Fusion Engineering and Design 56–57 (2001) 71–82

Materials research towards a fusion reactor

Karl Ehrlich *Forschungszentrum Karlsruhe, Institut fur Materialforschung I, Postfach 3640, 76021 Karlsruhe, Germany

Abstract

In this paper the expected goals for the development of structural materials for fusion reactors and importantnuclear-specific selection criteria are discussed. A short status of development is given and critical issues as well asnecessary investigations are identified for the three major materials groups, the ferritic–martensitic steels, ceramicSiC/SiC composites and vanadium alloys. Proposals for a long-term development strategy are developed, includingthe use of existing irradiation facilities and an intense 14 MeV neutron source. The International Fusion MaterialsIrradiation Facility (IFMIF), selected and recommended by the International Fusion Materials Community, is anappropriate and necessary test bed for the materials development. © 2001 Elsevier Science B.V. All rights reserved.

Keywords: Materials research; Fusion reactor; Ferritic–martensitic steels

www.elsevier.com/locate/fusengdes

1. Introduction

Energy generating systems based on nuclearfusion have the potential to provide a CO2 emis-sion-free, sustainable, safe and clean energy op-tion for the future. On the path of theirrealisation the choice of appropriate structuralmaterials and their behaviour under specificallyharmful loading conditions is a critical issue forthe technical feasibility and a safe and economicoperation. Favourable mechanical properties,good corrosion behaviour and compatibility, tech-nical maturity and the potential for low activationare the conditions for a pre-selection. Further-more, existing knowledge on material behaviourunder high fission neutron irradiation helps toexclude those materials which show relevantweakness under neutron irradiation.

The development of appropriate structural ma-terials follows at present two distinct lines: thefirst one is directed towards the construction ofthe next step machine, the international ther-monuclear experimental reactor, ITER. This facil-ity is—with regard to material issues—char-acterised by a strongly pulsed mode of operation,a very moderate neutron exposure and low opera-tional temperature. It is expected that these de-mands can be fulfilled by the use of an austeniticstainless steel of type 316 LN-IG [1], a material,which already had been successfully applied inconventional fission reactor technology.

The second developmental line aims for materi-als which can withstand high neutron fluence andtemperature ranges, typical for commercial fusionreactors, and which in addition have the potentialfor reduced or even low neutron-induced activa-tion. In the last two decades three major materialgroups have evolved which eventually can fulfil

* Tel.: +49-721-822-338; fax: +49-721-824-567.E-mail address: karl.ehrlich@imf.fzk.de (K. Ehrlich).

0920-3796/01/$ - see front matter © 2001 Elsevier Science B.V. All rights reserved.

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the requirements as reactor materials. These areferritic–martensitic steels, vanadium alloys andSiC/SiC ceramic composites. Recent assessmentshave shown that these alternatives have a differ-ent level of maturity and partially exhibit criticalissues which have to be investigated with priority[2–4].

In this paper the status of development isshortly summarised and necessary future stepsincluding the development of appropriate researchtools are discussed.

2. Performance goals for reactor materials

Structural materials will have to satisfy strin-gent demands for the construction of competitivefusion reactors: For instance they should ideallyguarantee a reliable and safe operation over aplant lifetime in the range of 30 full power years(FPys) with a high net plant efficiency in the orderof about 35–45%. Such ambitious targets are atpresent achieved with modern fission reactors(35%) and ‘clean’ coal fired commercial steampower stations (45%) and are the result of acontinuous development over several decades. Forinstance the optimisation of conventional 9–12%Cr steels for steam power plants changing fromunder-critical to supercritical steam parametersneeded a two decade lasting material developmentprogramme called COST 501 in Europe, and sim-ilar activities in USA and Japan [5].

It is clear that such performance goals can onlybe achieved at the end of a development whichhas recently been described in the Airaghi EU-as-sessment report 95–99 for Fusion [6]. It foreseesat least two more generations of machines beforea prototype reactor and finally commercial reac-tors can be built. This machine development hasto be accompanied by a strong material develop-ment and qualification programme. Hereby inaddition to the usual load conditions the exposureof first wall/structural materials to high energetic14 MeV neutrons in combination with low-energyplasma particles, high heat loads and electromag-netic radiation is the new, and most probably thekey material problem. For an assumed plant life-time of 30 years the following figures illustrate the

effect of irradiation on materials: a typically ex-pected average neutron wall load of 2–3 MW/m2

leads to an integrated exposure of 75 MWy/m2

after 30 years lifetime. This corresponds in Fe orsteel to about 750 displacements (dpa) for eachlattice atom, caused by elastic or inelastic interac-tions with neutrons. In addition about 7000 appmHe and 31200 appm H will be produced viainelastic (n,�)- and (n,p)-transmutation reactions.Displacement damage (dpa) and generation ofelements like helium, which are not soluble in theparent material, are known to be responsible formanifold radiation damage phenomena like mate-rial hardening/embrittlement and/or dimensionalinstability by void- and helium-driven swelling. Atypical example how irradiation-induced harden-ing/embrittlement can further limit the otherwiseallowable stress– temperature range by a shift ofthe fracture toughness curve towards a highertemperature level is presented for the ferritic–martensitic steel F82H in Fig. 1. Similarly, theupper temperature can be further reduced by cor-rosion effects as exemplified in a breeding blanketwith liquid lead– lithium as cooling/breedingmedium in comparison to a helium cooledcomponent.

The above mentioned final target of 750 dpa isextremely far beyond any existing experience withmaterials in nuclear environment. In most ad-vanced fission reactors like Fast Breedersaustenitic and ferritic–martensitic steels and Ni-alloys have been tested as cladding and wrappermaterials in high performance fuel elements up toabout 150 dpa. Hereby in austenitic steels exten-sive swelling and in Ni-alloys strong embrittle-ment effects have been detected, indicating adistinct performance limit. It is reasonable tosuspect that in the much harder neutron spectrumof a fusion reactor with at least one order ofmagnitude larger transmutation rates for He andH the material performance should be furtherreduced. The designed EU-test blanket modules(TBMs) for DEMO [7] limit the average neutronwall loading to about 70 dpa, which is well inaccordance with a target range of 80–100 dpa inother programmes. A reasonable intermediatetarget for material development for a prototype

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reactor lies in the range of max. 15 MWy/m2

which corresponds for most of the materialsgroups to about 150 dpa. Final goal of a materialdevelopment is to extend such performance limitsas much as possible in order to reduce the numberof replacements for plasma-near components likebreeding blankets to a minimum during the plantlifetime.

3. Materials requirements and nuclear-specificselection criteria

There are many requirements which have to befulfilled by structural materials. For the design ofFW/Breeding Blankets conventional material datalike thermophysical and mechanical propertiesand the compatibility with cooling media andbreeding/neutron multiplying materials areneeded. They determine the appropriate windowof application in terms of temperature and accept-able stress-levels for the envisaged lifetime. Such

design exercises, which are driven by the wish toget an optimum performance with regard toparameters like high tritium breeding, high netthermal efficiency, high power density etc. havebeen performed since more than 20 years. Someof them are compiled in Table 1 and they showthat only a limited number of combinations ofstructural with breeder materials and coolants arepromising. Mostly used structural materials arereduced-activation ferritic–martensitic steels (RA-FM-steels), vanadium alloys and ceramic com-posite materials of type SiC/SiC. More recentlyhigh strength tungsten alloys for very high heatwall loads have also been proposed. The alterna-tive designs operate at different temperature win-dows and offer a variety of heat load capacitiesand net efficiencies. But in view of partially exist-ing and identified material issues the real potentialof all these concepts has to be verified and demon-strated in the next phase of development.

As already pointed out, the response of materi-als to 14 MeV neutron irradiation is the impor-

Fig. 1. load-temperature window for the RA-FM steel F82H-mod in the design of the EU-HCPB- and WCLL test blanket modulesand possible further limits by corrosion and irradiation. Fracture toughness data from E.V. van Osch et al. NRG-Report20027/99.25387/P.

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Table 1Major breeding blanket concepts

BreedingCoolant Structural material Neutron Operation conditionsmaterial multiplier

Temperature Pressure (MPa)(°C)

He LiCeHe/LiCea/ F/M-steel Be 250–550 5–20 (8 MPa)FS/Beb

He LiCeHe/LiCe/SiC/ Ceramic composite 5–20Be 450–950SiC/SiCSiC/Be

Li Vanadium alloyLi/V LiLi 350–750 �1H2OH2O/Pb–Li/ Pb–Li F/M-steel Pb–Li 250–550 12–15 (15.5

FSc MPa)

a LiCe, lithium ceramic breeder materials: Li2O, Li4SiO4, Li2ZrO3 or Li2TiO3.b HCPB, helium-cooled pebble-bed blanket/EU; pressure data in brackets.c WCLL, water-cooled lithium-lead blanket/EU; pressure data in brackets.

tant criterion and has to be investigated withpriority. One possibility is to characterise thisresponse by appropriate primary damage parame-ters like the displacement rate and the relation oftransmutations versus dpa, i. e. He/dpa and H/dpa. Comparative calculations for a first wallposition in a DEMO with a wall loading of 2.5MW/m2 have been performed for SiC, V, Fe–9Crand W–5Re. The comparison in Table 2 showsthat the effective displacement rate is in a narrowrange of 7–9×10−7 dpa/s for SiC/SiC, vanadiumand steels and decreases to about one third for therefractory metal W. It has not yet been assessed,whether or not the lower displacement rate perunit fluence has a positive effect on the irradiationbehaviour of tungsten alloys. A larger variation ofdata exists for the production of helium which isresponsible for helium embrittlement and high-temperature, helium-driven swelling. Especiallyfor the ceramic composite SiC/SiC the dimen-sional stability of fibers and matrix could beadversely affected by the high helium generationrate. In general one has to say that the primarydamage parameters are not sufficient to predictthe materials behaviour under irradiation.

Much more advanced are the tools to analysethe irradiation-induced radioactivity and other ra-diological properties of materials, because formany elements sound experimental or theoreticalcross sections, activation and decay data (e.g.

EASY) and calculation codes (e.g. FISPACT) forthe interaction between neutrons and elements areavailable [8,9]. This is especially important, sincestructural materials for fusion are— in contrast tothe situation in fission reactors— the major sourceof radioactivity in a fusion reactor and play agreat role in the debate about a ‘clean’— fusionenergy. In Fig. 2a the decay heat of several mate-rials in a DEMO first wall position after 10MWy/m2 is plotted as function of the decay time.This figure is of importance for the case of animmediate shutdown of a reactor since a disper-sion of radioactivity by material volatilisationcould occur and could pose safety problems. AW–Re alloy presents the worst case, and it is stillopen whether or not much better thermophysicalproperties and strength data of W–Re can outbal-ance this weakness. In this respect SiC/SiC has byfar the greatest potential, not only because of the

Table 2Some primary damage parameters for comparison (DEMO-first wall position wall load 2.5 MW/m2)

To dpa/sMaterial appm He/dpa appm H/dpa×10−7

SiC/SiC 7.1 108 42.57.06 4.6V–4Cr–4Ti 19.56.6 9.8Fe–9Cr 37.81.8W–5Re 2.1 �10

K. Ehrlich / Fusion Engineering and Design 56–57 (2001) 71–82 75

Fig. 2. (a) The decay heat of structural materials in a DEMO-FW position after an exposure to 10 MW/m2 as function of the decaytime. (b) Surface-�-dose rate of structural materials after an exposure to 10 MW/m2 in a DEMO-FW position.

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lowest values in the very short term, but alsobecause of a faster decay for short and intermedi-ate decay times.

On the other hand, the level of long-term acti-vation defined by the �-dose rate after abouthundred years of intermediate storage will deter-mine the way of waste disposal and recycling. Allfour candidates have according to Fig. 2b after100 years decay time dose-rate levels in the rangeof low-level waste (�2 mSv/h) and can be recy-cled by remote handling methods. Especiallyvanadium alloys and low-activation steels have apotential to reach the so-called hands-on-level in�-dose rate (2.5 �Sv/h) within 100–1000 years (seeFig. 3 below). The decay heat is for all alloys inthe low-level region for such long decay times.

There are two possibilities to improve the radi-ological properties of materials: One is to substi-tute in complex multicomponent materials likesteels major alloying elements like Nb, Mo Ni.etc. by elements with an equal metallurgical func-tion, but lower radioactivity. This has been widelyexercised in the development of ferritic–marten-sitic steels with reduced activation [10] and also inthe proper selection of major elements for V-based alloys (e.g. V–Ti–Cr alloys). The secondpossibility is to limit or reduce the concentrationof unwanted tramp elements with high longtermradioactivity to a technically achievable mini-mum. How and at which decay times impurities

can affect the radioactivity in the three majormaterials groups is demonstrated in Fig. 3. Inthese plots the change in the �-dose rates arecalculated by introducing a set of specificallyharmful tramp elements in fixed wt–ppm andsub-wt–ppm-concentrations. The plots demon-strate, where relevant improvements in radiologi-cal properties can be achieved by producing‘super-clean’ alloys. The potential for an improve-ment is especially good for V-alloys in thelongterm activation regime and also attractive forSiC/SiC for an intermittent decay time betweenfew days up to 10 years.

4. Short status of material development, criticalissues and prioritary R&D-investigations

As mentioned above, advanced ferritic-marten-sitic steels, vanadium alloys and SiC-fibre rein-forced SiC composites have been selected asmajor materials groups. Their selection is mainlybased on favourable conventional properties and/or technical maturity, the potential for low activa-tion and /or promising results under fissionneutron irradiation. Opportunities for the consid-eration of alternatives could arise as a result ofmajor advances of other materials groups, e.g.through exploratory studies on chromium alloys,TiAl intermetallics or through aggressive design

Fig. 3. The influence of high-activating impurities on the �-dose rate of Fe, SiC and V after a wall loading of 12.5 MW/m2.

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concepts which need specific capabilities. The useof W–Re-refractory alloys in high power densitybreeding blanket concepts is an actual example. Abrief description of the status of development,identified (critical) issues and major R&D investi-gations for the three material groups follows:

4.1. Ferritic–martensitic steels

The ferritic–martensitic steels are furthest intheir development. There exists a well developedtechnology and a broad industrial experience withsuch alloys in fossil and nuclear energy technol-ogy. They show reasonably good thermophysicaland mechanical properties, a low sensitivity toradiation-induced swelling and helium embrittle-ment under fission neutron irradiation and goodcompatibility with major cooling and breedingmaterials. In recent years reduced activation ver-sions of these commercially deployed steels havebeen developed on a laboratory scale and testedwith equivalent or even improved properties[2,10]. In the meantime two industrial batcheshave been produced (F82H-mod. and Eurofer 97)in various semi finished products with reducedimpurity levels and good properties. A furtherstep towards superclean alloys to achieve the con-ditions for low activation (LA-FM) steels ispromising. A major issue of this material class isthe observed radiation-induced hardening/embrit-tlement and degradation of flow and fractureproperties below about 350 °C. The detailed un-derstanding of this material degradation in depen-dence of the microstructure and the relevantirradiation parameters like helium and displace-ments on the hardening/embrittlement and frac-ture and crack propagation mechanisms is ofutmost urgency and is one aim of internationalcollaboration under the auspices of the Interna-tional Energy Agency. A possible influence offerromagnetism on plasma stability and the ad-verse effect of magnetically– induced Lorenzforces on structural components is under investi-gation. To increase the attractiveness of breedingblanket components with FM-steels as structuralmaterial, the development of alloys strengthenedwith nano-scaled oxide dispersions (ODS-alloys)and precipitates is envisaged to improve the creep

Fig. 4. The influence of irradiation on the impact properties ofthe V–4Ti–4Cr alloy. Ref. [12].

rupture properties. It has the potential to expandthe upper operating temperature to 650 or even750 °C.

4.2. Vanadium alloys

Vanadium alloys based on V–Cr–Ti con-stituents have a favourable combination of ther-mophysical properties and high creep strengthand have in combination with lithium as self-cool-ing breeding material the potential for high heatloads and high operational temperature (750 °C).They exhibit the lowest longterm activation, pro-vided that the concentration of high-activatingimpurities can be reduced to a minimum in indus-trial heats (see Section 3). Major results of irradi-ation experiments in fission reactors regardingswelling and helium embrittlement are alsopromising [11]. Similar to the situation in ferriticsteels, the lower operation temperature is limitedby the propensity for brittle failure, induced byirradiation hardening, as has been suspected byformer investigations and has been recently confi-rmed by Rohrig et al. in Fig. 4 [12]. The forma-tion of dislocation loops and irradiation-inducedprecipitates causes this embrittlement. An even-tual saturation of this effect in the range of 10–20dpa can be concluded from recent experimentalobservations [13].

A major drawback and a possible feasibilityissue is the high solubility and permeability oftritium and the high solubility of the interstitial

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elements O, N and C in vanadium, which can leadto catastrophic embrittlement [3,14] Therefore,measures have to be undertaken to control andprevent their pick-up during fabrication, joiningprocesses and during exposure to the environ-ment. The development of self-healing and corro-sion-protective coatings, which at the same timeare electrically insulating and reduce the magnetohydrodynamic (MHD) effects in liquid-metalcooled breeding blankets to a minimum, has thehighest priority. A better understanding of theinfluence of these interstitial impurities on precipi-tation behaviour and mechanical properties, espe-cially their tendency for embrittlement is of greatimportance.

4.3. SiC/SiC composite materials

The development of this material group pre-sents the most difficult challenge of the threematerials groups [4,15]. They have potentiallyhigh payoffs in terms of very low radioactivityand decay heat at short and intermediate decaytimes and offer high operating temperatures inbreeding blanket components due to high-temper-ature strength (see Table 1). Primary feasibilityissues involved in their development are a princi-pal understanding of the interaction with neutronswith the complex fiber/interface/matrix structureunder the aggravating conditions of high crosssections for the production of insoluble heliumand hydrogen and their influence on properties.For instance a loss of thermal conductivity, reduc-tion of bending strength and enhanced swelling atthe upper operational temperature could limit theapplication. A key issue for a reasonable applica-tion of these materials lies in the development ofradiation resistant materials. Further investiga-tions are also necessary to explore their oxidationbehaviour and compatibility with different cool-ing/breeding media and other functional materi-als, e.g. beryllium. A serious drawback is the verylimited technology base for production, joiningand insufficient hermetic sealing capacity. Finallythe development of appropriate design rules forthe use of these innovative materials as structuralparts is necessary.

5. The need for an intense neutron source

The critical issue for all materials under investi-gation is their behaviour under fusion-specificconditions, i.e. under high-energetic neutrons witha peak intensity at 14 MeV. Since no appropriate14 MeV neutron source presently exists, materialperformance is mainly studied in irradiation facil-ities like fission reactors and ion accelerators,where fusion conditions can partially be simulatedby specific tricks. It is common understanding inthe Materials Community that such experimentsprovide very valuable information regarding spe-cific radiation damage effects and that they allowa pre-selection of promising material groups.However, a direct transfer of these results tofusion relevant conditions is not possible, sincethe deviations in relevant damage parameters areremarkable and their general influence on differ-ent radiation damage phenomena is not clear. Forinstance, the increased numbers for (n,p)- and(n,�)-reactions by typical one order of magni-tude—or even more for light elements like Si—cannot be translated by presently availabletheoretical models into quantitative predictionshow they would change irradiation phenomenalike embrittlement or hardening. Increasing inelas-tic events also contribute to the overall displace-ment damage by high-energy recoils and shift theenergy spectrum of primary knocked-on atoms(PKAs) to much higher values. This again raisessome concerns regarding the correct partitioningof defect clustering and single defects. A moredetailed discussion of the general problems arisingfrom high energetic fusion neutrons can be foundin [16]. Therefore, although good progress is beingmade in the area of modelling, the need to vali-date existing irradiation data from simulation ex-periments and to extrapolate such results to theconditions in a fusion reactor make a powerfultest bed for fusion materials studies indispensable[16,17].

There exist many proposals for an appropriatematerials test irradiation facility. They have beendiscussed very intensively in the last decade by aWorking Group under the auspices of the Inter-national Energy Agency. This group finally cameto the conclusion that an accelerator-based D–Li

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stripping source, denoted as the International Fu-sion Materials Irradiation Facility (IFMIF), is themost advanced and most suitable alternative [17].This facility, presented at all relevant materials andfusion technology conferences in the last years, canfulfil all essential users requirements; it meets therelevant radiation damage parameters, has a suffi-cient test volume in high-, medium- and low-fluxtest zones to perform necessary experiments forstructural-, breeding- and other materials and can,in a limited volume, achieve very high neutron wallloading conditions equivalent to 5 MW/m2. Incomparison to many other proposals, this facilityis based on proven technology with very moderatetechnical extrapolations and could hence be de-signed and constructed within 10 years after a finaldecision to build it [18,19].

6. Strategy for material development towards afusion reactor

Recently the elements of a common strategy formaterials development towards a fusion reactor

have been discussed by an IEA-Workshop onStrategy and Planning of Fusion Materials R&D.They have been summarised and further developedin a strategy paper presented at the 9th Interna-tional Conference on Fusion Reactor Materials inColorado Springs, CO [20]. In those countries(Europe, Japan and Russia), where a mission-ori-ented and hence time-driven strategy exists todevelop and achieve a magnetically confined fusionreactor, the material development has to go alongwith the general roadmap of development. In Fig.5 such a scenario has been projected by the AiraghiCommision, who recently assessed the progress ofthe Fourth EU Framework Programme in Fusion[6]. In this scenario the demonstration of a referencelow activation steel for DEMO target and thesearch for higher performance materials for aprototypic fusion reactor are explicitly requested asmajor achievements. This demand is very well inaccordance (also in the time scale) with the devel-opment path proposed by the Materials Commu-nity in [20].

For the ferritic–martensitic steels which are thefurthest in development and show the fewest areas

Fig. 5. Road map of achievements for the development of fusion reactors. Ref. [6].

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Fig. 6. The parallel development of structural and functional materials for the development of the EU-HCPB test blanket module.

of concern, a broad engineering data base needs tobe generated in existing test and irradiation facili-ties within the next decade to qualify a referenceRA-FM-alloy as first wall/structural material forDEMO test blanket modules (TBMs). This has togo in-line with a research and test programme forfunctional breeder and neutron multiplier materi-als and a technology and test programme forsub-module and mock up construction. Fig. 6shows, as an example for an integrated researchand test programme, the schedule for the develop-ment of the European HCPB-test blanket module.In the concept exploration phase the determina-tion of conventional properties for the envisagedloading conditions and the testing of irradiationbehaviour mainly in fission reactors up to the 70dpa fluence level is needed. A final confirmation ofthe reference material concepts needs, however, thetesting under real fusion neutron irradiation. Pre-requisite for this strategy is the timely availabilityof the high energetic and high intensity neutron

source IFMIF. In parallel, the testing of TBMs inthe next step machine, most probably FEAT-ITER, has to be performed in order to get infor-mation about the integral behaviour of suchcomplex components.

In principle also for other breeding blanketswith alternative structural materials a similar ap-proach has to be adopted. It is, however, unlikelythat all these options can be developed in parallel,not only because of limited manpower and finan-cial constraints all over the world, but also becausesome of the alternatives show feasibility issueswhich had been identified earlier in Section 4. Forsuch cases, therefore, a more selective developmentstrategy, in which potential key issues are resolvedwith priority, before a broad development pro-gramme is launched, is highly advised. For exam-ple the use of vanadium alloys in combination withliquid metal cooling/breeding media needs—asmentioned before— the development of self-heal-ing and stable, protective and insulating coatings.

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There are two options very often discussed aspromising strategies for material development andqualification. One is to confirm and validate mate-rial concepts for use in a fusion–specific neutronenvironment by component testing in the consecu-tive fusion machines like ITER, DEMO and PRO-TOTYPE reactors. The other one is to selectappropriate materials by a combined series ofstudies in existing fission reactors and subsequenttesting and validation in IFMIF in advance, fol-lowed by final component tests in ITER, DEMO,etc. where the integral and very complex loadingconditions—as mentioned above for the TBMdevelopment—can be studied. The first strategy isnot very flexible with regard to material develop-ment, does not take into account the longtermnature of material development and bears manyrisks for the licensing and safety of the fusionmachine development itself. Therefore, the materi-als community has followed since long the secondalternative. In Fig. 7 the inter link between themajor phases of material development, namely; (i)the material concept verification for test blanketmodules; (ii) the demonstration of the PCA-perfor-mance for DEMO operation conditions; and (iii)the search for high performance materials (refer-

ence and alternative materials for prototypic andcommercial reactors) with the general roadmap ofachievements, shown earlier in Fig. 5 is shown forthe case of a staged introduction of IFMIF. Withthis scenario a timely confirmation of the materialsconcepts and their qualification to the necessaryfluence targets can be achieved and the search forhigh performance materials can be intensified. Thisstrategy is further supported by the fact that—yearby year— the capacities for qualified irradiationexperiments in fission reactors decreases simply bydecommissioning of such facilities like fast breederreactors. Therefore, in conclusion, a concise andconsequent strategy for material developmentneeds an appropriate neutron facility like theIFMIF to support successfully the development offusion reactors.

7. Summary and conclusions

The selection of structural materials for futurefusion reactors is based on conventional data,nuclear specific primary damage parametersand experience in nuclear applications. It is alsodependent on the combination with breeding-,

Fig. 7. The inter link between material development, necessary irradiation facilities and fusion devices on the way towards a fusionreactor.

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coolant- and neutron multiplying materials inspecific components like breeding blankets.Expected performance goal for the DEMOBreeding Blanket development in the EU is 70 dpaand a realistic target for prototypic materials liesin the range of 150 dpa.

The present status of development of the majorcandidate materials is different. Whereas forferritic–martensitic steels an integrated researchand test programme is proposed to demonstratetheir feasibility in breeding blanket components,possible feasibility issues in vanadium alloys andin SiC-fiber reinforced SiC ceramics should beinvestigated with priority, followed by a generalqualification programme.

For the investigation of the material behaviourunder irradiation, which might be the criticalissue, the use of existing irradiation facilities ismandatory in the next decade. However, theinherent limits of such simulation experimentsnecessitate to confirm and validate the results inan appropriate 14 MeV neutron source.

The Materials Community favours anaccelerator-driven D–Li neutron source, denotedas IFMIF, which can fulfil the users requirements,is feasible with only moderate extrapolation ofpresent technology and could be realised in duetime by international collaboration. With such atest bed it seems possible to meet the essentialmilestones for a successful material developmentin accordance with the general roadmap for fusionreactor development.

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