back on line, the modules that have been removed can be serviced for later use a n d / o r prepared for recycling and waste disposal.

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M. Kikuchi et al. / Plasma physics and its impact on tokamak design

The FPC is cooled by helium at 10 MPa. The coolant first flows vertically (poloidally) in the plena located at the back of the blanket. The inlet plena feed the first-wall and blanket coolant tubes through which the coolant flows radially inward, then toroidatly across the module, and finally radially outward back into the outlet plena. This routing configuration was selected to provide adequate cooling of the blanket materials and to minimize the first-wall and blanket pressure drop. The total blanket-loop pumping power is 19 MW. An advanced Rankine power cycle is adopted because the coolant outlet temperature is sufficiently high (650 0 C). This cycle is similar to that planned for near-term, coal-fired power plant [33], and the predicted gross thermal efficiency is 49%. An 80 cm thick SiC-B4C shield is located behind the blanket to shield the superconducting coils over the 30 y lifetime of the plant. The shield also serves as a passive electrical conductor element for vertical-position stabilization of the plasma. The shield is constructed of forty 1 mm thick aluminum layers sandwiched between SiC panels. The aluminum layer is plasma sprayed onto the SiC-composite panels. To achieve toroidal electrical continuity, the aluminum layers are electrically connected between modules. The power density in the shield is low and the temperature is held low to account for the temperature limits of aluminum. The thermal mismatch between the aluminum and SiC-composite plates is accommodated by the higher ductility of aluminum relative to SiC. Most of the electromagnetic energy released during a plasma disruption is absorbed by these aluminum layers. The shearing loads during a disruption are transferred to the stiff and strong SiCcomposite panels, which have an estimated Young's modulus of 162 GPa. The vacuum vessel is made of steel and is located behind the shield. In ARIES-I, high recycling poloidal-field divertors in a double-null configuration are used. Each FPC module contains two upper and two lower divertor targets in wedge-shaped form for easy removal and installation. The target plates are fabricated from individual SiCcomposite tubes. Each tube receives the helium coolant from a supply manifold located in the private-flux region between the inboard and outboard strike-points. The plasma-facing side of the tube bank target is coated with a 2 mm thick layer of tungsten that would be plasma sprayed onto the SiC tube bank. This thin layer may be replenished by in-situ plasma spraying (e.g., for repair of disruption damage). The back of the tube bank (facing away from the plasma) is reinforced with additional SiC to strengthen the tube bank and to provide leak tightness.

The thermal hydraulic design of the divertor cooling system matches the inlet and outlet temperatures of the primary coolant system at the same pressure (350 and 650°C at 10 MPa). At the maximum surface heat loading of 4.5 M W / m 2. the corresponding maximum structural temperature is 850o( ', well below the maximum-allowable design value of 1100 o C. The pumping power of the divertor loop is 35.5 MW, but most of this power is recovered in the primary energy conversion system at 49% efficiency. 4.2. Magnet design

Operating the plasma in the first MHD stability regime using a high aspect ratio and relatively low plasma current leads naturally to a low toroidal beta and the need for high field magnet to achieve adequate fusion power density (which scales a s ]~2B4). The toroidal field at the plasma center is 11 T and the maximum field at the coil is 21 T. Currently available Nb3Sn alloys, produced by the powder metallurgy process, can generate fields up to about 21 T [34,35]. The conductor for the ARIES-I TF magnets is the same as that of the SSTR. The ARIES-I design, however, aims at a higher field strength by allowing for improvements in the superconductor coil manufacturing technology, by using a stabilizer which can carry structural load (CuNb), and by using two torsion caps (instead of shear panels) to counter the overturning moments. The conductor in each toroidal-field (TF) coil is graded, with Nb3Sn used for the intermediate- and high-field regions ( > 6 T), and NbTi used for low-field zones ( < 6 T). Detailed finite-element analysis of the TF coils and the support structure (bucking cylinder and structural caps) has been performed. The average vertical stresses in the throat of the TF coil on the mid-plane are - 700 MPa, the radial stresses are - 140 MPa, and the equivalent average stress on the mid-plane, with all loads added, is - 770 MPa. The loads on the bucking cylinder and the structural cap are shown in fig. 18. The toroidal stresses are 850 MPa. There is very little net vertical load and the radial loads are small. The maximum equivalent stresses in the bucking cylinder is - 900 MPa and can be lowered further by using a thicker structure. The reference structural material is Incoloy 908, one of the materials considered for the ITER magnets [36,37]. This rather low level of stress for a 21 T magnet is made possible by some unique aspects of the ARIES-I design: (1) high aspect ratio; (2) use of a bucking cylinder and two structural caps to support the out-of-plane loads; and (3) use of a copper-niobium (CuNb) high-strength stabilizer [38] that can carry

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

c

'

e

Fig. 18. Finite-element analysis of ARIES-I TF-coil support structure (bucking cylinder and structural caps). Contours of constant Von Mises stresses are shown and labeled (all in MPa) as A =110, B = 250, C = 395, D = 535, E = 675, F = 820 and G = 950.

structural loads. The design of the poloidal-field (PF) magnet system follows the ITER recommendations [37]. The peak field and the maximum stored energy in the PF system are 12 T and 17.8 GJ, respectively and most of the magnets have relatively low field and current density. 4.3. Economics, safety and environmental analysis

To estimate costs, a cost model is used that provides future "learning-curve" cost credits. Specifically, it is assumed that the ARIES-I unit costs are for a "tenthof-a-kind" plant. Such costs are taken to be - 50% of the cost for a "first-of-a-kind" ITER-like engineering test reactor costs [39,40]. Indirect costs are estimated to be 45% of direct costs. Standard assumptions [39,40] regarding construction time (6 y), plant availability (Pt = 0.76), economies of scale, and operation and maintenance (O&M) charges are used to estimate the constant-dollar (1988) cost of electricity (COE). The projected COE for the ARIES-I reactor is 65 mills/kWh. F o r comparison, the corresponding values for "median-experience" and "better-experience,, fission pressurized-water reactors (PE = 1100 MWe) are 78 and 46 mills/kWh, respectively, developed on the same cost-accounting basis. Coal-fired plant (PE = 2 X 550 MWe) costs are projected at 50 mills/kWh [25,26]. In

267

the cost estimate for ARIES-I, no safety-assurance cost credits [41] have been taken. The safety credits resulting from the elimination of the nuclear qualification (Nstamp) requirements on components could lower the reported COE by as much as 20 to 25%. Throughout the ARIES-I study, the design effort has been directed at maximizing environmental and safety characteristics of fusion by careful selection of materials and care in design. An extensive safety engineering analysis is reported in refs. [25,26]. The ARIES-I design is predicted to be passively safe. The site boundary dose (i.e., the early whole-body dose at a I km site boundary) is estimated to be 130 rein in a worst-case accident. Note that the tritium-breeding material in the blanket is lithium zirconate (Li2ZrO3) , a ceramic solid breeder. Lithium zirconate was selected as the reference solid breeder because the database demonstrates long-term and high-temperature stability. Extensive isotopic tailoring of zirconium would be necessary to reduce the level of decay heat and improve the waste disposal rating of the blanket. Laser isotope separation (LIS) is the only feasible technique for this tailoring since both light and heavy isotopes are to be removed, and the reference ARIES-I breeder is enriched to 99.9% 92Zr. In the worst-case accident, this breeder accounts for 91 rem of the total of 130 rem. Other solid breeders such as lithium oxide (Li20) and lithium orthosilicate (Li4SiO4) may be used as alternative breeder materials but require further data as noted above. Their use would lower the predicted worst-case site boundary dose to 39 rein, providing strong incentive to develop these alternative solid tritium breeders. Loss-of-flow and loss-of-coolant accidents have been analyzed. Low values of peak temperatures are calculated for the silicon carbide in the case of a loss-ofcoolant accident and low values of release fractions for induced activity are predicated. The coolant for the first wall, blanket, and divertor is helium. This removes concerns about coolant activation and chemical interactions between the coolant and blanket materials. In particular, the use of helium in the divertor (rather than water) eliminates concern about tungsten-steam reactions and the transport of the tungsten activation products from the reactor. The beryllium-neutron multiplier is also low activation, although recycling of beryllium will be necessary because it is not a widely occurring resource. All estimates reported here are calculated using impurities levels in the FPC materials as suggested in the ESECOM study [41]. The divertor plates contain a 2 mm thick coating of tungsten on the plasma-facing surfaces. A tungstencoated divertor collector plates qualifies for Class C

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waste after averaging with the silicon-carbide coolant tubes. The early off-site dose is estimated to be at most 57 rem for a worst-case accident. The reference design, however, uses an isotopically tailored tungsten that has lS3w enriched to an abundance of 90%. By using this tailored material, analysis shows that the prompt dose for a worst-case accident at 1 km from the reactor due to tungsten is reduced to about 11 rem. The tritium inventory in the reactor is predicted to be modest. The inventory in the first wall, breeder material, and divertor are estimated at 6, 1, and 10 g, respectively. The inventory in the fuel cycle is about 50 g. The total tritium production due to (n, 2n) reactions in the blanket is estimated at 1.1 kg per full power year of operation (fpy). Deposition of energetic tritium born in the breeder and deposited in the beryllium accounts for 1.4 kg/fpy. Recent data [42,43] indicates that the tritium is released from the beryllium above 610°C. Less than one quarter of the beryllium in the blanket will operate below 610 o C. It is assumed that as a part of a normal yearly shutdown, the entire blanket temperature will be raised to a value greater than 610°C, driving out the entire tritium inventory. As such, the tritium available for release is small. For example, a loss-of-coolant-accident (LOCA) in which the plasma remains operating for 15 s is predicted to release less than 700 g of tritium. The early off-site dose under this condition is estimated to be 7 rem.

5. Implications for fusion research During the last several years, steady progress has been made towards achieving fusion breakeven and reactor-relevant plasma conditions in the largest tokamak experiments. The degree of physics realism in conceptual commercial reactor designs, such as in the ARIES-I and SSTR studies, reflects the achievements in plasma physics research, particularly in the area of MHD stability, particle and energy transport, and special areas such as bootstrap current. Of course, the physics of burning plasmas must still be explored experimentally. Both SSTR and ARIES-I would have a high aspect ratio and a low plasma current, relative to the design of present experiments, in order to maximize the bootstrap current fraction and minimize the amount of current drive by external means. The trade-off between plasma current and aspect ratio allows these reactors to achieve the required confinement time. High magnetic field is used to compensate for the lower value of plasma beta that is expected for reactors designed in this part of

l p - B t - A space. Both designs operate at steady state, and minimization of the current drive power results in respectable net plant efficiencies (30% for SSTR and 39% for ARIES-I). Both designs also use high recycling divertors. The differences between these two tokamak reactor designs are due mainly to the design philosophy of each study. The SSTR effort is aimed at minimizing the degree of extrapolation from the present data base in technology while the ARIES-I study is aimed at maximizing the economic, safety, and environmental aspects of the reactor by assuming greater technological development. The contrast between the two designs can be used to identify high-leverage technology areas that are key to achieving safe, environmentally attractive fusion energy.

6. Conclusions Confirmation of the existence of the bootstrap current in the tokamak makes realistic the concept of a steady-state tokamak fusion reactor. The two reactor studies reported in this paper indicate that promising fusion reactors can be developed based upon reasonable extrapolations from our present knowledge of tokamak experiments. A high degree of safety and environmental attractiveness can also be achieved if there are significant advances in several technological areas, especially the development of low-activation materials for construction. The following physics items are found to be particularly important for the realization of these reactor concepts. (1) Reduction of the divertor heat load to an acceptable level ( - 4.5-6 M W / m 2) either by remote radiative cooling or by core radiation. (2) Maintenance of the He ash concentration around 5% in the D / T burning H mode plasma. (3) Demonstration of a Troyon factor g - 3 . 5 steady state, high bootstrap fraction, H mode plasma. (4) Assessment of the parametric dependences of the H mode energy confinement, especially with respect to the aspect ratio R / a and the toroidal field Bt. With respect to technology, there are different R& D items depending on the design, as discussed in sections 3 and 4. Both studies agree that safety must be pursued to obtain wide public acceptance of the potential merits of fusion. Demonstration of D / T fusion energy in T F T R / J E T . construction and operation of the experimental reactor and a successor DEMO reactor will provide the technical basis for the commercial use of the fusion energy.

M. Kikuchi et al. / Plasma physics and its impact on tokamak design

Acknowledgements The authors would like to express their gratitude to the members of the ARIES-I and SSTR design teams. Deep appreciation is expressed by M. Kikuchi and Y. Seki to Drs. T. A n d o and Y. Oliara who developed the designs of the high-field magnet and the high-energy neutral beam injector for SSTR. Collaborations with the University of Tokyo, Hitachi Limited, Kawasaki Heavy Industries, Mitsubishi Group, Toshiba Corporation and Sumitomo Heavy Industries were key to the SSTR effort. The ARIES-I study is supported by the United States Department of Energy, Office of Fusion Energy. Institutions participating in the ARIES effort, in addition to UCLA, are Argonne National Laboratory, General Atomics, Idaho National Engineering Laboratory, Los Alamos National Laboratory, Massachusetts Institute of Technology, Oak Ridge National Laboratory, Princeton Plasma Physics Laboratory, Rensselaer Polytechnic Institute, University of Wisconsin, Georgia Institute of Technology, Culham Laboratory, and Japan Atomic Energy Research Institute.

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[38] C.V. Renaud, E. Gregory and J. Wong, Production of high conductivity/high strength in-situ CuNb multifilamentary composite wire and strip, Adv. in Cryo. Eng. 32 (1985) 443. [39] J.G. Delene, K.A. Williams and B.H. Shapiro, Nuclear energy cost data base, US Department of Energy report DOE/NE-0095, September 1988. [40] J.G. Delene, H.I. Bowers and B.H. Shapiro, Economic potential for future water reactors, Trans. Am. Nucl. Soc. 57 (1988) 205. [41] J.P. Holdren et al., Report of the senior committee on environmental, safety and economic aspects of magnetic

fusion energy, Lawrence Livermore National Laboratory report UCRL-53766, June 1989; also Fusion Technol. 13 (1988) 7. [421 M.C. Billone, C.C. Lin and D.L. Baldwin, Tritium and helium behavior in irradiated beryllium, Proc. 9th ANS Topical Meeting on Technology of Fusion Energy, Oakbrook, IL, October 1990, to be published. [43] D.L. Baldwin, O.D. Slagle and D.S. Gelles, Tritium release from irradiated beryllium at elevated temperatures, Pacific Northwest Laboratory report PNL-SA-16998, November 1989.