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Fusion Engineering and Design 83 (2008) 893–902

Contents lists available at ScienceDirect

Fusion Engineering and Design

journa l homepage: www.e lsev ier .com/ locate / fusengdes

ivertor conceptual designs for a fusion power plant

rachai Norajitraa,∗, Said I. Abdel-Khalikb, Luciano M. Giancarli c, Thomas Ihli a,uenter Janeschitza, Siegfried Malangd, Igor V. Mazule, Pierre Sardain f

Forschungszentrum Karlsruhe, P.O. Box 3640, D 76021 Karlsruhe, GermanyG.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 0332-0405, USACEA Saclay, 91191 Gif-sur-Yvette, FranceConsultant, Fliederweg 3, D 76351 Linkenheim-Hochstetten, GermanyD.V. Efremov Institute, Scientific Technical Centre “Sintez”, 196641 St. Petersburg, RussiaEFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany

r t i c l e i n f o

rticle history:vailable online 16 July 2008

eywords:usion power plantPCS

a b s t r a c t

Developing a divertor concept for fusion power plants to be built after ITER is deemed to be an urgenttask to meet the EU Fast Track scenario. This task is particularly challenging because of the wide rangeof requirements to be met, namely, the high incident peak heat flux, the blanket design with whichthe divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incidentparticles from the plasma, radiation effects on the properties of structural materials, and efficient recovery

RIES-CSelium-cooled divertorEMJ concept

mpingement coolingigh heat flux

and conversion of the considerable fraction (∼15%) of the total fusion thermal power incident on thedivertor.

This paper provides an overview of the development of different conceptual divertor designs (water-cooled, liquid metal-cooled, and helium-cooled types); their advantages and disadvantages and expectedperformance are outlined and discussed. Emphasis is placed on summarizing the status and progress ofR&D associated with He-cooled divertor designs which have been proposed in most of conceptual plant

.

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models in Europe and USA

. Introduction

The development of a divertor concept for post-ITER fusionower plants (FPPs) is deemed to be an urgent task to meet theU Fast Track scenario [1,2], where electricity production by fusions to be achieved by 2030 and fusion power is to be commercial-zed by 2040. These goals can only be realized if several essentialhysics and technology issues are resolved in time for the designnd construction bases of a DEMO reactor to be established by 2025.

The divertor (Fig. 1) is one of the high heat flux (HHF) compo-ents of the fusion reactor. Nearly 15% of the total fusion thermalower has to be removed by the divertor which has to withstandhigh surface heat flux of up to 15 MW/m2. In addition, it serves

s a shield for the magnetic coils behind it. Developing a diver-or involves many uncertainties, including the physical boundaryonditions and properties of the candidate materials envisagedor the divertor concept. Many material properties are subject to

∗ Corresponding author at: Forschungszentrum Karlsruhe Gmbh, Institute foraterials Research III, P.O. Box 3640, 76021 Karlsruhe, Germany.

el.: +49 7247 82 3673; fax: +49 7247 82 7673.E-mail address: prachai.norajitra@imf.fzk.de (P. Norajitra).

bocfalpcdl

920-3796/$ – see front matter © 2008 Elsevier B.V. All rights reserved.oi:10.1016/j.fusengdes.2008.05.022

© 2008 Elsevier B.V. All rights reserved.

hysical limitations, which, in turn, limit the ranges of applicationf the materials. Therefore, many design requirements depend onuture achievements and can only be extrapolated from the presenttate of knowledge. The plasma-facing material must be carefullyelected to withstand the sputtering erosion caused by the inci-ent particles (energetic ions and neutral atoms) escaping from thelasma while taking into account radiation effects on the propertiesf structural materials. The choice of the divertor coolant dependso a large extent on the blanket design with which the divertor haso be integrated (see details below).

In the course of the EU power plant conceptual studies (PPCS)2], three near-term (A, B and AB) and two-advanced power plant

odels (C and D) were investigated. The near-term models areased on limited extrapolations both in physics and in technol-gy while more advanced ones use an advanced physics scenarioombined with advanced blanket concepts. The plant models dif-er in their plasma physics, fusion power, as well as blanketnd divertor technologies. Model A [3] utilizes a water-cooled

ead–lithium (WCLL) blanket and a water-cooled divertor with aeak heat flux (PHF) of 15 MW/m2. Model B [4] uses a He-coolederamics/beryllium pebble bed (HCPB) blanket and a He-cooledivertor concept (PHF 10 MW/m2). Model AB [5] uses a He-cooled

ithium-lead (HCLL) blanket and a He-cooled divertor concept (PHF

894 P. Norajitra et al. / Fusion Engineering and Design 83 (2008) 893–902

] (left)

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0 MW/m2). Model C [6] is based on a dual-coolant (DC) blanketlead–lithium self-cooled bulk and He-cooled structures) and a He-ooled divertor (PHF 10 MW/m2). Model D [7] employs a self-cooledead–lithium (SCLL) blanket and lead–lithium-cooled divertor (PHFMW/m2). This shows that helium-cooled divertor designs aresed in most of the EU plant models; it has also been proposedor the US ARIES-CS [8] reactor study.

An important aspect for the economic viability of a future FPPs the reduction of its cost of electricity compared to alternativenergy sources. Both the EU PPCS and the US ARIES studies havedentified the increase of the overall reactor efficiency (electricalower/fusion power) as a key issue to achieve that goal. For thiseason the engineering design and layout aim at efficient recov-ry and conversion of the divertor thermal power by maximizinghe coolant operating temperature while minimizing the pumpingower.

. The ITER divertor

The only existing example of an actual divertor design is the

TER divertor [9] (Fig. 2) which is a water-cooled type. It operates at.2 MPa inlet pressure and relatively low temperature (100–126 ◦Ct the outer vertical target (VT) and 127–141 ◦C at the inner VT),nd at a low neutron flux [10–12] (see Table 1). Each plasma-facingomponent (PFC) of the divertor comprises a number of elements

ig. 2. ITER divertor [9] (here: a central cassette) indicating the main components.

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f 20–44 mm toroidal width with water coolant flowing in chan-els in the poloidal direction. The reference design for the strikeoint region (lower part of the VT with 10–20 MW/m2 heat flux)ses carbon fibre composite (CFC) monoblock with an active metalast (AMC®) CFC/Cu joining, copper chrome zirconium (CuCrZr)eat sink, and a swirl tape insert in the coolant tube channel. TheMC® joint, which keys into the CFC, is obtained by casting pureu onto a laser-textured CFC with a Ti coating that aids wetting.he pure Cu is then joined to the Cu-alloy heat sink by brazingr HIP, with the additional option of electron beam (EB) weld-ng in the case of flat tile geometry. For the upper part of the VT5 MW/m2 heat flux) the selected reference design employs tung-ten tiles (10 mm × 10 mm × 10 mm) with a cast pure Cu interlayer,razed or HIP-ed onto a CuCrZr structural material (heat sink).

The use of the high thermal conductivity CuCrZr heat sinknables high performance of the divertor, on the one hand. Onhe other hand, its embrittlement at the high neutron flux, as wells a reduction of fracture toughness at a neutron damage dose of.3 dpa (available data), especially at elevated coolant temperaturesas reported in Refs. [13] and [14], respectively. This may well be

onsidered a drawback that causes certain doubts in the applicabil-ty of such a material in an operating FPP environment under higheutron flux when operating at high coolant temperatures.

. Early PPCS divertor studies (before 2002)

.1. Water-cooled divertor for PPCS-A

A water-cooled divertor (WCD) [15,16] has been selected for thelant model PPCS-A, which requires the lowest level of extrapola-ion for both physics and technology from ITER. An advantage ofhis divertor type is that the technology of water-cooling circuitss well established; experience from water-cooled fission reactorsmainly PWR’s) can be extrapolated to fusion reactor conditions.he WCD concept is strongly based on the ITER divertor referenceesign [9] taking advantage of the limited extrapolation requiredrom the technology developed and tested for ITER. The choice of

CD also fulfils the PPCS requirement of using the same coolanthroughout the reactor.

The initial WCD concept for PPCS-A assessed in 2001 [15]Fig. 3i) uses W-alloy monoblock (e.g. W–1% La2O3 (WL10), size20 mm radial × 18 mm toroidal), with a 3.5 mm thick sacrificial

ayer, a 2.5 mm deep lateral castellation for stress reduction, andn embedded CuCrZr water coolant tube (Ø 11 mm × 1 mm). The

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and Design 83 (2008) 893–902 895

ube material was selected because of its superior fracture tough-ess compared to other Cu alloys. Similar to the ITER divertor,edium-temperature water (inlet temperature 140 ◦C, inlet pres-

ure 4.2 MPa) is used and swirl tapes are placed within the tubeo enhance the maximum acceptable critical heat flux. Oxygen freeigh conductivity (OFHC) Cu is used as a compliant layer insertedetween the CuCrZr tube and the W-alloy monoblock. The thermo-echanic analyses for a water coolant temperature as in the ITER

ivertor show that this concept can withstand a maximum heatux of 15 MW/m2. All temperatures and stresses are within thellowable limits.

A more advanced WCD conceptual design [16] (Fig. 3ii) wasater introduced aiming at increasing the thermal efficiency byaising the water coolant outlet temperature to about 325 ◦C at5.5 MPa pressure. It is based on the use of a series of poloidallyriented EUROFER (the reduced activation steel developed in EU forusion application) coolant pipes (Ø 11 mm × 0.5 mm) which allowo increase the water temperature up to PWR conditions to allowood heat conversion efficiency. Each of the pipes is surroundedy brazed W-alloy monoblocks and fixed on a common EUROFERack plate. A sacrificial 5.5 mm thick W layer is assumed. A swirlape made of EUROFER is placed within the tube to promote tur-ulence. The temperature distribution was improved by includingcompliance layer of a soft-graphite material (“Papyex” 0.1 mm

hick) on each EUROFER tube and a thin layer of pyrolitic graphiteartly deposited on the front inner surface of the W monoblock,hich serves both as a heat flux repartitioning layer and a thermal

arrier thereby reducing the maximum heat flux and the corre-ponding temperature gradients. Its thickness varies gradually from.075 mm in the front region down to zero in the lateral sides.he analytical results confirmed that this concept can withstandn incident surface heat flux of 15 MW/m2. The use of heat fluxepartition and the thermal barrier made it possible to achieve aafety margin of about 1.28 on the critical heat flux while main-aining the EUROFER structure temperature below the admissibleimit of 550 ◦C. However, fabrication and irradiation issues for thisesign require future R&D to demonstrate the feasibility of such aonceptual proposal.

.2. Liquid metal-cooled divertor for PPCS-D

A forced-convection lead–lithium (Pb17Li)-cooled divertoresign [17] (Fig. 4) was chosen for the plant model PPCS-D. Theivertor has to handle a maximum peak heat flux of 5 MW/m2. Theivertor target plate consists of a number of poloidally orientedilicon carbide–silicon carbide (SiCf/SiC) composite square tubesFig. 4, right), with a 5.5 mm thick sacrificial layer of tungsten alloyrmor. A “T flow separator” is inserted in each tube which assumescomb form in the region nearest to the plasma, thereby creating

oroidal channels. The lead/lithium eutectic Pb17Li flows poloidallyn one-half of the tube (serving as an inlet header), then it is forcedo pass through the short toroidal channels to cool the high fluxegion through a very short path, and finally it is routed back to thether side of the poloidal tube which serves as an outlet header. Thehannel dimensions and the liquid metal velocity in the differentegions of the divertor are varied in order to adapt them to dif-erent heat loads. The poloidal tubes in the HHF region are 30 mmeep and 28 mm wide; the depth of the toroidal channels is 1.4 mm.ach divertor segment accommodates 22 poloidal tubes, each ofhich forms 32 toroidal channels. The velocity of the Pb17Li in the

oroidal direction ranges from 1.5 m/s in the front to 1 m/s in theear. The thickness of the poloidal SiCf/SiC tube varies from 1 mmn the region near the plasma (to lower the temperature gradientsnd stresses) up to 2 mm in the back and side walls (required toithstand the internal pressure).

896 P. Norajitra et al. / Fusion Engineering and Design 83 (2008) 893–902

F f the I[

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ig. 3. PPCA-A water-cooled divertor (q = 15 MW/m2 required) as an extrapolation o16] (right).

For the thermo-mechanical analyses of the PPCS-D divertor,t was assumed a surface heat flux of 5 MW/m2 with an inletb17Li temperature of 600 ◦C, the calculated maximum temper-tures at the channel outlet (W 1288 ◦C, SiCf/SiC 1016 ◦C) areithin acceptable engineering limits. These calculations were

ased on large extrapolations for the assumed material physicalroperties (e.g. SiC/SiC thermal conductivity of 20 W/mK), which,ogether with other open issues such as joining technology, neu-ron irradiation effect, and MHD, require significant and long term&D.

TTt

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ig. 4. PPCS-D liquid metal-cooled divertor [17] (q = 5 MW/m2 required). Left: cutout of a

TER design: (i) W/CuCrZr concept [15] (left), (ii) concept with RAFM steel heat sink

.3. Helium-cooled divertor for PPCS-AB, -B, and -C

Helium cooling offers several advantages including chemicalnd neutronic inertness and the ability to operate at higher temper-tures and lower pressures than those required for water cooling.

he drawback is its comparatively low heat exchange capability.his, however, can be enhanced in various ways, e.g. by promotingurbulence and/or by increasing the solid/fluid interface area.

A helium-cooled divertor (HCD) has been selected for the mod-ls PPCS-B, -C, and (after 2004) -AB which simplifies the balance

7.5◦ sector, right: cross-sections of the divertor target plate in the HHF region.

P. Norajitra et al. / Fusion Engineering and Design 83 (2008) 893–902 897

Table 2Divertor cooling design parameters

Heat flux (MW/m2) HTC (kW/m2K) Pressure (MPa) Tin (◦C) Tout (◦C) Reference

(a) The water-cooled divertor (WCD) and liquid metal-cooled divertor (LMCD)WCD (PPCS-A)

- CuCrZr type 15 ∼170a 4.2 140 ∼166 14- RAFM type 15 ∼200a 15.5 300 ∼325 15LMCD (PPCS-D) 5 b Hydro-static 600 ∼990 16

(b) Summary of data [20] of He-cooled divertor designsPorous medium 5.5 20 8 632 800 17Multi-channel 5 20 14 500 551 18Eccentric swirl 5 21 14 600 800 18Slot 5 14 14 600 800 18Modified slot 10 56c 10 640 712 20T-tube (ARIES-CS) 10 40c 10 600 680 26HETS 10 30 (55)c 10 600 669 22HEMP 10 35 (56)c 10 600 700 21HEMS 10 24 (43)c 10 634 713 23HEMJ 10 31 (57)c 10 630 700 25

a Indicative value.b Only conduction assumed.c Maximum local value.

F eferen( misin

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ig. 5. Some initial HCD designs: (a) porous medium concept [18] (q = 5 MW/m2), rq = 5); (c) modified slot principle [21] (q = 10): 1 reducing conduction paths, 2 maxi

f plant since the same coolant is used for all internal components,hereby allowing the power conversion systems to be well inte-rated. Additionally, for PPCS-B, it eliminates the risk of hydrogenormation from the water–beryllium reaction in the event of anccident. HCD investigations began in 1999 within the frameworkf the EU power plant availability study (PPA) and the following firsttage of PPCS in 2000. Several initial concepts [18,19] had been con-idered. Helium gas operating pressures of 10–14 MPa with an inletemperature of about 600 ◦C are typically assumed. The coolingarameters of these concepts [20] are summarized in Table 2b.

The unconventional design [18] (1999) (Fig. 5, left) uses a porousedium heat exchanger and can accommodate a peak heat flux of

–6 MW/m2. The porous medium provides a high surface area-to-olume ratio favourable for the heat transfer enhancement; it alsorovides an irregular coolant flow pattern favourable for turbulentixing. Typically, the effective heat transfer coefficient (HTC) is

early 20 kW/m2K. This design utilizes helium at 8 MPa with annlet temperature of ∼630 ◦C and an exit temperature of 800 ◦C,

hich is compatible with the operating temperature of the struc-ural material (TZM or W alloy). The helium is forced through a

lot at the top of the coolant inlet tube into a circular porous wickayer (void fraction ∼40%), flows sideward through the porous layerefore exiting through a bottom slot of the outlet tube.

The basic cooling principle behind the porous medium designas adapted to develop a simple slot concept [19] (Fig. 5, center)

btg

ce dimensions [mm]: r = 11, R = 14, wM = 36, t1 = t2 = 3; (b) simple slot concept [19]g HTC, e.g. by pin array, 3 maximising isolation.

hich relies on the heat transfer capability of the helium flowinghrough a narrow peripheral gap of 0.1–0.2 mm rather than throughporous medium. This approach simplifies the design and man-facturing of the coolant channel system by omitting the porousedium. A typical effective HTC of 14 kW/m2K could be achieved

sing the slot concept with 14 MPa and 600 ◦C helium at the inlet,hich results in a heat flux limit of 5 MW/m2. The same heat flux

evel could be reached by using multi-channel and eccentric swirloncepts in which the HTC is primarily enhanced by increasing theoolant velocity on the heated side of the coolant channel. The mod-fied slot concept (2001) [21] (Fig. 5, right) increased the heat fluximit to about 10 MW/m2 (Table 2). It uses either a narrow periph-ral gap of 0.1 mm thickness to increase the coolant velocity uponxiting the inlet channel or a pin array (with a larger peripheral gapf about 1 mm), through which the coolant passes before flowingnto the outlet channel. The maximum local HTC expected from thisesign is nearly 56 kW/m2K.

. Advanced PPCS He-cooled divertor studies (after 2002)

Since 2002, He-cooled modular divertor designs [22–25] haveeen pursued in the PPCS, since they offer the opportunity to reducehermal stresses. Two main conceptual designs have been investi-ated.

898 P. Norajitra et al. / Fusion Engineering and Design 83 (2008) 893–902

HEMJ

(

(5s

Fig. 6. PPCS modular HCD designs. Left: HETS concept [23]; right:

A) The HETS (high-efficiency thermal shield) concept (ENEA) [23](Fig. 6, left), initially developed for water, has been adapted foruse with He as coolant. It uses an axisymmetric cap geometryin which a single jet impinges on a curved heated surface uponexiting a Ø 7 mm nozzle. The coolant then flows down side-ways through a narrow channel of 1.8 mm height at the inletand 0.9 mm at the exit that is part of a connected manifold.It is capable of sustaining an incident heat flux of 10 MW/m2

when operating at 10 MPa, an inlet temperature of 600 ◦C (tomeet the ductile–brittle transition temperature (DBTT) limit ofthe proposed materials), and an outlet temperature of 800 ◦C.The average HTC for the HETS concept is predicted to be nearly30 kW/m2K. The HETS elements (single “dome” and “mush-room”) are arranged in modules of six elements arranged inparallel; a suitable number of modules can make a divertorplate. In order to keep the outlet temperature as high as possi-ble, the structure is made of W alloy, which requires joining bybrazing. The reference geometric element is hexagonal with awidth of 35 mm between flats (20 mm side).

B) The HEMJ (He-cooled modular divertor with jet cooling) andHEMS (with slot array) designs (FZK) [24–26] (Fig. 6, right)both employ small hexagonal tiles of tungsten (18 mm widthbetween flats) as a thermal shield and sacrificial layer (5 mmthickness). The tiles are brazed to a thimble (Ø 15 mm × 1 mm)

made of WL10, forming a cooling finger. Each finger is cooledwith He at 10 MPa and 600 ◦C/700 ◦C (inlet/outlet tempera-tures) supplied via a manifold system. The main differencebetween the two designs deals with the heat transfer mech-anism. In the HEMS design, a tungsten slot array is used to

filnt

Fig. 7. He-cooled T-tube divertor concept

[24] and HEMS [25,26] designs. All designs handle q = 10 MW/m2.

enhance heat transfer at the bottom of the thimble by brazingit to the cooling surface, thereby increasing the heat transfercapacity. This design evolved from the first design study with apin array (HEMP) [22]. The HEMJ is based on direct jet-to-wallimpingement cooling [24] with multiple helium jets. Theseare generated by a steel cartridge carrying an array of smalljet holes, which is placed concentrically inside the thimble.The cooling finger is fixed to the front plate of the support-ing structure made of oxide dispersion-strengthened (ODS)steel (e.g. ODS EUROFER or an advanced ferrite version of it).To compensate for the large mismatch in the thermal expan-sion coefficients of W and steel, a transition piece has beendesigned which is based on Cu casting (optionally, Co brazing)with a conical interlock. The helium inlet and outlet tempera-tures are determined by the DBTT of irradiated WL10 and thecreep rupture strength of the ODS steel structure, respectively.Both HEMS and HEMJ with a predicted average HTC of 21 and31 kW/m2K, respectively, are capable of withstanding an inci-dent heat flux of 10 MW/m2.

. Advanced He-cooled divertor concept in the ARIES-CStudy

The T-tube slot jet divertor concept [27] has been proposed

or the ARIES-CS study [8]. It is based on the impingement cool-ng by a helium jet issuing from a narrow slit (Fig. 7). Maximumocal HTCs in excess of 40 kW/m2K are predicted near the stag-ation point for operation with helium at 10 MPa and an inletemperature of 600 ◦C. The T-tube construction allows a certain

[27] for ARIES-CS (q = 10 MW/m2).

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P. Norajitra et al. / Fusion Engine

ending of the tube ends and contributes in this way to a reduc-ion of the thermal stresses. This design is capable of withstandingn incident heat flux of 10 MW/m2. For each T-tube module, theelium enters a concentric cartridge mid-way along its length.he helium exits the inner tube through a 0.5 mm wide slot alongts entire length with a velocity of nearly 200 m/s. It impingesn the inner surface of the outer tube; the stagnation pointow generated by the impingement of the rectangular jet onhe outer tube surface is used to cool the divertor. Downstream

f the stagnation region, the helium forms a turbulent wall jetlong the inside surface of the outer tube, and is then removedhrough two symmetrically located exit ports near the center of the

odule.

mIsp

ig. 8. Manufacturing studies (A) and successfully HHF testing (B) of 1-finger HEMJ andonfirmed design performance of both designs.

and Design 83 (2008) 893–902 899

. Status of He-cooled divertor development since 2006

.1. EU PPCS R&D work

Design verification and proof-of-principle experiments arendispensable elements in the development of a reliable divertoroncept. For this purpose, a combined facility (Fig. 8) consisting ofhe TSEFEY EB facility (60 kW at 27 keV beam energy) and a new

oveable helium loop (10 MPa, 600 ◦C) has been built at the Efre-

ov Institute, St. Petersburg, Russia under cooperation with FZK.

nvestigations have been performed to evaluate prototypical tung-ten mock-up manufacturing [28,29], i.e. fabrication of divertorarts, joining techniques, and material selection [30] for both the

HEMS mock-ups in a combined EB and He loop facility at Efremov; results have

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00 P. Norajitra et al. / Fusion Engine

EMS and HEMJ finger designs. In parallel with these studies, twolternative methods, viz. electro chemical machining (ECM) andowder injection molding (PIM), for structuring the W componentsave been investigated at FZK with the aim of mass production ofhe divertor parts; results of this effort have been reported in Refs.31] and [32], respectively.

Prior to the HHF tests, design-screening tests were performed ingas puffing facility [33] based on reversed heat flux method. Com-utational fluid dynamic (CFD) analyses [34] for the HEMJ designave shown that the jet diameter D has a substantially larger impactn the divertor thermal performance and pressure loss than theet-to-wall distance H (within the design range of 0.6–1.2 mm). Asresult, a reference HEMJ geometry option with 24 holes of D = Ø.6 mm and H = 0.9 mm was chosen for subsequent HHF testing. Aelatively large D was chosen at the expense of a reduced cool-ng capacity to avoid plugging problems that may be caused byontaminated helium gas.

Mock-ups of “single finger units” of both HEMJ and HEMSave been manufactured (Fig. 8) and tested within a heat fluxange of 5–13 MW/m2. The temperature cyclic loading was sim-lated by means of switching the beam on and off (e.g. 60 s/60 s,0 s/30 s). The helium cooling parameters are 10 MPa inlet pressure,500–600 ◦C inlet temperature and the mass flow rate varying in a

ange of ∼5–15 g/s. The mock-ups were then subjected to destruc-ive post-examinations, which revealed damage due to micro cracksresumably initiated during the fabrication processes. Neither brit-le failure nor re-crystallisation of the thimble was observed in any

ock-up. These studies have confirmed the ability of the HEMJ andEMS He-cooled divertor concepts to accommodate heat loads ofp to 10 MW/m2.

Efforts have also been made to verify the results of CFD codes; inooperation with Georgia Tech (GT), Atlanta, USA, an instrumentedEMJ mock-up has been designed for testing in the air loop at GTnd the helium loop HEBLO at FZK. Experimental data with air atrototypical Reynolds numbers [35] agree well with the CFD cal-ulations made using the Fluent code. Plastic FEM analyses of annnovative transition piece design for the W and ODS steel joint [36]

ere performed using the ANSYS code. The calculation results showhat a brazed butt joint can withstand 1000 thermocyclic loadingsetween 20 and 600 ◦C. A preliminary assessment of a He-cooledest divertor module (TDM) [37], which should be tested in ITER,as been performed.

.2. ARIES-CS R&D work

The high HTC predicted for the T-tube design (see above) haseen experimentally validated with an air coolant correspondingo the non-dimensional parameters anticipated for its heliumperating conditions [38]. The fabrication of complex tungstenomponents using electrochemical processing technique (EL-Form)as developed [39]. Using these techniques, impingement-cooled

nd straight bore tungsten heat sinks were fabricated and tested.reliminary high heat flux testing demonstrated the ability of aungsten heat sink with helium impingement cooling to reduce itsverage surface temperature by ∼20% as compared to the averageurface temperature for a straight bore tungsten heat sink underhe same heat flux.

. Conclusions and outlook

In the course of the EU PPCS, different divertor types (WCD, HCD,nd LiPb-cooled divertor) were investigated. The choice of divertorype is primarily governed by the desire to use the same coolantype as the blanket. Additionally, operation with a high coolant exit

auatf

and Design 83 (2008) 893–902

emperature is particularly important for a power plant in ordero achieve a high thermal efficiency in the power conversion sys-em. In the first PPCS stage between 1999 and 2001 basic conceptsf the above-mentioned divertor types were studied. Their designrinciples, advantages and disadvantages, and analytical results areutlined in Section 3.

The water-cooled ITER divertor, which is based on the use ofuCrZr heat sink, is the only conceptual design that has been fully

nvestigated. Although it has been designed to withstand a HHF ofp to 20 MW/m2, it was recognized that the CuCrZr heat sink isather suitable for use in a “mild” reactor environment in termsf low neutron flux and low operating temperature and wouldmbrittle at the much higher neutron flux expected in the FPP (seeable 1) and elevated coolant temperatures. Therefore, this divertorype is not recommended for a power plant application with moreemanding requirements.

Helium-cooled divertor designs have been favoured by mostower plant models because of the chemical and neutronic inert-ess of helium, while allowing operation at considerably higheremperatures and lower pressures than water-cooled divertors.uring the early development stages (1999–2001), the theoreti-al performance limit of the HCD plate designs was successivelyncreased from 5 to 10 MW/m2 using various cooling techniques.he thermal stresses encountered in the continuous plate designhich are resulted from suppressed bending of the plate struc-

ure by a strong mechanical support were reduced by utilizingmodular design. Since then, helium-cooled modular divertor

esign issues have been systematically investigated within the EUPCS. The current reference design (HEMJ) is based on impinge-ent helium jet cooling which is characterised by its high thermal

erformance combined with simple construction and easy fab-ication. Detailed design and fabrication studies as well as HHFxperiments have been made in a combined testing facility (TSE-EY EB device and moveable He loop) at Efremov for verificationf the design and proof of principle. The latest experimentalesults already confirm the divertor’s ability to accommodate

heat load of 13 MW/m2, well above the design target of0 MW/m2.

Based on the same jet impingement cooling principle, a sim-lar HCD, the so-called T-tube divertor, has been proposed in theS ARIES-CS study. For this design, experimental verification of theFD calculations was successfully performed; fabrication of com-lex tungsten components using EL-Form processing techniquesas developed; development of a more prototypical helium-cooled

ungsten heat sink will be a major future task.There are ideas to combine the “old” plate design [21] with the

et cooling employed in the modular “finger” concept [26] and tose transition pieces made of Ta-alloy for connecting the W platend ODS steel manifolds at both ends of the plate. Such a modifiedlate concept, with emphasis on fabricability, will be evaluated inhe present ARIES study.

Intermediate-term PPCS R&D issues include: development ofsuitable divertor structural material with an operating window

n the range 600–1300 ◦C, completion of a divertor test moduleTDM) to be proposed in the ITER (Fig. 9) which is to be pre-testedn a bigger helium loop, e.g. HELOKA, and irradiation experimentsf structural materials in typical neutron environments of fissionnd of the presently designed intense fusion neutron source IFMIFInternational Fusion Materials Irradiation Facility) with DEMO-elevant neutron fluence. A damage design limit of 70 dpa is

ssumed for EUROFER in the DEMO design, while an improvementp to 140 dpa is anticipated [10] for commercial FPP. Taking intoccount the average neutron wall load of FPP of ∼2.4 MW/m2 forhe FW and ∼1.7 MW/m2 for the divertor target, a neutron fluenceor a two-year service lifetime of the divertor foreseen corresponds

P. Norajitra et al. / Fusion Engineering and Design 83 (2008) 893–902 901

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Fig. 9. He-cooled divertor R&D

o about 3–4 MWa/m2 which is equivalent to about 30–40 dpa inteel or about one-third of it in tungsten.

cknowledgements

This work, supported by the European Communities under theontract of association between EURATOM, Forschungszentrumarlsruhe and CEA, was carried out within the framework of theuropean Fusion Development Agreement. The views and opinionsxpressed herein do not necessarily reflect those of the Europeanommission.

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