. z ha ni Energy Research & Development Administration a . sale ai " : i AS ties Division Of Magnetic Fusion Energy
| | ee oa July 1976
= 6
Ler eee 5
ogee
prewer
Re
ee ee ee oe ee ee ee ee ee ee Se
= *
5 } i f 1 ee
a x > } ‘5 % OHM WM IORI NIE
| * A * ines
aU | : a ae eat! nf i Fi , 4 ‘Bes 7 Sd , “ i7 g & {) R WPAN x 3 amy 2) . - e Si § - ; ae ' y
; Becteaee We Aeemertanoenad Si canck wie CATER OOO TNO TCE: eae Pee os
75 76 77 7 psa Sito Be 83 4.
DISCLAIMER
This Program Plan spells out the options available to the Division of Magnetic Fusion Energy under a range
of possible funding assumptions called ''Program Logics". As such, it represents a Division document and not necessarily the views of the top ERDA management. The Division believes, however, that the range of program logics and options presented will cover any eventual- ities which may arise in the course of setting program © strategies with ERDA management, OMB, and Congress.
Available from:
National Technical Information Service (NTIS) U.S. Department of Commerce
5285 Port Royal Road |
Springfield, Virginia 22161
Price: Printed Copy: 5$ 9,00 Microfiche: Siacea
ERDA-76/110/3 UC-20
FUSION POWER BY MAGNETIC CONFINEMENT PROGRAM PLAN
VOLUME III
FIVE YEAR PLAN
JULY 1976
Prepared by the
Division of Magnetic Fusion Energy U.S. Energy Research and Development Administration
Digitized by the Internet Archive in 2021
https://archive.org/details/ftusionobowerbymagO03usen
i
Liv
TABLE OF CONTENTS
GENERA U MENT RODLOELON sata dara e f.t tthe tat es Zaha Phot attest aot. Matera aa Hekete tate AUST OUBACVANLAPES . a ae tee ett C ee ol a od ahaldl Sats ETM ay Stat mart Be se PrOcramvitecoryve cae awan sare ate Sid Ovo bth LC ieiGind Ono Onl Ot WO Oubint 0 Mets atete C. Long Range Planning Projections ........ Pack ola chietal hd Perea md aatemate Dem Diviston, Organization.and Management «6 eeecn veces vow ee SOMadas Ai Ee Organization sof Reports ssssee27 222 SEER POP AEE ts Ha SPOS. APR: re BeIPOLaLCHsLeLOrts sees eect feof te ae PERE io at tar he ere Bee Gow PUGSELSOUMMIATY. Seated are re ce eee Oe ES Sa a aera cdeterete ere" aot cuete CONFINEMENT SYSTEMS ..... See SAT GATS RAGE Paes Go TS ae ee Meee Ar ACT OCUCTN OU ci ne ea Ties ase os 88 2a sd a te ate wane Patek, wield warans BreeerobLempArees tee Giese fe cree ase Shain ea cs oa sate chatted tet siete, ste Le Hokamaks. 6.24 2 t.chuc 3c a SIT ee SETS ROSTERS ate aren. ead oie is PPA PNCELC Mirrors SVS CEMS 0k cities oss ov 4 ee 0 3 6 6 Rialprerateiate or Bar 3. High Density Systems..... Pd iiud Sodas seer eso tiaeh ie Te ole aaa BEE a € C. Recent Achievements.........ce00. Eat Ae OA Se ea ee eee 1. -slokamaks2 4 fe 2825.5: 4 06.325 ee Lie eTaee eee E46 6 Sete mle 2. Magnetic Mirror :System$: is1s5254335%%> Teen Dore eal ats Bane 3. High Density Systems.......... CoA EEER TRADI ASS HS Me hems ; Dae Program’ stroctures and*Major* Milestones: : 225 5%5.0%s ssecele sie cae Te LORSMAK Sarat. ss oratsie he ee es Det Ey hte ee ee RR Loh Vetets-eeleks @ Zoeagnetic Mirror*Systemss 425555 5.4 2% otatedels shi el. Gta. aes SO. High Density Systems: .....%5%% 5s BRS ORES FF. DOS IR Me. RIBAS ; Eee DUC SCC SUMMATY <2. c's 5.0%: ee & vse lee o's 6 ok SS sale Mawes. Nat hetelcls a wer aes
Itt. TECHNICAL PROJECT SOFRICE ae etree oUcl shal olcrcnchoreneich i Meach cd Neretencnonces 569 53 A. Introduction, & i... ios we te eee eee eee teen on en ee ee nee mene 53 B Tokamak Fusion’ Test Reactor. «. celts st cere teed eee ° 54
A INCE OCUCHL OM in ers cree cite sel eres retro a ete ee ste en neem eer ene 54
2 -ODJECCIVES rs iy. Peete elerete soe, Sele Genet a ce Cae dey eee anne eee ee 54
3.. Principal Desion Features oo c.scee sie. oo cece) eee a yojoyetere 56
4.. Project=Miles tones ...0. 5 rc > oecaee te ene erate ate eee tener te eee 58
3. Projected Budgets. occ... asus cy ce sienees oh dees ofa) stele rere een ee 59
C. 9 Rotating Target, Neutron Source... acl ccc) cick cia ne 60 1. Introductions ¢ cya ose ecs +, ere, pele stomata ec rene eee ee eee 60
2. Objectives. <uscces osc ue oun. AAO eee cei. te std ehesene 60
3.. Principal Design Features)... a... +0 «cc: WoedeE tee eee ae 61
4, Project Milestones 2... cues ed has cee oe eect eee eee 61
Ds Projected Budsets 2. sone. ns oc ccc lan ote et eee eee 62
D. jIntensé Neutron Sources... 0 > ck cscs cel tele omr-tetees tee eee ee 62 I. Introduction... 4.4 «ces a4 sunt auto ele elite gels «a eee ee 62
2. ODJECELVES . c.. Faccle c.ccles cle cinta Cina hehe aie rete et een eee meen 63
3. Principal Desion Features... ove. e sce ee ee eee 64
4. Project (Milestones oo wauc ats so ccm flesiiet ors eileen tea eee 64
5. Projected Budgets 5.3 scans. scree cles cies etetetelons siete SA cen Ge dle 65
IV. DEVELOPMENT AND TECHNOLOGY PROGRAM ,........ AD SOO 0.00 DO Oo On0 On 66 AAS Introductions cies ccs ee oe eb ps ele Palos cies ere oiae oncte oleic eran enone 66 Bae Magnetic Systemsi... sos pce ooo 9 scp e cele ele ste tenale uae wonton eee eens ao T.. Masnetics. 566 borne: ecip ce See oc siete: oe ek ee ee ene eee B 74,
2. Energy Storace... coos «+c senIee tee eee eee are deonsasienekeney ios 81
CHE Plasma Engineering sous cus pcs heres Or elee stelle o-te tie tet ets eee ie 86 1. Neutral Beam Development ............2-ee+seeee és BES 440 tose ae 92
2. Alternate Heating Technology ..2...0.- +. stein ot APS cree 100
3. Direct Energy (Conversion wo. ee cs eee cis ete cierto een eee A 106
4. Vacuum Techno logy. sere seteatece ote sbecctey fetta sitet et nee mone emer en en wees ‘aan
5. Plasma Maintenance and Control ......--.csseeeeeserccoees o° 114
D. Fusion Reactor Materials yo. ere re ee en ot ee 119 1. Alloy Development and Irradiation Performance ......+..+.0. iba
2. Plasma-=Materials Interactton so... acme oes rene tee ee renner. P29
3. Special Purpose Materials Development .............+00¢ Bch 135
4. Damage Analysis: and. Dosimetry™.. «9:2 osc oe ee 138
5. Radiation Facilities Development and Operation ........... 141
ted
PAGE
Eee busion systems Eneineerinoeee een .ereme Lat «as. nsec ns ee Saye 147 PMA CVADCCCMD GST eri teee rye tien cts o oiois «ale cele aie Cae oS sco 4 on orale Bakes 150 Ths GSETET) GIG IGN IOS ometet St Se ee Se es Ae ea era ge reer 156 Sem Blanket and shield Eneineering, os... ..<+0sessissevsescvee: : 163 (emeetriciumerrocessing «anda Comerol ..5s.c60 ns cr obey see's voces oiere o's 172 Digececasmasoysitenset..c fee BAShELLS Onk emer. .Lusicm.sceese 179 Gren LAN EO YS Cl Serene: Porte ends ote cts eae pice ete alee eo ska Fics) Wl oMe svete o alee @ 186 VPM Vt ts OMMent SNnueoaALely meee sf eiek tee «Picts olecsteteie ele wale eshte ela alom 190 APPILATDEPLASMA’ Pity Si CStBPROGRAM! 9 Syit..<.2.- 000s Bee aiete eatieees eats, dro create ous 200 Ase Introductions:4 +) fe Eboeie cf. Foe) 40.6% Gstenriniin.seke cet 200 B. Problem Areas,....... Aah Se DB I ohms Poy cere we ee rE Arce 202 1 seFusion Plasma Theory.......... mie OR ILI ee PRR ee : 202 ZeepEaner iment ater) Asma; RESeCALche.. 4s sooo cls soles s ats o ae a > severe + acs 208 BeleComputer ServicessandsTechnologyc:s.:<.u%..ueckeeees 4% eh 213 CEM RCCENTSACHKLEVEMENESY., Circ src ca 6 cies cs.5 sicle's's AOD OO OR Siaotlaterere 214 i UiStonmPLASMaRTNEOLY scissile 6 ts 6 0.6 wiaiere cisicieie 6.0 0 a%.6) sieeve we 214 PeeeLADerimenLalerasma RESEGarCh :).4 tc sie 6cteels ses ofels 4c cmisieiee « 221 DraeCompupermuerviceswand Technology. cu. oo ssid cc gc wos ol e'e.c le 6 230 D. Program Structure, Budget Summary and Milestones.............. 232 ieee risione Pl aemamheoryh. Ret see 6 oc bos oe cs 000 6 ae oh c eee ave 232 DeoeExpenimentals Plasma RESEaLrChcg cs secs vcsceees cecre sabe ee vie 238 B28 ComputersServicesnandsTechnologysiiist4l...004.@8.¢00 00%. : 249 COGS LD) LOMUULUL ORE SDL SONOS wale occ 6 shh a cise eles 6 obs. sia's ote a ene 8 250 E. Applied Plasma Physics Operating Budget Simard etalee ots. toleetarncss 252
oie ip
. s © @ 6 se a & F.# - 7 os “40s ’ abr -Vinmuve Je_oul galrarego. eiuciausaleni ; jetty
.itier Devalogrant.@
sist eeninens aber
se - ea 6a 6
RPP OTTO TY eT asi
-ee e+ ere 4+ &©
ok
va a PT yee vee ‘aw Aes yee 88 r ya @ it ytnatge ete yarveentgor biside baw 3 4A SetEeetnee '. Toxo Y bas arlkessoxt a wi _ 44
enw vei vee yeaa vn ‘ »s ,2 ©6600 @ one 1 <324.), SES eines Ae! 09 SO ON Saen a ae siberaat Soe vena ane panes ty
Wet: prosdt ameatl® moles BS ig it eebasl san lt Thscout yeas ~ 7 . weolotttseT hea asahrreg ‘ystuqaed - nt ah Ae A ct ewer © '., SSrbasveRHse Beso. & 7) ; .. roan? aaeeTt aoreey - 7 ? foes ae onahlY Tasha toyed ,S weet nolon Ho@T baa aoaiviaee xetugmeo t cult brie worse Yo bell "state te mgout” a Levee ved ea ew VIOOOR "pemite wakes ak
ot | nel ‘aneely Tesnbir yogi ‘8 ne
<eoloniGey bre gesivae® Yelbqaed at.) 939% brsqxe Sau sUt” ale jc a P
* @
I ay, _
ona @e seeoe @eo= ips e esepeaat - sic abant i bi
+4 e0@ew 6 Oe § PS i
_" $ e¢*@ Bene and CerptE@h «ices cab 6 hee 4aver as 4 :
= 1. ¢ J aa 2 ‘paset@ne
ai ak ond Trre Lon on Perforwanes wa pate
’ ope as 66 €82e + ¥ matartials nye opment Poo a remy
ret 1 ae e*tt«ae6 oie 77
ij v 1 > 2
uo uv fe een |
“— at feet =e
ew SS + «we
I. GENERAL INTRODUCTION
A. Fusion Advantages
The presently-established program goal of the magnetic confinement fusion program is to DEVELOP AND
DEMONSTRATE PURE FUSION CENTRAL ELECTRICAL POWER STATIONS FOR COMMERCIAL APPLICATIONS. The potential
advantages of commercial fusion reactors would be:
1)
2)
3)
4)
5)
6)
7)
An effectively inexhaustible supply of fuel -- at essentially zero cost on an energy production scale;
A fuel supply that is available from the oceans to all countries and therefore cannot be interrupted by other nations;
No possibility of nuclear runaway;
No chemical combustion products as effluents;
No afterheat cooling problem in case of an accidental loss of coolant;
No uses of weapons grade nuclear materials; thus no possibility of diversion for purposes of blackmail or sabotage;
Activated materials that may be produced would have relatively low biological hazard,
and their handling and disposal can be readily managed.
B. Program History
The present program in magnetic confinement began in the early 1950's. During its early years, the research concentrated primarily on problems of plasma confinement and heating. By the late 1960's, solutions to these problems had progressed to the point where good plasma confinement was achieved
in a number of different magnetic field configurations. These conditions were achieved in relatively small devices with plasmas at temperatures low in comparison with the temperature necessary for signifi- cant fusion to occur. Nevertheless, to achieve these results, it was necessary to develop, to a highly sophisticated state, not only the basic science of plasma physics but also a variety of supporting
technologies including normal and superconducting magnetics, vacuum systems, particle beams, energy
storage and plasma diagnostics.
During the early seventies the program initiated an expansion to extend the results of the 1960's to larger hotter plasmas which ultimately would be capable of sustaining fusion reactions. Extension of the physics results of the 1960's involves new domains of physics and considerable technological development. Nevertheless initial results on plasma confinement in intermediate-size devices at moderate temperatures have inspired renewed confidence that the scientific and engineering/techno- logical problems of heating and confining a fusion plasma can be solved. Increased emphasis has
therefore recently been placed upon the longer range aspects of fusion reactors.
C. Long Range Planning Projections
The Long Range Projections are discussed in a separate volume (Volume II). In that volume five separate plans called ''PROGRAM LOGICS" are discussed which are constrained primarily by the level of funding. Within these program logics, examples of options are presented which follow from the degree of success in the physics and engineering. The present volume discusses the "near term"’
(next five years) of a single plan, the "Logic III Reference Option." nf
The Logic III plan is defined as LOGIC III. AGGRESSIVE The levels of effort in physics and engineering are expanded according to programmatic need assuming that adequate progress is evident. New projects are undertaken when they are scientifically justified. Many problems are addressed concurrently. Funding is
ample but reasonably limited. (This program would be aimed at an operating demonstra-
tion reactor in the late 1990's.) The following general assumptions were used in developing Logic III: e Tokamaks are presently the most promising approach to achieving commercial fusion power.
e One Experimental Power Reactor (EPR) will precede the Demonstration Reactor (DEMO).
The next step in the tokamak program will be based upon the physics and engineering/tech-
nology results available in 1979.
Of the Alternate Concepts, mirrors currently are the most advanced; the next step in the mirror program (MX) will be taken in FY 1978 and the next step beyond that will be based
upon the physics and engineering/technology results available in 1982.
Theta Pinches and other alternate concepts are being explored to identify more promising reactor concepts with respect to physics, engineering and economics. The toroidal theta pinch is the most advanced of these; other alternate concepts are less developed but do
demonstrate promise.
A decision on the first major DT facility for the theta pinch or other alternate concept
is based on the physics and engineering/technology results available in 1985.
Supporting theory and experiment is available throughout the program and supporting
facilities are built as required.
Costs are calculated assuming that the tokamak program proceeds to the DEMO and that one
Alternate Concept proceeds through an EPR.
The general features of the Logic III Reference Option are shown in Figure I-l. The present program consists of several small and medium-sized hydrogen experiments (most notably the ORMAK and Alcator Tokamaks, the 2XIIB Mirror, and the Scyllac Theta Pinch) and the larger PLT at Princeton which came into operation in December 1975. Two other large tokamaks, Doublet III at General Atomic and PDX
at Princeton, are in fabrication and scheduled to operate in early 1978. The first DT burning Tokamak, the Tokamak Fusion Test Reactor (TFTR) is scheduled to operate in mid-1981. A large mirror
experiment, called MX, has been proposed for operation in 1981.
Under a Logic III program each of these devices would be upgraded, primarily by adding more auxiliary heating power, to test physics scaling laws at higher temperature and higher power density (beta).
In the mid- to late-1980's, large device(s) would be built assuming good results are obtained on earlier facilities. The next step in the tokamak line is assumed to be either a Prototype Experi- mental Power Reactor or an Ignition Test Reactor (PEPR/ITR). TFTR would be upgraded. An engi- neering test reactor (FERF/ETR) is assumed, which could be a tokamak. By the early 1990's an Experimental Power Reactor (EPR) would be built, which makes net electrical power with high
reliability. This device would be followed by the Fusion Power Demonstration Reactor in 1998.
NEXT NEXT STEP STEP NEXT STEP
MEDIUM-
HYOROGEN UPGRADES FUSION POWER EXPTS DEMONSTRATION REACTOR
TOK AMAKS
UPGRADES PEPR/ITR
DT EXPT.
TFTR- TFTR uC
TRITUUM AND MAJOR NEUTRON BLANKET SUPPORT SOURCES & FACILITIES SHIELD
FACILITIES
SMALL- & MEDIUM- SIZE
LARGE (MX-SIZE)
MAGNETIC EXPT PEPR/ITR
MIRROR
HYDROGEN EXPTS
“SOME D-T SHOTS”
THETA PINCH
SMALL- &
AND LARGE MEDIUM- OTHER HYDROGEN
SIZE PEPR/ITR
ALTERNATE HYDROGEN EXPTS
CONCEPTS EXPTS
74 76 78 80 82 84 86 88 90 92 94 96 98 2000 02 04 06 0
Figure I-l1 General Features of the Logic III Reference Option
The Magnetic Mirror Program is assumed to evolve from the present small- and medium-size experiments, most notably 2XII at Livermore, to a larger device in which a limited number of DT shots would be possible. A major objective of this device would be to test confinement scaling for longer times, and to test methods for improving power balance, a prerequisite to the feasibility of a pure fusion mirror reactor. This would be followed by a PEPR device in the late 1980's. The FERF/ETR could be
a mirror. This could be followed by an EPR operating in 1996 and a DEMO around 2004.
For the other Alternate Concepts, larger hydrogen experiments, such as the Large Staged Scyllac, are assumed to operate in the mid-1980's, followed by a PEPR/ITR in the early 1990's. Next could
come an EPR in the late 1990's and a DEMO around 2007.
For costing purposes of Logic III, 3 PEPR/ITR devices; 1 FERF/ETR, 2 EPR's, and 1 DEMO are assumed.
D. Division Organization and Management To manage the program, the Division of Magnetic Fusion Energy is organized into the following
four interrelated programs:
1)
2)
3)
Confinement Systems (CS) within which experiments are fabricated and operated to model many of the features of fusion reactors, in order to determine practical methods of achieving the conditions necessary for fusion reactors Included within the Confine- ment Systems Program are all of the major tokamak, mirror and theta pinch experiments (with the exception of the construction of the Tokamak Fusion Test Reactor) and two
other highly developed Alternate Concepts: the Z-pinch and the Elmo Bumpy Torus;
Technical Projects Office (TPO) which supervises the construction of the Tokamak Fusion
Test Reactor and two neutron sources (RINS and INS);
Development and Technology (D&T) within which solutions to the problems associated with
the design and construction of the next generation of plasma confinement devices are developed and a broad technological base is developed in areas important to practical fusion power reactors. Included within the program are neutron radiation damage studies, superconducting magnet development, development of auxiliary heating systems and power
supplies, systems studies and environmental safety;
4) Applied Plasma Physics (APP) within which theoretical and experimental studies of fusion-relevant plasmas are conducted, seeking the body of knowledge required to understand and predict the behavior of fusion experiments and the operating char- acteristics of fusion reactors. Applied Plasma Physics supports all of the Division's theoretical work including management of the computer facilities, the basic smaller experiments, diagnostic development, and the smaller and newer
exploratory concepts.
The personnel of the Division are listed in Figure I-2.
E. Organization of Report
The present report is structured along the lines of the four programs discussed above. Each major section of the report is devoted to one of the four programs. In Section II the Confinement Systems Program is discussed. Problem areas are described followed by a discussion of recent achievements for Tokamaks, Magnetic Mirrors and High Density Systems. The present program structure and major
milestones are then presented, followed by a five-year budget summary.
Administration
J. R. Young, Assistant Director L. Routzahn, Secretary
J. Anthony R. Rosselli W. Burrier Ree Zane
R. Cunningham
A. Brown, Secretary Valentine, Mail & Records
ry
Technical Projects
J. N. Grace, Assistant Director
Program Management Branch Die lee MEGOmEas Chie: Js dhehest
Systems Engineering Branch R. J. Impara, Chief
W. Marton C. Smedira
Reactor Engineering Branch K. G. Moses, Chief
L. Price
A. Dixon, Secretary
Office of the Director
Ee B.) Kimtnesy, . Weller, Secretary
Director
. Mason, Special Assistant
aq ESF, HORM es.
Confinement Systems
S. O. Dean, Assistant Director
A. M. Sleeper, Scientific Coordinator
R. A. Watkins, Secretary
Tokamak Systems Branch
N. A. Davies, Chief
R. Blanken
De Lena
J. Willis
T. Hsu
Pe die Shower, sSeckecany
PaeAw Diickencsta tram oechebany
High Density Systems Brnnch W. R. Ellis, Acting Chiet A. Kadish
E. Oktay
L. K. Worden, Secretary
Open Systems Branch
Rw Elis? ‘Chile
V. George
D. Johnson
L. Gilbert, Secretary
eee
Figure
10
Senior Scientific Advisor
R R R. Bingham, Coordinator for Plans G S Secretary
Development & Technology
J. M. Williams, Assistant Director L. Bogart, Technical Assistant Je Haas, secretary,
Mat'ls & Radiation Effects Branch K. M. Zwilsky, Chief
Cohen
a Daliden:
Finfgeld
Reuther
Ruby, Secretary
AHOnSs
Systems & Applied Studies Branch F. E. Coffman, Chief
R. Kostoff
M. Murphy
J. Baublitz
J. Neff
B. Twining
li) MOSEL peCheltaiay,
Plasma Engineering Branch Jin We Beall, Chick
H. Cullingford
Se oeaeen
C. Smith, Secretary
Magnetic Systems Branch C. D. Henning, Chief
D. Beard
1s Zaehesas
D. Wilt, Secretary
B=?
Applied Plasma Physics
J. F. Decker, Acting Assistant Director R, Stevens, Secretary
Fusion Plasma Theory Branch Re H. Price, Chiles
O. Manley
D. Priester
W. Sadowski
Jin) Hund secretary
EXp. Plasma Research Branch J. F. Decker, Chief
W. Dove
G. Mischke
P. Stone
Computer Serv. & Tech. Br. G. R. Ingram, Chief
C. McCoy J. Esworthy, Secretary
In Section III the activities of the Technical Projects Office are discussed. Three major construction projects are treated: the Tokamak Fusion Test Reactor (TFTR), the Rotating Target
Neutron Source (RTINS) and the Intense Neutron Source (INS).
In Section IV the Development and Technology Program is discussed. Major development programs are described in four areas: Magnetic Systems, Plasma Engineering, Reactor Materials, and
Fusion Systems Engineering.
In Section V the Applied Plasma Physics Program is discussed. Problem areas followed by a discussion of recent achievements are described for Fusion Plasma Theory, Experimental Plasma Research Computer Services and Technology. The present program structure and major milestones
are then presented along with five-year budget requirements.
F. Foreign Efforts Vigorous programs to develop fusion reactors also exist in the Soviet Union, Europe and Japan. The Soviets are planning a superconducting tokamak (T-10M) the size of PLT and a DT burning
tokamak, the T-20, which is much larger than the TFTR.
LL
The Euratom countries have formed a joint fusion development program, which is completing the design of a large experiment called the Joint European Tokamak (JET). The Japanese are planning a TFTR- sized Tokamak (JT-60). It is scheduled to operate in 1980. Mirror research is conducted in the Soviet Union, Japan and Sweden. There are high density programs in Britain, Germany, Japan, and Russia. High Density facilities projected are the High-Beta Stellarator, HBS-II, at Garching, and the Toroidal Z-Pinch, HBTX, at Culham. The U.S. program is approximately 1/3 of the total world
effort, based upon total manpower.
G. Budget
A summary of the budget requirements for the years FY 1978-82 is shown in Figure I-3. FY 1976 and FY 1977 are shown for comparison. Budget detail is presented in Figure I-4. It should be noted that prior to the FY 1976 budget most major experiments in this program have been built using operating and equipment funds rather than as construction line items. This is reflected in the FY 1976-77 budgets, which include the fabrication of PLT at a cost of $14 million, the PDX at a
cost of S18.8 million and the Doublet III at 4 cost of S27. 7emil tion.
b2
Operating Expenses ..\...:....
Confinement Systems ....... PeCiOL CALS ELOACCUS ies cee s\s!s5 Development & Technology .. Applied Plasma Physics ....
EQuUEpMCTIC oc wres-ew ce EOE ee as COnstrucEVoOn mec. ses sca cue sie ss
TOTAL Magnetic Fusion Energy
ONOF WEE WO
Fiscal Years
LOT6L
3750 182.0 247.1 280.0 L9<3 ge 6 110.0 1250.0 0.8 L020 18.0 12-0 9i2 58.0 75,0 9520 Pace, 34.4 44.1 48.0 4.6 23.0 a Za0) 44.6 Died oF a) 13050 105.0
47.1 BOZaD 409.1 429.6
Figure I-3
Summary Budget Projections
13
CEO Ors OC Or@ ©
SO oOLOowre
Total
et
oO’ fo) ie]
Cais [SS (@e) je) (e=)
Figure [I-4
Budget Detail Fiscal Years ($M) Total
1976 1976T Loe 1978 1979 19380 1981 usley: 78-82 Tote WeOperats nw. piece. <a ok 120.0 Sis 182.0 DET « Il 280.0 327/50 346.0 37/60 US76e Confinement Systems ........ 61.9 £933 L926 VLOF0 125-0 140.0 140.0 150.0 665.0 Dolatia los.) Sato ota a ae 4362 1358 Sit L050 Bain 0 80.0 80.0 80.0 335.0 NAPE ays CGMS a cin eile eis 10.9 SS) 14.0 2329 36.4 41.1 50n9 40.7 18 .0 High Density Systems ..... Taste 20) Bic 2 Lie il LST6 To8 ab ew | DO) & 96.0 TeChnLeal wen [eCes aes ecco. 0.4 0.8 10.0 18.0 orl 22.0 ee) 25.0 102.0 TR Riss csc 2 cu eae een ois «eee cE 0.4 0.8 10.0 15.3 £023 20.0 23.0 75.0 93.6 LNG teste chee iste actor etree es 0.0 0.0 0.0 2 0.0 0.0 0.0 0.0 Le TG os iret cl eranat ogress. shah, keh eenas 0.0 020 0.0 15 Law Fh), 2.0 0.0 Whe? Development & Technology .. 33.4 942 58.0 75.0 95.0 110.0 126.0 146.0 55250 Magnetic Systems... e255 % 7.4 149 D7 od 19.0 25.0 28.0 3254 Sue 2 141.3 Plasma Engineering ....... LOR Bae WOR7 Dib o® 250) 26730) 29758 BA 5 3 ORS Fusion Reactor Materials . Org Leg 9.3 16.0 ZOO 23.0) ZOeS SORE ibibsyete} Fusion Systems Eng. .....- 8.6 esis) 10.3 18.0 C30 30.0 34.4 39.9 ie yobs) ENVirOn. sand esalety coy sme. 0.4 Ort He 1.0 220 3.20 324 399 1373 Applied Plasma Physics ..... 24 3 Tel BO 44.1 48.0 55.30 D0 sey) 2 Lok Fusion Plasma Theory ..... L2 «3 326 13.6 Sar 21°50 25.0 YAY) 25.0 Lis27 Experimental Plasma Res. 8.8 Syail IL Bho bod RS 2050 20.0 20.0 O)5 he 7! Comp. Services & Tech. ser 16 50 920 Say 10.0 10..0 1.0.0 47.7 PQ DING Le merase as ee eat sy) tel eee Tho 4.6 B30 3220 44.6 2k.) 45) 2 Sail DSM D Confinement Systems ........ re 255 Cae 10-2 17.4 Pj 18a5 20.4 89 .8 Technica MPro jectsee ar. eae Ot On 1gt 2.3 365 3.0 Lee 0.0 10n2 Development & Technology ... Le HE) aa7 8.5 HESS) 16.0 192 20.0 LIQ Applied Plasma Physics ..... 4.4 1eo 8.0 5.0 8.4 19.0 6.1 1487 53.2 ConSicuct tone asmms cr caine Lo Di Vis D 130.0 (LODeO iene) 2750 Bae. 961.0 TE TR whe 30: i le eal 28s Sevres hohe ie Lon0 5s 80.0 95.0 ee LD5e0 Som) 35.0 Zim) PRER (LT Rivtietoe vans aie sareekes as 0.0 0.0 0.0 0.0 $5.0 50n0 80.0 105.0 250.0 High Field Neutron Source .. 0.0 0.0 0.0 TORO 15.0 20.0 20.0 10.0 75.0 TRLCi nig PACiVEy er aens sel es. 6 0.0 0.0 0.0 0.0 5.0 10.0 20.0 10.0 45.0 Blanketcand ‘Shield ie. oon. =. 0.0 0.0 0.0 (ia) 0.0 520 10.0 2020 3520 UM Sie exer clare atendte sromeoreue Sieeetere ts Zw 0.0 es 0.0 0.0 0.0 026 0.0 0.0 TINSIs, sii atten spe teceeiaret ante ara lereiaes Ont 0.0 155.0 10.0 0.0 0.0 0.0 0.0 10.0 Engineering Test Facilities 0.0 0.0 0.0 0.0 0.0 10.0 3220 39.0 81.0 MRE Aelcier a teetonen somo icusuotcdeteisloner els. ORO 0.0 0.0 5) 30 35),0 35)50) ESO) 0.0 100.0 ESS rte aero aa iilincel anole arora ete 0.0 0.0 0.0 0.0 O<0 2.0 15:0 35.10 50.0 LHX (2) seeseeeeees te teeeeee 0.0 0-0 0.0 0 ee 47.1 302.5 4097.0 429.6 D215 648.2 Use x 270953
TOTAL S560
ee -
II.
ING Aljayeiereye (eye eal yoyo)
CONFINEMENT SYSTEMS
The Confinement Systems subprogram is responsible for solving the experimental problems connected with
the confinement of fusion plasma by magnetic time confinement of high temperature plasmas
the plasma physics aspects of fusion reactor
The principal approach to the confinement of
time, moderate-density device In addition,
fields. The goal of this program is to demonstrate long at power-producing reactor conditions and to optimize
systems.
plasma is the tokamak which is a donut-shaped, long pulse
strong efforts are maintained in two other magnetic
confinement concepts. These are magnetic mirror systems, including both open and toroidally linked
mirror systems, and high density short pulsed systems, including the toroidal theta pinch, the
straight theta pinch and the Z-pinch. Each of these approaches is believed capable of contributing
to the major goal of a power producing, economic electrical power plant and/or to one or more of
several other possible applications of fusion, e.g., materials testing reactors, fusion-fission
hybrid reactors, fission product burners, production of fissionable material, etc.
The following sections present discussion of
problem areas for each of the three approaches, recent
achievements, program plans and budget requirements.
15
B. Problem Areas Each of the three confinement systems approaches has unique problem areas, and each consists of experi-
ments aimed at solving these problems.
1. Tokamaks The major problem areas of the tokamak physics program are: e Heating e Transport and Scaling @e Plasma Shape Optimization ® Impurity Control and Boundary Effects
e Fueling
Heating refers to the process of producing the plasma temperatures necessary for a fusion reactor. Fusion reactors require ion and electron temperatures of about 10 keV. Electron temperatures of 1-2 keV and ion temperatures of 0.5-1.5 keV are typical of today's plasmas. Initial heating is provided by ohmic heating, which is produced by passing an electric current through the plasma.
Additional heating by other methods is required to raise the temperature to that which will be
16
required for a reactor. Two methods are being tested: injection of beams of energetic neutral atoms and application of radio frequency power. Fusion reactors may require beam or rf powers of ~ 100 MW. Present day neutral beam experiments are performed with injected powers of ~ 0.3 MW, and a 4 MW beam system is being built for the Princeton Large Torus (PLT). The Tokamak Fusion Test Reactor (TFTR), scheduled for operation in 1981, will have an injected beam power of ~ 20 MW and will have some plasma compression capability as well. Heating with RF power is being tested
at the 0.2 MW level in advance of a decision to proceed to higher power levels.
Transport and Scaling refers to the development of the physical laws which describe the measured transport of plasma energy in present experiments and the development of scaling laws to predict plasma behaviorin larger, higher temperature devices. This area is, therefore, closely related to the heating program, and research on the two is conducted simultaneously. Fusion reactors are expected to have plasma radii of about 2-4m and plasma currents of about 10 MA. PLT, having begun operation in December 1975, is designed to operate with a plasma radius of ~ 0.5m and a plasma current of ~ 1 MA; it will therefore provide an operating point between existing smaller plasmas and those of a reactor. Present theory predicts a significant change in the plasma transport as electron temperatures increase above 1-2 keV. The PLT experiment will explore
the physics of this important regime. TFTR will have a plasma radius of ~ lm and a plasma
current of ~ 2 MA. (L7
Configurational Stability or Plasma Shape Optimization addresses the possibility, predicted by theory, that non-circular plasma shapes can be confined by lower strength magnetic fields and thus lead to lower fusion power plant costs. The techniques required are in use today on the Doublet IIA experiment, and definitive tests are scheduled on reactor grade plasmas in the Doublet III, beginning in 1978. Slightly elongated plasmas can also be studied on the Poloidal
Divertor Experiment, beginning also in 1978.
Impurity Control and Boundary Effects refers to problems resulting from the interaction of the plasma with its material boundaries. These interactions can result in an influx of non-hydrogenic (impurity) atoms into the plasma, possibly cooling the plasma core directly and/or cooling the plasma edge, causing the plasma to shrink and become unstable. One method of reducing boundary effects will be tested in the Poloidal Divertor Experiment in which additional magnetic fields near the plasma edge will carry escaping particles away from the walls into special pumping
regions.
Fueling refers to problems associated with replenishing plasma fuel in reactors with long burn
times. This is a long range problem, but initial experiments are planned on PLT, PDX and ORMAK.
18
2. Magnetic Mirror Systems In the Magnetic Mirror program, the major areas of investigation are open systems (minimum-B
configurations) and toroidally-linked mirrors (EBT).
a. Minimum-B Mirrors
The major problem areas for the minimum-B (open) configurations are: e confinement scaling @e Q-enhancement
e steady-state operation
Confinement scaling refers to the dependence of the confinement time T, or more generally the Lawson parameter nt, on experimental variables such as the amount of warm plasma stream necessary to stabilize the drift cyclotron loss cone (DCLC) mode, the plasma dimensions (in units of the ion gyroradius) R/p, and L/o,> the plasma beta, the angle of neutral beam in- ay : Sy) pepsi
jection, etc. Classical theory predicts nt ~ qT, in the absence of a stabilizing stream,
as is predicted in the case of large experiments (R/p; z 40). This prediction needs to be
tested experimentally at ion energies = 50 keV. Present experiments are operating at ion
energies up to 13 keV and with R/p,; ~ 2-3. MX will have ion energies ~ 50 keV and R/o,; uo
ibe)
Q-enhancement refers to methods for improving the power balance in mirror fusion reactors (Q is a plasma quantity, defined as the ratio of thermonuclear power output to neutral- beam heating power input). Recent mirror reactor designs have been based on classical values of Q in the range of ~1.0-1.1. These low Q values require stringent reactor engi- neering measures to yield net electrical power. A factor of 2-3 or more improvement in Q
would greatly ease the technology requirements and reduce capital costs for mirror reactors.
Steady-state operation refers to the achievement of at least multi-second, high throughput vacuum pumping capability, neutral beam sources, and neutral beam power supplies. The multi- second operating regime is important because on this time scale the neutral particle reflux from the walls is predicted to reach an equilibrium rate. The ultimate goal of mirror fusion
reactors is steady state operation.
b. Toroidally-linked mirrors (EBT)
The major problem areas for EBT are: e plasma stability e microwave heating e confinement scaling
20
Plasma stability refers to the stability of the complex EBT confinement configuration of a
toroidal loop within annular rings. MHD stability has been observed in EBT with low
toroidal plasma density (2-6 x ingore,
densities (~ Loe) needs to be investigated.
). However, the stability of projected operating
Microwave heating at millimeter wavelengths (ECR) is the method proposed for plasma heating in EBT. A clear understanding of ECR heating at higher densities (> Orr ey is thus an
important goal of the program.
Confinement scaling refers to the dependence of confinement time on plasma parameters (e.g., density, beta), the frequency and power of the microwave heating source, and the aspect ratio of the device. Ina steady-state device, such as EBT, the particle and energy con- finement times will depend on equilibrium transport properties. The critical-density limit set by the microwave frequency is important in this regard, since even at 120 GHz (n, o~
14 =3 : P 2x10 cm), rather long energy confinement times (~ 1 sec) are required to exceed the Lawson criterion.
21
3. High Density Systems In the High Density Systems Program, the major areas of investigation are the theta pinch and the Z-pinch. Theta pinch research is concerned with both linear and toroidal devices. Z-pinch
research is presently conducted only in tori. a. Toroidal Theta Pinch
In the toroidal theta pinch program, the major problem areas are:
e plasma confinement
e staged heating
Plasma _ confinement refers to the problem of creating a stable toroidal equilibrium for the theta-pinch-like plasma column in Scyllac. Equilibrium requires that the plasma column be formed initially in an approximate force balance near the center of the toroidal discharge tube. The plasma column should have the appropriate equilibrium surface distortion required by the presence of higher order multipole fields (2 = 0,1,2) which, in addition to the usual theta pinch magnetic field Bo? provide the toroidal force balance. Stability requires that
the equilibrium plasma, once formed, remain confined in spite of small perturbations in
22
position. In the Scyllac experiments, this stability is achieved by means of a fast feed- back stabilization system which drives % = 2 multipole windings. A more efficient stabili- zation technique, wall stabilization, is planned for theta pinch reactors, and will be
tested in the Staged Theta Pinch (STP) experiment and on Staged Scyllac.
Staged heating refers to the separate application, or staging, of the two phases of heating in theta pinches: implosion (or shock) heating and adiabatic compression. Projected toroidal theta pinch experiments, such as Staged Scyllac and LSS, require separation and control of these two heating phases to achieve greater implosion heating and less adiabatic compression. The resulting 'fat'’ plasma has a large ratio of plasma radius to wall radius, which is both economically advantageous for reactors and essential for effective stabiliza-
tion of the m= 1 (sideward) mode by the wall.
b. Linear Theta Pinches
In the linear theta pinch program, the major problem areas are: e end loss
e high field operation
23
End _ loss refers to the loss of both particles and energy (in the form of heat conduction) from the open ends of linear systems. Without some form of end-stoppering, plasma ions
will stream out of the ends at roughly the ion thermal velocity. Without steps to correct this situation, a fusion reactor based on the linear theta pinch would be impractically
long (many kilometers). A variety of flow barriers, both material and electromagnetic, are currently under investigation. Axial thermal conduction by electrons along field lines is
a potential source of heat loss to the central plasma column, even in the absence of particle
end loss, and studies to reduce this effect are in progress.
High field operation refers to the practical necessity of operating a linear fusion reactor at rather large values of the magnetic field. Since reactor length scales as Bas) an in- crease in the magnetic field from, say, 50 kG to 500 kG, can yield dramatic reductions in
reactor length requirements. o, 2=finch In the Z-pinch program, the major problem areas are:
e heating
e profile optimization
24
Heating refers to the process of achieving plasma temperatures relevant to a fusion reactor, i.e., > 5 keV. Z-pinches are, in principle, capable of being heated to ignition by joule heating alone, without the addition of auxiliary heating techniques such as neutral beams or r.f. Shock heating may also play an important role in reaching ignition temperatures in
Z pinches and is being studied.
Profile optimization refers to the tailoring of pressure and magnetic profiles to achieve MHD stability in the Z-pinch. Non-ideal MHD mechanisms, such as plasma transport in the form of diffusion and heat conduction will modify the programmed profiles. Confinement time, which has been diffusion limited in ZT-1, is predicted to scale with the square of
the minor radius.
Summary
A summary of the critical problem areas and the key experimental programs addressing these problems
is shown in Figure II-1. Note that many of the experiments are designed to address more than one
critical problem.
25
Figure II-l
Critical Problem Area
Tokamaks
e Heating
e Transport and Scaling
e Plasma Shape Optimization
e Impurity Control and Boundary Effects e Fueling
Minimum-B Mirrors e Confinement Scaling
e Q-enhancement e Steady-state Operation
Toroidally-Linked Mirrors
@e Plasma stability e Microwave heating e Confinement Scaling
Toroidal Theta Pinch
e Plasma Confinement e Staged Heating
Linear Theta Pinch
e End Loss e High Field Operation
Z-Pinch
e Heating e Profile Optimization
Key Experiments
ORMAK, PLT ORMAK, Alcator, PLT
Doublet IIA, Doublet III, PDX Aléator, JSXk; PDX PLT, PDX, ORMAK
2XSTIBREBES, Lae 2X-IIB, BB, LITE, FRTP BB, LITE
EBT-I EBT-I EBT-I, EBT-S
Scyl lac; ir STP STP, IHX
Scylla IV-P, Scylla-IC Scylla IV-P
TAH hy VAS Hives Vetkens
Critical Problem areas and Key Experiments addressing these problems in the Confinement Systems
Program.
26
C. Recent Achievements In the past year, important progress has been made toward achieving the plasma conditions and gaining the understanding of observed phenomena necessary for the development of fusion power. These achieve-
ments are outlined below, by confinement approach and problem area.
1. Tokamaks
Heating. Major progress in the area of tokamak heating has been achieved with two methods:
1) injection of energetic neutral beams of hydrogen and 2) heating with radio-frequency waves. Heating by neutral injection was demonstrated at 30 kW and 60 kW power levels during FY 1973 on the Adiabatic Toroidal Compressor (ATC) at PPPL; in August, 1974, neutral beam power of 100 kW was injected and the ion temperature increase continued to scale with injected power;
in August 1975, 250 kW was injected, resulting in an ion temperature increase of up to 300 eV.
In FY 1974 neutral beam injection was demonstrated on the Oak Ridge Tokamak (ORMAK) at ORNL, at the 200 kW power level. These experiments demonstrated some of the subtle effects of particle orbits on injection heating. The effects were found by comparison of the results
from experiments in which neutral beams were injected in the same direction as the plasma
27
current and in the opposite direction (co- and counter-injection). The experimental observa- tion stimulated theoretical analysis leading to improved plasma simulation codes. (The codes are predictive models to aid in designing future experiments.) More recent experiments at higher toroidal magnetic field and higher plasma current have confirmed the predictions of this theoretical work and injecting 250 kW against the current now results in ion heating, in agreement with theory. ORMAK is presently operating with a total injection power capa- bility of 300 kW, about half the ohmic heating power. The result is a trebling of the ion
temperature, to 1.5 keV, which is even higher than the electron temperature.
Significant heating with radio-frequency waves was achieved during FY 1974 on the Symmetric Tokamak (ST) at Princeton. Power levels of 100 kW were coupled into the plasma with 90% efficiency and resulted in a doubling of the ion temperature (to 230 eV). A similar system
was used on ATC at the 200 kW level with corresponding results.
Transport and Scaling. The area of tokamak physics transport and scaling addresses the physics questions relating larger plasma size, higher magnetic fields and higher currents to plasma
temperature, density and confinement time. On ORMAK a modification to the magnet cooling
28
system has permitted the device to operate at approximately 25 kG (an increase from 18 kG). This higher field has allowed ORMAK to operate at higher current and has provided some improvement in the effectiveness of the injection heating as discussed above. A new power supply recently
installed will make possible operation at ~ 30 kG.
In this area, however, the major experimental device is the Princeton Large Torus (PLT) which began operation in December 1975. In early experiments operation at 600 kA for one full second has been achieved at 35 kG, with electron temperatures exceeding 2 keV. A decision to operate at fields up to 50 kG will be made by early 1977. A 4 MW neutral injection system being built at ORNL will be installed during the period Nov. 76 through Mar. 77 and will allow studies of
plasma confinement at ion temperatures in excess of 2 keV.
The Alcator at MIT has operated with high toroidal magnetic field (up to 70 kG with a plasma; up to 100 kG without plasma) and large plasma currents (up to 200 kA) in a relatively small plasma cross-section (9 cm radius compared to PLT's 45 cm). The resulting large current densities have produced no deleterious effects on plasma confinement and, in fact, have allowed high density and relatively high temperature operation. Experiments are now being carried out over a wide density
i range, from less than 10%; to greater than 10 : Per TiGteeye tin the observed linear dependence of
*es)
confinement time on density has been confirmed with experiments on ORMAK, ATC and Pulsator at the Max Planck Institute for Plasma Physics in Garching, Germany. Operation at 75 kG has re- sulted in a density-confinement time product (nt) of pobecut aes at a temperature of 1 keV.
(For a reactor an nt of 3 x AG den csee is required at a temperature of 10 keV.)
In the French TFR, plasma currents of 400 kA have been obtained with a toroidal magnetic field of 60 kG, resulting in 3 keV electron temperatures and 1 keV ion temperatures. Neutral injec-
tion of 450 kW has increased the ion temperature to between 1.5 and 2.0 keV. An upgrade of this device to the 600 kA current range is planned for 1976.
In general, confinement time increases with plasma density and in any given tokamak the upper limit on density seems to be given by the total power density, be it from the plasma current,
neutral beams, or some combination thereof.
Plasma Shape Optimization. In the area of tokamak plasma shape optimization, experiments are carried out using non-circular plasma shapes in the expectation that vertical elongation of the plasma will result in a reduction of the required toroidal magnetic field strength and will ultimately lead to lower fusion power plant costs. By October 1973 preliminary results on
Doublet II at General Atomic Company indicated that a magnetic field one-third that required
30
for tokamaks with circular plasmas was sufficient to obtain similar plasma parameters in a kidney- shaped "doublet" plasma. This device, now called Doublet IIA, has been modified to allow the production of elliptical and circular plasmas as well as doublets. Initial attempts to center
the plasma and stretch it to a two-to-one ellipse have been successful; a comparison of circular and 1%-to-1 ellipses indicates that an improvement in confinement is achieved. In 1976, effort
is concentrated on a systematic study of the doublet configuration. A new power supply will
allow operation at 15 kG (cf 7.5 kG at present) by the end of 1976. DIIA is providing data and experience used in the design and planning for Doublet III, now in fabrication and scheduled to
operate in early 1978 as a major test of non-circular cross-section physics.
Impurity Control and Boundary Effects. Experiments in the area of impurity control and boundary effects address the problems of the origin of impurities, i.e., the interaction of the plasma with its boundaries, as well as the transport of impurities within the plasma and the detection
and control of impurity levels.
In ATC, the volution of impurities from the walls and limiters has been substantially reduced during the initial plasma buildup phase by evaporating titanium to coat the boundary surfaces before each plasma discharge. The effective Z, or average ion charge (it would be unity for
a pure hydrogen plasma), is reduced from twoto x l.
31
In Alcator, careful attention to high-vacuum techniques and high-power discharge cleaning seems
to provide clean boundary surfaces and result in clean plasma discharges.
There is also the problem of preventing diffusing plasma and neutral particles from interacting with the boundaries during longer plasma discharge times. The magnetic divertor, in which judiciously arranged magnetic field lines near the walls sweep charged particles away into pumping regions, is one solution to the problem of impurities in a "steady state'' discharge.
The FM-1 demonstrated divertor action effectively in a low-temperature plasma, with 80-90% of the plasma diffusing toward the walls being swept into the divertor region. These experiments confirmed the theoretical basis for the design of PDX, now in fabrication and planned to operate
by spring 1978 as a major test of divertor physics.
Fueling. A longer range problem is that of fueling. A steady state reactor requires that the DT fuel be replenished as fast as it is lost through fusion interactions and plasma losses. Prospective fueling methods include gas puffing with neutral beam injection and pellet injec- tion. Very preliminary experiments on ORMAK indicate that pellets do penetrate into the plasma. Further experiments, on ORMAK, PLT, and especially PDX, where recycling of gas from the walls
should be greatly reduced, will be very important to the tokamak reactor program.
52
2. Mirror Systems
The 2X-IIB and BB-IIT experiments at the Lawrence Livermore Laboratory and the LITE experiment at the United Technologies Research Center constitute the major elements of the open-ended mirror- confined fusion plasma research program. The EBT experiment at the Oak Ridge National Laboratory
is the major experiment in the toroidal mirror-confined fusion plasma research program.
In July 1975 the 2X-IIB experimental program successfully achieved the near-term goals assigned to it in November 1972 (see WASH-1299): high density (e = wg! e4 >\100)2?nT = 167 -enee sec, T, 2
10 keV, and a plasma sustained by neutral beams. Some of the important performance figures are
shown below.
2X-II Performance 2X-IIB Performance Parameter October 1972 Uy eon a aeeerot- Pecos ao T, (keV) 6 13 T (ms) 0.4 zi nt(em > + 8) ag 7x 10°?
33
In October 1975 it was demonstrated in 2X-IIB that the cold-plasma stream used to stabilize the plasma provides a suitable target plasma for build-up of the hot plasma by neutral injection. This demonstration of build-up without the previously used pulsed compression to generate the target provides a solution to a major technical problem for mirrors, i.e., plasma start-up in a
steady-state magnetic field.
In February 1976 the density in 2X-IIB was further increased to 1.2 x LOpIcHhe.s and the peak
value of beta (8 = 87 nt, /B°) was found to exceed unity, where B is the central vacuum field:
B le 2eton lL. 6
peak
B
oe 0.7 (diamagnetic loop).
Perhaps more significant is the fact that the 2X-IIB results are quantitatively consistent with the theoretical prediction that warm plasma would suppress the dangerous drift cyclotron loss cone (DCLC) mode. These results are also in agreement with observations from the PR-6 and PR-7 experiments in the Soviet Union. The experiment has confirmed that nt increases with T, - Present confinement times are limited by the low electron temperature which is believed to be due to losses on the streaming plasma. Theory suggests that higher electron temperatures and
hence larger nt should result from increased plasma size.
34
The theoretical stability questions for conventional mirror systems now seem to be well delineated. Three basic modes of instability are important---drift cyclotron loss cone (DCLC), convective loss cone, and Alfven ion cyclotron (AIC). A good base of theory appears to exist for all of these (studies in the AIC mode behavior are somewhat less developed than the others). Nevertheless,
the problems posed by these modes are complex enough that only experiment will determine the precise compatability with reactor parameters. To test this scaling, a physics experiment sub- stantially larger than the 2X-IIB facility appears necessary. The MX device is projected for
this purpose, and is scheduled for operation in 1981.
The BB-IIT experiment is investigating both a means of starting up a mirror machine by neutral beam injection in a steady-state magnet field, and confinement under steady-state conditions. In April 1976 two-sided laser irradiation of 100u diameter NH. pellets injected into the BB-ILIT chamber was achieved using a 300J co, laser. Near term plans are in progress to supplement the 50-amp, 20-keV, 10-msec neutral beam line with three identical units for start-up experiments
and two 35-amp, 50-keV, 50-msec beams for plasma sustenance experiments.
33
The LITE experiment is designed to address the target plasma buildup approach to a steady-state mirror fusion reactor. In September 1975 they demonstrated electrical capture and suspension
in hard vacuum of 100u LiH target pellets and began two-sided pellet irradiation experiments with a 100J Nd-glass laser. A Heavy Ion Beam probe is being developed to measure the plasma potential. This measurement requires a knowledge of the complex Yin-Yang magnetic field geometry and will
produce both spatially and temporally resolved measurements.
The EBT device at ORNL is a non-axially symmetric torus composed of linked magnetic mirrors. EBT has produced a continuously operating stable plasma. The application of modest amounts of micro- wave power (30 kW) has produced electron temperatures of 300-400 eV. In November 1975 EBT received a very favorable review. The microwave heating power was recently doubled, and experiments are
in progress. Experiments designed to test the scaling of the EBT concept with power and size are
currently projected for EBT-S and EBT-II.
3. High Density Systems
Feedback experiments began in December 1975, on an 8.4-m (20. arc), derated (reduced magnetic
field and temperature), sector of the Scyllac Torus at LASL. Scyllac is designed to investigate
36
plasma equilibrium and stability in a toroidal theta pinch. Full torus experiments have pre-
viously demonstrated plasma production and heating, in the presence of toroidal curvature, and 1 as
£ = 1,0 equilibrium magnetic fields, to (T, + T;) ~w 1.3 keV, densities of ~ 3 x 10 poe! 3 and
beta values in the 0.5-0.9 range.
Confinement times of 8-10 us were achieved in the full torus, being limited by plasma-wall contact. The need for external feedback fields to control this plasma motion has been recognized as an important element to the Scyllac program since its beginning. A high-power feedback system em- ploying optical detection has been developed and installed to overcome this problem. Results with a smooth-bore tube show success in stabilizing plasma motions perpendicular to the major radius, less success in stabilizing plasma motions along the major radius (in or out). The problem in the toroidal plane is attributed to an imperfect initial equilibrium, and steps are being taken to remedy this problem in future Scyllac experiments. These steps include the use
of helically shaped, toroidal quartz discharge tubes, and £ = 1,2 equilibria which do not involve the axially-dependent % = 0 magnetic field component used in earlier experiments. The plasma confinement time in the smooth-bore tube feedback experiments has been extended to the 25-30 us
range.
o/
Scylla IV-P is a 5-meter linear theta pinch capable of producing reactor-grade plasmas (T, ~ several keV, n ~ few x tee 8B ~ 1). Scylla IV-P began operation in January 1976 at LASL, and will be used to study specifically linear physics problems, such as particle end-loss, axial thermal conduction, and high field operation. Initial experiments are concerned with time resolved measurements of the plasma parameters along the length of the device. End-
stoppering plans call for the study of material end plugs during 1976 and multiple mirrors
way LL TH
ZT-S is an axisymmetric, high beta, toroidal Z-pinch experiment at LASL whose primary mission
is to investigate the scaling of plasma confinement time as a function of minor radius (compared
to ZT-1) and current density. It is an attractive alternative to the tokamak with MHD stability
at 8 ~ 40% and correspondingly high power density. ZT-S is a modification of ZT-1 to incorporate a larger minor radius and improved electrical circuit. The Z-pinch program received a favorable
review in December 1975. The ZT-S results will be reviewed in July 1976 to determine the
future course of the Z-pinch program. The larger ZTI-P and ZT-II experiments are currently
projected as follow-on experiments to ZT-S.
38
D. Program Structure and Major Milestones
This section discusses the general structure of the three confinement systems programs, as pres- ently conceived for the next five years. The major milestones of the program are presented below; detailed budgets and milestones for each project are given in Volume IV. Budgets consistent with
these programmatic aims are summarized in the next section.
1. Tokamaks
The tokamak program is the most highly developed of the Confinement Systems Programs. The anticipated contribution of each of the tokamak devices to each problem area is shown in the flow chart of Figure II-2. The five critical problem areas now being addressed are shown
across the bottom of the chart. These problem areas subdivide into problem subareas as follows:
© Transport and Scaling @® Heating Scaling with size and temperature Compression Scaling with density Injection RF e Configurational Stability ® Boundary Effects Current Distribution Surface Interactions Non-Circular Cross-Section Divertor e Fueling
Gas Blankets Pellet Injection Se)
KEY
Ootme’
CALENDAR YEARS
SS
2a S woe foo} m = LW oa je)
WAN KK . \ q SO \ SY MAO MOY Qn DWDM PES \ WN
CQQH
Wy
P L a U Pp G R A D E ZG j y y Z Z Z | ]
Energy May 1976
netic Fusion En
e DIVERTOR EFFECTS PREPARED BY THE US ERDA Oo Mag
BOUNDARY
SURFACE INTERACTIONS
GAS Walia BLANKETS FUELING
Figure II-2 Tokamak Flow Chart
CURRENT DISTRI BUTION
rie ale A saa BN ° PT Se Bd SONNE
CROSS SECTION ONFIGURATIONAL STABILITY
O aE: Re
<<) Ea
qr-O
TOKAMAK SYSTEMS
——__ en aor
TD oeSay saGedou a
|
|
SIZE AND
qsaodrtoOc oO
TRANSPORT AND SCALING
ESL TEMPERATUR
[n= | '
Slim SO jie
40
Major milestones for the Confinement Systems tokamak program have been established as follows:
Program Area
Heating - Neutral Beams
Heating - RF
Transport and Scaling
Milestones
Inject 100 KW into ATC (PPPL)
Inject 250 KW into ATC (PPPL)
Inject 500 KW into ORMAK (ORNL)
Operate 4 MW injection system on PLT (PPPL) Operate 2 MW injection system on ORMAK Upgrade (ORNL)
Operate 4 MW injection system on DIII (GA) Operate 30 MW injection system on TFTR (PPPL)
Test 200 KW of RF in ATC (PPPL)
Decide whether to proceed with fabrication on 2.5 MW RF supply for PLT (PPPL)
Test 2.5 MW of RF on PLT (PPPL)
Begin operation of ORMAK at higher field (~ 25 kG) with higher installed neutral beam power (~ 200 KW) (ORNL)
Begin operation of PLT (PPPL)
Evaluate operation of Alcator at high field (# 90 kG) (MIT)
Begin operation of ORMAK at 33 kG with higher plasma current (~ 200 kA) (ORNL)
Evaluate plasma results from PLT (PPPL)
Evaluate 33 kG operation of ORMAK (ORNL)
Evaluate 4 MW heating experiments in PLT (PPPL)
41
Target Date
Aug. 1974 (achieved)
May 1975 (achieved Aug. 75) Dec. 1976
March 1977
Och, 1970 July 1979 July 1981 May 1975 (achieved)
Sept. 1976 July 1979
Dec. 1974 (achieved) Dec. 1975 (achieved)
Janel e7 Jul veuk9 7.6 Sept. 1976 Feb. 1977
Sept. 1977
Problem Area Milestones Target Date
Plasma Shape Optimization 1. Evaluate plasma shape control in Doublet IIA (GA) July 1975 (achieved, Aug. 75) 2. Evaluate results from Doublet IIA (GA) July 1976 3. Evaluate current distribution in PLT (PPPL) Sept. 1976 4. Operate Doublet III (GA) Feb. 1978 5. Evaluate Doublet III results (GA) Dec. 1978 6. Evaluate stability in D-shaped plasma in PDX April 1979 Impurity Control and Boundary Effects 1. Decide whether to proceed with ISX (ORNL) July 1975 (achieved) 2. Evaluate Impurity Effects in PLT (PPPL) Sept. 1976 3. Operate PDX (PPPL) April 1978 4, Evaluate impurity control in ohmically heated PDX (PPPL) Dec... 1978 Fue Ling l. Test first generation pellet fueling device on ORMAK Sept. 1975 (achieved) 2. Test 200ym pellet injector from U. of Illinois on ORMAK Nov. 1976 3. Test 3 x 10°cm/sec pellet injector on PLE June 1978 4. Operate pellet injector on PDX June 1979
2. Mirror Systems The outstanding success of the recent experiments in 2XIIB,which have investigated start-up,
stabilization, heating, and fueling of mirror systems by neutral injection into a target
42
plasma, have increased the fundamental understanding of mirror confinement physics. The mirror program underwent three intensive reviews in the first half of 1976. The Mirror Senior Review Panel reviewed the program in March 1976 (ERDA 76-74) and recommended an expanded national program in mirror research with emphasis on Q-enhancement experiments, and construction of an MX-scale device for confinement scaling studies. The Fusion Power Coordinating Committee (FPCC) reviewed the program in April 1976, and issued the following
position statement:
"]1) The national magnetic mirror program should be strengthened. The program should build vigorously on the positive results of the past year, to enhance and accelerate progress towards the national goal of producing a practical pure fusion power reactor.
"2) A magnetic mirror device with the general goals of the proposed MX should be authorized as a line item in the FY 1978 budget, with an initial operating date of 1981. A review panel should be convened to consider the MX proposal and define, with LLL participation, the appropriate device, considering both technical objectives and cost, required to clarify the physics basis for mirror fusion reactor designs, as the next logical step in research on this confinement method. The recommended device should have sufficient flexi- bility to allow Q enhancement experiments.
43
'"3) A program plan for increasing Q in magnetic mirrors should be prepared. This plan should include testing, with high priority, under leadership of LLL,
methods for increasing Q in existing devices or in new devices of modest size,
coupled with substantially increased theoretical work." A technical review panel was convened in May 1976, at LLL to conduct an engineering review of the MX proposal, address general questions of Q-enhancement, and define the nature of the next major step in the mirror program. The project data sheets which are given in Volume IV for the mirror program represent an expanded funding level over that supported under the previous "Open Systems" program, and reflect an increased national involvement. Individual milestones for existing experi-
ments, 2X, BB, and LITE in particular, will be subject to further changes as the national mirror
program plan evolves with added emphasis on Q-enhancement studies.
Major milestones for the magnetic mirror program have been established as follows:
Problem Area Milestones Target Date
Minimum-B Mirrors
Confinement Scaling 1. DMFE decision on MX proposal July 1976 2. Extend nq (Ty) scaling from 13 keV
to 20 keV in 2XB Oct. 1976 3. Increase radial dimension of plasma
from R/ 9; ~ 2-4 to 5-10 in 2XB March 1977
4. Begin plasma experiments in MX Sept. 1981
44
Minimum-B Mirrors (continued)
Q-enhancement le Steady-State the Zs
Toroidally-Linked Mirror
Plasma stability, micro- wave heating and con- 2 finement scaling
tr
Go
3. High Density Systems
Milestones
Develop DMFE program plan for Q-enhancement investigations
Preliminary evaluation of quasi-steady- state (2 0.5 sec) operation in BB mirror system
Begin plasma stabilization investigations in high density, injection-sustained LITE plasma
Evaluate results on EBT-I Begin plasma experiments in EBT-S DMFE decision point for EBT-II
Larget Date
Oct. 1976
March 1978
June 1978
April 1977 August 1977 Sept bo/7
August 1977 is the DMFE decision point on whether to proceed further with the toroidal
theta pinch program at LASL.
This decision will involve a review of results on plasma
confinement and staged heating experiments on Scyllac and STP, and an evaluation of
future prospects for this high density system.
If an expanded effort for the toroidal
theta pinch program is approved, the lead experiment at LASL will be Staged Scyllac,
which is envisioned as the proof-of-principle experiment for the toroidal theta pinch
45
concept. The toroidal Z-pinch and linear theta pinch experiments would serve in the capacity of back-up concepts, and would be funded accordingly. If the toroidal theta pinch program should be terminated, the program emphasis at Los Alamos would shift to the Z-pinch line of
experiments, with the linear theta pinch approach continuing to serve as a back-up.
The project data sheets which are given in Volume IV for high density systems reflect scenario 1 (i.e., the toroidal theta pinch-Staged Scyllac - line of research) is the lead program at LASL. Detailed project data sheets and program plans have also been prepared for scenario 2 (Z-pinch is the lead program and the linear theta pinch is the back-up program); however, project data sheets for this alternate program scenario are
not given in Volume IV.
Problem Area Milestone Targets Dace
Toroidal Theta Pinch Plasma confinement 1. Evaluation of feedback stabilization in Scyllac and initial wall stabili- July Lone.
zation in Staged Theta Pinch
Staged Heating 1. Evaluate staging experiments in Staged Theta Pinch
46
Problem Area Milestone Target Date
Linear Theta Pinches
End loss 1. Evaluate end-stoppering experiments in Scylla IV-P Jan. 1978 High field operation 1. Begin high field operation in Scylla IV-P Jan. elL978 Z-Pinches
Heating and profile
optimization 1. Review of ZT-S results and ZT-P proposal Sept. 1976
4. Future Experiments a. Tokamaks Two experiments not discussed above are expected to be proposed in FY 1977-78. At Princeton Plasma Physics Laboratory a small toroidal device is under consideration. It would be specially designed to provide a controlled and localized ripple in the toroidal field which would be large enough (70-80 kG) to provide a high density (3-6 x 10!4¢em73) plasma. This device would then be used to test vertical neutral
beam injection, or ripple injection, a concept which offers the possibility
47
of beam penetration and heating at lower beam energy and with less access than
required for present-day tangential injection systems.
At General Atomic a small superconducting tokamak is under consideration. Its focus would be rf heating and it would also test the possibility of driving a
steady state plasma current by use of rf power.
Dae MiBnor Systems
The major experiments projected in the mirror program are MX at LLL, MRTX at
UTRC and EBT-II at ORNL. The major objective of the MX experiment is to provide
data on open systems confinement scaling laws in an nT range near 1012em™3sec.
A secondary objective of MX, and the primary objective of MRTX, is to extend the
operating regime for mirrors into the multisecond regime to reach an equilibrium
state for wall-reflux neutrals. The EBT-II device is intended to provide heating and confinement information for toroidally linked mirrors in a significantly
larger facility than EBT-I.
48
c. High Density Systems
Under the toroidal theta pinch scenario, the primary experiment planned at LASL
is Staged Scyllac, which involves an ~ $10M modification of the existing Scyllac
bank. For the Z-pinch program, the ZT-P and ZT-II experiments are projected.
If end-stoppering research is successful on Scylla IV-P, a Long Linear Experiment §)
(LLX) is planned, which will extend the confinement physics into the nt~ 1013cm7~’sec
range.
E. Budget Summary
The budgets necessary to achieve the objectives discussed in Section D are displayed in Figure II-3.
49
A. Operating
Confinement Systens War. oe. s
Tokama Sie een PPPL
PLT Normal Op.
Pit habe PISTSNB Merit ees lib due oad oo PDX Normal Op
PDX? Fa bi.) gaeeer kel aia «beso esate
PDKXeNB # race PDX Upgrade .
coe ee eee eee oo
oe eee eee eo ee oO
eee ee ee eo eo eo we ow
oe eee eee eee ee
ORMAICONOTMma Ui Op. gl. cicjetnin
ORMAK Fab. . ORMAK Upgrade ORMAK Upgrade ORMAK Upgrade ORMAK Upgrade
ISX Normal Op.
eS Xeehia Dene: SKE RR aor cate PLT NB Lines
coor oe ee eee eo
Fab. a. eae
eceeve eee oo eee eee ee ew © 8 8
coe eee ee ee eo eo oo
CONFINEMENT SYSTEMS BUDGETS
LOZOE 19,300
135230
1,900 0
300
0 1,350 2,700
650
Dollars in Thousands
Fiscal Years
1907 1978
79,600 110,000 57,400 | 70,000
8,100 | 11,400
0 0
250 0
2,300 3200
200 10,000
55.00 2,200
650 2 5200
0 0
0 0
2,700 0
0 0
1,500 4,500
3,300 aL OO
0 1,000
0 0
2000 2, 100
600 0
0 0
500 0 Figure II-3
50
1979 125,000
75,000
11,000 0 1,500 2,500 10,000 0
0 0 0
0 0 7,000 0 2,000 0 2,700 0
300
0
1980
140,000 80,000
11,000 0 3,500 0 10,000 0 5,000 3,000
1982
150,000
80 , 000
11,000 0 0 0 11,000
Qe) S=)
(2) @&)
7,000 4,000 3,000
0 0
Total 78-82
665,000 385,000
55,400 0 8,000 6,000 51,900 2,200 9,200 5,000 0
0 0
32, 500 3,700
3,000
8,000
14,400
2,300 0
GA Deter Oia LOD so vs le ce dies) 06 (DY Gabel ad SY eel ett, Mace RCRD Phe | 2 sald A pa Maat, by Sed WAN bil gli eye ots Ye ie era naar DEPAPNOLMAL “OD. ccc cs 5.6
De ye Ve Rees a PC ee CMa Outed Aa US LOM RAE, 55 Meee os eer atehes MIT ALeator A Normal Op. ~.--... Alecator7C Normal Op. %..... POC AWOGe Ce MACe. cass sce oe Alcator C Power Supply ... PUG COS Gm RE eg cle iie cs 6 6 9 ence 0 A UeCareOrmeC UD EDA Cn. 5 ogstsnale« Other Noma Op. 9 Ok a... - eee. sk. Be DRG Oak ba cpeddbhade ht +. apa ab bbsechs
MLFYOr SyStems coo. je weieeeis «ois Da IR. SS. PRAY Eee, Ge ae
DDC, QE Pea CREO Les OP
BB ee oe Apienie 6 wuss oe e+
LIZ ES Re 9s Aainoieerae ¢ 0 tae
Ae dae ee ce ea a
Dollars in Thousands Fiscal Years
0 0 0 0 2,600 600 400 0 100 50 1,500 350 0 500
200 250
0 0
0 0
0 0
0 0
0 0
10,900 3 , 560
5,700 1,560
0 0 0 150 2,150 500 0 475 0 0 1,650 425 0 0 0 0 1,400 350 0 100 0 0
Figure II-3 (continued)
avi
28,900 6,100
1,600 3,100 1,800 2,300 1,500 2,100 1,000 2,600 2,700 2,900 1,200
9,000 3,000 2,500
500
1,100 3,600 3,000 5,000 1,000
4,000 7,000
500
3,000
1,000
6,000 4,000
41,100
7,900 6,400
5,200 2,000
500 5,200 2,200 1,500 9,000 1,200
6,000 7,000
500
9, 500
10,000 5,000
40,700
3, 500 0 15,000 0 5,200 2,000 0 3,000 5,000 0 7,000 0
31,000 17,000 6,200 0
2,500
1,000 13, 500 1,500 2,500 2,000 11,000
25,000 19,000
184,000 28,800 1,600 42,600 1,800 21,900 9,000 4,500 12,900 16,900 7,700 32,400 3,900
Dollars in Thousands Fiscal Years
Total 1976 1976T 1977 1978 1979 1980 1981 1982 78-82 High Density Systems ........ 7,800 1290 8,15) LL5100 13 ,600 18,900 23,100 29,300 96,000 BOy LUA GC. siiy<ccuna aaks +o eet S\nho0 950 1,205 900 0 0 0 0 900 Staced ocv lL Lacan. tal ore 0 0 80 5,000 7,000 8,000 7,500 5,500 33,000 Large Staged Scyllac ..... 0 0 0 400 1,200 1,200 4,100 6,800 13,700 SES CeCe INE CASE TNC r.: acum. aes 1,400 300 1, 100 530 0 0 0 0 530 ZL aie, vekel ate ghadaader al caoh em cteicas larson 850 300 530 300 0 0 0 0 300 PAWN SER: ha ere ee; oR re 0 0 ROLE 2,140 2,300 2,500 1,000 0 7,940 Ek A aE as ete ve cs, ot Potabes cs 0 0 0 0 800 4,500 7,500 11,300 24,100 Scy bia We P ey os Sse oan wees 1650 415 1,670 P00 2,200 2,500 2,500 3,200 12), 100 LiL et aM aris aa 4 ea ee oo 0 0 0 100 100 200 500 2,500 3,400 Conceptual Studies ....... 150 25 60 30 0 0 0 0 30 B. Equipment Confinement Systems .......... teal 20 2 450 8,200 $5.60 17,400 Li 260 18,490 20 OU 88,660 OK Aims Meee ter on ee ete eters 4,458 1,400 5,800 82500 9,900 9,700 9,800 9,800 47,750 MLYTOURSVSEeMS i. .aets. . Avr. ts 2 S205 790 435 7,020 7,400 5,060 5,090 6,420 31,490 High Density Systems ....... i S047 260 965 1,190 1,600 2,500 3,100 4,130 12 20 C. Line Item Construction Dollars in Millions Confinement Systems .......... 0 0 0 SE.0 60.4 2 ak LOO Sid 2004 Tokaniaks ca" Pane et. BES 0 0 0 0 0 0 0 0 0 METrorioySteMmss ty. antes ote 0 0 0 Sly. 60.4 DSicah 0 0 Ld folk: MX CH BS SOC honor) Reet cele 0 0 0 31.0 48.0 A a8, 0 0 100.0 MRI, eR Sn ROR 0 0 0 0 1256 Git 0 0 iby ea High Density Systems ./..... 0 0 0 0 0 0 10.0 136 83:40 Large’ StagedMSeyl lac @aa..5 0 0 0 0 0 0 0 53956 39.6 Scyllac Fusion Test Reactor 2M ia 2 it. 3s AAG: 0 0 0 0 0 0 0 0 0 LongMLinear WExpte . << Aten gets 0 0 0 0 0 0 0 19.0 19.0 ZU Tele Oe Sc Bechtel aout chee ane 0 0 0 0 0 0 10.0 15,0 25.0
Figure II-3 (continued) 52
III. TECHNICAL PROJECTS OFFICE
A. Introduction
The Technical Projects Office (TPO) is responsible within MFE for the management of the design and construction of large complex line item fusion facilities, including development programs in direct support of the projects. During FY 1976 TPO has provided program management of the Tokamak Fusion
Test Reactor (TFTR) at Princeton Plasma Physics Laboratory. PPPL is the prime contractor for the design and fabrication of the TFTR device and the associated hardware, much of which will be engineered and fabricated in industry. The ERDA Princeton Area Office is responsible for administration of the project, including prime contracting for the conventional facilities. PPPL will operate the experi- ment and facilities when construction is complete. Within MFE, the Confinement Systems Subprogram
will manage the experimental program after completion of construction.
In addition, the Technical Projects Office manages the design and construction of facilities capable of producing intense sources of high energy neutrons required to determine the effects of high energy neutrons on reactor materials. The Rotating Target Neutron Source (RTNS) embodies known technology
and is scheduled for completion in FY 1978. The Intense Neutron Source (INS), authorized for funding in FY 1977, is scheduled for completion in FY 1981.
53
Conceptual design studies performed in the Development and Technology subprogram will identify specific goals and design criteria for future large fusion facilities. Responsibility for the final design and
construction will be transferred to the Technical Projects Office.
B. Tokamak Fusion Test Reactor
Pe sintroduction
The Tokamak Fusion Test Reactor (TFTR) will be the nation's first magnetic confinement fusion device to experimentally demonstrate the release of fusion energy from the deuterium-tritium reaction under conditions projected for future experimental power reactors. TFTR will represent an intermediate step between present, relatively small zero-power physics experiments and future experimental reactors planned for the mid-1980's. The TFTR will be located at the Princeton Plasma Physics Laboratory
(PPPL) near Princeton, New Jersey. The construction project should be completed in mid-1981 at a
total cost of $228M, including escalation.
2. Objectives
The TFTR has major objectives in both physics and engineering. The principal objectives are: a. To demonstrate fusion energy production from the burning of deuterium and tritium (DT)
in a magnetically confined toroidal plasma system.
54
b. To build a neutral beam heated tokamak in which hydrogen, deuterium and DT plasma can be inserted in order to: (1) Study the physics of large tokamaks, and (2) Verify advanced engineering concepts for DT tokamak systems, c. To experimentally demonstrate that sufficient physics and engineering understanding of large
fusion systems exists in advance of construction of an Experimental Power Reactor.
The TFTR will serve as an intermediate step to bridge the gap between current, relatively small, hydrogen plasma confinement experiments and the first Experimental Power Reactor. The unique features required are its DT burning capability, its size which permits physics experiments in the EPR range of interest and some of its engineering features, not heretofore tested. The experience to be gained in design, construction and operation, and the information to be gathered in physics and engineering will provide
a sound foundation for EPR design and construction.
The specific objectives of the TFTR project are: e Attain reasonable pure hydrogenic plasma conditions at 5-10 keV temperature, approximately
1 - 1 = 10 con ? density, and provide stable confinement with nT, equal to or greater than 10 ees rene
55
e Provide a neutral beam injection system capable of injecting into the plasma 20 MW of 120 kev D° beam, for at least 0.5 sec.
e Provide a toroidal magnetic field of about 5 tesla (on vacuum chamber axis), for at least 3-sec plattop time, with a 5-min. repetition rate.
@® Develop plasma handling techniques and provide hardware capable of initiation, control (including feedback control and major radius compression) and dissipation of tokamak discharges up to 2.5 MA.
e@ Provide a vacuum chamber of adequate size (2.7 m major radius and 1.1m minor radius), equipped for high-power discharge cleaning and capable of achieving base pressures below 5 x i0°° Torr.
e Provide capability for routine pulsed operation with H-H; D-D; Dene or DT plasmas, with safe
and reliable gas handling and support systems.
\6 Principal Design Features
The principal design features of TFTR are: Size
e 2./ m major radius
56
Plasma
e 2.5 MA maximum plasma current e 1.7 meter diameter
3 e Target plasma conditions nt ~ 10 ; qT, ~ 5 keV
Neutral Beam Injection
@ 120-150 kev
Magnets
e Water-cooled copper @ 50 kG centerline field
e Provision for adiabatic compression (c = 1.5)
Power Supply e 4500 MJ, 660 MW pulsed e AC motor/generator/flywheels
sy)
Special Features
e Tritium supply and containment systems @ Shielded magnets and auxiliary systems @e Remote handling capability
@e Neutron and plasma diagnostics
Energy Release
e 1-10 MJ with D-T
4. Project Milestones
Major milestones for the TFTR project are:
e Order long lead material (Toroidal Field coils) Dec. 1976 e Order Motor-Generator-Flywheel System Dec. 1976 e Start Site Construction Fepeeo i
58
@ Complete Tokamak System Final Design Review Aug 1977
e Initiate Fabrication of Vacuum Vessel May 1978 e Initiate Fabrication of Toroidal Field Coils Oct 1978 e Order Neutral Beam Systems Production Unit Apr 1979 @e Delivery of first MG set Oct 1979 e Complete TFIR Complex Construction Jan 1980 e Start Final Assembly of Tokamak May 1980 e Initiate Testing of Neutral Beam Line #1 Jul 1980 e Complete Site Construction Nov 1980 e Hydrogen Plasma Operation Jun 1981 5. Projected Budgets ($M) FY78-82 1976 1976A 1977 1978 1979 1980 1981 1982 Total Construction: (TEC $228.0) 15.0 ae 80.0 94.4 Saad! 0 0 0 PZT Operating (Total) 0.4 0.8 10.0 15.3 10.3 20.0 23.0 25.0 93.6 Component Dev. 0.4 0.8 ee) 1259 S10) 2.6 PaaS. 0 20.2 Exp. Research 0 0 0.25 1.6 2.8 5.6 6.2 10.0 26.42 Facilities Oper. 0 0 0.25 1.4 4.5 11.8 14.5 15.0 47.2 Equipment 0.15 0.05 1.0 2.0 Zed 2.9 ibys! 0 rein)
oe)
C, Rotating Target Neutron Source
1. Introduction
The Rotating Target Neutron Source (RINS-II) will be the first high-energy, high-intensity neutron irradiation facility dedicated to the fusion reactor materials program. The RINS-II will provide the neutron sources and support facilities required to provide the "pure" 14 Mev neutron energy component necessary as a base line for displacement and transmutation studies. This facility will be located at the Lawrence Livermore Laboratory (LLL), Livermore, California. LLL is the prime contractor for the design and fabrication of the facilities. LLL will operate the facility when completed in 1978, at a total construction cost of $5M, including estalation. The ERDA San Francisco Operations Office is
responsible for administration of the project, including cost, schedule and technical monitoring.
2. Objectives This 14 MeV source is intended to provide: e high energy damage information at low fluences required for verification of theories of fission data extrapolation; e surface and defect data for comparison with high energy spectra such as those arising from the Be(D,n) and Li(D,n) stripping reactions, which have tentatively been identified as possible reactions upon which to base a higher intensity, larger volume neutron source; e cross-section measurement data; e synergistic effects on materials due to the interaction of neutron damage with other plasma radiation; and
e comparison with ion simulation. ap
The design requirement and overall performance specification for RTNS-II is for two source strengths of
4x 107 n/sec. The source is based on the D-T reaction produced by impinging an accelerated deuterium
beam on a solid, rotating, titanium tritide target.
3. Principal Design Features Number of Sources Beam Current (400 kev) Spot Size Source Strength Maximum Flux Target Size Target Speed Target. Lifetime
Test Volume
4, Major Project Milestones
Completion of Preliminary Proposal Start Excavation
Initial operation of accelerator prototype
61
2 150 MA 1 cm
4x 1023 n/sec
Tex ‘ee wie -sec 46 cm
5000 rpm
100 hours
1 ce
May 1976 Sep 1976
Mar 1977
Complete site construction activities Aug 1977 Complete final target design Sep 1977
Operate first source, project completion Feb 1978
5. Projected Budgets
The Technical Projects Office budget provides operating funds for the RINS through FY 78,the year of project
completion. Funding for subsequent years is provided in the Development and Technology budget.
Fiscal Years
Millions Total 1976 1976A 1977 1978 1979 1980 1981 1982 78-82 Construction: (TEC $5.)M) 25 0 PBS) 0 0 0 0 0 0 Operating 0.0% 0.0% 0.0% de 0. 0 0 0 Te? Equipment 0.0% 0.0% 0.0 0.11 0 0 0 0 0.11
D. Intense Neutron Source
1. Introduction
The Intense Neutron Source (INS) is the second high-energy, high-intensity neutron irradiation facility dedicated to the MFE reactor materials program, the first being RINS-II. The INS will provide higher 14-MeV neutron source intensities and a larger experimental volume than its predecessor, the RTNS-II.
This facility will be used to study the behavior of candidate materials for fusion devices (such as EPR-I)
—— eee 6 *Carried under D&T Funding Budget for FY 76-77.
under radiation-damage conditions similar to that anticipated in large fusion reactors. The facility
will provide prototype neutron flux levels and will have expanded volumetric capability to investigate microstructures, to perform initial screening for mechanical properties of candidate materials, and to examine blanket moderated spectrum effects. This facility will be located at the Los Alamos Scientific Laboratory (LASL), Los Alamos, New Mexico. LASL is the prime contractor for the design and fabrication of the facilities. LASL will operate the facility when completed in 1981, at a total construction cost of $25.4M, including escalation. The Los Alamos Area Office is responsible for administration of the
project, including cost, schedule and technical monitoring.
2. Objectives This 14 MeV source is intended to provide:
e high energy damage information at high fluences required for verification of theories of fission data extrapolation:
e microstructural and mechanical property data for comparison with high energy spectra such as those arising from the Be(D,n) and Li(D,n) stripping reactions, which have tentatively been identified as possible reactions upon which to base a higher intensity, larger volume neutron source;
e neutronic studies in tritium breeding blankets;
e mock fusion reactor first wall life tests;
63
® neutron cross-section measurement data; and
® data on tritium behavior and handling.
The design requirement and overall performance specification for INS is for two source strengths of
a Be) 1 x 10 “n/sec. This source is based on the DT reaction produced by accelerating a tritium beam to
react with a supersonic deuterium gas target.
Se Principal Design Features
Number of Sources Beam Current (270 keV) Jet Target Volume Source Strength Maximum Flux
Test Volume
4, Major Project Milestones
Title I begins Title II begin
Title III and construction begins
64
2
1.1A
lecc
ex 1Ogeniaes
eps Ee lOseicnoee
Sece
Oct. 1976 Julyol977
Feb. 1978
Construction complete Apr 1980
Facility operational Jan 1981
5. Projected Budgets The Technical Projects Office budget provides operating funds for the INS through FY 81, the year of
project completion. Funding for subsequent years is provided in the Development and Technology budget.
Fiscal Years
Millions Total 1976 1976A 1977 1978 1979 1980 1981 1982 78-82 Construction: (TEC $25. 4M) - 100 0 14.4 10.9 0 0 0 0 10.9 Operating Ox Ox 0 x reo Alpe o7/ 2.0 Ze 0 Theo Equipment O* O* 0 * 0.19 0.8 0.1 0.1 0 1.19
* Carried under D&T funding budget for FY 76-77
65
IV. DEVELOPMENT AND TECHNOLOGY
A. Introduction
In 1973, the Development and Technology program (D&T) was created within the Division of Magnetic Fusion Energy (then the DCTR) to provide near term engineering/subsystems support to existing and proposed experiments and, in the longer term, to provide development of the technology which will be necessary for fusion energy to become commercial reality. These both remain as the fundamental objectives of the D&T program; only the scope of the program objectives has expanded as the DMFE
enters into the Fusion Reactor Engineering phase of Fusion Power RD&D.
Development and Technology program activities presently are organized in five technical problem areas:
e Magnetic Systems - This activity sponsors research and development of large superconducting magnet. systems needed for Fusion Reactor Engineering experiments within the next ten years. Examples include toroidal field magnets and ohmic heating coils for tokamaks, yin yang coils for mirrors, and superconducting inductive energy stores for theta pinches. These systems will
be necessary for both plasma confinement and energy storage.
66
Plasma Engineering - This activity is directed principally at the development of efficient plasma heating systems (neutral particle beams, radiofrequency waves, and electromagnetic plasma implosion systems) that are essential for all approaches to commercial fusion energy
by magnetic confinement. Additional responsibilities include direct energy conversion,
vacuum component development, and plasma maintenance and control systems.
Fusion Reactor Materials - The purpose of this long lead-time activity is to develop the high performance reactor materials required for the economical generation of energy from the fusion process. The principal focus is on materials that will be placed within the first ten centimeters or so of the plasma where the fusion radiation environment imposes the most difficult materials requirements. Other areas of effort include development of those materials to a fusion reactor design (e.g., both hot and cryogenic insulators, special structural materials). Fusion Systems Engineering - This activity focuses principally on the next generation and longer term fusion power reactor designs. Specifically, a major activity is the support of necessary reactor design efforts to enable the DMFE to assess the desirability of alternative fusion prototype experimental power reactor configurations presently planned to operate in 1985 or 1986. Other Fusion Systems Engineering activities include systems studies of fusion appli-
cations and economics, blanket and shield engineering, tritium processing and control, plasma
67
systems, and plant systems design and test. A major near-term activity will be the development and prototyping of engineered tritium processing and control systems that will be required for tritium burning experimental fusion reactors.
e Environment and Safety - The purpose of this activity is to pursue technical programs which will assure that fusion power reactors will operate with no potential for hazard either to the environment or to plant personnel and nearby populations. At this time, Environment and Safety efforts focus on environmental impact analysis, facility safety analysis, and reactor safety
research.
Development and Technology activities, introduced above and discussed in detail in the following sections, will grow and evolve to fit the requirements of an integrated Magnetic Fusion Energy research and development program. A key element in the structuring of these activities is our current percep- tion and anticipation of the requirements for engineering development and technology for future
fusion reactors. Thus the following plan, which is designed to pursue the next five years in the
Logic III long range projection, represents our best judgement of what must be accomplished during
the next few years. As technical progress in physics and engineering occurs, our plans will be
modified as needed.
68
In preparing the current portion of the Development and Technology five-year plan, the following
assumptions were made with regard to overall Logic III milestones.
Tokamak confinement continues to be the main approach, but mirrors and other advanced systems (principally high beta theta-pinch) will continue as alternate confinement concepts.
Mirrors are the most attractive alternate confinement concept at this time. MX is planned to operate in 1981.
The next major facility, a Tokamak Prototype Experimental Power Reactor/Ignition Test Reactor (PEPR/ITR) is planned for operation in 1985/86 with Title I funding to begin in 1979. This reactor is also referred to as the TNS.
Development and Technology activities must maintain pace with the PEPR/ITR, but also must meet the requirements of nearer term facilities (Doublet-III, PLT, PDX, ORMAK-Upgrade, MX, and
TETR
Table IV-1 presents the budget summary of the Development and Technology program through FY 1982.
69
Table IV-1
Development and Technology Budget Summary
Fiscal Year (Millions)
PY 76-67 FY 1976. FY .19/76T FY.1977. FY 19/8, FYs1979@ EY eLOSO ye Sime ece7 Total
MAGNETIC SYSTEMS Se eso ed Were Sa sas) $19.0 S250 $28.0 Sep acal $3152 $141.3 PLASMA ENGINEERING TO aL Ow 19.70 2. O 25.0 26.0 29:56 34.5 L360 FUSION REACTOR MATERIALS 6.87 SS TPS) RES 16.0 20.0 23.0 26.23 30%5 ES.S FUSION SYSTEMS ENGINEERING 8.67 22.01 1Os3 1820 25.0 300 34.4 3929 145.3 ENVIRONMENT & SAFETY 45 onal “95 13.0 20) 0. 3.4 Bug Los}
TOTAL OPERATING $33.45 So. 43 $58.05 S72 30 $95.0 $110.0 $1260 $146 .0 S552 50
TOTAL EQUIPMENT 3.18 2 LUA® S2.66 5 9.00 SLO oo $ 16.0 Salo SuZ0. 0 Tel)
70
The following sections describe the major activities of the Development and Technology program in some detail. Each section presents the objectives and scope, technical/management approach, notable recent achievements, and the near term major milestones and funding requirements for the D&T
program activities briefly described above.
a
B. Magnetic Systems
As magnetic fusion experiments approach reactor size, energy and cost considerations force a departure from conventional copper magnets and power supplies. Superconductors offer the potential for meeting all necessary requirements, but the materials, magnets, and energy storage devices must be substantially
improved and enlarged for even a Prototype Experimental Power Reactor.
The purpose of Magnetic Systems is the engineering design and component development of large superconducting magnet systems for future tokamak and mirror experiments, to develop superconducting ohmic heating coils
and power supplies, and to develop fast-discharge energy storage systems for pulsed high beta concepts. Large (approximately three meter diameter) coils are being developed for assembly into a Compact Torus,
and high field superconducting materials developed for mirror experiments. These major activities are supported by numerous efforts in the base technologies of magnet design, cryogenics, materials development and systems analysis. In addition both magnetic and homopolar energy storage devices are being developed
for high beta plasma experiments and tokamak ohmic heating systems.
The management approach is to identify a magnet performance requirement for current or future projects authorized by DMFE management. A review of the present technology is conducted to arrive at any assessment of what developments are necessary, and proposals encouraged from national laboratories,
universities and industries. The detailed development schedule, cost and program integration are
72
planned with field participation and funding provided for implementation. The evaluation of development
progress is accomplished through visits, reports and technical review meetings.
Management is striving to engage and to develop industry in appropriate areas of MFE development, and to transfer basic manufacturing technology from the national laboratories to the industy. This includes not only hardware fabrication, but design and development of reactor components and systems as well. As a result, the NbTi and Nb35n superconductors are being developed and produced in industry, as are energy storage magnets. Recently, the conceptual design of EPR-I magnets was contracted to three industrial firms from which large prototype coils can be constructed. Table IV-2 summarizes the current and
projected funding requirements for Magnetic Systems.
1s)
Table IV-2
Fiscal Years
Millions FY 78-82 Too 1976A EOT? 1978 LIS 1980 1981 1982 Total
Magnetics
Operating 5.20 1.40 15.45 16.40 217, 00 23.00 25.10 29.20 114.70
Equipment Ou OFZ7 1.60 RaW) 4335 4.90 6995 7995 26935 Energy Storage
Operating Phe AY) 0.45 2.50 2.60 4.00 25.00 7.00 8.00 26.60
Equipment 0.24 0.05 OFZ5 0.50 0.50 0.70 0.70 1.00 3.40 Total Magnetic Systems
Operating 7.40 13385 NPE SHES) 19.00 23400 28.00 32.10 J1620 141.30
Equipment dee2 Oey 4 ios Pri h@) 4.85 ey(e)@ Fie) 8.95 295
1. Magnetics
Advancing the size, reliability and performance of present day laboratory magnets involves the development and integration of many interdisciplinary tasks. Engineers, physicists, and metallurgists must all combine talents to make such a hybrid technology satisfy the fusion magnet requirements of
8-16 T fields with up to 20 meter diameters in a hostile pulsed-field and radiation environment.
74,
Objectives and Scope
The objective of the magnetics activity is to develop the large, reliable superconducting magnets and magnet systems that will be necessary to confine plasmas for the tokamak, mirror, and other magnetic fusion concepts. Tokamak reactor studies project the need for "D'" shaped niobium-titanium and possibly high field niobium-tin superconducting magnets with inside dimensions that may be as large as 20 meters, which, when arranged in toroidal geometry, will produce a central field of 4 to 6 Tesla and a peak field between 8 and 12 Tesla. The mirror concept utilizes a Yin-Yang magnet enclosing a sphere of 8 to 10 meters in diameter; the superconducting windings would be niobium-tin and produce
a vacuum central field of about 5 Tesla, a field in the mirror region of 10 Tesla and a peak field
of about 12 Tesila.
Technical Approach
In order to develop such large superconducting tokamak magnets, the Large Coil Project was initiated
at ORNL. Three industries are now preparing conceptual designs for the EPR magnets and recommendations for a subsize prototype to demonstrate the technology. Six prototype magnets will be constructed to the same performance specification by industry, ORNL, and through an IEA collaboration and assembled
into a Compact Torus. By having multiple contractors a variety of magnet designs can be tested in
75
realistic toroidal array. All of the coils in the test configuration will be EPR relevant, so that
performance and cost comparisons can be made and the best selected.
In conjunction with the Large Coil Project, ORNL will develop basic technology in niobium-titanium superconductors, systems design, magnet winding and fabrication for use by the industrial subcontractors. Thus it will be unnecessary for each industry to duplicate the expensive test facilities and breadth
of expertise centered at ORNL.
Testing the large coils for the Compact Torus requires an extensive test facility including cryogenic, vacuum and heavy crane facilities. Both industrial and government-owned facilities are being evaluated for the facility location. Most attractive are those at NASA Plumbrook, Boeing, NASA Houston and ORNL. A decision will be made in the fall of 1976 on the location, so that construction
can begin.
Ohmic heating coils promise to be a most difficult task. The coils must cycle from +8 Tesla to -8 Tesla in less than two seconds, while transferring about 1 gigajoule of energy in a high radiation environment. Development begins with the measurement of ac losses and resolution of the various loss
mechanisms. As funds permit, sample conductors and model coils will be constructed.
76
Mirror magnets clearly need high field conductors. While presently the MX superconductor is being developed for 7.5 Tesla operation with niobium-titanium, future mirror experiments will strive to operate at 16 Tesla. Accordingly, multifilamentary niobium-tin superconductors are being pursued
for operation in the next generation fusion experiments.
Recent Achievements
As a result of the LLL magnetics program, a large advance in multifilamentary niobium-tin superconductors was made at AIRCO. Three superconductors were developed: one containing
67,507 filaments in 3.2 mm diameter wire, a second with 259,369 filaments in a 9.3 mm by 3.9 mm conductor, and a third with 664,411 filaments in 13.75 mm by 6.0 mm conductor. Each of these conductors had filaments of 5 um in diameter and were operated at or above the design values of
1 kA, 3.5 kA and 10 kA, respectively at 12 T. In conjunction with the conductor development a tensile tester was constructed for simultaneously subjecting superconductors to loads of 50,000 pounds, 10,000 amps, and 12 T. An 11 inch bore coil was fabricated from the 1 kA @ 12 T design conductor and has performed beyond the conditions expected from short sample tests. While tests at ORNL show an early onset of resistance in the conductor at nominal strain values, this effect can be
corrected by further development,
LT,
Another method of mitigating strain effects in such high field brittle superconductors is through the recent development of the MIT-ORNL force-cooled, cabled conductor. Here a relatively open and loose cable is enclosed in a structural tube which supports most mechanical loads. High current densities are possible with full magnetic stability because of the large surface to volume ratio of the cable, and pulsed field losses greatly reduced by transposition of the cable strands. Recently a small coil of one meter length has been tested under forced-flow conditions in a 9 T field and has recovered from quench conditions according to an improved analysis including axial conduction and a more rigorous specific heat formulation. So attractive is this result, that ORNL
adopted the design for the EPR design.
A further conductor improvement was made by Battelle Columbus with hydrostatic extrusion of niobium-titanium and niobium-tin superconductors. Extrusion ratios were extended from the current commercial practice of 10:1 to 167:1, which could greatly reduce the manufacturing costs and increase the quality of bronze matrix niobium-tin superconductors. When completed, this technology will be
transferred to industry.
While large coil constructions have not yet begun, ORNL has installed 3 meter magnet winding equipment,
designed four prototype EPR conductors, and is ready to wind coil segments. A helium liquefier has
78
been ordered, capable of supplying 866 watts of refrigeration at 3.5 K for support of the Compact
Torus and smaller laboratory tests.
Tension-free tokamak coils have been optimized by ORNL to reduce the amount of structure required in the EPR, and quench protection analysis and simulations completed. In addition Cornell has analyzed mechanical instabilities in toroidal field sets and demonstrated the phenomenon with
small superconducting magnets.
Plans for the ORNL Large Coil Project are complete and three industries have been contracted for EPR magnet designs. These designs will form the basis for the three meter diameter coils to be
constructed for the Compact Torus to support the DMFE FY 1979 decision for the next major step in
fusion.
Major Milestones and Funding Summary
The pacing item in magnetic systems is the Large Coil Project at ORNL, which will develop the technology to support a major superconductor tokamak experiment decision in FY 79 for operation by FY 85. This project will construct six coils from industry and ORNL, and possibly from IEA, for
assembly into a Compact Torus. An extensive test facility is required soon, so that single coil
The)
tests can be made as the coils are produced. As soon as funding permits, a prototype magnet will be constructed for the TNS. Other near term milestones include the conductor development and magnet
design for the MX experiment at LLL.
Fiscal Years
Millions FY 78-82 1976 1976A 1977 1978 1979 1980 1981 1982 Total Operating: 5320 1.40 13.05 11356 13.41 17.20 19.84 2120 83521 MDF: - - 2.40 4.84 7.59 5.80 5.26 8.00 31.49 Equipment : a0) O27 1.60 2420 4.35 4.90 6.95 7.95 26.35
Major Milestones:
1. Industry to complete conceptual desien of EPR® magnet.) 151-0. ce emenec nee eo OG Le mmnt 71 2«, Start industrials design-andsgconstruction,of sCompact,Torusscoils) pe asas emu LCU os 3. Complete MX conductor and magnet design .. .. . . . « « \EsmiCls Wit Mews Re BOSD sme ono 4. .DMFE decision. to build INS,andsprototype, magnet -.6. «Wis sneyemeue) «mel omen a Loo 5. Completionvof CompactiTorus éoiladesigns?.[ ia) saved af2 Gi. eeeseve.a.0em.ee ce Mar, m1 odc ban eGompletesfabrication of Compact Torus coils 0...) mss) uec ee eenn een ae ee ee Lm 7 Startveonstruction off INS! prototypes! 7372 veto? 7Otoubeorteqva .tofem.s.37onge0cts gia79 8. Complete single coil tests for Compact Taree ere ery eee Se ee ys Moh be eR Ie 9.) aTestAfulleCompactaTorusysg.ycU20. ben OtIRaSGs GOVE Rte cele. 29essenee sk. tee peer SoeDE m1 50 0. Complete INS prototype ~ << 3s 65955 a ee eeu we te es cs
|
80
2. Energy Storage
Presently, large capacitor banks are used to power high-beta experiments such as Scyllac. As the stored energy requirements exceed 100 MJ, it is clear that cost and reliability considerations require alternate approaches. In addition, if a theta pinch reactor is to succeed, it must have a highly efficient energy storage and transfer system which can be provided by superconducting magnets and/or homopolar generators. These same technologies will be necessary for ohmic heating in tokamaks as
well.
Objectives and Scope
The energy storage activity is directed toward providing efficient, reliable electromagnetic and electromechanical devices that shape, switch, store, and reclaim the energy flowing through electric and magnetic subsystems of the various fusion concepts. Flywheel motor-generator sets are currently used to provide pulsed power for the larger fusion experiments, and an assortment of battery banks, capacitors, and line-rectifying sources are used for the smaller experiments. In the longer tern, more efficient and cost-effective energy storage and switching components must be developed. Principal efforts underway include ohmic heating coil development, superconducting magnetic energy
storage systems, advanced inertial (flywheel) motor-generator systems, and efficient fast-switching
81
components. The program over the next five years will culminate in the demonstration of both pulsed superconducting magnetic energy storage systems and inertial homopolar motor-generator energy storage
systems.
Technical Approach
Because superconducting magnets can store energy inductively without loss and have a reasonable transfer efficiency to the compression coils, they are a logical choice for advanced theta pinch concepts. Some losses still occur due to hysteresis in the superconductor and eddy currents in the conductor matrix material. However, by finely dividing the superconductor the hysteresis losses are controlled, and adding high resistance barriers to the conductor matrix decreases matrix eddy current
losses to an acceptable level.
Based upon these principals three industrial magnets have been constructed in addition to one by LASL. These coils store 300 kJ for discharge in 1 millisecond at 40 kV. The LASL coil has performed as expected and the others are now being tested. Subsequently the coils must be housed in a non-conducting
dewar for a complete systems test.
82
For longer discharge times (30 msec to 1 sec) a homopolar generator with superconducting windings
is attractive. These can be used to power current tokamak toroidal field coils, as developed by Texas, or as a second compression stage in a theta pinch experiment. Alternately they may be useful in a tokamak ohmic heating circuit to act as a transfer capacitor. No other known alternate can provide such efficiency and low cost potential. A significant ancillary need for both magnetic
and homopolar energy storage is fast, high voltage power switching.
Recent Achievements
At LASL a 300 kJ superconducting magnet has been designed, fabricated, charged to 10 kA and discharged in 1 msec; the average power delivery during discharge was 300 megawatts! A second coil from the Magnetic Corporation of America was tested to half rating when it suffered some insulation damage.
Testing will continue with the magnets from Intermagnetics General Corporation and Westinghouse.
Commercially available AC vacuum circuit breaker switches, which are generally used for one shot circuit protection by utilities, have been successfully tested and utilized in DC operation to
50 kilovolts and 30 kiloamperes. One switch has been mechanically cycled over 70,000 times. These types of switches are economical and adequate for applications in near term fusion experiments. In the long term, the trend will probably be toward solid state switching for reliability and lifetime
reasons.
83
At Texas a 5 megajoule homopolar motor-generator has been designed and fabricated, and is presently undergoing performance testing. Homopolar machines are characterized by high amperage, low voltage outputs. From half speed (2800 RPM) this machine has produced over 500,000 amperes at 30 volts, and the inertial rotor was stopped in less than one second. Besides ultimate use in fusion systems, these types of machines appear to have characteristics that could be utilized in attractive ways
for resistance welding of large structures, as has been recently demonstrated with the 5 MJ machine.
Major Milestones and Funding Summary
In setting the major milestones and funding priorities, one must assume that the theta pinch experiments will progress as projected, but possibly on a longer time scale due to recent problems with feedback stabilization on Scyllac. Accordingly the superconducting magnetic energy storage coils will be completed and tested in a reduced system. Because tokamak ohmic heating coils use a similar technology, some LASL effort will be continued in superconductor development for this purpose. In addition a superconducting homopolar is planned for future high beta experiments to b2 later
modified for tokamak ohmic heating.
It is assumed that the TNS will not require a superconducting OH system. Large motor-generator sets
will be available instead from the TF power supplies of tokamaks to be phased out around 1985.
84
However EPR II will definitely need superconducting OH coils and an efficient energy transfer system, incorporating a homopolar as a transfer capacitor. In the meantime support of alternate high beta
concepts will require power supplies as alternates to the present capacitor banks.
Fiscal Years
Millions FY 78-82 1976 1976A Led. 1978 Nees, 1980 LOSE 1982 Total
Operating: 220 On45 19:0 ae) 35, 4.5 Gr 6.0 75 VT MDF: - - 0.40 Ons Onl, WIGS) 15,0 Za10 4.9 Equipment : 0.24 0.05 On25 OS) OS) Ons f/ Oy 1.0 3.4 Major Milestones: 1. Design, fabricate, and test commercially ened ie 300 kJ ALi rae
sctorace colils 24h ¥ 2m : ere e297 .Hohseobedes .ow Oct , 1976 2. Demonstrate a Bitereonductdne SFTR Tor Onn eeted (1 oie Ar Go anineeat clicaeADmret975 3. Complete conceptual design of tokamak OH system..... co 9c astiueas cot Septcg 19/9 4. Design, fabricate, and test a superconducting fast ddeennecd (O. 03. sec. 5)
10 MJ homopolar motor-generator (cooperative effort with EPRI) ......... May 1980 yam Complete: evaluation) ofysuperconductors) for: pulsed 0H coils ©...) ey si. se teers aJune» 1980 6. Fabricate and test scaled model of OH system (10-50 MJ coil,
reversible power supply system, and associated switching) ............ Sept. 1982
85
C. Plasma Engineering The Plasma Engineering subelement is actively pursuing the engineering design and component development of fusion plasma support subsystems for near-term fusion experimental devices and ultimately for fusion
reactor systems. Within the Plasma Engineering subelement five, major technical areas are being pursued:
e Neutral Beam Development e Alternate Heating Technology e Direct Energy Conversion e Vacuum Component Development
e Plasma Maintenance and Control Systems
Within each of the five, major technical areas the principal objectives are to provide for the definition, development, design, fabrication, testing and operation of engineering subsystems which will satisfy the particular operating requirements for fusion device applications. In order to meet these objectives, the Plasma Engineering subelement will operate test and development facilities; design, construct and operate prototypical systems; and provide particular components and subsystems to be used on fusion device
experiments. Furthermore, the Plasma Engineering program will define those components and/or subsystems
86
which require further development and will initiate, expand or redirect appropriate developmental efforts. Combined with the other ongoing physics and technology programs which will test the efficacy of the components and subsystems, the Plasma Engineering program will provide the total subsystem design,
technology and operating basis.
As defined, the activities within the Plasma Engineering program are primarily user oriented. Therefore, a secondary objective of the development program is to insure close coordination between the user and developer. The users must get the components and subsystems that perform to specification, and the developers must be involved in the design of the confinement device. This is particularly true for questions regarding interfaces where the components and subsystems and the confinement device come together. The user-developer coordination requires a two-way communication and a joint responsibility
for the development, fabrication arid operation of the complete subsystem.
The final, overall objective of the Plasma Engineering subelement is to maintain close coordination with the various activities of the Systems and Applications Studies subelement. This will help to ensure that the designs, components and subsystems will be developed with realistic requirements and are directed
towards specific, long-term applications such as TNS and EPR.
87
The Plasma Engineering subelement is organized to utilize the expertise of selected institutions which can contribute towards reaching the overall goals. At present the National Laboratories are the principal contributors to the development program. These National Laboratories are carrying through most of the actual development effort as well as participating in design, fabrication, and testing of
components and subsystems for use on near-term confinement experiments.
DMFE strives to involve industry in the technical development of fusion reactor subsystems as well as
in the area of manufacturing of specific components. This is particularly true in areas where it is possible to utilize a unique industrial capability. Several areas have been identified in which industry is particularly well-suited to participate. These are areas such as: high voltage power supplies; switch tube and klystron development; cryogenic equipment; electrode manufacturing; and other ion positive and negative seurce concepts. Efforts are underway or are being initiated at appropriate industrial
contractors.
Several universities presently are participating in support of the overall development program. The areas of university work are, in general, those of basic research, measurement and analysis, and small-scale
design and testing. However, the greatest contribution of university effort is in the area of training
88
of students. When their training is completed, these students are then available to work in the rapidly expanding fusion power program. The development program will expand in the area of university
involvement as additional needs and expertise become identified.
Each of the five technical areas within the Plasma Engineering subelement is discussed in the following
sections. A budget summary of the Plasma Engineering Branch activities is shown in Table IV-3.
89
TABLE IV-3
Plasma Engineering Budget Summary
Fiscal Years
Millions
FY78-82
1976 1976T 1977 1978 1979 1980 1981 1982 Total
Neutral Beam oper. 8.39 3.170 14500150 LO 2065 24.0 95.60
equip. Zine 0.360 AO fi 2810 aS 4.0 320 330 16.60
Alternate oper. 1.49 0.450 S00 2180 Be5 3) 4.5 5. 0 19.30
Heating equip. 0.07 0.025 Wi2s OerS 0 1.0 20 1.0 S75
Direct Energy oper. 0.20 0.050 0.40 0.50 0.8 0.8 1.0 L0 4.10
Conversion equip. On:02 0.005 0.08 0.05 O52 Ons OS Ord 1.55
Vacuum oper. 0.03 -0- 0.50 0.70 Lard ihe ib 5S) 18 6.00
Components equip. -0- -0- OO, 0.08 OLS WS) Dig WES) .28 Plasma Main-
tenance & oper. SWS 0.040 130 thes, 2.0 23 ZeS 3.0 Lie 30
Control equip. -0- OR025 0.10 OF12 OWS: OW On? 0.8 2,02
Totals oper. 10.1 32740 19% 70; me2e00 250 26.0 29.8 3435 136.30
equip. Wyse IAs Dn See TOMO NS 6.1 6.5 5.6 27.80
90
Key milestones in the five major areas of the Plasma Engineering subelement which are required for TNS
are summarized below:
e Neutral Beam Development
- Define beam requirements and development plan for TNS Aprils1977 - Operate 150 kV, 1A, 10 msec. negative ion beam system June 1977 - Operate 200 kV, 25A dc test facility Oets 61977
e Alternate Heating Technology - Define rf heating system technology requirements for TNS Octwatl9 7.6
- Complete rf heating system preliminary design for TNS Octegel 980
e Direct Energy Conversion - Operate direct converter in conjunction with 100 keV, 1 MW ion beam June 1977
- Define beam-line direct converter requirements for TNS Oct ih 9.7 7
e Vacuum Component Development - Define vacuum pumping requirements and development plan for TNS April 1977 - Construct and operate large scale cryocondensation vacuum system
on a neutral beam test facility Oct. LO
oi
- Test cryogenic vacuum systems in radiation environment June 1978 - Simulate and operate TNS vacuum pumping system etwas 1979
e Plasma Maintenance and Control Systems
- Define diverter requirements and development plan for TNS Oe telo77 ~ Define fueling requirements and development plan for TNS Octsee1977 - Simulate and test TNS plasma maintenance and control system June 1980
1. Neutral Beam Development
The Neutral Beam Development effort is devoted to the development of neutral particle injection systems for specific tokamak and mirror applications. At present, neutral beam injection is the principal method of auxiliary plasma heating for these confinement methods. Near-term experimental fusion devices will require appropriate and fully integrated neutral beam injection systems. Furthermore, it is probable that longer range fusion reactors will also require appro- priate neutral beam injection systems. The neutral beam development program will provide for the development, design, and prototype systems for neutral beam injection systems with the required
characteristics of beam energy, power, and pulse length.
a2
Objectives and Scope
The principal objective of the Neutral Beam Development Program is to develop neutral beam injection systems for use on near-term confinement devices such as PLT, Doublet-III, MX and TFTR. In addition to direct, confinement device support, the development program must provide a basic development effort in order to meet longer term requirements for TNS, EPR, and beyond. The development program, therefore, is primarily user oriented, wherein the developers, users, designers and DMFE, working jointly, determine the specifications and parameters required for a particular neutral beam injection system. These specifications and parameters along with the
project schedules then determine the specific goals and timing required of the development effort.
In order to support present and future confinement devices, a vigorous basic development program is required. This is particularly true at the present time in the area of negative ion source technology. A number of methods for the production of negative ions have been suggested, and several of the more promising concepts are being pursued. Other areas of basic, supporting beam development which are being pursued as funding permits include: alternative neutralization
schemes; beam energy recovery; and atomic and molecular species control.
23
In order to provide complete, reliable neutral beam systems, close attention must be given to
the technological and engineering developments necessary for the various subsystems of a neutral beam injector. This includes such items as the components in the electrical, mechanical, diagnostic and control subsystems which are necessary for the operation of the neutral beam
system. Items in this area include: cheaper and more reliable power supplies; high speed vacuum systems; and non-intercepting neutral beam diagnostics. Neutral beam control components to be designed and specified include closed loop, active feedback techniques to time and operate multiple
beam lines in a coordinated manner.
The final objective of the program is to provide for manufacturing support and technology transfer to industry for the overall neutral beam program. The particular goal is to identify, educate and qualify manufacturers who ultimately will produce reliable neutral beam system hardware to stated
specifications and in the required quantities.
Technical Approach
The Neutral Beam Development Program is a broadly based technological effort aimed at providing suitable neutral beam injection systems for use on plasma confinement devices and fusion power
reactors. As such, the development program is a multidisciplined one which encompasses not only
94
the engineering and technology aspects but also some of the plasma physics and systems engineering aspects as well. Therefore, it is necessary that the structure and content of the development program be clearly formulated to assure that the necessary coordination and communication is main-
tained so that the program goals may effectively and efficiently be met.
A number of technological problem areas must be explored and solutions obtained before large, reliable, and efficient neutral beam systems can be designed, fabricated, and operated routinely. Specific development efforts in some of the major technological problem areas are to provide:
high quality and high efficiency ion sources; improved ion optics and beam focussing; more
efficient and more vacuum compatible neutralizers; high power density beam diagnostics and targets; improved electrode and insulator materials for operation in a radiation environment; high-speed, high-throughput vacuum systems; safe and reliable high power electrical systems; and instrumentation and control systems to operate multiple beam systems. All of these technological problem areas
are addressed within the various elements of the development program. In this manner complete engineering designs can be provided for neutral beam injection systems which are appropriate to the
particular confinement device application.
25.
Several types of key programmatic milestones have been established for the development program. One is the operation of required test and development facilities; a second is the design, fabrica- tion, and testing of particular neutral beam hardware for confinement experiments. In addition,
milestones relating to the establishment of particular technologies have been identified.
One of the key items within the development program is the availability of adequate and timely test and development facilities. These test facilities are the pacing milestones in the development program. Although must of the development work can be and is performed on low-
power, small-scale test and development facilities where it is technically reasonable and cost effective, it is essential that full-scale, full-power testing of neutral beam system modules be included in the development program. Testing of such modules must be performed on a timely basis. That is, sufficient testing and development time must be allocated within the development schedule for test stand and prototype operation before it is needed for operation on the confinement devices. This time is necessary in order to allow for several iterations and turn-arounds before the design
of the neutral beam system is fixed.
96
Next generation neutral beam systems will require beam energies in the range of 80-150 keV. These neutral beams are required for confinement experiments such as TFTR, MX and Doublet-III. Therefore, the two large test stands presently under construction have been designed to cover this range. Specifically, the maximum operating parameters of these test stands are:
ORNL 150 kV, 50 A, steady-state
LEL 200 kV, 25 A, steady-state, or
100 kV, 100 A, 5 sec.
As the development program proceeds to higher energies and higher powers, these test stands become Larger and more complex. This in turn means that the test facilities become more costly and take longer to bring on line. Therefore, it is essential that the specifications for the required neutral beam systems be determined in a timely manner; i.e., in the conceptual design stage, the neutral beam parameters must be adequately specified so as to define the appropriate
development program.
Beam systems for experimental devices to be operational in the 1985 time frame are presently thought to be in the 300-400 keV range. Therefore, neutral beam test and development facilities will be provided for this extrapolation by early 1981. It is anticipated that one of the existing large test
facilities will be upgraded for this effort.
eM
The ongoing development effort in plasma sources, accelerator structures and subsystems is the general basic effort of improvement and optimization of neutral beam systems. All of the sub- tasks in these basic efforts are brought together and are tested on the particular test and development facilities. Therefore, this development effort is an iterative one in which new scalings, new ideas and new concepts will be operated on the test facilities and results will be fed back to the appropriate sub-task area. In this manner, the test facilities serve as a focus
for the ongoing development effort.
Recent Achievements
e Operate neutral beam injection systems on ATC Aug. 1974 e Operate 12 neutral beam injectors on 2XII-B Sept. 1974 e Operate 4 neutral beam injectors on ORMAK Jans 1975 e Operate 1 A direct extraction negative ion source May 1975 e Operate 60 kV, 60 A neutral beam development test stand Sept. 1975 e Operate pulsed 80 A ion source at 40 kV Sept. 1975 e Completed final conceptual designs for TFTR Neutral Beam Systems Sép tel 9:75 e Complete first phase of neutral beam I&C design and development package Jan 20119:46
98
Initiated detailed design and fabrication of neutral beam lines for PLT Feb.
Operated first computer control/automatic conditioning of neutral beam lines
at LBL
Funding Summary and Major Near-Term Milestones
1976 oy ET.
Base Program
Operating Be 2.00 LOEZO Major Device
Fabrications =e) Ley. 3.80 Total Operating 8.39 3.74. 00 Equipment DD SHES 0.36 1507, Milestones
Operate first PLT neutral beam system
Operate 80-150 kV, 20A, 0.5 sec. test facility
Operate 150 kV, 50 A, dc test facility
Deliver PLT injection systems
Fiscal Years
Millions 1978 IMTS) Ze 14.5 oP) 30 LS sare) Deel: Ne?)
Define beam requirements & development plan for TNS
Operate 150 kV, 1 A, 10 msec. negative ion beam system
o9
1980
leas
20
18.1
4.0
ale ksgil
1355
Oct.
1976 March 1976
FY 78-82
1982 Total PP) $0) {333} 5 db De) 255 Dan O 95.6 a0 16.6 July 1976
July 1976
Oct. 1976 1976—March 1977 April 1977
June 1977
e Select TFTIR injection system design June 1977
@e Operate 200 kV, 25 A de test facility Cereal 77, e Operate prototype TFTR injection system Decsat 1973 e Deliver Doublet-III beam injection systems 1979 e Deliver TFTR beam injection systems 1981 e Deliver MX beam injection systems 198i
2. Alternate Heating Technology
The Alternate Heating Technology program is concerned with providing for the development of plasma heating techniques and technology which will be an alternative or supplement to neutral beam heating. At present, this activity is concentrated on the development of radiofrequency and micro- wave technologies suitable for plasma heating experiments on present and next generation fusion experiments. Also, this activity includes the development of shock and compression heating and
other technology such as magnetoacoustic heating suitable for pinch devices.
Objectives and Scope
The objectives of the Alternate Heating Technology program are to provide the technological develop- ment and design expertise necessary to design, fabricate, and operate radiofrequency, microwave,
shock and compression heating systems on experimental fusion devices. The focus of this effort is
100
on the present and next generation experiments such as EBT, Scyllac, PLT, PDX, ISX and possibly TFTR. In this manner, a viable alternate heating technology base will be maintained as a backup
or supplement to neutral beam heating.
The Alternate Heating Technology program will provide the basic operating techniques for an alternate plasma heating system for an Experimental Power Reactor (EPR). This will lead to well- defined conceptual designs of alternative plasma heating systems for ultimate fusion reactor applications. For example, present thinking indicates that a radiofrequency plasma heating system for a tokamak EPR will require in the range of 50-70 MW of steady-state power. The coupling structures, waveguides, and transmission lines will be exposed to the radiation fields present in a reactor. Therefore, engineering designs and components must be developed beyond the present
state-of-the-art for these systems.
Technical Approach
Radiofrequency (RF) heating of plasmas offers certain advantages and is a major alternative or
supplement to neutral beam heating. Development work on RF heating presently is being carried out at
101
the lower hybrid frequency (LHF), the ion cyclotron frequency (ICF), and the electron cyclotron
frequency (ECF).
The engineering advantages of working at the LHF are two-fold. First, there is a significant amount of power available from commercial tubes in the frequency range that is required for fusion © research devices or reactors. For example, klystrons are available in this frequency regime with an output power in the 500 kW range. Second, for some conditions, power at this frequency can be transmitted through the vacuum vessel wall by open-ended wave guides. The use of wave guides eliminates the need for high-voltage insulators at the first wall near the plasma. Such insulators
would be required with inductively- or capacitively-caupled radiofrequency systems.
Radiofrequency heating experiments are being conducted at the lower hybrid frequency on the ATC device at the 250 kW level. Plans are in progress to incorporate LHF heating into the next generation fusion device experiments such as PLT, PDX, and ISX. Radiofrequency systems for
applications on these experiments will be in the multimegawatt regime.
The principal engineering advantage of operating at the ICF is also that large blocks of RF power
are readily available from commercially available power tubes. The principal disadvantage is that
102
antennas or similar coupling structures are probably required inside the vacuum chamber in the
plasma environment.
Radiofrequency heating experiments are being conducted at the ion cyclotron frequency on the ATC device. These experiments will serve to corroborate earlier results as to the efficiency of plasma heating using this technique. Plans are being prepared to incorporate ICF heating into the next generation fusion device experiments such as PLT or PDX. Radiofrequency systems for these applica- tions will also be in the multi-megawatt regime. Design and development of radiofrequency systems for applications at both the LHF and the ICF are major milestones for the Alternate Heating Technology program. The results of this design and development coupled with the basic theory and
experimental results will point the way toward RF heating of fusion reactors such as EPR.
A similar effort to develop and test plasma heating via electron cyclotron resonance heating (ECRH) is being conducted and includes testing on the EBT device.. The EBT currently is operating with
60 kW of combined 10.6 and 18.0-GHz microwave power. The development of this technology represents a significant milestone in the alternate heating effort. The ECRH technique can, in principal, be used to heat the plasma in bulk or as an adjunct to another heating mechanism for additional plasma
control.
103
The long-range ECRH effort encompasses the development of multi-kilowatt, millimeter wave-length microwave power sources and coupling circuits. A major milestone of the Alternate Heating program is to develop a high power (200 kW), high frequency (120 GHz), steady-state microwave power source. The design and development of this microwave power source is to be performed by industry to
laboratory specifications.
Shock heating and compression heating systems are required for theta pinches, toroidal Z pinches, and possibly high-beta tokamak experiments. The principal effort in this task is development of components and subsystems which will satisfy the heating system requirements. Materials such as ceramics which can withstand both the high neutren flux and the plasma particle flux must be developed if these systems are to be successful. In addition, this program will pursue the general development of high voltage, high power, high coulomb components. For example, LASL is developing a Marshall coil system and a power crowbar system for the next generation of experiments. These
components will have utility in many areas of MFE.
Other potential methods of plasma heating such as lasers and electron beams are in the research stage
and may be selected for hardware development if technical progress warrants. The program plan
104
assumes that plasma heating techniques other than or in addition to neutral beams will be required. Therefore, the program includes technology development and large scale testing of appropriate
devices in order to assure their quality, performance, reliability and lifetime.
Recent Achievements
@ Operate 100 kW ICF system on ST Jan. 1974 e Operate 125 kV spark gap and capacitor components Jana) 19/5 e Operate 30 kW microwave system on EBT April 1975 e Operate 250 kW LHF system on ATC Sept, 1975
Funding Summary and Major Near-Term Milestones
Fiscal Years
Millions
FY78-82
1976 1976T 1977 1978 1979 1980 1981 1982 Total Operating: 1.492 0.450 5-50) 2.80 BAS) oS) 4.5 55.0 19.30 Equipment : 0.970 03025 On23 O375 0 LO 2.0 0 Sas)
Milestones
e Evaluate LHF and ICF plasma heating on ATC Judy) etl.9'7.6
e Define radiofrequency heating system technology requirements for TNS Oct. 1976
105
e Fabricate Marshall Coil Nov. 1976
e Test components for 200 kV shock heating circuit June 1977 @e Evaluate power crowbar system Sept. 1977 e Operate prototype 25 GHz microwave source for EBT Oceaseloy 7 e Install and operate multi-megawatt rf heating system on PLT/PDX Oct. 1978 e Complete rf heating system preliminary design for TNS Oct. 1980
3. Direct Energy Conversion
The Direct Energy Conversion activity is concerned with the recovery and conversion of the energy of charged particles in fusion plasmas and ion beams directly into electricity. The ability to utilize such an energy conversion system would lead to increasing the overall efficiency of fusion power plants and neutral beam injection systems. Thus, there are both economic and efficiency
incentives for the development of a direct energy conversion technology.
Objectives and Scope:
The principal objective of the Direct Energy Conversion program is to develop the components and subsystems required to operate direct energy conversion systems of fusion device experiments and
neutral beam systems. In addition, a secondary objective is to develop the design experience and
106
capability with which to implement direct energy conversion systems on new fusion experiments or different applications. In this manner, the Direct Energy Conversion program is an auxiliary
developmental effort in support of other technological programs.
The Direct Energy Conversion program is a relatively small development effort in support of mirror fusion reactor programs and neutral beam development programs. However, at least conceptually, direct converter technology could apply to other fusion concepts such as pinch devices or tokamaks.
For example, a direct converter may be an integral part of tokamak diverter system.
The scope of the Direct Energy Conversion program will be maintained so as to provide at least small- scale hardware tests and conceptual design efforts. In this manner, the program will maintain its
viability to provide engineering design data and system design concepts.
Technical Approach
The reason for developing direct energy conversion techniques is to raise the efficiency of the power handling equipment for fusion reactors and neutral beam injection systems. The basic principles of direct energy conversion have been demonstrated in laboratory experiments, and the
program now is turning to the development of practical direct conversion systems.
107
Three applications for direct energy recovery have been identified for both near- and intermediate-
term applications:
(1) recovery of the energy of unneutralized ion beams to make neutral beam systems more efficient and thus extension to higher energies with neutral beams, production of less waste heat and improvment in charged particle beam dump;
(2) recovery of the energy of plasma fuel ions which leave the plasma in a fusion reactor; and
stele ; (3) recovery of the energy of fusion products (e.g., He ) in the exhaust of a fusion reactor.
All three of these applications will increase the operating efficiency of the overall fusion reactor
system.
Ion beam direct conversion experiments will begin in the second half of 1976 on the LBL 80-150 keV neutral beam test facility. In this facility direct conversion equipment can be operated for beam powers of approximately 1 MW for pulse times of up to 0.5 sec. The goal of this effort is to develop beam direct converters as fast as the neutral beam systems will be developed so that overall system
designs can be made. This work will be closely coordinated with the neutral beam development efforts.
108
The direct conversion facility will be modified so as to test plasma direct converter concepts at full voltage, full power density and for steady-state conditions expected in reactor applications.
These tests will demonstrate:
(1) in-situ voltage holding; (2) adequacy of radiation cooling; and
(3) materials testing under mirror reactor-like conditions.
This effort will be closely coordinated with the ongoing fusion reactor experimental and advanced
design efforts.
Recent Achievements e Operate multi-stage direct converter at predicted efficiency of about 85% July 1974 e Operate two-stage direct converter at predicted efficiency of about 657% JAN oy nel IVD
e Operate direct converter in conjunction with ion beam system Jan. 1976
109
Funding Summary and Major Near-Term Milestones
Fiscal Years
Millions
FY78-82
1976 1976T TOF 1978 ihe yes 1980 1981 1982 Total Operating: 0.200 0.050 0.40 0.50 0.8 0.8 1.0 i.0 4.10 Equipment: OnOL> 0.005 0.08 0.05 ORZ 0.3 Ge. Oss: Anes
Milestones
e Design beam-line direct converter for 120 keV beams at 5 Wien Oct. 1976
e Operate direct converter in conjunction with 120 keV, 1 MW ion beam June 1977
e Define beam-line direct converter requirements for TNS Octeg me Osy:
@ Operate plasma direct converter on mirror device (BB-II) Oct. 71978
e Assess direct converter technology for tokamak DEMO June 1979
e Complete design of a plasma direct converter for MX Oct. . 1980
e Deliver MX beam-line converter June 1981
e Deliver MX plasma direct converter June 1982
110
4, Vacuum Component Development
Fusion reactors and several reactor subsystems will require high-speed, high-throughput clean vacuum pumping systems. In addition, these vacuum systems must be compatible with the plasma system and also operate reliably in the radiation environment. The Vacuum Technology program will provide the basic engineering design data and systems tests to specify near-term fusion experiment
and ultimately fusion reactor vacuum systems.
Objectives and Scope
The objective of the Vacuum Component Development program is to provide the engineering design and component development of vacuum systems to meet the requirements of the fusion reactor systems. A fusion vacuum system must handle various gas species at the required pumping speed and throughput under
either pulsed or steady-state operation. Such a system must perform reliably in an environment
of energetic and charged particles. Furthermore, size, space and geometry requirements may be
imposed by the interfacing systems or components of a fusion system.
The Vacuum Component Development program will provide the required engineering design data for qualifi- cation of components and systems for both near-term experiments and long-term fusion reactors. The program will be responsive to new requirements and be closely coordinated with other engineering design efforts.
Pot
Technical Approach
The Vacuum Component Development program is focussed on meeting the design vacuum requirements of magnetic
confinement experiments and neutral beam systems. This program provides for the development of
vacuum technology to meet the needs of the post-TFIR fusion reactors operating under
long-pulse or steady-state conditions.
It is projected that all fusion concepts will require demanding vacuum systems and components for reactor operation. Specifically, vacuum systems capable of pumping gas loads on the order of
100 torr-litre/sec of hydrogen isotopes at pressures of Tome to 10a, torr must be developed.
Gas loads in this regime will be encountered in the next generation neutral beam injection systems. Additional pumping problems will be encountered when DT reacting plasmas become a reality. For example, fusion plasmas will produce a helium gas load due to alpha particle production. There are only a limited number of methods available for pumping helium; and the attendant deuterium gas load will compete in all of these methods for pumping capacity. Furthermore, reacting DT plasmas also
will increase irradiation of the pumping surfaces.
The Vacuum Component Development program will design, fabricate, test and evaluate small-scale and/or
prototype vacuum systems and components in order to meet the program objectives. The present program
Dl
focusses on wall pumping techniques such as cryocondensation, cryosorption, cryotrapping, and bulk
gettering for very high gas loads. The program will investigate these pumping techniques and
will develop the basic design data, capabilities, and limitations of each.
Development and
documentation of procedures such as leak hunting, degassing, and regeneration also will be
carried out as part of the development program. Development and testing of other vacuum hardware
such as valves, flanges, seals, etc. will be performed to support specific vacuum system designs
as needed.
Recent Achievements
e Operate 1.5 x 10° litre/sec. D, cryocondensation pump
2 e Operate 1.2 x 10° litre/sec. H, cryocondensation pump manufactured
by industry
e Operate cryocondensation pump on high energy beam line
Funding Summary and Major Near-Term Milestones
Fiscal Year Millions
1976 1976T 1977 E978 1979 1980 1981
Operating: 0.03 - 0 : ibe Equipment: -0O0- -0O0- 0.07 0.08 On
(=) [3 Lo Wo
2 bis) 3 0.3
Jan. 1976
May 1976
June 1976
FY78-82
1982 Total 15 6.00 ORS the He!
Milestones
e@ Define vacuum pumping requirements and development plan for TNS April 1977 e Provide basic design data for cryocondensation and cryosorption
vacuum systems Oct. 1977
@ Construct and operate large scale cryocondensation vacuum system on a
neutral beam test facility Octo ilo7 7 e Test operation of titanium bulk getter pump Jan. 1978 e Test cryogenic vacuum systems in radiation environment June 1978 e Assess vacuum technology for tokamak June 1979 @ Simulate and operate TNS vacuum pumping system Octru L979 @ Operate test model for steady-state vacuum pumping system for MX Jan. 1980
5. Plasma Maintenance and Control
The Plasma Maintenance and Control program includes the engineering design and component development of plasma fueling and impurity control systems. As such, this program provides the engineering development of those systems and components necessary to sustain and control the plasma in a fusion
reactor.
114
Objectives and Scope
If a fusion power reactor is to be operated for burn times significantly longer than the average particle confinement time, it will be necessary to refuel the reactor while it is operating. Because only a small fraction of the fuel will react before it is lost, methods are required to divert and collect these non-reacting plasma ions as well as the fusion reaction products. The principal objective of the Plasma Maintenance and Control program therefore is to develop and test
components and systems for plasma fueling and sustenance for a fusion reactor.
For plasma fueling technology, the Plasma Maintenance and Control program provides for the develop- ment of particular fueling hardware, provides an assessment of the various fueling techniques, and provides engineering designs and tests in order to determine capabilities and limitations for the various fueling approaches. To meet these objectives the program now focusses on the development of fueling hardware with capabilities to meet the needs of present and next generation fusion reactor experiments. This development and testing will generate the engineering data base and design
experience required to provide fueling systems for ultimate fusion reactors.
Similarly for diverter systems, the Plasma Maintenance and Control program provides engineering
design data as input into reactor conceptual designs. Furthermore, small-scale tests and diverter
ie
assembly prototypes will be performed as they become necessary. In this manner, the engineering data base and design expertise is generated so as to provide adequate diverter systems for fusion
reactors.
Technical Approach
Within the Plasma Maintenance and Control program, one of the sub-tasks currently under investiga- tion is the plasma fueling effort. Several techniques for plasma fueling have been suggested, such as: solid pellets, neutral particle beams, cluster beams, plasma guns, liquid jets, and gas blankets. At present, solid pellets of frozen deuterium and tritium are expected to meet the
fueling requirement for tokamak fusion reactors.
The present and near-term development program for solid pellet fueling systems is directed toward developing controlled and reliable methods of pellet formation. In addition, various candidate acceleretion techniques include the gas dynamic or jet technique, rotating mechanical systems, electrostatic accelerators, and laser acceleration systems. These techniques will be studied using small-scale and prototype hardware in order to determine the operating characteristics, capabilities,
and limitations associated with each technique.
116
Also, it is necessary to determine where the fuel will be deposited within the plasma. This in turn requires experimental knowledge of the pellet ablation rate versus pellet speed and size. Prototype pellet injection systems will be built and installed on appropriate fusion device experi- ments in order to determine the detailed plasma fueling mechanisms. This work will be closely
coordinated with the ongoing confinement experiment programs.
If diverter systems are required for fusion reactors, such systems will place severe requirements on the particle collection and vacuum pumping devices for such a system. In addition, these diverter systems will be required to operate in a high radiation environment. Conceptual designs for such diverter systems will be prepared, and small-scale tests and prototypes will be performed
as necessary.
Recent Achievements e Operate first generation deuterium pellet injector July 1975
e Test pellet fueling device with hydrogen on ORMAK Sept.1975
PL7
Funding Summary and Major Near-Term Milestones
Operating: - 0 - 0.040 1730 é koala 8 Equipment: -0O-
Fiscal Year
Millions
1976 1976T 1977 1978 1979 1980 1981 1982 20 Pd \ 75S) 356,
0.025 0.10 0.12 0.5 ORS On? 0.8
Milestones e Test prototype pellet injection device on ORMAK e Define diverter requirements and development plan for TNS e Demonstrate rotating mechanical pellet accelerator e Establish limits for electrostatic pellet acceleration e Define fueling requirements and development plan for TNS @e Operate pellet injection device on PLT and ISX or ORMAK-UG e Operate pellet injection device on PDX and/or Doublet-III e Assess technology of plasma refueling and impurity control for tokamak DEMO e Simulate and test TNS plasma maintenance and control system
118
FY78-82 Total bl Eire (0) De Oe LEY. | ANSE! OCt emai, OCEe me oe Cet cee) Sir? OC Gemma June 1978 Sarees June 1979 June 1980
D. Fusion Reactor Materials
Materials represent one of the major long-lead time problem areas for power producing fusion systems. Fusion reactors present their own special materials problems related to, but often separate from, those of LMFBR's and other systems. The most severe of these deals with the effects of 14 MeV neutrons and other high energy particles generated in fusion reactors.
This radiation not only results in increased bulk radiation effects (displacement damage, void formation, helium embrittlement) but gives rise to a whole new set of surface related effects including neutron and charged-particle sputtering, blistering, and chemical sputtering. Such effects are known to exist, but there is only a modest knowledge about their magnitude, reproducibility, variability with materials, and energy dependence. Other areas that require investigation include potential coolant and blanket materials, compatibility with structural components, welding and fabrication techniques, nondestructive testing, high temperature design, and adequate electrical insulators. A part of the Reactor Materials Program is the development
of radiation facility concepts for testing of materials and components.
The objectives of the fusion Reactor Materials Program are to develop the materials and
materials technology for commercial fusion power generation. This includes the development
119
of new radiation resistant first wall/structural alloys, as well as the development and
testing of other materials such as insulators, advanced superconductors, moderators,
breeding materials, and materials for heat conversion systems.
The Reactor Materials Program is divided into five research and development areas as
follows. These areas are a major departure from previous Five Year Plans not only in
title but in the work to be accomplished under each technical area.
5%
Alloy Development for Irradiation Performance Plasma - Materials Interaction
Special Purpose Materials Development
Damage Analysis and Dosimetry
Radiation Facilities Development and Operation
The management approach is management by objectives with major objectives defined by the
Materials and Radiation Effects Branch relative to the prime DMFE goals, Detailed
milestones and program plans will be developed by technical area Task Groups composed
of laboratory and contractor personnel, both from the materials research and engineering
communities as well as from the design community. The Task Groups will make programmatic
recommendations to DMFE for implementation at the laboratory and contractor organizations,
120
and the contractors in turn work toward specific milestones which are defined in the Task Group Program Plans and incorporated into the 189a documentation. Four Task Groups in the following subject areas will be convened: First Wall Alloy Development, Plasma-Materials
Interaction, Radiation Damage Analysis and Fundamental Studies, and Special Purpose Materials.
An overview of the funding requirements for the next five years for the Fusion Reactor
Materials Program is shown in Table IV-4.
nis Alloy Development For Irradiation Performance
Objectives and Scope
The objective of the Alloy Development for Irradiation Performance subprogram is to provide the materials development and application for those structural
materials which are subject to significant radiation damage.
The prime technical objective is the development of a structural material for the first wall and structural elements for the blanket and shield of a commercial
fusion power reactor.
Secondary objectives, intermediate in time and technical difficulty are specific
tasks directed towards the prime objective that will provide a series of material
development and optimization steps designed to satisfy the data requirements for intermediate fusion power systems,
ial
TABLE IV-4 FUSION REACTOR MATERIALS BUDGET SUMMARY
Fiscal Years
Millions 78-82 1976 1976T 1977 1978 1979 1980 1981 1982 Total Alloy Development Operating L389 0.36 2.96 Diao 6.8 8.7 1170 P3220 45.3 for Irradiation Equipment 0.28 OF03 0.17 Oa335 O29 1.0 a Les BS Performance Plasma-Materials Operating 1.63 G739 De Oe 2.8 Ble 4.5 530 a0 20.9 Interaction Equipment 0.41 0.08 OF37. 0.43 O35 OF ORS 0.6 2.63 Special Purpose Operating Lely 0.26 OF 81 130 Vez ete: 2a 225 9.1 Materials Equipment 0.24 OF 02 Ome 0.08 Oe O52 OFZ Gz3 0.88 Development Damage Analysis Operating @..85 0.26 0.61 1.0 2.0 292 205 235 10 <2 and Dosimetry Equipment 0.18 0.04 O2L0 0.03 OeZ O22 O23 0.3 L038 Radiation Source Operating 1.83 0.48 Hie Wi ay 6.4 5.8 se paz 30:53 Development Equipment 0.29 0.04 0.30 bs03 Lee Aw OW 0.8 4.43 and Operation Total: Operating 6.87 Le) one IS 16.0 20:20 2350 26253 s0n. 11526 Total: Equipment 1.40 Oa21 1.06 1.90 209 2.6 3.0 Boe. Lacey
The scope of these activities includes (a) the definition of material requirements that are needed to satisfy fusion power system design goals, including the Experimental Power Reactor (EPR), the Demonstration Plant (DEMO) and Commercial Fusion Reactors, (b) the evaluation of existing materials to meet those goals, (c) the development of new materials where they are required to achieve those goals, and (d) the development of the materials
radiation data base required for design, construction and operation of fusion power systems.
Technical Approach
Several major paths for alloy development of a first wall/structural material are outlined below. The importance of achieving the objective of a radiation resistant alloy and the relatively long lead time to commercial fusion power make it both rational and imperative that multiple development paths be followed. Neutron irradiation of materials leads to displacement damage and transmutant production which affect the engineering properties of structural materials. A fusion reactor spectrum contains about 25% of neutrons above 10 MeV as compared to fission reactor spectra where the number of neutrons above 2 MeV is very small. This results in greater damage to materials in the fusion spectrum and hence more severe
service conditions.
123
A four-path program that is focused on the commercial fusion power goal, and that will
make maximum use of both broadly innovative concepts and relatively well established
materials technologies, will be followed.
The paths contain interim objectives that reflect aggressive but realistic development and use of irradiation test facilities, In addition, at least two of these paths have interim objectives that will provide alloy selection and data for the intermediate
fusion power goals of experimental and demonstration power reactors,
Path A is directed to development of the optimum class of austenitic alloys. Path B is aimed at the higher strength, higher temperature alloy development in the Fe-Ni-Cr alloy system (i.e. superalloys), Both A and B can be tested to end-of-life damage
levels in fission reactors with confirmation in d(Li,n) neutron test facilities,
Path C will address the reactive and refractory alloy option. The initial phase on this path will be a screening activity designed to screen materials and eliminate those that are unsuitable through use of fission reactors, available neutron sources, and simulation tools. Initial development of prime candidates will be accomplished using d(Li,n) facilities. Final alloy optimization engineering data acquisition will be done
in a Materials Fusion Test Reactor as outlined in DMFE's Long Range Planning document,
124
Path D will be established through open solicitation of innovative material development and application concepts. This could include minimum activation materials, composite
or graded materials or other imaginative ideas that offer the possibility of reduction to engineering practice. The timing of the overall alloy development program is given in
the section entitled Milestone and Funding Summary.
Recent Achievements:
- Stainless steel alloys 316 and 316 Ti modified have been examined after irradiations
in HFIR to displacement damage levels up to 60 dpa and transmutant levels of more
than 4000 appm He. These levels exceed approximately 5 and 10 year exposures, respectively, in a fusion reactor with a l MW/m? wall loading. Both titanium and cold work reduce the
adverse effects of irradiation on the properties and behavior of the alloys.
These preliminary results indicate that a cold worked stainless steel would still have
4 useful engineering properties after about 5 MW-years/m in the fusion reactor environment.
- Niobium and vanadium alloys irradiated in HFIR show little loss of ductility
(at relatively low He levels) and no swelling.
25
- Results from RINS irradiations of copper and niobium indicate that, for threshold phenomena, 14 MeV neutrons create about ten times as much displacement damage as an equivalent number of fission reactor neutrons; however, displacement damage theory
is able to account for these differences to within a factor of two or less.
Irradiation of an EBR II experiment was initiated. The test assembly contains
an array of candidate alloys that have had helium added by alpha particle injection.
126
Major Milestones and Funding Summary
Fiscal Years - Millions
Budget: Alloy Development for FY78-82
Irradiation Performance 1976 1976T LOT 1978 1979 1980 4981 1982 Total Operating 1.39 0.36 2.96 Ss) 6.8 Saiae LicOe 823.0 ees: Equipment 0.28 C203 0. Liaes0 533 0.9 1.0 Als a ee} 4.73 Milestones:
- Establish Task Group on First Wall Alloy Development.........ceceeeeeeeeeeee August 1976 - Establish preliminary target goals to meet the materials
requirements of commercial fusion power...... wievale lars leelersclele sels aiersra sles ee sfelei ste March 1977 - Complete alloy development to meet materials requirements of commercial fusion power. eeeeeeoeeeveeereeeeeeeeeeeseeeeeeeteeoeeeeeeeeeeeeee eeee#e 1993
Detailed alloy development milestones follow to illustrate the approximate timing of each
development path.
a) Path A Austenitic Alloys
Complete Initial Optimization of Composition and Structure (OPT-1) 1983 Establish Maximum Performance Level of OPT-1 1988 Complete Final Alloy Optimization 1988 Establish Maximum Performance Level of Alloy Class 199%)
127
Bb) Path 8
¢) Path ¢
d) Path D
High Strength/High Temperature Fe-Ni-Cr Alloys
Complete Initial Optimization of Composition and Structure (OPT) Establish Maximum Performance Level of OPT-1
Complete Final Alloy Optimization
Establish Maximum Performance Level of Alloy Class
Reactive/Refractory Alloys
Complete Initial Screening Phase - (Eliminate losers using simulation testing) Select Prime Candidate Alloys (via d(Li,n),DT and fission reactor irradiation) Initiate Irradiation Testing in Fusion Materials Test Reactor(FMTR) Complete Initial Alloy Optimization (OPT-1) (via d(Li,n)DT sources,and FMTR) Establish Maximum Performance Level of OPT-1 Complete Final Alloy Optimization
Special and Innovative Materials Concepts
Minimum Activation Concept
Complete Initial Irradiation Test Plan Complete Candidate Screening
(via d(Li,n) ,DT,and fission reactors) Evaluate Concept Feasibility
128
1983 1988 1988 1o9
1983 1988
1990 1993
1998 Uae pee,
1983
1988 1990
Composite, Graded, and other Innovative Materials Concepts
Issue Request for Proposals
Complete Initial Study Phase
Complete Initial Irradiation Test Plan
Issue RFP for Engineering Development
Complete Initial Engineering Development
Complete Initial Irradiation Testing of Candidate Concepts
Complete Phase II Irradiation Testing of Candidate Concepts
Select €oncept
2. Plasma-Materials Interaction
Objectives and Scope
One of the major objectives of the Plasma-Materials Interaction subprogram is
to investigate the interaction of the plasma with the wall material in terms of the
impurities injected into the plasma. A part of this objective which is longer range is the development of materials' structures which are less susceptible for impurity
injections and which maintain structural integrity of the wall for long times.
Specific objectives are:
- To treat surface effects from the standpoints of plasma contamination, wall
erosion, and device efficiency
- To evaluate surfaces as part of a completely integrated system consisting of the plasma and its perimeter, the external blanket, subsystems for vacuum
1978 1981 1985 1985 1989
1993
1997 1995
pumping, fueling and ash removal, and possibly bumpers or divertors
129
> Fo develop new materials resistant to surface damage in concert with overall alloy development tasks for commercial fusion power
Technical Approach Surface work is carried out in the following technical areas: i) Desorption by a) photons, b) electrons, c) ions and neutrals and d) thermally
+ ii) Sputtering by a) hydrogen ions, b) residual gas ions (Onn N , etc. ),3c)shelium, d) metallic ions (self sputtering), and e) neutrons
iii) Hydrogen-wall interaction by a) absorption, b) diffusion, c) surface migration and d) reflection
iv) Blistering, flaking and surface erosion by a) incident hydrogen, and b) helium from the DT reaction, (n,a) reactions, and tritium decay
v) Chemical reactions involving a) hydrogen and carbon, b) residual contaminants, c) carbon and metals (e.g. NbC) and d) decomposition, principally of insulators
vi) Electron emission
Synergistic studies of two of these processes occurring at the same time are now beginning, and will be extended to studies investigating the effects of three or
more processes. These studies are designed to simulate the effects of a burning
DT plasma.
130
Recent Achievements
- Neutron gputtering and chunk emission An upper limit of 5 x 10e2 can now be placed upon the total yield resulting from neutron bombardment, including both atomistic sputtering and the emission of chunks, if any. Participants in the round-robin sample exchange are completing evaluation
of their results, and a joint report summarizing this work is expected in the fall
of 191763
- Blistering and flaking Investigation of these processes has addressed means to suppress them, and also sought an improved fundamental understanding of them. Progress has been made in both areas with identification of optimum temperature ranges for blister Suppression, and the types of materials in which blistering is least serious. The relative roles of gas pressure buildup in blister formation and stress
gradients resulting from swelling in the near surface region are being clarified.
- Sputtering by hydrogen ions Recent yields were reported in the USSR and it appears that the results are in reasonable agreement with those reported in the US. In general, yields for
physical sputtering appear to range downward from 1 x TOne for most candidate
138
materials. Although this does not assure freedom from plasma impurity problems, it makes serious wall erosion by this process improbable.
Fluxes to first walls
Data has recently been provided by members of the plasma physics community which give estimates of the energy distribution of fluxes from plasmas to first walls in operating tokamaks and also in proposed beam driven and ignition tokamaks. This data will be of great use in identifying those energy ranges which are most
important for surface experiments.
Honeycomb walls
Experimental work to provide data for comparison with calculations on the relative yields of planar and honeycomb walls is in progress, and it is expected that
an assessment of the actual effectiveness of this concept will be available by
the end of the fiscal year. Suppression of impurity influx by a factor of 4 is
theoretically predicted.
£32
Attention is turning to two new areas to further reduce impurity introduction. The first is the study of sputtering of charged and neutral particles, with the objective of increasing the fraction of particles which are sputtered in a charged state, thereby permitting their deflection back to the wall by the toroidal field. The second is consideration of surface treatment by ion implantation, coatings, or other means, as
a way of producing and sustaining within the exposed surface of an operating tokamak,
the composition which is optimum for impurity suppression.
ibe Ke)
Major Milestones and Funding Summary
Fiscal Years -— Millions
Budget: Plasma Materials FY 78-82
Interaction 1976 19767 ae L977 1978 1979 1980 1981 1982 Total Operating 163.08 0.439 2.05 2.8 3.6 gece dD ei 360) 20.9 Equipment 0... 4.e9s0,508 O37 O03 oe 205 pe0. 5.5, ma. © 0.6 2.03 Milestones
- Establish Task Group on Plasma-Materials Interaction.....cceccccccecceceveeceeee August 1976
- Establish preliminary target goals to meet the plasma-materials interaction requirements of commercial ‘fusion powers... «<+cscs ces ce sicie ce celecisisieceiemide se emeMACCHiML ous
- Develop detailed milestones that complement the alloy development program....... May 1977
- Operate surface diagnostics packages on Alcator and ORMAK......ceceeevesseee-eee October 1977
>. Begin surface experiments in ISX. on. cslc oclcc.0a 0 0 uw clas cles ale nicts state seie cuiienie ieee CCEMDC TEL Ua, « Operate Multiple Beam, Radiation Facility (MBRE) 2.6. ...ccnscvelcs ve ciicile pacts cuss smOeCDLEMDeTel oss « Complete single beam experiments... i ..s 0 00s vsiclercis sles sieve sissies siereisicleicierciecieereie een a Caro OU
~» Begin surface diagnostics in WIFIR.. 26 ci ccs. ws sew cece sels sis saree ceeuinies ieee mC LOD ETE Od
134
3%
Special Purpose Materials Development
Objectives and Scope
This technical area covers the development of materials other than first wall/structural
materials described previously. Included are the following materials applications:
1)
ii)
iii)
Insulators for superconducting magnets
The lifetime of superconducting magnet insulators is limited due to neutron and gamma radiation damage. Since the properties of insulators deteriorate at fluences lower than those impairing the performance of the pucerecnanc ror insulators are generally regarded as limiting the overall magnet performance at this time. Radiation damage of superconducting magnet wire
The current carrying capacity of superconductors is impatred by neutron irradiation, and performance. Characteristics have to be defined as a function of accumulated damage.
Insulators for electrical and structural applications
The theta pinch first wall is the best example of an insulating requirement for a structural component. Bulk and surface dielectric strength and resistivity requirements have to be met subsequent to each cycle of irradiation and in the presence of high
temperatures, chemical attack and variable stresses.
USNS
Many other examples of insulator requirements exist, among them those for neutral
beams where voltage holding capability is all important, and sputtering and
evaporation of electrodes may become problems as neutral beam energy is increased, iv) Moderator and breeding materials
Several schemes for neutron moderation and breeding have been identified in
systems studies. All of them raise questions regarding radiation damage to
moderating materials, compatibility, and chemical and mechanical behavior of
solid breeding materials.
v) Heat transfer and energy conversion systems Intermediate heat exchangers and steam generators present traditional problems in structural design, compatibility and long term reliability which need to be addressed within the context of specific fusion systems, vi) Other applications Many other materials requirements will become obvious with time and will be
addressed as they become identified.
136
Technical Approach The approach to solving technical problems in this subprogram varies with the particular
problem area. The Task Group on Special Purpose Materials will be asked to address
individual areas over the next 12-18 months.
Recent Achievements The one area that has received attention in the past two years is the development of insulators for structural purposes. The following milestones have been achieved: - EBR-II irradiations of different ceramics were initiated. The first capsule was removed in January 1976 at a fluence of 2 x 10°*n/em"(E_ > 0.1 MeV).
Two other capsules are scheduled for removal in December of 1976 after irradiation
to "2x TO-on/ene (En > 0.1 MeV).
An Insulator Workshop was held at LASL in May of 1976 to assess progress made in the past 18 months and to define the requirements for fusion reactor insulator
applications.
Major Milestones and Funding Summary Fiscal Years - Millions
Budget Special Purpose FY78-82 Materials Development 1976 1976T 1977 2.1978 1979 1980 1981 1982 Total Operating eRaglhy/ 0.2659 .0.81 51.0 ae 1.8 Ze5 PAS OL Equipment 0.24 OD. 02m sl len. Sem Os 1 nie On2 On 0.88
P37,
Milestones: - Evaluate superconducting magnet insulator needs and initiate program..........October 1976
- Evaluate special purpose materials problem areas and develop detailed milestones to meet requirements: .si0 << occ ciete swale cle'eiele sale sis clsiein eerste ete eine UCT G /,
4. Damage Analysis and Dosimetry Objectives and Scope
The objectives of the Damage Analysis and Dosimetry subprogram are to characterize available irradiation test environments and to establish a basis for predicting materials performance under irradiation in a fusion reactor environment. This will
be accomplished by materials irradiation data obtained in fission reactors, accelerator based neutron test environments (RINS, INS, d(Li,n), d(Be,n), etc.) and charged particle
irradiations.
The scope of the Damage Analysis and Dosimetry subprogram includes development of
the methodology and nuclear data base required for characterization of the reactor
test environment. Included are the development of fundamental radiation damage models,
the development of methods to account for interactive phenomena in the evaluation of damage
structures, the relationship of structure and damage parameters to property changes, and
138
the application of the above to the prediction of material performance in fusion reactor
systems.
Technical Approach
Damage analysis relies on a relatively well established methodology for the fission program which needs to be extended to the fusion environment. Included are calculations, and wherever possible measurements, of damage by high energy neutrons both for displacement and transmutant production. Included also are calculations of pulsed versus steady state effects. Ion irradiations and high energy neutron experiments will be conducted to support the calculational
efforts.
Recent Achievements
A precursor to the Radiation Damage Analysis and Fundamental Studies Task Group was convened by DMFE in February of 1976. This group, called the DCTR Radiation Damage Assessment
Panel was asked to "evaluate test environments for fusion reactor materials." The preliminary Panel report, published in May 1976, contains an extensive state-of-the-art review of damage analysis, en extensive description of fusion reactor and neutron irradiation test environments, and a series of recommendations on the application of damage analysis
principles to the evaluation and utilization of radiation test environments.
139
Major Milestones and Funding Summary Fiscal Years - Millions
Budget: Damage Analysis FY78-82 and Dosimetry 19765 “9197 6Tot41977. aal9 7840519795 S1980se 61981 Be 19s2 Total
Operating 0285, 60.26 0.6L 44,0 2.0 Lee 2nD Ziad 19032
Equipment 05185640.04 0.106590. 03e550.2 One O03 0.3 1e03
Milestones:
- Convene DMFE Task Group for Radiation Damage Analysis and “Fundamental Studies). cs. .ccmescenscvelscce ses e sees so sles ste sneer ate canes Ce DCeMiC am a6
- Complete initial detailed Program Plan for Damage Analysis.........+.+++++- March 1977 - Complete BCC Charged Particle Intercorrelation Program........scscesceseee JUly 1977
. Characterize d(Be,n) and d(Li,n) in support of Radiation Facilities Development Program. .. <<... ss 6 sic sisiesw sissies cleiv sie sieele siete ie ieee DOC ele let t7 a
- Establish intercorrelations among low fluence neutron irradiation test environments for model materials in support of Paths A, B and C of the Alloy Development for Irradiation, Performance, Task, sisi. ewsielcle ascleaieis sle's o cisveislerereisratee « sienelenensientts ciciensmenn a OU
. Establish intercorrelation at high damage levels for model materials in a number of trradiation test enyitronments,....-0cne see eso
140
- Establish high fluence damage correlation for engineering alloys ipetaesionmreactor andid(Li, n)Mfort Path) AvandyPathe Bens che bias wee vie eet £985
5. Radiation Facilities Development and Operation Objectives and Scope
The objectives of this subprogram are to define the radiation environment of fusion reactors and to pursue the development of neutron and plasma sources to simulate this environment for materials testing. Since fusion reactors are not now available for testing, high energy neutron and plasma sources are needed to develop materials for
commercial fusion power.
Technical Approach
High energy neutron sources are based on the deuterium-tritium (DT) reaction to produce 14 MeV neutrons and on the d(Li,n) and d(Be,n) stripping reactions to produce a broad neutron spectrum of high energy. Two DT neutron sources authorized for construction
by Congress are the Rotating Target Neutron Source, RINS, being built at LLL and the Intense Neutron Source, INS, being built at LASL. The RINS is designed for 14 MeV neutron fluxes on the order of 2 x SOME nae ce at the specimen, while the INS is
designed for 14 MeV fluxes of up to LOmeny cneoeec at the specimen. Proposals for a
141
d(Li,n) neutron source of much larger yolume and higher flux are under evaluation, This source will have a distributed spectrum which has been shown in the past year to be an excellent simulation of the fusion reactor damage spectrum. D(Be,n) sources
for specimen irradiations to relatively low fluences exist at several laboratories.
Fission reactors are used in the program for testing of nickel-bearing materials because of the two-stage reaction for helium production in mixed fast and thermal reactor spectra. Thus, fission reactors permit simulation of helium/dpa damage accumulation similar to that which occurs in fusion reactors. Unfortunately, this statement is true only for nickel-bearing alloys and high energy neutron sources are needed for all other
materials.
The above sources will be used to accumulate neutron radiation data for materials development as well as to contribute to an understanding of damage analysis and extrapolation of fission reactor data to the high energy fusion reactor spectra. The program is currently evaluating the need for a multi-beam facility to simulate the effects of a plasma on the first wall. Materials studies to date have only
addressed this problem by evaluating the effects of single neutron or particle
142
beams on materials surfaces. Multiple beam evaluations including neutrons as well
as charged or neutral particles need to be undertaken in materials evaluation and development to more accurately simulate reactor types of plasmas. Although confinement devices such as ISX are necessary steps in the understanding of the plasma-wall interaction,
they do not simulate reactor conditions.
All facilities when constructed will be used by DMFE contractors on a priority
system to be established by DMFE. The cost of operation of all facilities as well as development costs prior to construction authorization (except for d(Li,n) where the total estimated development is shown), plus the cost of facilities built under
MDF Procedures are included in this subprogram.
Recent Achievements |.
The RINS and INS were authorized by Congress for construction in FY 1976. During
the third quarter of FY 1976, DMFE convened an ad hoc Neutron Source Proposal Technical Evaluation Panel to evaluate proposals from four ERDA laboratories for a d(Li,n) neutron
source. A decision is expected to be made during the transition quarter following FY 1976.
143
A second Panel on Damage Assessment was conyened in February 1976 to evaluate potential radiation environments for simulating fusion reactors. One of the conclusions of this Panel was that "...the transmutant damage parameters of the (d,n) stripping sources
more closely match those of the first wall of a fusion reactor than any of the other
neutron environments considered".
144
Major Milestones and Funding Summary
Fiscal Years - Millions
Budget: Radiation Facilities FY 78-82 Development and Operation 1976 1976T 1977 1978 1979 1980 1981 1982 Total
Operating Details
MDF on Multibeam
Facility -- -- -- INES; O79 -- “+ = Lae Multibeam Facility
Operation -- -- -- -- -- 0.4 0.4 0.4 De Fission Reactor
Operation == -- -- 073 0.6 Oa Oy On 3.0 D-Be Operation -- -- -- 0.2 0.2 On2 0.2 O72 1.0 D-Li Development
and Operator Training -- -- Cae ae Pay 2) 2.4 1.5 ats 11.4 RINS Upgrade and k
Operation Wealth Oashr) = 0.4 io 2.0 2.0 220 Sez INS Upgrade and i
Operation 02800516 -- -- -- -- 0.6 Pag 8) 209 Miscellaneous 200m. 0S OFD7 OL Ot Can Cel O.L O55 Total Operating 1705 0.46 Dede Sl 6.4 5.8 Deo cee 30.3 Total Equipment 0.29 0.04 0.30 ikl@ey aby 0.7 Oe) 0.8 4.43
* Funding and Technical Responsibility Transferred to TPO
145
Milestones:
- Decision to develop and build high energy neutron sources with fluxes aga the 1013-1014n/cm2-sec TANGSC cee veer eevrcecvvneeevecesveseevesvesesveevevece 1975 Achieved
- Decision to develop and build a larger volume, higher flux (101>n/cm2-sec) neutron source facility. Select from competing
CEST ENS «oc 550 esis ces 6 66 see sce sista seca ers sida ie statetaielece erehenetere cnelaieteraenscelcdererctetetaneeanere October 1976
- Decision to build multibeam test facility incorporating DT-plasma components for surface radiation testing. = o.% cs. cles oo os Sieislclcicl oc sicicls s cle mNOVCTIDG LANG G
- Operate Rotating Target Neutron Source (RINS) facility at LLL.............. 1978
« Operate Intense Neutron Source (INS) facility at LASL.. 2.2. sscssseeescee er LOOL
146
E. Fusion Systems Engineering
The principal objective of Fusion System Engineering is to provide the technical framework for the evaluation, assessment and guidance of the Magnetic Fusion Energy R&D program. Program goals and milestones are established and evaluated on the basis of maximizing the probability of progress
success within a minimum practical program cost and with an acceptable level of risk.
The Fusion Systems Engineering program is presently divided into six categories: Advance Design Systems Studies Blanket and Shield Engineering Tritium Processing and Control Plasma Systems
Plant Systems
Advance Design provides conceptual design studies for each major D-T burning facility in support of agency requests for authorizations. Systems Studies provide the identification, characterization, and
assessment of timing of the technological achievements that must be reached to meet each of the major
147
program milestones. Blanket and Shield and Tritium Processing and Control tasks provide for reactor subsystem development, and testing to establish the engineering design basis for fusion power systems. Plasma Systems encompass detailed systems analyses of plasma behavior which provides guidance and design criteria for component development programs which are intimately associated with plasma maintenance and control. The Plant Systems area includes work in areas such Jas Reactor Instrumentation and Control (1&C) and Balance of Plant. An overview of the funding requirements for the next five years for the
Fusion Systems Engineering program is shown in Table IV-5.
148
TABLE IV-5
Fusion Systems Engineering Budget Summary
Fiscal Years Millions
LD 76F eh GA relD:27) aal978 L979 1980
Advance Design Operating: 4.30 939893540 8.00 9.50 1050 Equipment: 503 - - - - - Systems Studies Operating: 2201. .69 3.50 4.00 4.80 biesys: Equipment: 03 - - - - - Blanket & Shield Engineering Operating: -49 ong 39 12.80 2280 4.90 Equipment: Oz 20 36006 20 0.30 0.50 Tritium Proces- sing & Control Operating: 0 Olen ORES ee) 2.30 362% 3590 Equipment: el2 -03 54 Q425 0.30 0.50 Plasma Systems Operating: SJOMEEO SOG aye) isa) 30 2.00 Equipment: 70% - = 0.10 0.10 0.10 Plant Systems Operating: e135 0.02 0.30 0.80 vos Sh, |e) Equipment: Ou: - 0.00 .09 0.10 0.30 TOTALS Operating: 8. GHerlo2sObe 10430 Pavk8i00 #52300 30.00 Equipment: e22 06 26 - 64 0.80 1240
149
1982
16.0
FY 78-82 Total
56.00
26.25
20.00
19R6Z Zo
10.00 0.70
13.43 1 09
145.30 6.24
1. Advanced Design
Objectives and Scope
The objective of Advance Design is to identify major fusion facility plant characteristics and performance requirements. This includes design goals and alternatives, which are needed to achieve commercial fusion power in the late 1990's. Advance Design serves to focus the engineering and technology necessary to produce fusion power. Technology requirements are determined in scoping and conceptual design of each major D-T burning facility. The scope includes Tokamak, Mirror and other confinement configurations of interest. These advance designs provide a basis for
(1) selecting detailed design alternatives (2) evaluating development and construction costs and schedules and (3) deciding to request authorization to proceed with detailed facility design, construction, and operation. Conceptual design studies provide focus and direction for technology
development programs supporting reactor components and systems.
Technical Approach
These objectives are accomplished in four stages. First, scoping studies are performed within the Systems Study category. Systems requirements and available technology are identified to assess the feasibility of proceeding with conceptual design. Second, preliminary conceptual design is performed to develop a "baseline design'' where compatible design alternatives are determined and technology
requirements are documented. Third, a detailed conceptual design of a facility is completed to
150
optimize the design, to determine cost and technology requirements and to support requests for congressional authorization. The fourth stage, advanced A-E work (interim funding), is sometimes
necessary to determine accurately project cost and schedule prior to Title I construction
authorization.
FY-1976 EPR conceptual design efforts continued at ANL, ORNL and GA and included engineering analysis on many plant features of which the following are examples: a) basic design alternatives (e.g., divertor vs. non-divertor, circular vs. non-circular cross-sections) b) basic design characteristics to mitigate plasma physics uncertainties (e.g., alpha-particle heating, fueling problems, impurity control, etc.) c) reactor subsystem design problems (e.g., first wall, blanket and shield, toroidal magnet set, ohmic heating and equilibrium coil systems, energy supply and conversion, reactor maintenance,
tritium handling and auxiliary systems).
During FY-77, an evolution of the EPR-1 studies will occur such that preliminary conceptual designs will be oriented toward a prototype experimental power reactor with a main objective of reaching
ignition. A separate study will continue on the EPR.
fot
From 1978 through 1981, Advance Design activities will concentrate on the Tokamak devices. Beyond 1981, these efforts will expand to provide for Mirror-PEPR/ITR design and for subsequent devices
under alternative concepts.
The devices to be designed conceptually under this category include Prototype Experimental Power Reactors (PEPR) which achieve ignition of the plasma and provide significant plant systems experience; Experimental Power Reactors (EPR) which generate significant net electrical power; Demonstration Reactors (DEMO) which demonstrate the performance of a fusion power system on an electrical grid; and Fusion Experimental Research Facility (FERF) which provide a capability to perform extensive materials
and engineering tests in support of major devices.
The pace and direction of the designs are closely coordinated with the total Development and Technology,
the Applied Plasma Physics, Confinement Systems, and the Technical Projects Programs.
Recent Notable Achievements The TFTR conceptual design studies by ORNL and PPPL/Westinghouse were completed in FY 1975 and provided the technical basis for the request for authorization to proceed to the Title I phase
of the IFTR project.
152
- EPR scoping studies were performed at ANL, ORNL, and GA in FY-1975. They serve as the focus for current R&D efforts and have been essential for near term planning of the Development and
Technology, Applied Plasma Physics and Confinement Programs.
- During FY-1975, LASL initiated a study on the Scyllac Fusion Test Reactor. This work continued
in FY76. A baseline design is to be documented.
153
Major Milestones and Funding Summary
Funding Requirements: 19-769 1976004 19°77 P1997 C8 2 L079
OPERATING 4.30 soo 3.4 8.0 9.5 EQUIPMENT 03 = = = —
Major Milestones:
PEPR/ITR (10/85)
Milestone Begin conceptual design Begin preliminary design and Title I Begin operation
Major Device Upgrade
Milestone
Identify Key PLT, D-III, TFTR Upgrade Alternatives - Initiate TFTR Upgrade Design
Mirror (EPR) Milestone
Begin Prel. Conceptual Design Begin Conceptual Design
154
19804190 14 M1992
10.50 91220991650
Date
10/76 10/79 10/85
Date
10/76 10/81
Date
10/86 10/87
FY 78-82 TOTAL
56.0
Major Milestones:
Fusion Engineering Research Facility (1991)
Milestone Date Begin detailed conceptual design 10/79 pemitnttiate Title wd 10/82 - Construction Complete 10/89 Experimental Power Reactor (1991) Milestone Date Begin conceptual design 10/82 minitTateslatile . 10/85 . Construction completed 10/91
oD
2. Systems Studies Objective and Scope
The objective of systems studies is to provide the basis upon which program managers make choices about future courses of actions within DMFE. The end product of system studies is program action. The scope of this information base that is developed includes three distinct areas: (1) Major Power System Analyses, (2) Component and Subsystem Analyses and (3) Socio-economic/benefit risk analyses. This scope then covers all areas from program cost-benefit analyses to individual component comparative
analyses.
Technical Approach
The approach to achieving these objectives is to maintain an iterative procedure of (1) systematically examining program objectives and the alternative strategies for achieving them, (2) identifying key parameters against which program strategies can be evaluated, e.g. cost, technical risk, schedule and
(3) developing and comparing the quantitative data for the key parameters.
Presently, systems studies are being performed in five closely related areas: . Fusion Reactor Reference Design
. New Systems/Fusion-Fission Concepts
156
- Parametric Studies - Econometric Studies
. Systems Analysis/Technology Assessment
Fusion reactor reference design studies of specific fusion power reactor concepts integrate all elements of fusion reactor design into a self-consistent plant design and define the technological and engineering requirements and limitations and important performance characteristics of each
major subsystem.
New Systems and Fusion-fission Concepts studies are performed to determine the extent to which the fusion process can be utilized in forms other than pure fusion reactors. Also included in this area are screening studies of non-mainline fusion confinement methods which may be either a new approach or an old approach previously discarded but having potential based on new findings. Non-mainline approaches, such as shock tubes, laser heated solenoids, and relativistic E-beam heated solenoids are possible candidates for the breeding of fissile fuel and may be studied on a limited basis. Other non-mainline approaches, such as steady state toroidal reactors (bumpy torus and stellarator) and the reverse field Z pinch, if appropriate, also may be studied for both pure fusion and fusion-fission
applications.
17
Parametric studies of fusion systems are performed with computer codes which mathematically describe the operation of each reactor subsystem and the integration of these subsystems into a self-consistent reactor configuration. They will provide analysis of the sensitivity of overall system performance and economics to various technical reactor parameters and thus help to define optimum power systems
and provide basis for evaluating promising fusion reactor concepts.
Econometric studies are analyses of a broad nature which help define the role of fusion both for the United States and the world in the context of available alternative energy sources such as fission, fossil, solar and geothermal systems. At present these analyses are based on linear cost/benefit static and dynamic codes at BNL and PNL which prescribe fusion reactor performance, costs and
implementation schedules to compete with alternative power sources.
As more data become available from other systems studies (reactor design) and systems analysis (PERT programming), these codes will be modified to provide more refined parametric examination of sensitive independent and dependent variables such as plant capital and operating costs, resource problems, fuel
costs and market entry dates.
isk)
System Analysis has two elements. One is the production of necessary tools which can be used to accomplish overall system studies objectives. The work at the University of Texas on the Advanced Optimization Code is an example of this. The second area is the one of Application Studies which
supports both the New Systems Concepts and the Econometric Studies.
Applications Studies are performed to characterize more clearly the role of fusion power systems in
providing benefits in addition to central electric power stations.
Current applications studies focus on Energy Park configurations for fusion and fusion-fission.
Materials processing (chemical, agro-industrial, etc.) and synthetic fuel production also are being
studied.
Technology assessments have been performed on the effects of impurities on tokamak reactor dynamics and of the engineering feasibility of various combinations of blanket and shield materials. Presently, the focus of these efforts is on defining the requirements for fusion reactor subsystems and assessing
the state of technology required to satisfy these requirements.
E59
Recent Notable Achievements Commercial reference reactor designs performed at the University of Wisconsin and PPPL, and provided preliminary data relative to the economic potential of fusion reactors (i.e., plant
capital costs and $/KWH).
. Preliminary studies completed in FY 1975 for mirror and linear theta-pinch hybrid reactors indicated that these applications of fusion power may be potentially attractive; however
additional effort is not warranted at this time.
. Initial parametric systems analysis codes have been developed for all three magnetic confinement
concepts at ANL in FY-1975.
. Applications studies performed at BNL in FY 1974-1975 have identified a number of non-electric goods that could be produced by fusion reactors. These include synthetic fuels, industrial
chemicals, and a variety of primary metals.
160
Major Milestones and Funding Requirements
Funding Requirements:
1976 1976A 1977 1978 1979 1980 1981. 1982 OPERATING Peete h 69 35.9 4.0 A Sedge «opee oS oye O EQUIPMENT 603 go— - - - - - - Major Milestones: Systems Studies Milestone Date Initiate Mainline DEMO studies 1/ TEES Initiate Tokamak commercial reactor studies (Circular & Doublet)— and EPR 10/77 Scoping study Assess Tokamak Commercial Reactor Designs 10/78 Initiate Alternate Concept commercial reactor designs 10/79 Initiate SSTR-EBT DEMO design 10/80 Fusion-Fission/New Systems Milestone Date . Initiate TCT Tokamak and Mirror Fusion-Fission scoping/commercial reactor design 7475 study Initiation of low Q, non-mainline electric breeder scoping/commercial studies 10/75 Initiate Steady-State Toroidal Reactor--EBT (SSTR-EBT) scoping/commercial Tie reactor design study Initiate EPR/DEMO fusion-fission design for TCT Tokamak; Mirror; and Low Q, 9/77 Non-Mainline concepts Evaluate design studies for SSTR-EBT 1017s Evaluate TCT Tokamak; Mirror, Linear Theta-Pinch; and Low Q, Non-Mainline fusion 10/79
fission concepts
1/ Mirror & Theta-Pinch design teams to focus on EPR (Advance Design)
161
ae
FY 78-82 TOTAL
2Oc5
Major Milestones:
Parametric Analysis
Milestone Date Provide parametric codes for Tokamak, Mirror and Theta-Pinch fusion reactors 9/77 - Provide parametric codes for Tokamak, Mirror, Linear Theta-Pinch, and SSTR-EBT 9/78 fusion-fission reactors and for SSTR-EBT Fusion Reactor - Provide Global parametric code 9/80
Econometric Analysis/Systems Analysis/Technology Assessment
Milestone Date - Initiate assessment of stainless steel activation problems TS Initiate economic assessment of mainline fusion-fission energy systems thy ls
- Fusion Fission Technology Assessment 10/77
- Fusion Power Technology Assessment 9/79
- Fusion Technology Assessment 10/77
162
3. Blanket and Shield Engineering
Objectives and Scope
Blanket and Shield Engineering provides neutronics data, calculational techniques, and benchmark testing in support of pure fusion and fusion/fission reactor blanket designs. Blanket designs are developed to show strengths and weaknesses of various materials combinations and to identify critical development needs. Development programs are established to answer key design related questions and, later in the program, experimental blanket modules are to be fabricated and tested in a fusion reactor environment. This blanket design and testing activity draws upon basic materials property and radia-
tion damage data developed in the DMFE Materials and Radiation Effects program.
Technical Approach
The Neutronic Data Development activity provides data for fusion reactor blanket and shield design. Data assessment is performed as blanket and first wall designs are produced and also through more ordered sensitivity analyses. Measurements and/or nuclear model calculations then are undertaken in response to identified needs. Most cross section measurement needs will be satisfied by programs in the Division of Physical Research. However, the DMFE will maintain a capability to support measure-
ments which are critical to the program and which could not be provided soon enough under the DPR
163
program. As new measurements become available they will be evaluated (including error files), provided to the ENDF library, processed into forms which can be used in MFE design analysis, and incorporated into the MFE data file. Finally, this activity will provide some support for nuclear data centers so
the fusion community has ready access to the latest available data and calculational tools.
The Neutronics Methods Development activity will provide for the development of calculational techniques and codes required for design of CTR blankets and shields. Included will be such items as 1) extending transport theory and Monte Carlo codes to toroidal geometry, 2) improving nuclear data processing codes (Kerma factors, radioactivity, gamma production, gamma interaction, covariance matrices, self shielding factors, etc.), and 3) combining transport and Monte Carlo calculations for efficiently analyzing detailed blanket and shield designs. Development of techniques for sensitivity analysis and calculation
of recoil spectra for materials damage studies is also included.
The Neutronics Benchmark activity will support small scale calculational efforts aimed at scoping neutronic problems of particular interest (e.g., the effect of penetrations in MFE blankets). It also will allow limited study of blanket and shield optimization from a neutronics point of view. Results
from these studies, aimed at maximizing tritium production, heat generation, shielding effectiveness,
164
etc., will be provided to those doing Advanced Design and Systems Studies. Benchmark experiments will be performed to gain confidence in nuclear data calculational techniques. The first priority
will be testing in support of PEPR or EPR.
The Blanket and Shield Engineering and Methods Development activity will look in detail at specific designs for liquid lithium, molten salt, solid, minimum activity, and fusion-fission blanket designs. The designs will consider: materials compatibility; the impact of pulsed operation on the tritium breeding, the power cycle, and thermal fatigue of the structure; tritium extraction; the effect of coolant and structure materials on magnetic fields; maintenance and replaceability; economics; etc. From these studies will come an assessment of technology needs (heat transfer, thermal hydraulics, structural analysis, thermal energy storage, fatigue data, etc.) to support each of the general blanket concepts as well as a determination of the designs with the highest potential for success. Methods development tasks (thermal hydraulic analysis, structural analysis, and reliability/safety analysis) will be undertaken to satisfy some of these technology needs. Experimental needs will be addressed under Blanket Fabrication and Testing and Coolant Chemistry and Purification tasks, but
design of a Blanket and Shield Test Facility will be carried out by this task. Basic materials data
165
will come from the Fusion Reactor Materials Subelement. Finally, designs of prototypic blanket modules for DEMO and commercial plants will be developed for extensive testing. The blanket engineering activity must look at near term PEPR/ITR needs and must explore all viable concepts until the most promising approach for long range fusion reactor application is found. It is expected that
at least two basic concepts will be selected and carried to a point where DEMO plant design decisions
are made.
The Blanket Fabrication and Testing activity will respond to the technology needs identified in the above blanket design studies as well as to those identified in specific reactor studies. Included will be testing of liquid metal pumping and heat transfer in high magnetic field, first with NaK or sodium and then possibly with lithium. Molten salt loop experiments may be performed if design studies indicate this to be a viable concept and to the extent that Molten Salt Reactor data are not applicable. Data on high temperature/radiation behavior of solid breeder materials (sintering, creep, swelling, and tritium release) will be obtained. Experimental blanket and shield modules (pure fusion and fusion-fission) will be fabricated and tested to verify manufacturability, thermal hydraulic and
structural design adequacy, and maintainability. Finally, blanket and shield modules of differing
166
design will be tested under near prototypic irradiation conditions in PEPR or FERF/ETR to verify the
~
complete blanket design.
The Coolant Chemistry and Purification tasks will explore means of measuring and controlling impurities in liquid metal, molten salt and helium coolant systems. There will be a close interface with work on
tritium extraction from coolant.
Recent Notable Achievements
. A Blanket and Shield Workshop was held at BNL to investigate fusion reactor design problems
. "A Comparative Study of the More Promising Combinations of Blanket Materials, Power Conversion
Systems, and Tritium Recovery and Containment Systems for Fusion Reactors" was performed and
published as an ORNL report (TM-4999)
. The nuclear data requirements for TFTR have been evaluated and a LASL report published
(LA-6118-MS)
167
. The MFE multigroup library has been issued by RSIC for testing by the MFE neutronics community
. A decision was made to initiate a neutronics mockup experimental program in support of PEPR/ITR.
168
Major Milestones and Funding Summary Funding Requirements:
19] Ono Ohm eo Ome 7 See LOO LI gee
OPERATING ~49 -14 ae) I Orie 2nG eid oe last. 50) EQUIPMENT 407? -03 -06 EAN ue SIO) » 20 - 60
Major Milestones:
Nuclear Data Development Milestone
Complete assessment of TFTR data needs . Multi-group cross-section library for MFE use tested and available for general use . Complete assessment of EPR/FERF/PEPR data needs . EPR/FERF/PEPR neutronics data is available to designer . Complete EPR /DEMO assessment . Neutronic data for EPR /DEMO is available
Neutronic Methods Development Milestone Complete update of nuclear heating code Variational (sensitivity) methods are adapted for parametric and optimization studies
eee) o,ecode available for MFE design . 3D Monte Carlo code available for general MFE use
169
Dio -60
Date
12/75 6/76 1/78 1/80 9/82 9/85
Date
6/77 9/78
O/ 77 9/79
FY /8-82 TOTAL
Major Milestones:
Neutronics Benchmark Calculations and Experiments*
Milestone Date
Start calculations to scope specific MFE neutronics problems and plan early 10/76 experiments
Begin blanket and shield benchmark experiments in support of PEPR/ITR 10/77
Complete initial PEPR/ITR benchmark program 6/79
Complete engineering mock-up of PEPR/ITR 9/81
Initiate tritium breeding experiments 10/82
Blanket and Shield Engineering and Methods Development
Milestone Date Complete technological assessment of current pure fusion blanket designs 6/76 Start methods development to support blanket design 10/76 Identify detailed needs for structural analysis, fabrication techniques and 9/77 maintenance technology development (from design efforts) Start preliminary design of Blanket and Shield Test Facility 10/78 Idenitfy most likely blanket for EPR 9/79 Initiate Title I construction of Blanket and Shield Test Facility 10/80 Complete design of test modules for PEPR 9/82
Blanket Fabrication and Testing** Milestone Date
Start tests on pumping and heat transfer of liquid metals in strong magnetic fields 10/77
Start feature tests on molten salt or He blankets as indicated by design choices 10777 Complete initial phase of liquid metal pumping tests 9/79 Initiate mechanical tests on PEPR first wall and shield 10/80 Complete feature testing on promising blanket designs 9/82%* Initiate Testing in Blanket and Shield Facility in support of EPR 10/82
170
Major Milestones:
Coolant Chemistry and Purification
Milestone Date
- Start standardization of analytical chemistry methods for liquid lithium LO fete,
Begin to identify and develop techniques for impurity monitoring in other 10/78 coolants and breeder materials
Start loop tests on impurity monitoring and liquid lithium purification 10/79
Data to support PEPR breeder module design is available 9/81
*The budget allows only for experiments relevant to one machine. Additional funding would be needed for a Mirror PEPR, a FERF or Fusion Fission
**The budget shown in Table 25 does not provide sufficient funds for this task. Also funding is not sufficient for fusion fission.
byl:
4. Tritium Processing and Control Objectives and Scope
Tritium Processing and Control includes (1) the analysis of the projected performance of individual components of tritium handling subsystems; (2) identification of those alternative components and subsystems which require experimental work, component fabrication, pilot plant demonstration, etc., to provide reliable and safe tritium subsystem performance; and (3) research and development necessary to demonstrate all tritium processing and control subsystems required for each major fusion facility. Tritium processing and control is comprised of the following major research areas: tritium fuel cycle, blanket tritium extraction systems, tritium isolation in coolant/
energy conversion subsystems, and tritium primary and secondary containment/cleanup systems.
Technical Approach
The near-term program for development of tritium handling technologies for major DT burning facilities is paced primarily by the schedule and technology requirements for PEPR/ITR. A relatively good defini- tion of the PEPR tritium processing and handling system should be provided in the preliminary conceptual design study reports to be prepared by ORNL, ANL and General Atomic Co. at the end of FY 1976. Nearer
term facilities such as the Rotating Target Neutron Source and the Princeton TFTR can be designed and
L7i2
operated by adapting existing technology and experience from the ERDA nuclear weapons program and the commercial light water reactor industry. The LASL Intense Neutron Source may require some development in the areas of cryogenic distillation and cryosorption pumping. This work will be highly relevant to
the tritium fuel cycle eventually needed for fusion reactors.
It is worth noting here that a substantial tritium handling technology exists in association with security classified ERDA weapons program activities. ML and PNL are attempting to extract relevant unclassified data from the classified literature to provide better access for DMFE interests. Efforts are also underway to declassify any tritium technology which has direct application to the DMFE
Development and Technology program.
The current plan for conduct of the tritium processing and control program may be divided into the
following phases:
FY 1974-1975: Comparative surveys and parametric studies have been performed in parallel at several
laboratories to identify the more promising tritium handling system configurations for application
L73
to EPR-I or a prototype EPR, with emphasis on identifying and prioritizing the near-term technology
development needs.
FY 1976-1977: Conceptual designs for PEPR level tritium handling systems will be conducted. Comparative surveys and assessments will be phased out; being replaced by bench-scale testing and verification of more promising techniques for extracting, purifying, isotopically separating, recycling, storing and containing tritium in PEPR. Development of new methods will be carried out in areas where there are significant technical uncertainties associated with the primary techniques. During this period, "National Centers of Expertise" will be identified (on the basis of quality, pertinence and creativity of work performed for MFE and other organizations) to take the technical lead role in developing selected portions of the tritium handling technology (e.g., the tritium fuel cycle, emergency clean-up systems, etc.) for PEPR and subsequent major DT burning facilities. Classified tritium technology directly applicable to MFE programs will be
declassified to the extent practicable.
FY 1978-1982: Test loop and sub-size prototype facilities will be operated to test and qualify
tritium handling components and systems for a prototype experimental power reactor. In subsequent
174
years efforts will be directed toward upgrading tritium technologies to permit safe, reliable operation
of the EPR in the early nineteen-nineties.
In association with decisions to construct a FERF/ETR or other PEPR's, laboratories will adapt
then-existing technology to the tritium handling requirements.
Recent Notable Achievements - ORNL has completed a survey of alternative tritium extraction methods from lithium blankets. The primary purpose of the study was to identify the extraction methods which have technical and
economic potential for application to PEPR, EPR-I and subsequent facilities.
. ANL has completed a preliminary examination of alternative oxide barrier materials and bimetallic metal barrier combinations for inhibition of tritium permeation through reactor heat exchangers
and piping.
. Several laboratories (PPL, ML, ORNL, LASL) have completed designs of the tritium handling systems for FTR's. The relative merits of each design have been compared and it is anticipated that some
optimization of TFTR tritium handling systems will result.
5
. BNL has completed a preliminary examination of the extractibility of tritium from minimum activity solid blankets containing lithium aluminate, lithium silicate and lithium-aluminum alloys. Early
results indicate that acceptable extraction rates can be achieved.
. ANL has preliminary data which indicates that molten salt extraction offers promise as a method of
recovering tritium from liquid lithium.
176
Major Milestones and Funding Requirements Funding Requirements: LOT EL oTOR 1977 9 7S) 1979" 1980" 19ST age?
OPERATING -61 15 shes 3 2PSS9 B27 Be BOO} Rie 2
EQUIPMENT Belk - 03 4 20 a3 oD Hai 0) - 60 Major Milestones: Fuel Cycle Reprocessing Milestone Date . Initiate design of facility to Prototype EPR fuel cycle and containment/cleanup 10/76 systems assessment of alternative tritium fuel cycles systems - Begin construction of prototype loop facility ead Initiate operation in facility with hydrogen/deuterium 10/78 Begin operation of facility with tritium 10/79 Deliver technology for tritium systems in the PEPR 10/80 . Begin testing optimized components for later D-T reactors 10/80 Blanket/Coolant Extraction Milestone Date Initiate design of bench scale lithium loop for tritium extraction tests 1/76 . Complete construction of extraction loop 9/77 Initiate operation of loop with flowing salt and advanced extraction systems 9/78 Demonstrate bench scale feasibility of extraction from solid breeding materials 9/78 and molten salts Initiate design of prototype extraction loop for most promising breeding material 7/79 . Begin operation of some systems in the extraction loop 10/81 Provide technology for EPR tritium extraction systems 9/84
hd
FY 78-82 TOTAL
Jad ey Pe IS)
Major Milestones:
Routine/Accident Containment & Cleanup System Design
Milestone
e
Initiate experimental tests of candidate barrier systems for tritium handling systems which must operate at elevated temperatures
Begin construction of prototype clean up system test facility (in conjunction with fuel cycle loop facility)
Begin fabrication of a subsize model low permeation heat exchanger
Complete testing and demonstration of alternatives for safe reliable tritium containment and cleanup systems applicable to PEPR/ITR
Identify EPR tritium containment/cleanup system technology development requirements, initiate requisite work to upgrade technology developed for PEPR
Begin testing of subscale low permeation heat exchanger
178
Date 1G ORS Tl,
10/77 10/80
10/80
1/79
5. Plasma Systems
Objectives and Scope
Plasma Systems is responsible for the analysis of systems which will be necessary to create and sustain a D-T plasma to significant burn-up. Experimental and theoretical data on plasma behavior is developed within the Applied Plasma Physics and Confinement Systems Programs of DMFE. Data on the behavior of materials in a reactor plasma environment (e.g. swelling, sputtering, embrittlement) is developed by the Fusion Reactor Materials subelement to allow determination of the effects of reactor first walls on plasma performance. By analyzing plasma systems in greater detail and over a wider range of parameter space, than possible in reference designs, this category serves as a focal point for coupling materials and plasma physics research data to the design of fusion reactor plasma maintenance systems. The purpose is to define reactor subsystem interface requirements and provide design criteria for Development and Technology Programs and to formally identify basic and experimental research needs to
the Applied Plasma Physics and Confinement Systems Programs.
Technical Approach
Primary emphasis within plasma systems is directed towards providing information necessary for the
design and proper operation of the PEPR and EPR. Analysis of confinement concepts considered highly
iby
likely for use as FERFs will be conducted as appropriate. In each of the task areas, all confinement concepts require substantial analysis and development; however, at this point in time, alternate confinement concepts have not progressed to the stage where separable engineering design of reactor
plasma maintenance systems would be warranted.
At present the critical areas of fusion plasma operation roughly overlap as follows: 1) Start-Up/ Shut-Down, 2) Burn Cycle Dynamics, 3) Fueling/Pumping/Impurity Control, and 4) Reactor Plasma
Simulation.
Plasma Start-Up/Shut-Down Systems
DMFE Confinement Systems and Applied Plasma Physics programs will provide much information concerning this task. As the plasma start-up/shut-down requirements are determined, D&T will assume major responsibility to oversee the engineering design of poloidal field coils, related power supplies, instrumentation and control and other start-up/shut-down systems. Questions concerning reactor diagnostics for plasma control during start-up/shut-down and steady state also must be answered and
Applied Plasma Physics is sponsoring work in this area to identify long range research needs. This
180
area will be explored to determine a proper sequence of systems analysis and engineering
development.
Burn Cycle Dynamics
Efforts in this area will be directed at establishing design criteria for plasma support systems such as neutral beam, RF and ohmic heating, gas blanket and pellet fueling, and first walls as a function of varying reactor plasma performance requirements. Plasma codes will be used to determine
the overall plasma operating cycle, in detail, for devices such as PEPR, EPR-I and DEMO.
Fueling/Pumping/Impurity Control
Efforts in this area will be directed at establishing interface requirements among the three