Overview of the ARIES-RS reversed-shear tokamak power plant study

Fusion Engineering and Design 38 (1997) 3–25

Overview of the ARIES-RS reversed-shear tokamak powerplant study

Farrokh Najmabadi a,*, The Aries Team: Charles G. Bathke b, Michael C. Billone c,James P. Blanchard d, Leslie Bromberg e, Edward Chin f, Fredrick R. Cole g,

Jeffrey A. Crowell d, David A. Ehst c, Laila A. El-Guebaly d, J. Stephen Herring h,Thanh Q. Hua c, Stephen C. Jardin i, Charles E. Kessel i, Hesham Khater d,V. Dennis Lee g, Siegfried Malang a,j, Tak-Kuen Mau a, Ronald L. Miller a,Elsayed A. Mogahed d, Thomas W. Petrie f, Elmer E. Reis f, Joel Schultz e,

M. Sidorov e, Don Steiner k, Igor N. Sviatoslavsky d, Dai-Kai Sze c, Robert Thayer k,Mark S. Tillack a, Peter Titus e,l, Lester M. Wagner g, Xueren Wang a,

Clement P.C. Wong f

a Fusion Energy Research Program, School of Engineering, Uni6ersity of California, San Diego, La Jolla, CA 92130, USAb Los Alamos National Laboratory, Los Alamos, NM 87545, USA

c Argonne National Laboratory, Argonne, IL 60439, USAd Uni6ersity of Wisconsin, Madison, WI 53706, USA

e Massachusetts Institute of Technology, Cambridge, MA 02139, USAf General Atomics, San Diego, CA 92186, USA

g McDonnell Douglas Aerospace, St. Louis, MO 63166, USAh Idaho National Engineering Laboratory, Idaho Falls, ID 83415, USA

i Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USAj Forschungszentrum Karlsruhe, Gmbh, Karlsruhe, Germanyk Rensselaer Polytechnic Institute, Troy, NY 12180, USA

l Stone and Webster, USA

Abstract

The ARIES-RS tokamak is a conceptual, D–T-burning 1000 MWe power plant. As with earlier ARIES designstudies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics andengineering models. Detailed analyses of individual systems together with system interfaces and interactions wereincorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest costsystem. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A=4.0).The plasma current is relatively low (Ip=11.32 MA) and bootstrap current fraction is high ( fBC=0.88). Conse-quently, the auxiliary power required for RF current drive is relatively low (�80 MW). At the same time, the average

* Corresponding author.

0920-3796/97/$17.00 © 1997 Elsevier Science S.A. All rights reserved.

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F. Najmabadi et al. / Fusion Engineering and Design 38 (1997) 3–254

toroidal beta is high (b=5%), providing power densities near practical engineering limits (the peak neutron wallloading is 5.7 MW m−2). The toroidal-field (TF) coil system is designed with relatively ‘conventional’ materials(Nb3Sn and NbTi conductor with 316SS structures), and is operated at a design limit of �16 T at the coil in orderto optimize the design point. The ARIES-RS design uses a self-cooled lithium blanket with vanadium alloy as thestructural material. The V-alloy has low activation, low afterheat, high temperature capability and can handle highheat flux. A self-cooled liquid lithium blanket is simple, and with the development of an insulating coating, has lowoperating pressure. Also, this blanket gives excellent neutronics performance. Detailed analysis has been performedto minimize the cost and maximize the performance of the blanket and shield. One of the distinctive features of thisdesign is the integration of the first wall, blanket, parts of the shield, divertor and stability shells into an integral unitwithin each sector. The maintenance scheme consists of horizontal removal of entire sectors. Prior to the initiationof the ARIES-RS study, a set of top-level requirements and goals for fusion demonstration and commercial powerplants was evolved in collaboration with representatives from US electric utilities and from industry. The degree towhich ARIES-RS reached these requirements and goals and the necessary trade-offs are described and thehigh-leverage areas and key R&D items are presented. © 1997 Elsevier Science S.A.

Keywords: ARIES-RS; Reversed-shear; Tokamak power plant

1. Introduction

The ARIES-RS tokamak power plant studycontinues in the tradition of the ARIES Programto establish the economic, safety and environmen-tal potential of fusion power plants, and to iden-tify physics and technology areas with the highestleverage for achieving attractive and competitivefusion power in order to guide fusion R&D. TheARIES Team is a US national effort with partici-pation from national laboratories, universities andindustry, and with strong international collabora-tions. The Team performs detailed physics and

engineering analyses using the most current anddetailed models available, and then uses the re-sults to perform optimization and trade studiesvia a cost-based systems code.

Prior to the initiation of the ARIES-RS study,a set of top-level requirements for fusion demon-stration and commercial power plants wasevolved in collaboration with representatives fromUS electric utilities and from industry [1–4].These requirements were framed in a quantitativeway to help establish the minimum necessary fea-tures of a fusion power plant that would lead toits likely introduction into the US and world

Table 1Top-level requirements and goals for commercial and demonstration fusion power plants

DemonstrationElement Commercial

Must use technologies to be employed in commercial power plant Yes YesNet electric output must be greater than N/A75% CommercialCOE must be competitive (in 1992 mill (kWe h)−1) 80 (Goal) 65 (Goal)

90 (Reqmt) 80 (Reqmt)No evacuation plan required for any credible accident:

B1 remTotal dose at site boundary B1 remGenerate no rad-waste greater than Class C Class CMust demonstrate public day-to-day activity is not distrubed YesYesMust not expose workers to a higher risk than other power plants YesYes

YesYesMust demonstrate robotic maintenance of power core1/10Must demonstrate routine operation with less than (x) unscheduled shutdowns per year 1

including disruptionsDemonstrate a closed tritium fuel cycle Yes Yes

50%50%Must demonstrate oepration at partial load conditions at

F. Najmabadi et al. / Fusion Engineering and Design 38 (1997) 3–25 5

electric energy supply. These requirements consti-tute a common basis for quantitative evaluationsof any candidate fusion concept. A similar set ofcriteria has been developed by the EPRI fusionworking group [5]. These criteria and associatedtop-level requirements and goals (Table 1) can bedivided into three general categories: (1) cost; (2)safety and environmental features; and (3) reli-ability, maintainability, and availability.

Top-level requirements and goals for cost ofelectricity (COE) were adopted based on the esti-mated cost of competitive source of electricity atthe time of introduction of fusion in the marketplace [1,6,7]. These requirements and goals forCOE are also in line with projections of futurepower plant costs based on energy forecastingmodels [8]. Safety and environmental require-ments are included to circumvent the difficultiesexperienced by fission and, to some degree, ex-pected to be faced by fossil fuels in the future.Fusion should be easy to license by the nationaland local regulating agencies, and be able to gainpublic acceptance. Fusion power plants shouldonly generate low-level waste (i.e. waste storagetime less than a few hundred years, such as Class-A through C under US regulations). Realizationof the full safety and environmental potential offusion will also help fusion to achieve a costadvantage over other sources of electricity. Fusionpower plants can be designed to achieve thesecriteria only through the use of low-activationmaterials and care in design. However, these re-quirements result in stringent constraints on thesub-system choices and design. Lastly, it shouldbe demonstrated that Demo and commercialpower plants can achieve the necessary degree ofreliability. The conceptual design studies can par-tially address this issue (e.g. by including mainte-nance considerations in the design). This criterion,to a large degree, should be addressed in thedevelopment path of fusion power. Today’s exper-iments are not intended to provide detailed engi-neering data to support the design, construction,and operation of a power plant.

With the top-level requirements in hand, anassessment of various tokamak plasma operationmodes and engineering design options was made[1,3,9]. In each area, this assessment was aimed at

investigating: (1) the potential to satisfy the top-level requirements and goals; and (2) the feasibil-ity (e.g. critical issues) and credibility (e.g. degreeof extrapolation required from present data base).

Five different tokamak plasma regimes wereconsidered: (1) steady-state operation in the first-stability regime, e.g. ARIES-I [10]; (2) pulsed-plasma tokamak operation, e.g. Pulsar [11]; (3)steady-state operation in the second-stabilityregime, e.g. ARIES-II and ARIES-IV [12]; (4)steady-state operation with reversed-shear profile;and (5) low-aspect ratio tokamak (spherical toka-mak). Several ‘physics figures of merits’ wereidentified in order to assess the extent of theplasma physics data base for each option [1,13].

An assessment of the five tokamak physicsregimes was made based on economic perfor-mance and maturity of the data base [1,9,14]. Thefirst-stability pulsed-plasma and steady-stateregimes are closest to the present data base. Ofcourse these regimes should be demonstrated atlong-pulse discharges with burning plasmas. Onthe other hand, the economic performance ofpulsed-plasma operation is poor. First-stabilitysteady-state did not achieve the economic require-ments for the Starlite project. High-field toroidal-field coils can improve the attractiveness of thisregime of operation. The second-stability regimehas better economic performance but the experi-mental data base for this regime is very small. Thedata base for spherical tokamaks is not matureyet and, in addition, many critical issues remain.Detailed design studies are needed before the truepotential of spherical tokamak power plants canbe assessed. The reversed-shear mode of operationoffers the best economic performance. The database for this regime, while small, is growingrapidly. Based on the superior economic perfor-mance and the growing experimental and theoret-ical data base, the reversed-shear plasma mode ofoperation was chosen for detailed design phase.

Various classes of engineering design optionswere examined and their potential to meet thepower plant requirements assessed [1,15,16].These options included material choices for thestructure, breeder and coolant. The design spacefor an attractive tokamak fusion power core isnot unlimited; previous studies have shown that


advanced low-activation ferritic steel, vanadiumalloy, or SiC/SiC composites are the only knownviable candidates for the primary in-vessel struc-tural material. In order to provide a framework forthis assessment, these three material classes wereused to distinguish engineering design choices, todetermine if they meet the top-level requirements,and to characterize the remaining technical issues.

Silicon-carbide composite has a unique potentialfor safety and high performance, but the databaserequires substantial improvement in order to iden-tify and develop a satisfactory material composi-tion. Significant improvements are needed in thebasic properties, and several key material issuessuch as joining and hermeticity must be resolved.Ferritic steel has the largest database, and hencethe smallest uncertainty in its performance. How-ever, as with all materials, behavior after long-termoperation in a complete fusion environment ishighly uncertain. The features of ferritic steelwhich cause the greatest concern are its restrictedtemperature window, which limits the maximumachievable thermal conversion efficiency, and itspotential loss of ductility under irradiation. It isnot clear how one would operate a high-powerdevice with pressure-bearing structures made ofpotentially brittle materials. In addition designoptions have to be developed to address the lowmaximum temperature capability of ferritic steelsso that an acceptable power conversion efficiencycan be achieved [15]. Vanadium offers significantadvantages in its high temperature and high ther-mal performance capability, and its low activation.The combination of V-alloy structure and Licoolant appears particularly unique in materialchemistry, offering good compatibility up to veryhigh temperatures. The primary concerns withvanadium alloy are high cost and liquid metalMHD effects (including the need for and feasibilityof insulating coatings). Minimization of vanadiumin the shield and external systems is an importantcost-reduction strategy. The absence of an estab-lished industrial base is also an important consid-eration, but not one of the top-level requirementsas elaborated above. Due to its greater ultimatepotential for attractive commercial power plantsand a development path which appears practicalwithin a 25 year time-frame, the combination of Li

breeder/coolant and vanadium-alloy structure waschosen for full system design and analysis.

While there is no unique design concept guaran-teed to succeed, the ARIES Team choices (areversed-shear plasma with in-vessel componentsmade of high-performance vanadium-alloy struc-tures cooled by lithium) represent trade-offs be-tween a reasonable database and an acceptableperformance. For the detailed engineering design,the blanket and shield concepts were based onARIES-II [12]. The choice to develop an existingdesign, rather than explore entirely new concepts,was made in order to evolve a higher level of detailand sophistication in the design, which then allowsa much firmer basis for evaluating the attributesand a clearer understanding of the issues and R&Dneeds.

This paper provides an overview of the ARIES-RS design including physics (Section 2) and engi-neering (Section 3) analysis, design, and trade-offs.Safety and licensing issues are discussed in Section4. Section 5 examines the extent to which theARIES-RS has met the top-level requirements anddescribes the key R&D issues. The details of theARIES-RS study can be found in [17–23] in thisissue and in the ARIES-RS final report [24].

The ARIES-RS tokamak is a conceptual, D–T-burning 1000 MWe power plant. As with earlierARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner usingthe best available physics and engineering models.Detailed analyses of individual systems togetherwith system interfaces and interactions were incor-porated into the ARIES system code in order toensure self-consistency and to optimize towardsthe lowest cost system.

Fig. 1 shows the fusion power core cross sec-tion and Table 2 summarizes the key parametersfor the final design point of ARIES-RS. Thedesign employs a moderate aspect ratio (A=4.0).Preliminary observations indicated that the cost-of-electricity (COE) variation within the range35A54 was not significant. Engineering designconsiderations which could not be quantified accu-rately for systems-code analysis led to the choice ofA=4. The plasma current is relatively low (Ip=11.32 MA) and bootstrap current fraction is high( fBC=0.88). Consequently, the auxiliary power


Fig. 1. The ARIES-RS fusion power core.

required for RF current drive is relatively low. Atthe same time, the average toroidal beta is high(b=5%), providing power densities near practicalengineering limits (the peak neutron wall loadingis 5.68 MW m−2). The toroidal-field (TF) coilsystem is designed with relatively ‘conventional’materials (Nb3Sn and NbTi conductor with 316SSstructures), and is operated at a design limit of�16T at the coil in order to optimize the design point.

Following the ARIES-II/-IV [12] and PULSAR[11] designs, a level of safety assurance (LSA) rating[25,26] of LSA=2 has been assigned, and appro-priate cost credits (J.G. Delene, LSA cost factorsfor fusion reactors, Oak Ridge National Labora-tory, private communication, December 1990) havebeen taken. Table 3 summarizes the final set ofeconomic parameters.

2. Plasma physics

The ARIES-RS design operates in the reversed-

shear mode which has a considerable potential forproviding the basis of an economically competitivefusion power plant. The benefits of this configura-tion are that it achieves both high bN and high b,it obtains large bootstrap current fractions withvery good current profile alignment, and appears toprovide the transport suppression necessary tosustain the pressure profile that is consistent withthe higher b and high bootstrap current [27–29].

Like all configurations with bN values whichexceed the first stability regime this plasma requiresa conducting wall to stabilize low-n external kinkmodes. For ballooning modes the reversed shearplasma is in second stability in the region inside theminimum safety factor, and in first stability outsidethis region. An additional advantage of the reverseshear configuration is that the hollow currentprofile matches very closely with the natural shapeof the plasma generated bootstrap current profile.This allows one to significantly reduce the amountof external current drive required.

Recent experiments [30–32] have demonstrated


transiently the improved MHD stability and sup-pression of plasma particle and energy transport.The experiments each observed strong transportsuppression in differing degrees for energy andparticles. Both experiments have determined thatthe plasma core is in the second stability regimefor ballooning modes. In addition, both experi-ments obtain b limits consistent with ideal MHDstability predictions, and these are expected to beresponsible for the discharge terminations. Theglobal energy confinement times obtained in theseexperiments have H-factors over ITER89-P scal-ing of 2–3. In addition, bootstrap current frac-tions between 50 and 75% have been inferred.Complete demonstration of all the favorableproperties simultaneously will require steady stateconditions, which are not accessible in presentexperimental devices.

The reference plasma configuration is givenin Fig. 2 and Table 4. The operating b va-lue is reduced to 90% of the maximum stablevalue to provide some margin to the stabilityboundary.

Table 3Reverse-shear power-plant economic parameters (1992 $)

Account title CostAccountnumber ($ M)

Land and land rights 10.420.Structures and site facilities21. 331.0Reactor plant equipment 1390.122.First wall, blanket, and reflector22.1.1 74.3

22.1.2 168.0ShieldMagnets22.1.3 274.0TF coils22.1.3.1 163.4PF coils22.1.3.3 110.6Supplemental heating systems22.1.4 164.2Primary structure and support22.1.5 53.4

22.1.6 Reactor vacuum systems 163.5Power supplies22.1.7 55.3Impurity control22.1.8 13.6

22.1.9 Direct energy conversion system 0.022.1.10 4.3ECRH breakdown system

Reactor equipment22.1 970.722.2 Main heat transfer and transport 258.3

Turbine plant equipment23 284.4Electric plant equipment24 110.6

25 Miscellaneous plant equipment 56.211.1Special materials26

Total direct cost90 2193.891 Construction services and equipment 263.3

Home office engineering and services92 114.1Field office engineering and services93 131.6

94 Owner’s costs 405.496 Project contingency 524.5

Interest during construction (IDC)97 600.198 Escalation during construction (EDC) 0.0

Total capital cost 4232.999

Constant$

[90] 2193.8Unit direct cost, UDC ($ kWe−1)Unit base cost, UBC ($ kWe−1) 3632.8[94]

4232.9[99] Unit total cost, UTC ($ kWe−1)Capital return (mill (kWe h)−1) 61.46

9.16[40–47,51] O&M (1.4%) (mill (kWe h)−1)[50] Blanket replacement (mill (kWe h)−1) 4.63

0.50Decommissioning (mill (kWe h)−1)Fuel (mill (kWe h)−1) 0.03[02]

75.79COE (mill (kWe h)−1)

Table 2Operating parameters of the ARIES-RS tokamak power plant

Aspect ratio 4.005.52Major radius (m)

Minor plasma radius (m) 1.381.70Plasma vertical elongation (x-point)11.32Plasma current (MA)0.88Bootstrap current fraction81Current-drive power (MW)

Toroidal field on axis (T) 7.9816Peak field at the TF coils (T)0.05Toroidal b

Average neutron wall load (MW m−2) 3.96Natural lithiumPrimary coolant and breeder

Structural materials Vanadium & steelCoolant inlet temperature (°C) 330

610Coolant outlet temperature (°C)Fusion power (MW) 2170Total thermal power (MW) 2620Net electric power (MW) 1000

0.46Gross thermal conversion efficiencyNet plant efficiency 0.38Recirculating power fraction 0.17Mass power density (kWe ton−1) 66.70Cost of electricity (mill (kW h)−1) 75.79

2.1. MHD stability and bootstrap current

Studies were done to determine the impact ofplasma aspect ratio, triangularity, elongation,pressure and current profiles, and the kink stabi-


Fig. 2. The reference plasma equilibrium parallel current profile as a function of poloidal flux, the plasma flux surface contours, andtoroidal current density as a function of the major radius.

lizing wall position. The plasma current and pres-sure profiles were determined to provide a highstable b and large well-aligned bootstrap currentsimultaneously. This includes maintaining the min-imum value of the safety factor, qmin, above 2.0everywhere, maintaining q95 above 3.5, moving thelocation of qmin as close to the plasma boundary aspossible (to increase the high pressure plasmavolume), maintaining a degree of negative sheargiven by qo−qmin]0.3, and maximizing the matchbetween the bootstrap current profile and thedesired MHD stable current profile. The referencecase plasma profiles are given in Fig. 3.

The plasma geometry plays a significant role inMHD stability as well as the overall power plantdesign. The aspect ratio variations showed the wellknown trend that b increased as the aspect ratiowas lowered, however, the current drive powerincreased as well. System analysis indicated that theCOE variation within the range 35A54 was not

significant. Engineering design considerationswhich could not be quantified accurately for sys-tems-code analysis led to choice of A=4. Scans ofthe plasma triangularity from 0.2 to 0.6 showedthat a triangularity above 0.4 was necessary forhigh b and that b increased further with increasingtriangularity. For ARIES-RS, the triangularity waslimited to 0.5 because of the inboard divertorrequirements (mainly space requirements for suffi-cient neutron shielding).

Plasma elongation improves b, but makes theplasma unstable to vertical motion. The maximumelongation is limited by the passive-stabilizationconducting structure and the feedback controlsystem. The conducting structure was designed toprovide a stability safety margin ( fs=1+tg/tL/R)of 1.2, and was made of tungsten 4.0 cm thick. Thetungsten structure was made electrically continuousin the toroidal direction and located in the gapbetween the reflector and shield. The plasma elon-


gation at the separatrix (and 95% surface) waslimited to 1.9 (and 1.7). Feedback control simula-tions for the plasma vertical position showed that15 MVA was required if the control coils werelocated in the gap between the reflector andshield, and 40 MVA was required if they werelocated behind the shield (note that most of thispower is reactive power).

It is important to emphasize that a conductingwall is required to stabilize low-n kink modes forthese configurations to obtain high b values. Sinceno second stability regime (similar to that for then�� ballooning modes) has been found for thelow-n kink modes, it is necessary to utilize aconducting wall to access the higher pressureswhere a fusion power plant can become signifi-cantly more attractive. Recent experiments andtheory indicate that a resistive wall can stabilizethe low-n kink modes if plasma rotation and adissipation mechanism in the plasma are present.

Predictions from ideal MHD theory indicatethat the plasma must rotate at about 5–10% of

the Alfven speed, which for ARIES-RS is 2.8–5.6×105 m s−1. Neutral beams are capable ofproviding such plasma rotation speeds. On theother hand, ARIES-RS uses only RF currentdrive. The highest plasma rotation speeds fromRF were observed for ICH at 3×104 m s−1.Whether these RF driven plasma rotation speedswill be sufficient for kink mode stabilization is notclear, and this is presently an active area of re-search. In principle, low-power beams can beadded to the system for plasma rotation if RFsystems cannot provide sufficient rotational speed.The theory also gives some criteria for the wallconductivity and thickness. For the ARIES-RSdesign, the criteria is D/h\1.5×104 V, where Dis the thickness and h is the wall resistivity. Scansfor the distance of the wall from the plasmashowed that bN was limited to 5.35 for walllocations closer than 0.25a from plasma separatrix(a is the plasma minor radius). This was due tothe bootstrap current becoming too large, eventhough the b could actually be increased as thewall was made closer. For ARIES-RS, the kinkwall is integrated in the blanket design by slightlythickening the structure behind the first wall cool-ing channels.

2.2. Current dri6e

In the ARIES-RS reference equilibrium, thebootstrap current fraction is 0.88, using densityand temperature profiles which are characteristicof reversed shear plasmas. External non-inductivetechniques are generally required to drive currentsat the plasma center and off-axis near the shearreversal region. A number of non-inductive cur-rent-drive techniques have been considered for thereversed shear power plant. Depending on thereference equilibrium, a combination of thesetechniques is required to drive currents in differ-ent locations within the plasma. Because of theirnatural tendency to propagate radially towardsthe plasma center, ICRF fast waves are mostsuitable for current drive on the magnetic axis. Onthe other hand, with a proper launched spectrum,the lower-hybrid wave can be made to drive cur-rents off-axis at the shear reversal region. How-ever, lower-hybrid waves typically cannot

Table 4Reference reversed shear plasma configuration

Plasma current, Ip (MA) 11.37.98On-axis toroidal field (T)

Major radius (m) 5.52Minor radius (m) 1.38Elongation k95 1.70Triangularity d95 0.50Poloidal b 2.28a

4.96%aToroidal b

Toroidal b* 6.18%a

Normalized bN 4.83%a

Maximum bN 5.35%a

Bootstrap current (MA) 10.0Driven current, ICD (MA) 1.2On-axis safety factor, qo 2.80Minimum safety factor, qmin 2.49Location of qmin, c. (qmin) 0.69Edge safety factor, qe 3.52

2.37Edge safety factor, q*0.42Internal inductance, li(3)1.36Density profile peaking factor, n0/�n�1.98Temperature profile peaking factor, T0/�T�2.20Pressure profile peaking factor, p0/�p�0.095bLocation of kink-stabilization wall

a The ARIES-RS operates at 90% of maximum theoretical b.b Distance fom the plasma edge normalized to minor radius.


Fig. 3. The reference plasma equilibrium profiles for the safety factor, pressure, temperature, and density as a function of poloidalflux.

penetrate deeper into the plasma and the choice ofanother RF technique for off-axis (mid-plasmalocation) current drive is necessary. Several candi-dates have been considered, including high-fre-quency fast waves (HFFW), mode conversioncurrent drive (MCCD), minority heating currentdrive (MHCD), and electron-cyclotron waves. Allthese techniques have the potential capability oflocalized off-axis current drive in a fusion plasmacore, each using a distinct physical mechanism.

For the ARIES-RS design, HFFW was chosento aid the lower-hybrid wave for off-axis currentgeneration. Other candidates may be equally at-tractive, but have not been fully explored in thisstudy mainly because of the lack of resources withinthe project. Indeed, seeking the most attractivetechnique to drive currents off-axis in the plasmacore is presently a topic of intense research interest.

A large number of current drive calculations

were performed to scale the current drive efficiencyas a function of plasma aspect ratio, temperature,and Zeff for the reversed shear plasma configura-tions. Shown in Fig. 4 is the reference plasmaequilibrium with the various current drive compo-nents. All three systems were used in this particularcase, even though in some instances, only two oreven one system may be sufficient. All requiredwave spectra have values of N that can be launchedby reasonable launcher designs. In this particularcase study, the aggregate current-drive efficiency isfound to be gB=1.28×1020 A W−1 m−2, and thecurrent-drive system details are given in Table 5. Aradiative divertor was utilized to disperse the trans-port power and keep the heat-flux at the divertorregion to an acceptable level. The impact of theradiative divertors on the choice of MHD equi-librium and current-drive power constitutes a newfeature of the ARIES-RS physics analysis and isdescribed in Section 2.3 below.


Fig. 4. Matching of the driven (dotted line labeled T) currentdensity profile to the target profile (solid line labeled E), usingICRF fast wave (FW), high-frequency fast wave (HFFW), andlower-hybrid wave (LH) power, for a reversed shear equi-librium similar to the ARIES-RS design point.

and main plasmas, often referred to as ‘radiatingdivertor’ and ‘radiating mantle’ methods, respec-tively. In the former, excessive power flowing intothe divertors is dissipated largely by radiation in thedivertor channel. In the latter, this power flow is ra-diatively dissipated at the edge of the plasma beforeit passes into the scrape-off layer (SOL). Both met-hods require injection of high-to-medium Z impuri-ties into the plasma (which increases Zeff) and oper-ation at a high edge density (nea/ne0]0.2). Analysisindicates that increased Zeff severely degrades theoverall current-drive efficiency. In addition, thesubstantial edge density implies that a larger frac-tion of the current-drive power will be in the lowerhybrid system, which has the lowest system effi-ciency and the highest unit power cost among thethree RF systems. The lower-hybrid power fractionincreases at higher Zeff and lower temperatures.

Analysis showed that the ‘radiating mantle’ op-tion (which requires a higher Zeff and/or a higheredge density) leads to systems with unacceptablecurrent-drive power requirements. Therefore, aradiative divertor option was chosen which re-quired a lower edge density, nea/ne0�0.2 and alower Zeff�1.7.

For the radiative divertor case, the impuritychosen was neon, since this atom radiates effi-ciently in the cooler scrape-off and divertor re-gions. While the higher Z impurities can alsoradiate efficiently, their impact on the core plasmais more severe than the lower Z neon. Thesestudies indicated that adding neon to the plasmasystem, under conditions consistent with otheranalyses (i.e. MHD stability, current drive, boot-strap current, and transport), would radiate a

2.3. Di6ertor physics

Because of the high power density of the ARIES-RS design, the heat fluxes on divertor structures canbe very high. Material surfaces under such highheat loads are subject to severe thermal-relatedstructural damage. Thus, to avoid compromisingthe structural integrity of the divertor it is necessaryto reduce the heat fluxes to more manageable levels.

To avoid overheating the divertor surfaces, thepower flow into the scrape-off can be dispersed overa wider area. One way this can be done is byenhancing the radiated power from the divertor

Table 5Current drive requirements for ARIES-RS-like equilibrium

FrequencySystem Launcher position (°)aPower (MW)N

2.0 5.491.0 MHzICRF fast wave 151.0 GHz 2.3 21.4High-frequency fast wave 0

27.0 01.0 GHz 2.5Lower hybrid wave −154.6 GHz 1.9 9.5

−154.6 GHz 2.1 2.856.42.33.5 GHz

5.13.5 GHz 52.53.5 GHz 510.53.0

a In poloidal angle above (+) or below (−) outboard mid-plane.


Fig. 5. Plasma operating contours of the auxiliary power (inincrements of 10 MW) required for steady-state plasma powerbalance displayed in ion density and temperature space for thefollowing assumptions: ITER-89P confinement scaling with amultiplier H=2.339, Ip=11.3 MA, Ip/apBT=1.028 MAmT−1, ap=1.38 m, fNe=0.0055, and tp*/tE=10.

tion (i.e. bremsstrahlung and line radiation) hill atlow temperature and a transport hill at high tem-perature that results from increasing power levels,decreasing confinement times, and increasing ashconcentration with increasing temperature. Theminimum-power trajectory requires a maximumof �15 MW of auxiliary power at Ti:7 keV andni:6×1019 m−3. The start-up trajectory to theARIES-RS operating point requires four timesmore power, but the maximum auxiliary poweroccurs at the final operating point, because theoperating point resides on the side of the high-temperature transport hill.

As was mentioned before, plasma transport isnot explicitly modeled. Plasma temperature anddensity profiles are only made to roughly agreewith theoretical and experimental observations forreversed shear configurations. The global plasmaenergy confinement time is determined by thesystems code to provide plasma power balance.The energy confinement time is 1.34 s, whichcorresponds to an ITER-89P scaling multiplier of2.34. This value is within the present experimentalbounds for reverse shear plasmas, as well as otherplasma operating modes.

The particle confinement time in the mainplasma is expected to track the energy confi-nement time. Experiments on helium removalwhich used significant active pumping indicatethat the ratio tp*/tE is in the range of 7–15 (tp* isthe effective particle confinement time includingrecycling). For the ARIES-RS design a value oftp*/tE=10 is used.

Controlling the edge plasma density, n(a) at theseparatrix boundary is important for avoidingdisruptions. For ARIES-RS density limits are pre-dicted to be in the range of 0.6–1.75×1020 m−3.The edge density ratio n(a)/n0 of 0.2 satisfies thelower limit and is consistent with the divertor,ideal MHD stability, and current drive solutions.

3. Fusion power core engineering

The ARIES-II power core [12] has been used asa starting point from which additional designdetail and improvements can proceed. TheARIES-RS design process was driven by the need

sufficient amount of power in the SOL and diver-tor regions to make a feasible divertor solution. Inthis case, the sum of particle and radiation heatflux does not exceed 6 MW m−2.

2.4. Plasma operating regime

A Plasma OPerating CONtour (POPCON) plotprovides a rudimentary indication of a start-uptrajectory and thermal stability of the steady-stateoperating point. The contours of the auxiliarypower required for steady-state plasma power bal-ance as a function of fuel density and temperatureare displayed in Fig. 5. Ignited plasmas are possi-ble for fuel temperatures in the range of 95Ti515 keV and fuel densities ni\2×1020 m−3. Inregions outside of the ignition region transportand radiation losses exceed the fusion power, andauxiliary power is required for steady-state opera-tion. An ARIES-RS-like operating point is main-tained with an auxiliary power of �81 MW(current-drive power).

The start-up trajectory to fusion burn that re-quires the minimum auxiliary power follows thevalley floor (in auxiliary power) between a radia-


to meet the top-level system requirements. Meet-ing economic goals while maintaining the safetyand environmental attractiveness possible with afusion energy source was a major thrust. Forexample, minimization of the use of vanadium toonly those areas where its use is necessary forperformance reasons made a significant impact onthe COE.

Plant availability is another significant factor inthe COE. Modern fission plants are aiming at80% plant capacity factors [33], implyingavailability of 85% or more. Availability is espe-cially important with high capital-cost facilitiessuch as a fusion plant. The layout of the powercore, as well as the detailed design of the internalcomponents, was continuously influenced by thegoal to reduce the down-time for sector repairsand replacement to about 1 month.

Finally, attempts were made to develop designsolutions which integrate both physics and engi-neering constraints. For example, the require-ments for both passive and active stability systemswere incorporated into the sectors from the initialstrawman and revised as necessary during thedesign process. The divertor is another area whereefforts to develop physics solutions consistentwith engineering limitations, and engineering so-lutions consistent with physics constraints, weremade throughout the design study.

In the following subsections, the overall designfeatures of the power core are summarized andthe principle features and conclusions from theindividual components are described. These in-clude the first wall and blanket, radiation shield-ing, divertor, magnet systems, and current drive,heating and fueling systems.

3.1. Configuration and maintenance

Achieving a high plant availability goal was aprimary influence on the overall configuration andon the major power core design decisions. Thisrequirement is met by emphasizing: (a) rapid re-placement time (1 month is the goal) via configu-ration design; (b) component reliability via designsimplicity and operating margins; (c) long lifetimevia material selection; and (d) horizontal mainte-nance of full sectors is performed through large

ports (Fig. 6) using a rail system together withtransporter casks. Port doors on the back of theshield and at the cryostat prevent the spread ofradioactive contamination to the confinementbuilding. All coolant connections are made in theevacuated port area where the radiation field islow. This design allows rapid disconnections ofthe piping, which could be either mechanicallysealed or cut and rewelded.

One of the distinctive features of this design isthe integration of the sectors. The first wall, blan-ket, parts of the shield, divertor and stabilityshells form an integral unit within each sector(Fig. 7). The integrated sector construction elimi-nates in-vessel maintenance operations and pro-vides a very sturdy continuous structure able towithstand large loads. Cavity loads are supportedat the bottom through the vacuum vessel. Sectorsare disassembled and reusable parts maintained inhot cells after the plant returns to operation. Norewelding is needed for elements located withinthe radiation environment.

Fig. 6 highlights the significant impact of thehorizontal maintenance of a complete sector as aunit on the reference design. The outboard TF-coil leg must be displaced 2.56 m from the back ofthe vacuum vessel to provide the necessary spacefor single-piece removal of the breeding blanketand structural ring out through the access tunnelbetween adjacent TF coils. As the TF coil loca-tion is moved further from the vacuum vessel toaccommodate the horizontal maintenance scheme,the PF coils are also displaced further from theplasma, thereby increasing the PF-coils currentand mass. To minimize cost penalties associatedwith a larger TF coil (i.e. increased TF and PFcoil costs), the TF coil was deformed from aconstant-tension shape, as is shown in Fig. 6.Although the TF coils are supported by a cap, thedeformation is limited by the magnitude of thebending stress that can be tolerated across theTF-coil structure.

The most severe penalty of single-piece sectorsis the increased size of both the TF and PF coilsystems, needed to allow adequate space for sec-tor removal. With an optimized design, the cost ofthe increased size of the TF and PF coils is �2–3mill (kW h)−1, which is substantially smaller thanthe cost savings due to the increased availability.


Fig. 6. The elevation view of the ARIES-RS fusion power core.

3.2. First wall, blanket and shield

The ARIES-RS design uses a self-cooledlithium design with vanadium alloy as the struc-tural material. It was judged by the ARIES teamthat this blanket has the best potential to fulfil theplant requirements with a moderate extrapolationof today’s technology. The V-alloy has low activa-tion, low afterheat, high temperature capabilityand can handle high heat flux. Also, this blanketgives excellent neutronics performance. One mainconcern with liquid-metal cooled systems is theMHD pressure drops. In typical tokamak powerplant designs, the MHD pressure drop for theinboard regime of a tokamak results in pressurestresses beyond the material limits [12]. To reducethis MHD pressure drop, some form of insulatingwall is required, either with a direct coating orwith a sandwich-type structure.

With the assumption of reliable insulating coat-ings, the MHD pressure drop is no longer a majorconcern. The design of the first wall, blanket, andshield can be optimized to improve heat transferand to simplify the configuration. The first walland breeding blanket use a simple box-like struc-ture, with lithium coolant flowing in poloidalpaths. The outboard cross section is shown in Fig.8. The design uses an insulator coating such ascalcium oxide on all coolant channel interior sur-faces. This can be achieved by adding 0.5% Ca inthe flowing Li. In principle, this insulator can bereplaced by any other self-healing insulator with-out any impact on the overall performance of thedesign. The development of insulating coatings isat a very early stage and much more R&D isrequired. The improvements and design flexibilitythat coatings provide for self-cooled liquid metalblankets make this a very high leverage item.


Fig. 7. The elevation view of an integrated ARIES-RS sector.

Table 6 summarizes the heat loads and peaktemperatures in the blanket and divertor. Multipleflow passes in the blanket provide the capabilityfor removing up to 1 MW m−2 of surface heatflux, although the radiative divertor mode of op-eration of ARIES-RS results in a peak of only 0.4MW m−2. The full coolant flow is passed firstthrough the front zone, where the surface heatflux creates large temperature gradients, and thenthrough the back zones were the bulk temperaturecan be raised by volumetric heating without ex-ceeding any structure temperature limits. Segmen-tation of the shield into a hot and a cold zoneallows partial utilization of the heat deposited,and also provides further capability for superheat-ing the coolant away from the high heat fluxregion.

An important feature of the blanket and shieldis the radial segmentation into four zones: twoblanket zones and two shield (high- and low-tem-perature) regions to maximize the lifetime of thestructures, reduce the replacement cost, and mini-mize the waste stream. Scheduled replacement

occurs after 2.5 full power years (FPY) when theV-alloy reaches 200 dpa (for ARIES-RS, thistranslates to �15 MW m−2). At that time, thefront portion of the blanket is disposed, but therear portion of the blanket can be used until itreaches its own radiation lifetime, at about 7.5FPY. This rear portion also serves as the struc-tural ring, which provides poloidal continuity tothe sectors and attachment points for the innerblanket segments. Location of the coolant connec-tions outside the vacuum vessel allows for easydisassembly of the segments. Radial segmentationcreates safety concerns, since radial heat transportpathways are critical in loss-of-coolant scenarios.Design solutions have been proposed; however,blanket response to coolant loss remains an im-portant concern.

Detailed shielding analysis has raised no seriousdifficulties with ARIES-RS shields. A dedicatedeffort was devoted to the bulk shield in particular,as it represents a major cost item for advancedtokamak designs. Significant savings in shield costwere obtained by these strategies: (1) Expensive V


Fig. 8. The cross section of the ARIES-RS outboard blanket and shield.

alloy is limited to those regions where high tem-perature operation is absolutely necessary. (2)Shield is segmented into high temperature andlow temperature zones. The nuclear heat de-posited in the high-temperature shield (�20% oftotal heating) is recovered as high grade heat toenhance the overall power balance of the ma-chine. The low-temperature shield, vacuum vessel,and other external components (having low levels

of nuclear heating) can employ cheap stainlesssteel as the main structural material, instead of V.(3) Cheap steel filler (rather than V filler) in theshield reduces the cost tremendously. Fillers haveno structural role and thus have lower unit costscompared to structures. (4) Highly efficient, ex-pensive materials, such as WC and B4C, are usedonly in the space-constrained inboard side to re-duce the overall size and cost of the machinewhile less efficient, cheaper materials is used in thedivertor and outboard sides.

The ARIES-RS design uses both active andpassive stabilization systems for vertical displace-ment and kink-mode stabilization. These systemswere integrated into the sectors (Figs. 1 and 7).Radiatively-cooled tungsten shells are located be-tween the blanket and shield on the inboard andoutboard sides to provide passive stabilization ofvertical displacement modes. They are augmentedby actively-cooled coils which are placed out-board between the low-temperature shield and thevacuum vessel. A thickened vanadium ‘secondwall’ behind the first wall cooling channel wasshown to provide sufficient conductivity to stabi-lize kink modes.

Table 6Power flows and peak temperatures

Multiplied neutron heating (MW) 2092431Total transport power (MW)

Bremsstrahlung power (MW) 56Core line radiation (MW) 25

341Power reradiated in divertor (MW)165First wall surface heating (MW)348Divertor total surface heating (MW)88Divertor particle power (MW)

Blanket bulk outlet temperature (°C) 610610Divertor bulk outlet temperature (°C)

�700Peak V temperature in first wall (°C)Peak V temperature in divertor (°C) 681


Fig. 9. The ARIES-RS divertor.

The design of the primary loop and powerconversion system are critical to the attractivenessof the power plant. This system is responsible forefficient power conversion, isolation of radioac-tive products within the nuclear island, and reli-able operation of the power plant. The choice ofan advanced Rankine cycle conservatively offers46% gross thermal conversion efficiency. Highthermal efficiency is desirable to partially offsetthe high capital cost of fusion. A double-walledintermediate heat exchanger (IHX) with an Nasecondary loop is used to isolate the activated Liprimary coolant from the stream side. The IHX isalso the location where the transition from V tostainless steel is made. The piping which connectsthe blanket to the IHX uses a double-walledstructure with a thin V liner to minimize theadded cost of vanadium.

The tritium recovery process proposed here isbased on cold trapping. For the liquid lithiumsystem, the hydrogen solubility at the cold traptemperature of 200°C is 440 appm, which is farabove the design goal of 1 appm. For this reason,cold traps have not been considered previously asa candidate process for recovering tritium fromlithium. However, the concept developed here ismodified to add protium in the lithium so that the

total hydrogen concentration in the lithium ishigher than the 200 appm saturation value. At200°C, Li(H+T) will be supersaturated and pre-cipitate out together. The Li(T+H) can be sepa-rated from lithium using a ‘meshless cold trap’process which was developed by the breeder pro-gram to separate NaH from Na by gravitationalforce. The Li(T+H) can then be heated up to600°C for decomposition. The hydrogen streamwill then be fed to the main isotope separationsystem (ISS) to separate tritium from protium.

3.3. Di6ertor

The divertor region of the sector, highlighted inFig. 9, consists of two principal parts: the targetplates and the structures. The structures fulfilseveral essential functions: (1) it provides for me-chanical attachment of the plates through ad-justable screw-bolts which provide for modulealignment and offer lateral flexibility for thermalexpansion together with strong support againstEM events (e.g. disruptions); (2) it shields themagnets; (3) it provides coolant routing paths forthe plates as well as the inboard blanket andreplaceable shield; (4) divertor-plate coolant isrouted through this region and is superheated.


Low outlet temperature of divertor plate coolantallows optimization of surface heat removal whilea high outlet temperature for the divertor systemis maintained; and (5) it contributes to the breed-ing ratio, since the coolant is Li.

The target plates include three pieces: inboard,outboard, and ‘dome’ plates. The plasma flowsthrough the scrape-off layer and enters the diver-tor, where enhanced line radiation from injectedneon impurity allows much of the power to bedistributed along the plates and also partiallyredirected out to the first wall. Most of the unradi-ated particle energy strikes the outboard plates,but the peak heat flux (Combined Radiation andTransport) has been maintained below 6 MWm−2. The strike points are located close to thecoolant inlet in order to maintain the vanadiumstructures below 700°C.

A 2-mm thick castellated W coating is appliedto the coolant channel front surface, which is only1-mm thick V to satisfy temperature limits. Ther-mal stresses are reduced by using a relatively thicksolid back on the target plates. The plates areconnected to the rear zone via strong adjustablescrew-type attachments. These attachments can bedesigned to react to the full force of disruptionsand also accommodate thermal expansion. Theyalso permit precise alignment to adjacent surfacesand removal of individual plates in the hot cells.

Vacuum pumping ducts are placed behind thedome near the strike points for efficient exhaust.Radial channels then direct the gas to a single setof cryopumps at the bottom of the machine.Sufficient conductance between top and bottomdivertor pumping ports are provided by using theinter-sector vessel volume underneath the TFcoils.

3.4. Magnet systems

The ARIES-RS toroidal field (TF) coil set con-sists of 16 coils using multi-filamentary Nb3Sn andNbTi superconductors with a peak field of 15.8 Tat the coil. The TF coils are flattened verticallyaway from a constant-tension D-shape (Fig. 6) inorder to reduce the size of both TF and poloidal-field (PF) coils and the peak fields. The designoptimized the superconductor, copper, helium, in-

sulation, and structural ratios by using fourgrades of conductor. Advanced magnet designtechniques maximize the utilization of structuralmaterials by using structural forms with groovesinto which the conductor is wound. Rather thanwinding all of the materials in the magnet, onlythe conductor requires winding.

Support of out-of-plane loads has been pro-vided without intercoil structure in the outer legsof the TF coils, using caps and outer straps. Thismakes full-sector maintenance possible. The capand strap structures also have been shown toaccommodate off-normal events (e.g. a single coilshort during a dump) and the bending stresses.

The PF coil set consists of 22 coils: eight formthe center stack, and the remaining 14 elongatethe plasma, provide equilibrium and form thedivertor magnetic configuration. An attempt hasbeen made to keep the field at the PF coilsB8 T,such that less expensive NbTi conductors can beused. This is accomplished by shaping the TF coilsand by allowing a larger number of closely-spacedPF coils.

3.5. Heating and current dri6e systems

The three RF current-drive systems make use ofICRF fast waves (98 MHz), high-frequency fastwaves (1.0 GHz), and lower hybrid waves (3.0–4.6GHz) to drive currents in the on-axis, off-axis(inside the reversed-shear region), and edge re-gions, respectively. The power from these systemscan be used to heat the plasma from start-up to itsfinal operating conditions. The total power deliv-ered to the plasma from these systems is 102 MW,of which 81 MW is used in driving useful currents.The wave launcher and transmission systems aredesigned and configured so as to minimize intru-sions into the periodic blanket and shield struc-tures. All RF launchers can fit with a singleblanket sector occupying 0.58% of the first wallarea. As a result, the engineering impact are mod-est, and the effects on shielding and waste disposalare manageable. The launchers are located aroundthe outboard mid-plane. The transmission linesand waveguides are routed behind the shieldingstructures to minimize neutron streaming and irra-diation of ex-vessel components.


4. Safety and licensing

The ARIES-RS safety design and analysis fo-cused on achieving two key objective: (1) theavoidance of sheltering or evacuation in the eventof an accident; and (2) the generation of onlylow-level waste, e.g. Class C [34]. The achieve-ment of the first objective requires that the doseper event not exceed 1 rem at the site boundary.The achievement of the second objective dependson the limits adopted for Class C waste.

Detailed activation analysis was performed toidentify the safety, environmental and rad-wastecharacteristics of the ARIES-RS power plant. Therad-waste of the different regions of the ARIES-RS were evaluated according to both the NRC10CFR61 [34] and Fetter [35] waste disposal con-centration limits. The analysis reveals that allcomponents would qualify for near surface shal-low land burial as Class C low-level waste orbetter (Class A waste). For example, Table 7 liststhe waste disposal ratings for the blanket. Thewaste disposal rating is defined as the sum over allisotopes of the ratio of the concentration of aparticular isotope to the maximum allowed con-centration of that isotope. If WDR is less than 1,the waste can be disposed of as low-level waste inthat class. This table shows values for Class Aand Class C, and includes both uncompacted andcompacted (in parentheses) cases.

A major advantage of the ARIES-RS vana-dium structure is that it generates low levels ofintermediate and long-lived radioactivity com-

pared to other metallic structural materials. Thus,safety concerns for the formation of highly ra-dioactive isotopes in the blanket and shield aregreatly eased. The radioactivity generated inARIES-RS is only slightly higher than an all-vanadium system because of the steel filler(Tenelon) used in the reflector and shield.

Regulatory limits require that the radiationdose at the 1 km site boundary be less than 1 remto ensure that the impact of an accident on thesurrounding population is sufficiently small sothat no evacuation plan is necessary. To deter-mine the ability of the ARIES-RS design to meetthis goal, an assessment was made of the impactof a loss-of-coolant accident (LOCA) on theARIES-RS device and its environment. This par-ticular accident was selected because it is poten-tially one of the most harmful. It was assumedthat the plasma quenches simultaneously with theonset of a LOCA. The baseline blanket designwas shown to exhibit peak temperatures in thefirst wall as high as 1200°C (due to afterheat).Knowing the material compositions and tempera-ture histories of the various components enablesan evaluation of the elemental release rates of thevarious materials. Radioactive inventories wereevaluated based on 2.5 FPY of neutron irradia-tion resulting from an average first wall loading of4 MW m−2, so as to allow for a significantbuildup of activation products. An assumption ismade that all coolant is lost, either through abreak in the system or because the coolant wasdumped as part of the response to another off-normal event. The results have confirmed that theworst-case accident scenario does not exceed the1-rem evacuation limit (no credit for a confi-nement building is taken).

The ARIES-RS design meets the safety crite-rion that the dose be less than 1 rem at the siteboundary during a severe LOCA. No active safetymeasures are required to meet this criterion, andno evacuation plan should be necessary for thisextreme scenario. However, because of the sever-ity of this accident scenario, and the assumptionsmade in the analysis, it would be reasonable toassume that this is a bounding case and that anycredible accident occurring during operationwould result in an even lower off-site dose.

Table 7First wall and blanket waste disposal ratingsa,b

OutboardInboardWaste-disposal rating

Class A (10CFR61) 0.3 (1.77)0.2 (2)94Nb, 14C94Nb, 14C

0.02 (0.2)Class C (10CFR61) 0.032 (0.19)94Nb, 14C94Nb, 14C

0.016 (0.16) 0.029 (0.17)Class C (Fetter)94Nb, 26Al 94Nb, 26Al

a Waste disposal rating is defined as the ratio of the concentra-tion of a particular isotope to the maximum allowed concen-tration of that isotope (then summed over all isotopes).b For uncompacted and compacted (in parentheses) waste.


5. Design evaluation and R&D needs

Prior to the initiation of the ARIES-RS study,a set of top-level requirements and goals for fu-sion demonstration and commercial power plantswas evolved in collaboration with representativesfrom US electric utilities and from industry (Table1). These requirements were framed in a quantita-tive way to help establish the minimum necessaryfeatures of a fusion power plant that would leadto its likely introduction into the US and worldelectric energy supply. These requirements consti-tute a common basis for quantitative evaluationsof any candidate fusion concept. The degree towhich ARIES-RS reached these requirements andgoals, and the necessary trade-offs are describedbelow. In addition, a major by-product of studieslike ARIES-RS is the identification of high-lever-age areas and key R&D items. These are de-scribed later in this section.

5.1. Design e6aluation

The top-level goals and requirements can bedivided in three general categories: (1) cost; (2)safety and environmental features; and (3) reli-ability, maintainability, and availability.

5.1.1. Cost of electricityTargets for the life-cycle cost of electricity were

determined in consultation with our Utility Advi-sory Committee: a requirement of 80 mill (kWeh)−1 and a goal of 65 mill (kWe h)−1 (in constant1992 dollars). Using best estimates for all ele-ments of the cost breakdown, it was determinedthat the ARIES-RS design point at 76 mill (kWeh)−1 can meet the requirement of 80 mill (kWeh)−1, but does not attain the goal of 65 mill (kWeh)−1, thought to be representative of the futurecompetition. Thus, further improvements and in-novation in the ARIES-RS design should con-tinue to be sought.

5.1.2. Safety and licensingSafety and licensing remain top priorities for

potential operators of fusion power plants. Priorto the design phase of the project, top-level safetyand licensing requirements were established for

fusion power systems and a strawman pathwaywas proposed for the development of fusion regu-latory policy and requirements [36–39]. Two ofthe dominant concerns are avoiding high-levelnuclear waste and reducing the on-site hazardpotential to the point where a worst-case accidentwould not require evacuation of off-site person-nel. Through analysis of activation products andLOCA accident scenarios, all components of theARIES-RS design were shown to meet Class-Cwaste disposal guidelines, and the worst-caseLOCA radioactivity releases were demonstratedto result in off-site exposures below 1 rem, suchthat public evacuation should not be necessary.

5.1.3. Reliability, maintainability, and a6ailabilityLack of a data base and engineering experience

with fusion power systems precludes any assess-ment of reliability and availability of ARIES-RSdesign. This is an issue that, to a large degree,should be addressed in the development path offusion power as today’s experiments are not in-tended to provide detailed engineering data tosupport the design, construction, and operation ofa power plant. Conceptual design studies canpartially address this issue through design by min-imizing the downtime of the plant during mainte-nance. Horizontal replacement of an integratedsector has been used in ARIES-RS as an effectivemeans to minimize the downtime. The integratedsector construction eliminates in-vessel mainte-nance operations and provides a very sturdy con-tinuous structure able to withstand large loads. Inorder to improve plant reliability, engineering andphysics margins were used when prudent. Forexample, ARIES-RS operates at 90% of maxi-mum allowable b to minimize the frequency ofdisruptions.

Achieving the top-level requirements involvestrade-offs. Design safety factors were imposed onseveral key parameters such as the plasma b limitand peak stresses in the structures in order toimprove system reliability but they increase theCOE. The ARIES-RS maintenance scheme re-quires an increase in the size of the TF coils andnecessary modifications in the TF coils and PFsystems. All these items increased the overall sys-tem cost but were judged to be acceptable in order


to achieve reliability and availability goals. TheARIES-RS blanket and shield was radially seg-mented to reduce component cost and minimizethe waste stream. This radial segmentation led toan increase in the maximum structure temperatureduring the LOCA. While the safety goal of ‘no-evacuation plan’ was realized, a passive safetyloop had to be introduced in order to ensure thatcomponents are not damaged during a LOCA.

5.2. R&D needs

In the design of ARIES-RS, an attempt wasmade to avoid physics and technology extrapola-tions beyond credible limits to be expected for a20–40 year R&D program. For example, designsafety factors were imposed on several keyparameters such as the plasma b limit and peakstresses in the structures. At the same time, ad-vanced physics and engineering were assumed inareas of high leverage. In order to assure theselevels of performance in ARIES-RS, critical R&Dtasks must be implemented in both physics andtechnology. Some of the more critical R&D needsidentified in this study are highlighted below.

5.2.1. Demonstration of a stable and controllableoperating point

The benefits of the reversed-shear plasma oper-ating mode are that it achieves both high bN andhigh b, it obtains large bootstrap current fractionswith very good current profile alignment, and itappears to provide the transport suppression nec-essary to sustain the pressure profile that is consis-tent with the higher b and high bootstrap current.Recent experiments have demonstrated transientlythe improved MHD stability and suppression ofplasma particle and energy transport. Completedemonstration of all the favorable propertiessimultaneously will require steady state condi-tions, which are not accessible in present experi-mental devices. In addition, it should bedemonstrated that plasma parameters (e.g. den-sity, temperature, position, etc.) can be controlledprecisely as large variation in fusion power causesignificant problems for the designs of fusion corecomponents. Essential to this regime is control ofcurrent and pressure profiles. Understanding of

transport barriers may be a way to control thepressure profiles necessary for advanced tokamakmodes.

5.2.2. Ignition physicsPhysics of burning plasmas and a-particle dy-

namics remains unresolved issues for fusion re-search. Obtaining advanced tokamak modes suchas reversed-shear, in the presence of dominanta-particle heating is a critical issue that can onlybe addressed in a long-pulse, burning-plasma ex-periment.

5.2.3. Di6ertor and edge physicsDivertors remain one of the most difficult

physics and engineering challenges. Radiativemodes which result in dispersing most of theplasma transport power on the first wall anddivertor plates can help keep the heat load on thedivertor plates to a manageable level. However,the increased edge density and particularly thehigh Zeff caused by impurity injection drasticallyincreases the current-drive power. Compatibilityof these techniques with MHD, bootstrap, andcurrent-drive in a steady-state discharge is a criti-cal item which require further analysis and exper-imentation.

5.2.4. Disruption a6oidanceOne of the most serious problems with the

tokamak is the possibility of a major plasmadisruption requiring plant shutdown. Besides thepotential for damage to the surrounding struc-tures and the possibility to initiate events leadingto chemical and radioactivity release, unplannedshutdowns are an unacceptable operating charac-teristic for a power plant. Aside from the impacton plant availability, it is likely that extendedoutages would follow an unplanned plasma shut-down to identify the source of the problem andensure that it could not happen again. In ARIES-RS, the maximum allowable b was set to 90% ofthe predicted b limit in order to avoid suchevents, however further studies are needed todetermine operating points with very low proba-bility of disruption (less than one disruption peryear) and/or reliable active measures of disruptionavoidance.


5.2.5. Controlled start-up, shutdown, and loadfollowing

Non-inductive start-up has been assumed forARIES-RS with little help from the PF systems.This approach minimizes (or eliminates) eddy-cur-rents in the cold structure of the TF coils andsimplifies the TF coil design considerably. Whileanalysis indicates that non-inductive start-up,plasma shutdown, and load following can beachieved without major difficulties, these tech-niques must be experimentally demonstrated.

5.2.6. De6elopment and qualification of V-alloysIn the technology area, the development and

qualification of low-activation materials remains akey issue. In addition to irradiation testing ofsmall samples, the development program shouldaddress many issues such as the response of powercore components as a material system (not indi-vidual samples), manufacturing, reliability, mate-rial lifetime, cost, etc. In the absence of a neutronsource, substantial programs in some of theseareas can still be undertaken in a non-radiationenvironment.

5.2.7. Component reliabilityOne of the principal reasons for choosing the

self-cooled blanket configuration is simplicity,which is hoped to lead to a more reliable system.Nevertheless, there are many uncertainties in thecomponent and materials responses, including butnot limited to the V-alloy and insulating coating,which make failure rate predictions impossible.Since the plant safety and availability are so de-pendent on component failures, and because thequantity of data required to establish the neededlevels of availability is high, it is imperative tobegin a program of engineering testing and tocapitalize on all planned and existing irradiationfacilities. Valuable data can be acquired by testingcomponents in non-radiation environments aswell.

5.2.8. Material lifetimeComponent lifetime is also important in deter-

mining the availability, replacement cost andwaste stream. Large uncertainties exist due to thelack of high-fluence data. In general, the sched-

uled replacement time should be no shorter than1–2 years, implying a fluence level of the order of150–200 dpa. Current estimates predict such lev-els are achievable, but there is insufficient data todemonstrate acceptable material properties. High-fluence irradiation testing is expensive and time-consuming research, which is difficult to sustainunder the current funding levels.

5.2.9. CoatingsThe choice of Li coolant is generally conceded

to require the development of radiation-resistantelectrically insulating coatings in order to over-come the problems associated with MHD pres-sure drop in a high-field environment. At present,no acceptable coating has been demonstrated un-der relevant service conditions, even in the ab-sence of neutron irradiation. Coatings are ageneric issue for fusion, having been proposed fortritium control in several blanket design concepts,and even for isolation of otherwise incompatiblematerials. Surface modification and system chem-istry required to maintain the coatings is a major,as yet poorly understood, area of fusion research.

5.2.10. High heat flux remo6al combined withhigh performance

The simultaneous requirements of high perfor-mance under extreme heat and particle loads,together with extended lifetime, maintainability,safety, and radioactive waste pose a difficult prob-lem for divertors and high heat flux components.The lack of predictive capabilities for the edgeplasma exacerbates the problem. From an engi-neering point of view, design solutions and R&Dprograms are difficult to implement when thegoals are unknown.

References

[1] F. Najmabadi, The ARIES Team, The Starlite project-as-sessment phase report, Report UCSD-ENG-005, Univer-sity of California, San Diego, CA, 1996.

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Documents

Overview of the ARIES-RS reversed-shear tokamak power plant study