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806 Chapter 6.2: The Development of Ultra-High-Vacuum Technology

factor of 2 increase in size, which would yield a device capable of reaching the

Lawson condition. These ambitious plans languished for two years until the Arab

oil embargo of 1973 pumped significant funding into any energy research. The

Princeton Large Torus (PLT) [63], a tokamak with a 6,000 L vacuum vessel and

0.5 m plasma radius came on line late in 1975. With the use of high-current,

hydrogen-neutral beams developed at Oak Ridge National Laboratory, PLT be-

came the first fusion device to exceed the Lawson criterion minimum temperature

of 10 keV in the summer of 1978 [64].

Scientific and technical progress came at an exhilarating pace as the fusion

budgets and resulting activity grew at 30-40% a year for most of the decade fol-

lowing the threshold year of 1973.

6.2.12

PLASMA IMPURITIES AND VACUUM TECHNOLOGY

The achievement of the high plasma temperatures in the PLT tokamak did not

come without several key developments in controlling how the plasma interacts

with the vacuum vessel wall. Spectroscopic and plasma resistivity measurements

on the first generation of Western tokamaks (ST and ATC at Princeton, Ormak

at Oak Ridge, Pulsator in Germany, and TFR in France) were indicating that

plasma impurities were a persistent problem. The ATC tokamak, which was built

at Princeton shortly after the Model C conversion to test a magnetic compression

scheme of heating plasmas [65], achieved in 1974 the first pure hydrogenic

plasma that was undiluted with carbon and plasma impurities desorbed from the

vacuum vessel wall. This result was achieved by evaporating titanium onto a large

fraction of the vessel wall prior to a plasma discharge to suppress the impurities,

combined with hydrogen gas injection during the discharge [66].

In 1975, the first in a series of compact, high-field tokamaks built at MIT, Al-

cator A, also achieved pure, "Z^^ =1" plasmas [67], by a combination of UHV

vessel design, programmed fueling of the discharge with hydrogen gas injection,

and pioneering use of a pulse discharge cleaning technique to condition the vac-

uum vessel wall before exposure to high-temperature plasmas. The discharge

cleaning technique [68] was developed by Robert Taylor and rapidly adopted by

all the existing tokamak experiments. Ironically, as a combination of the dis-

charge cleaning and Ti-gettering techniques was applied to the PLT device, low-Z

(carbon and oxygen) impurities were reduced to the point that these ions no

longer cooled the edge plasma by radiation. The resulting edge plasma was suffi-

ciently energetic to sputter excessive quantities of metal impurities from the tung-

sten "limiter" structure that defined the plasma radius and protected the vacuum

vessel wall. A switch of limiter material to a high-purity graphite solved the prob-

lem for PLT [69] and helped launch a development program for low-Z refractory

6.2.12

Plasma Impurities and Vacuum Technology

807

materials that would be needed for "first-wall" structures in the succeeding gen-

eration of tokamaks [70].

The problems of impurity generation and impurity suppression in tokamaks

were deemed of sufficient importance to the development of fusion that signifi-

cant portions of experimental programs on many machines and several dedicated

devices were devoted to impurity studies. The Oak Ridge Impurities Studies Ex-

periment or ISX device at Oak Ridge tested a prototype of the PLT graphite

limiters [71], developed a Cr-gettering technique [72], and investigated unpolar

arcing as an impurity generation source [73]. Impurity transport and improved

hydrogen injection techniques were studied at the Doublet tokamak at General

Atomics [74]. Large-scale tests of magnetic diverter structures for controlling the

transport of both hydrogenic and impurities species were performed on the PDX

tokamak at Princeton [75] and the ASDEX tokamak at Garching, Germany [76].

The latter two tokamaks, which became operational in the early 1980s, required

large vacuum vessels (~36 m-^) to contain the internal magnetic structures for the

diverter and a ballast volume for pumping the plasma particles that were neutral-

ized on the diverter target plate (Figure 8). The diverter volumes were equipped

Fig.

The vacuum vessel of a diverter-pumped tokamak (PDX, 1980), Princeton Plasma Physics

Laboratory.

808 Chapter 6.2: The Development of Ultra-High-Vacuum Technology

with large arrays of titanium evaporation sources to provide up to 2 X 10*' L/s of

hydrogenic pumping capacity. The PDX device was also used to test a prototype

nonevaporable getter array [77] (based on the Zr-Al alloy) and a large-area,

toroidal limiter fabricated from graphite tiles [78]. Because of the large size of the

PDX and ASDEX vacuum vessels compaied to the plasma volumes (6:1), both

machines became test beds for developing hydrogen glow discharge techniques

for cleaning complicated vacuum vessel structures [79,80].

6.2.13

TOWARD THE BREAKEVEN DEMONSTRATIONS

The rapid pace of the fusion program in the mid to late 1970s compressed and

overlapped the design and operational schedules of three generations of ma-

chines. This accelerated schedule was most evident in the history of the large

tokamaks built to reach the Lawson criterion for energy breakeven. Before the

first plasma dischaige was fired in PLT (or in a similar-sized device, the T-10

tokamak at Kurchatov), preliminary designs were in place for three tokamaks that

would be sufficiently large to reach energy confinement times near one second,

and include sufficient plasma heating and fueling capability that the remaining

Lawson conditions of plasma temperature and density could be simultaneously

achieved. These three large devices became the flagship experiments of

the

United

States, European and Japanese fusion programs: the Tokamak Fusion Test Reac-

tor (TFTR) [52] at Princeton began operating in 1982; the Joint European Torus

(JET) [53], built next to the Culham Laboratories in the United Kingdom, began

operating in 1982; and the JT-60 [81] tokamak in Japan began operating in 1984.

All three machines are still operating and have had their original programs ex-

tended to the mid- to late 1990s. TFTR and JET were both designed to operate

with D-T mixtures, and therefore have the complication of dealing with radioac-

tive gas handling [82,83] and the additional activation of the vessel [84] by the 14

MeV D-T neutrons. JT-60 operations were confined to hydrogen plasmas, and a

recently completed upgrade of the machine (JT-60U) has allowed the use of deu-

terium fueling [85]. The complications and expense of D-T fueling have post-

poned significant D-T experimentation to the last operations period for both

TFTR and JET. Late in 1992, the JET physicists fired a single D-T fueled dis-

charge that produced a megawatt of fusion power for a fraction of a second [86].

Unraveling the plasma physics in these devices is difficult with a single discharge;

however, this short-lived experiment did achieve a long-standing program goal of

producing a significant quantity of fusion power, and the subsequent exercise of

tracing the tritium inventory and decontaminating the vessel first-wall compo-

6.2.13 Toward

the

Breakeven Demonstrations

809

nents was very informative [87]. As presently scheduled, JET will return to an ex-

tended series of D-T experiments in 1996. TFTR began a two-year series of D-T

fueled experiments [88] in 1993, and subsequently the machine will be decom-

missioned. Before these long-awaited D-T experiments in JET and TFTR, both

machines achieved impressive plasma parameters on the Lawson chart with the

use of 20-30 MW of plasma heating [89,90] (with both neutral beam and ion

cyclotron RF sources) and frozen hydrogen pellet injection systems [91]. Plasma

parameters within 80% of

the

minimum Lawson requirements have been achieved

in addition to pushing the plasma temperature records to thiee or four times the

achievements of PLT [89]. Recently, experiments in JT-60U [92] and the D III D

[93] device (at General Atomics) have also achieved these high-temperature

regimes.

These impressive results in the large tokamaks have required continued devel-

opment of first-wall materials, first-wall structures, and the conditioning tech-

niques for these materials needed for edge-plasma particle and impurity control.

The TFTR vacuum vessel is protected with a toroidal array of graphite of over

2000 tiles [94]. Special discharge cleaning techniques were developed to remove

the large quantities of H2O that are absorbed by this structure during vents to at-

mosphere [95]. The highest performance discharges were achieved in TFTR after

the graphite first-wall was induced to behave like a plasma pump by condition-

ing the structure with a regimen of special He discharges [96]. Before initiation

of D-T discharges on TFTR, extensive studies of tritium trapping and removal

mechanisms were developed to recycle tritium trapped in the vessel [97]. After

operational experience was gained initially with the use of fine-grained graphite

for first-wall armor, JET replaced graphite in high-flux areas with Be [90], and

the TFTR, D III D, and JT-60U devices upgraded to carbon-fiber-composite ma-

terials [98]. After pioneering applications of a large-scale vapor deposition pro-

cess successfully coated the entire first-wall of the TEXTOR device in Germany

with thin layers of boron-carbide [99], the technique was applied to other toka-

maks (ASDEX, TFTR, D III D) as a means of in situ modification of first-wall

properties.

The careful control of the surface properties of the first-wall structures in these

large tokamaks would not be possible if high-reliability "second-wall" or vacuum

containment structures were not maintained. The complicated vacuum vessels

must satisfy many constraints, including UHV standards for leak-tightness de-

spite hundreds of access ports for diagnostic instrumentation and plasma heat-

ing equipment, and structural integrity to withstand cyclic loading due to bakeout

cycles and large eddy-current-induced forces

[100].

The need for high-capacity,

noncontaminating vacuum pumping systems for these large vessels drove the de-

sign of large (>3500 L/s) turbo pumps

[101],

and the associated high-power neu-

tral beam heating systems drove the design of large He cryocondensation pumps

with speeds exceeding 10^ L/s for hydrogenic species

[102].

810 Chapter 6.2: The Development of Ultra-High-Vacuum Technology

6.2.14

THE NEXT STEP IN FUSION

So where will the fusion program go next after the D-T demonstration experi-

ments on JET and TFTR are complete? The international fusion research com-

munity and their supporting government agencies have been debating the next

step for more than a decade. An era dedicated to demonstrating the scientific fea-

sibility of fusion energy is drawing to a close—after a considerably longer time

and larger investment than was envisioned by the scientists who launched Project

Sherwood. Some argue that the next step must address engineering feasibility of

a tokamak reactor if the goal of producing affordable fusion energy is to remain in

the foreseeable future. Others argue that this next step is impractical and

unaf-

fordable, and the program should explore different physics regimes that might re-

sult in a more appealing reactor design.

The next-step fusion device that is feeding this debate is a megaproject called

ITER, the International Tokamak Engineering Reactor (Figure 9). This device is

currently entering the engineering design phase after a three-year conceptual de-

sign effort was completed by teams from Europe, Japan, Russia, and the United

States. The goals of the machine include demonstration of a 10^ second plasma

burn with power extraction from a Li-blanket system. Effective first-wall protec-

tion, diverter pumping, tritium recovery, and remote maintenance systems would

have to be demonstrated. The machine would probably cost somewhere between

the estimated costs of the LHC ($4B) and the SSC ($11B). Clear thinking by the

scientific community and sensible planning by the government-industry partner-

ships who will build the next-step fusion device (and the next large accelerator)

are needed to keep these important fields moving forward.

ACKNOWLEDGMENTS

The author acknowledges the able assistance of the following colleagues with the

preparation of this paper. C. Sinclair at Jefferson Laboratory, D. Carroll at Prince-

ton Plasma Physics Laboratory, S. Wofford at Lawrence Livermore National Lab-

oratory, E. Johnson at Brookhaven National Laboratory, K. Wilson at Sandia Na-

tional Laboratory, and A. Mathewson at CERN. This work was supported by U.S.

DOE Contract No. DE-AC05-84ER40150.

Acknowledgments

811

Fig.

lOH

s^H

S-i

ioH

15-f

Bucking

Cylinder

Poloidai

f=fddCoil

Central Sdenoid

Vacuum

Pumping Oud

Poioidai

Fteid Coils

0 5 10 15 meters

ITER 1993 - Elevation View

The ITER vacuum vessel (1993), Princeton Plasma Physics Laboratory.

812 Chapter 6.2: The Development of Ultra-High-Vacuum Technology

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1040.

Index

Ablation mechanism, pulsed laser deposition,

698-699

Abrasion resistance, 786

Abrasives, for cleaning, 570

Absorption, and construction materials, 447

Abundance sensitivity, mass analysis, 299

AC tungsten inert gas welding, 547

Acceptance ellipse, 316-317

Acetone, 571

Active storage, 591

Actuation of valves, 400-401

manual, 401

remote, 401

sealing of actuating motion, 400

Adhesion, loss of, causes, 786

Adsorption, 40-41

and construction materials, 447

cryogenic pumps, 160-162

heat of, 42

sticking coefficient, 43-44

surface adsorption isotherms, 45-47

Air firing, for cleaning, 576

Air pollution measurement, mass analysis, 291-

292

Alconox, 520

Alkaline cleaners, 572-574

Almeco 18,520

Alodyne, 454-455

Alpha ionization, 307-308

Alternating Gradient Synchrontron (AGS), 478

Aluminum

chamber wall thickness requirements, 521,

533-535

chemical finishes, 520

cleaning, 569

in construction of vacuum chamber, 453-455

corrosion, 538-541

cylinders, 533-534

demountable seals, 512-518

flat

plates,

533

joint design, 546-547

mechanical finishes, 518-519

and outgassing, 509-512

peak stresses/strains, 535

protective coatings, 541

radiated heat, absorption of, 536-537

relative thickness, 534-535

repairing defects, 548

as seal metal, 481

thermal gradients, 536

thermal time constant, 537-538

weight/thermal expansion, 535

welding, guidelines for, 541-548

Aluminum alloy

chemical composition limits, 531-532

mechanical properties of, 525

and outgassing, 486

polishing, 559

types/designations, 521-525

Alutone, 520

Amchem, 520

Amklene, 520

Amplified sputtering, 626

Anhydrous alcohol, 585

Anisotropic etching, 635-638

Anodized aluminum, cleaning/sealing of, 560

Apiezon, 451

Appendage pump, 221

815