current status of beryllium materials for fusion blanket applications

CURRENT STATUS OF BERYLLIUM MATERIALS FOR FUSION BLANKET APPLICATIONS PAVEL VLADIMIROV,a* DMITRY BACHURIN,a VLADIMIR BORODIN,b VLADIMIR CHAKIN,a MARIA GANCHENKOVA,c ALEXANDER FEDOROV,d MICHAEL KLIMENKOV,a IGOR KUPRIYANOV,e ANTON MOESLANG,a MASARU NAKAMICHI,f TAMAKI SHIBAYAMA,g SANDER VAN TIL,d and MILAN ZMITKOh a

Karlsruhe Institute of Technology, Karlsruhe, Germany National Research Centre ‘‘Kurchatov Institute,’’ Moscow, Russia c National Research Nuclear University ‘‘MEPhI,’’ Moscow, Russia d Nuclear Research and Consultancy Group, Petten, The Netherlands e A.A. Bochvar Research Institute of Inorganic Materials, Moscow, Russia f Japan Atomic Energy Agency, Rokkasho, Japan g Hokkaido University, Sapporo, Japan h Fusion for Energy, Barcelona, Spain b

Received November 28, 2013 Accepted for Publication April 3, 2014 http://dx.doi.org/10.13182/FST13-776

Beryllium is a promising functional material for several breeder system concepts to be tested within the experimental fusion reactor ITER and, later, implemented in the first commercial demonstration fusion power plant DEMO. For these applications its resistance to neutron irradiation and the detrimental effects of radiogenic gases (helium and tritium) is crucial for fusion reactor safety, subsequent waste management and material recycling. A reliable prediction of beryllium behavior under fusion irradiation conditions requires both dedicated experiments and advanced modeling. Characterization of the reference and alternative beryllium pebble grades was performed in terms of their microstructure and tritium release properties. The results are discussed with respect to their application in fusion blanket systems. The outcomes from the HIDOBE-01 post irradiation experiment (PIE) are discussed to highlight several interesting features manifested by beryllium irradiation at fusion relevant temperatures.

Titanium beryllide is presently developed as a possible substitute for beryllium pebbles as it shows better oxidation resistance, higher melting temperature and tritium release efficiency. Pebbles consisting predominantly of Be12Ti phase were successfully fabricated at Rokkasho, Japan. Recent advances in modeling provide new insights on the production of point defects and the behavior of helium and hydrogen impurities in beryllium, improving understanding of the mechanisms of primary damage production, hydrogen’s effect on the size and the shape of gas bubbles, and tritium removal from the pebbles. The relevance of the experimental and modeling results on irradiated beryllium for the design of a fusion demonstration reactor is evaluated, and recommendations for future R&D programs are proposed. KEYWORDS: beryllium, breeding blanket, irradiation Note: Some figures in this paper may be in color only in the electronic version.

*E-mail: [email protected] 28

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STATUS OF BERYLLIUM MATERIALS FOR FUSION BLANKET APPLICATIONS

I. INTRODUCTION

II. OPERATING CONDITIONS AND MATERIALS REQUIREMENTS

One of the important goals in the design of future fusion reactors is the development of a closed fuel cycle, so that a fusion reactor would be able to produce sufficient amount of fuel for itself. For this purpose, at least one tritium atom should be produced per neutron generated in fusion reaction. Since some neutrons are lost before creating tritium, an efficient neutron multiplier is needed to increase neutron flux, thus maintaining the number of tritium atoms produced per neutron (tritium breeding ratio or TBR) slightly above one. Metallic beryllium is especially suitable for this purpose as it produces two neutrons instead of the incident one with high probability. The additional neutrons can be used for tritium production, thus closing the fuel cycle. In the European strategy the first commercial demonstration fusion power plant, DEMO, is suggested to follow on from the ITER project to bring fusion energy to the commercial market. The ITER project aims at building a fusion device demonstrating the scientific and technical feasibility of fusion power. Testing of Tritium Breeder Blanket concepts is one of the ITER missions being an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Six breeding blanket concepts can be simultaneously tested in three dedicated equatorial ports of ITER (Ref. 1). Some of these concepts are using beryllium as neutron multiplier material. Europe has developed two reference tritium Breeder Blankets concepts that will be tested in the ITER Test Blanket Modules (TBMs): i) the Helium-Cooled LithiumLead (HCLL) which uses the liquid Pb-16at%Li as both breeder and neutron multiplier, and ii) the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Both concepts are using the EUROFER-97 reduced activation ferritic-martensitic (RAFM) steel as structural material and pressurized helium technology for heat extraction. The HCPB design is based on the concept of a pebble bed with interchanging layers of lithium ceramic and beryllium pebbles enclosed into breeder units, and separated by helium-cooled plates. Lithium ceramics pebbles are used for tritium production, while beryllium pebbles serve primary as neutron multipliers. Tritium generated by neutrons in lithium and beryllium pebbles is collected by low pressure (0.1 MPa) purge gas consisting of helium with small amount of hydrogen (0.1 vol.%), which is added to facilitate tritium pick up through the isotope exchange reaction. Pebbles are used to shorten diffusion pathways of tritium between material bulk and surface, while still having sufficient pebble bed density for keeping TBR slightly above 1.1 and allowing efficient cooling of pebble beds. Other TBM concepts utilizing beryllium pebbles as neutron multiplier are proposed by Japan,2 China3 and Korea.4 FUSION SCIENCE AND TECHNOLOGY

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The development of beryllium pebble beds and, to a limited extent, their testing in the ITER TBMs should demonstrate that the pebbles are able to ensure safe and reliable operation fulfilling the requirements formulated in Ref. 5. It is foreseen that normal operating temperature of beryllium pebble bed should be 300–650uC. Higher operating temperature would be beneficial to facilitate tritium release from beryllium. However, the upper temperature is limited by irradiation-induced swelling and breakaway oxidation of beryllium. Another important requirement limiting its operating temperature is compatibility between beryllium and surrounding structural material. Unfortunately, this interaction is not yet studied in full detail, especially under irradiation. Beryllium should withstand severe operating conditions for a long time without significant changes of thermo-physical and mechanical properties thus reducing the frequency of TBM replacement. A prospective pebble material should possess low tritium retention and low activation under neutron irradiation for easier and less expensive waste recycling and management. These properties are of superior significance in the view of large amounts of Be-pebbles required for the DEMO reactor: about 1 ton of beryllium pebbles is necessary for ITER TBM, while more than 300 tons will be necessary for DEMO. Therefore an efficient and affordable fabrication method for the production of beryllium pebbles on industrial scale needs to be developed. Hydrogen and passive heat release during accidental conditions should be kept below allowable safety limits during Be/steam/water interaction. These issues are crucial for the water-cooled concept of blanket (WCCB) pursued in Japan.2 The proposed Be and Be-alloy pebbles should present good compatibility with structural material and purge-gas flow. The use of helium as coolant alleviates the safety concerns associated with the high chemical reactivity of beryllium with water vapor and/or air and the possibility of hydrogen production. Actually, the use of beryllides has been proposed in order to mitigate these interactions. Tritium inventory in In-Vessel components should be minimized in order to avoid significant burst release of tritium during temperature excursions. Indicative values of tritium accumulation in beryllium (about 0.5 kg for DEMO) assume that the major part of generated tritium should be released during operation. This assumption, however, needs to be confirmed. As far as much more helium than tritium is generated in beryllium, the latter is trapped mainly by vacancy-helium clusters5 and within helium filled bubbles.6 The lack of numerical and experimental data stipulates large uncertainties in the design calculation of the end-of-life tritium

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inventory in beryllium under DEMO/FPP conditions. Possible design solutions to reduce this inventory (e.g., by increasing the operating temperatures of the beds or baking of the blanket in the reactor at about 650uC) cannot be supported by calculations as far as models accurately accounting for tritium trapping and release still have to be developed. Progress has been made recently in better understanding the physics of the phenomena,7–11 but a reliable code to support the design choices does not exist yet. Irradiation campaigns to obtain data on beryllium at 3000 and 6000 appm of helium within temperature range of 425–750uC have been successfully performed at NRG, Petten. The results of the HIDOBE-01 post irradiation examination will be shortly discussed later. The data obtained would allow a progress in the modeling and, complementarily, an empirical extrapolation to the end-oflife DEMO conditions (18 000 appm) could be attempted. Alternative Be-grades (small grain Be or the Be-Ti intermetallic compound) have been proposed for the use in fusion blanket. Titanium beryllide promises better tritium release, reduced swelling, lower chemical reactivity and better mechanical properties at high temperatures. Long lifetime of the components increases the availability of the fusion reactor by reducing frequency for replacement of its components. Presently, the lifetime of blanket is limited by the maximum irradiation dose acceptable for its major structural material EUROFER-97 (80–150 dpa). It is therefore highly desirable that functional materials do not reduce this limit significantly. With this respect the results of beryllium pebble irradiation campaigns HIDOBE-01 and -02 are very important. However, an irradiation experiment up to the full lifetime conditions is highly desirable.

III. RECENT PROGRESS IN THE DEVELOPMENT OF Be-BASED FUSION BLANKET MATERIALS III.A. Characterization of Alternative and Reference Be Pebbles

Industrially fabricated 1 mm Be pebbles produced using Rotating Electrode Process (REP) by NGK Ltd., Japan are presently the reference neutron multiplier material for the HCPB TBM (Ref. 12). Being produced by rapid cooling of melted droplets, the NGK pebbles possess a peculiar microstructure. Optical microscopy of their cross section shows several grains separated by large-angle grain boundaries and often a shrinkage hole near the centre of a pebble. Often small cracks along grain boundaries, sometimes connected with the shrinkage hole, can be found. As far as scalability of this method is limited, alternative production routes are being explored to reduce fabrication costs and increase the yield of pebble production. 30

One of the tested methods consists in crushing of Be-billets with subsequent rounding in a ball mill. Such pebbles have non-spherical shape and their microstructure resembles mainly that of the source beryllium billet. Three batches with different grain sizes were produced by Bochvar Institute, Russia13 and characterized with respect to their chemical composition, density of individual pebbles and packing density of pebble beds, mechanical as well as microstructural and tritium retention properties at KIT, Germany.14 Another possible method consists in spreading melted material through a nozzle to produce droplets which solidify to spherical beryllium particles. Successful pilot fabrication of beryllium pebbles was reported by Materion Corp., USA at the BeWS-11, Barcelona (2013). These pebbles are not yet available for characterization. The pebbles produced by Bochvar Institute, the commercially available impact ground (IG) vacuum cast grade (Materion Corp.), and the reference NGK REP 1 mm pebbles were tested with respect to their tritium retention.14 The pebbles were soaked in the gas mixture (H2z500 appm T2) at 1123 K (850uC) for 6 hours under the pressure of 4 bar. After that tritium thermal desorption experiments were performed using constant temperature ramping of 5 or 7 K/min and holding at the maximum temperature of 1373 K (1100uC) for 2.5 hours with consequent cooling down to room temperature (see Fig. 1). All Bochvar pebble grades were tested at 5 K/min, whereas NGK and IG pebbles at 7 K/min. At least three major release peaks can be identified on the release curves. The peak (marked in Fig. 1 as I) at about 450 K was found for all investigated materials. Its position on the temperature scale is essentially independent from the beryllium grade. The following broad peak (II) is observed in a wide temperature range from 550 K up to about 1050 K. For all samples, except the impact ground (IG) material and Be-pebble grade with the largest grain size (w 100 mm), this peak provides the main contribution to the total tritium release. The high temperature peak (III) is observed for Be pebble grades with grains of 10–30 mm and 30–60 mm as a shoulder of the main peak (II), but, in the cases of IG material and NGK pebbles, it is significantly delayed and shows up after achieving the maximum temperature (indicated by shadow in Fig. 1). In theory, no gas release peaks should be observed after the end of temperature ramping. Therefore, the peaks in the shadowed area are due to delay: either due to long diffusion time of tritium released from the bulk traps or due to the long travelling time of tritium in the purge gas system (e.g. low pumping rate of purge gas or long piping between oven and proportional counter). The observed delay of about 35 minutes cannot be explained by the second reason. We tentatively attribute this delay to the slow diffusion of tritium through the surface beryllium oxide, so that the third peak is not independent but is the repetition of the second one.


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Fig. 1. The rate of tritium release as a function of temperature (bottom left), continued with time dependent release after reaching the maximum test temperature of 1373 K (bottom right), while the temperature ramping is shown in the top panel.

If oxide layer would cover the surface of the pebble completely only the third peak will be observed. After formation of thick oxide layer, the stresses induced by a mismatch in the lattice parameters of beryllium and beryllium oxide results in formation of cracks and partial peeling of the oxide layer. In this case tritium reaching clean beryllium surface forms the second peak, while tritium diffusing through the beryllium oxide layer is delayed and forms the third peak. More experimental details can be found in Ref. 14. Release characteristics of various grades are usually assessed by comparing positions of the major release peaks (II) on the temperature scale. As can be seen from Fig. 1, the position of the major peak (II) is shifting to the right with increasing the average grain size of material. Bochvar pebbles and the impact ground material (with grains larger than 100 mm) follow this tendency, while the reference NGK pebbles not. After tritium release from the bulk traps it should reach grain boundaries which are accelerated tritium diffusion paths and, finally, desorb from the material surface. During this process the temperature ramping continues. The longer is the way to the surface, which is proportional to the grain size if one neglecting short diffusion time along grain boundaries, the higher will be temperature corresponding to the desorption peak. Therefore the position of the peaks is shifting to the higher temperature with increasing grain size. Presently it is not quite clear why this tendency does not hold by the NGK pebbles. It is commonly accepted that materials with smaller grain size should be preferable for reduction of tritium FUSION SCIENCE AND TECHNOLOGY

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inventory. Although the positions of the release peaks are changing on the temperature scale, the binding energies of the bulk traps responsible for tritium inventory remain the same. In fact, the observed time delays are irrelevant comparing to the lifetime of beryllium pebble bed. With this respect, tritium inventory in beryllium after a longterm irradiation at elevated temperature should be essentially independent from the grain size. Therefore, other factors as, e.g., structural integrity and resistance to neutron irradiation should be more important for the selection of beryllium multiplier material. III.B. Main Outcomes from the HIDOBE-01 PIE

High Dose Beryllium fission reactor irradiation program, consisting of two steps: HIDOBE-01 and 02, is the first irradiation of modern beryllium grades at blanket relevant temperatures (425–750uC) where irradiation dose (18 and 40 dpa) and helium production (3000 and 6000 appm He) are comparable with that for DEMO (Ref. 15). Several batches of pebbles with diameter of 0.5, 1 and 2 mm produced by REP (NGK Ltd., Japan) were irradiated. Some drums were completely filled with pebbles and slightly compressed (‘‘constrained’’ pebbles), thus imitating pebble bed irradiation.16,17 Other pebbles were loosely packed, so they will be called ‘‘unconstrained’’ or ‘‘free’’ pebbles. It is important to note that the irradiation temperatures given below are target temperatures defined during the design of irradiation rigs.15 The real temperature history

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monitored by several thermo-couples is rather complicated: on average the measured temperature is by about 20–50uC lower than the target one (see Table 1 in Ref. 18). Pebble size distribution was measured before and after irradiation for each grade and irradiation temperature.16 Figure 2 represents the relative changes of the average pebble diameter after irradiation at various temperatures, while the error bars correspond to the standard deviation calculated for the measured population. In most of the cases the change of pebble diameter and, hence, swelling of the constrained pebbles is higher than that of free pebbles. This conclusion should be, however, taken with caution as only small subsets from the constrained pebbles were investigated (due to the impregnation of the pebble beds and high activity of the larger number of pebbles), while all unconstrained pebbles were measured before and after irradiation. Unconstrained pebbles irradiated at temperatures below 723 K (v 500uC) (Fig. 3, top row) show largeangle grain boundaries characteristic for the microstructure of unirradiated pebbles and no visible damage. On the contrary, low-angle subgrain boundaries formed within large grains can be clearly seen in the constrained pebbles irradiated at the same temperature (Fig. 3, bottom row). The central grain shows chains of small bubbles, probably, along subgrain boundaries or dislocations, whereas no visible damage is observed in the other grains. Above 723 K (500uC) the constrained pebbles show developed porosity in the contact zones between two pebbles. Surprisingly, often only one of two pebbles in contact is damaged, while no visible voids are observed in the other. At higher irradiation temperatures the

unconstrained pebbles show formation of large faceted voids along previous large-angle grain boundaries and no visible damage inside grains. At the same temperature voids in the constrained pebbles are more homogeneously distributed over the grain volume and have much smaller size. Chains of bubbles with the size above average can be seen along large-angle grain boundaries and newly formed subgrain boundaries. Substantial, often asymmetric, thickening of some previous grain boundaries is observed as formation of denuded zones free of bubbles. Excess of material transferred to grain boundaries protrudes outside of the spherical pebble shape (Fig. 3, see constrained pebble irradiated at 730uC). The former edges of pebbles near such zones do not match together indicating plastic deformation under strong internal stresses induced by the combined action of thermal expansion and swelling. Physical mechanisms responsible for the formation of the features described above are to be discussed in details elsewhere. Probably, subgrain boundaries are effective in capturing gas-vacancy complexes thus resulting in more homogeneous distribution of bubbles. These results indicate that the compression stress characteristic for pebble bed significantly affects microstructure and pore/bubble morphology and these effects should be thoroughly investigated in future irradiation campaigns. Tritium release experiments on the irradiated pebbles performed at NRG, Petten and KIT show only one distinguished desorption peak in the range 880–950uC (Refs. 11, 18, and 19). Systematic comparison of the positions of tritium release peaks for the ‘‘constrained’’ and ‘‘unconstrained’’ pebbles (see Fig. 4) suggests that in six cases from seven tritium release from the ‘‘constrained’’ pebbles occurs slightly earlier than from the ‘‘unconstrained’’ both for 0.5 and 1 mm pebbles heated either with a rate of 1 or 7 K/min. Figure 5 presents the fraction of tritium retained in the pebbles with respect to the amount of generated tritium as a function of irradiation temperature. Released tritium activity drops down sharply above the irradiation temperature of 650uC. At the maximum irradiation temperature of 750uC the residual tritium activity is 6–8 times lower than at other irradiation temperatures. It is natural to suppose that in this case tritium was released from Be pebbles already during irradiation. The observed strong decrease of the residual tritium activity in the irradiated beryllium pebbles with increasing irradiation temperature is of outstanding importance for a sound evaluation of the integral tritium inventory remaining in highly neutron irradiated HCPB pebble beds. III.C. Development of Beryllides and Their Properties

Fig. 2. Relative change of the average pebble diameter for 0.5 mm (top), 1 mm (middle) and 2 mm (bottom) unconstrained (cyan) and constrained (blue) pebbles irradiated at 425, 525, 650 and 750uC. 32

Beryllium intermetallic compounds are considered as a backup solution for advanced neutron multiplier of future DEMO reactor. These materials shall further improve tritium release, reduce reactivity with air


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Fig. 3. Optical microscopy of cross sections of unconstrained (top row) and constrained (bottom row) pebbles irradiated at various temperatures.

or water vapor, and swelling at blanket operating temperatures. Here we report on the recent activities in the frame of Broader Approach in support of the ITER Project and an early realization of fusion energy performed at Rokkasho, Japan20,21 and at KIT, Germany.22 Beryllides are supposed to be used in the form of pebble beds. The same method as for fabrication the reference Be-pebbles, namely, rotating electrode process (REP), was selected for production of pebbles from Be12Ti. At the first step beryllide rod should be fabricated. Powder metallurgy route was tested for beryllide rod fabrication with a consequent consolidation by plasma sintering, hot isostatic pressing or hot extrusion. The Japanese group has checked casting and hot isostatic pressing, but finally selected plasma sintering method due to the lower impurity level and the lack of oxide layer in the final product.21 KIT successfully tested beryllide rod production using hot extrusion method.22 Intermetallic compounds are known to be brittle, so to avoid fracture of rods during pebble production by REP some amount of residual beryllium is favorable to increase ductility. The next production step consist in granulation, that is, in production of pebbles using rotating electrode process followed by annealing to obtain homogeneous one-phase material.23 It was recognized that larger diameter of the chamber used for REP is favorable for FUSION SCIENCE AND TECHNOLOGY

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increasing the yield of pebbles and their quality as far as in larger chamber liquid droplets have more time for solidification before they collide with the chamber walls. On cooling, the titanium rich Be17Ti2 phase nucleates and grows first. At lower temperatures Be12Ti phase is growing on the periphery of the Be17Ti2 particles due to diffusion of beryllium atoms into titanium rich phase. If time is not sufficient for complete diffusion of beryllium, pure beryllium is left outside the titanium phases. Additional annealing (homogenization heat treatment) allows beryllium to diffuse and to complete the phase transformation resulting in formation of essentially onephase Be12Ti with some residual porosity. Stability of titanium beryllide to irradiation was tested on many occasions. Be12Ti produced at KIT and irradiated in HIDOBE-01 shows at least one order of magnitude lower swelling than beryllium irradiated at the same temperature.24 Even the implantation of 2.5 at% He does not deteriorate material hardness.25

IV. MODELING AND DEVELOPMENT OF PREDICTION TOOLS FOR THE BLANKET DESIGN

The results of the HIDOBE-01 PIE reveal complicated and not always easily rationalised features of

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Fig. 4. Temperature of tritium release peak from unconstrained (cyan) and constrained (blue) pebbles irradiated at 425 and 750uC and tested with temperature ramping of 7 or 1 C/min.

beryllium performance under irradiation. The modeling of material behavior at a broad range of length and time scales, from atomistic to macroscopic, is potentially an important tool for understanding the basic mechanisms of beryllium response to severe environmental impact and for developing predictive tools to justify and guide the DEMO blanket design. One of the important tasks for the beryllium material development is to assure the pebble structural integrity under combined action of high temperature, mechanical stresses and irradiation. With this respect the radiationinduced swelling of pebbles requires clarification of the radiation damage kinetics, starting from the primary damage production and up to the development of physically-based predictive laws of pebble size and density evolution. Especially important is to clarify the effects of transmutation gas accumulation [tritium (Refs. 8 and 9), He (Ref. 5)] and impurities, because in their absence a pure Be is not expected to swell.7 Among the recent developments in the modeling of the defect and gas performance in beryllium one can mention simulation of the defect production in cascades in Be (Ref. 26). The overall cascade structure in beryllium is very different from that observed in more heavy metals applied in fusion devices (e.g. steels). Radiation defects are mostly created in subcascades along the projectile trajectory starting already from the PKA (primary knockon atom) energy of only *3 keV. As a result, in contrast to other structural materials, the damage production even exceeds the predictions based on the binary collision code SRIM (Ref. 27), indicating that the standard beryllium atom displacement energy of *15 eV can be an overestimation. On the other hand, at the average nearly half of the created defects are eliminated during the correlated recombination inside the cascade area. 34

Fig. 5. Fraction of retained tritium after irradiation at various temperatures with respect to the generated amount of tritium.

A progress is reached in the understanding of relevant hydrogen effects potentially important for simulation of tritium release and retention. An interesting prediction of the recent first-principles calculations28 is that hydrogen adsorption on free surfaces reduces the surface energy of beryllium, the magnitude of the effect being sensitive to the surface crystallographic orientation and hydrogen surface coverage. Therefore, hydrogen adsorption on the surface of helium bubbles can change a relation between the surface areas of various crystallographic planes of faceted helium bubbles thus affecting their shape. Another achievement is the improvement of our understanding of mechanisms controlling tritium desorption from beryllium pebbles. In the fusion reactor design, the tritium produced in pebbles by transmutation reactions is assumed to reach eventually the pebble outer surface, from which it can be removed by a purge gas pumped through the pebble bed. Since hydrogen can leave beryllium surface only in molecular form, it is currently believed that hydrogen molecules from the flowing gas dynamically dissociate and recover on Be pebble surfaces, so that tritium is removed mostly as HT molecules. Recent ab initio calculations have revealed, however, that this picture can be oversimplified.29 The tritium atoms adsorbed at clean beryllium surface possess excess charge and, thus, repel each other preventing formation of molecules and suppressing desorption at low temperatures until hydrogen surface coverage reaches a critical concentration (about half a monolayer surface coverage). Any additional hydrogen atom cannot avoid being the first neighbor to the others and forming a molecule which immediately desorbs without thermal excitation. At elevated temperatures, however, desorption is expected to occur at hydrogen coverage lower than the critical one. This has important implication for the development of predictive tools, implying that simple equations


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describing tritium desorption in standard predictive codes (e.g. TMAP code30) should be improved.

V. CONCLUSIONS

Four test breeder blanket concepts from six presently suggested for testing in ITER are considering beryllium as neutron multiplier. All concepts foresee beryllium pebble beds, which should demonstrate packing density necessary for sustainable tritium breeding as well as thermal conductivity and structural integrity for assuring stable operation in the defined temperature range. After testing in ITER a promising breeder blanket concept will be selected for the implementation in DEMO. If this concept foresees the use of beryllium, the large amounts of beryllium pebbles required in the DEMO reactor blanket will impose severe limitations on the Be-pebble fabrication process. The fabrication route has to be scaled to the industrial level for production of several hundred tons of beryllium pebbles with acceptable yield of the final product. In addition, reduction of the content of impurities controlling beryllium radioactive inventory should be performed to reduce nuclear waste recycling and utilization costs. Availability of the production route meeting these requirements and affordability of the final product in necessary amounts are crucial for the decision in favor of the beryllium-based tritium breeder concept for DEMO reactor. Close contacts between the major world beryllium manufacturers, material scientists and breeder blanket designers are necessary to meet these ambitious goals. Although, at present, the 1 mm Be pebbles remain the reference multiplier material for the EU HCPB concept, alternative scalable production routes of Be pebbles should be more actively explored to increase efficiency and to reduce costs of the pebble fabrication. Properties of the produced material should be controlled and the final product has to be qualified for the use in ITER and, later on, in DEMO. Material characterization should include investigation of chemical and phase composition, microstructure and mechanical properties, interaction with air/steam and tritium release properties. In addition, properties of pebble beds, such as packing density, thermal conductivity and mechanical response, should be studied. To assure acceptable radiation resistance of Bepebbles in the defined operating temperature range, new irradiation campaigns in fission reactors are crucial. The first important results on the microstructure, swelling and tritium release were delivered by the HIDOBE-01 post irradiation experiment and will be supplemented by the HIDOBE-02 PIE, which will start next year. Fission reactor irradiation up to ITER relevant dose (*2 dpa) is foreseen by F4E for qualification of beryllium materials for their use in ITER TBMs. FUSION SCIENCE AND TECHNOLOGY

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Several notable results obtained in the HIDOBE-01 PIE have important implications for further material development and characterization. Since pebbles irradiated in the constrained pebble beds show significantly different microstructure (more homogeneous spatial distribution of gas bubbles within grain, while bubbles on the grain boundaries are smaller than in the case of free pebbles) and higher structural integrity, irradiation of the constrained pebble bed mock-ups should be suggested for the future irradiation campaigns. Irradiation of beryllium single crystals stressed along different crystallographic orientations would be highly interesting for advancing modeling of irradiation creep and the effect of crystal anisotropy on point defect behavior. Although tritium retention drops down rapidly with the increase of irradiation temperature above 650uC, more than 50% of the generated tritium still remains trapped at this irradiation temperature (see Fig. 5). It should be checked if a short time annealing at temperatures above 650uC is sufficient to reduce tritium inventory without loss of pebbles’ structural integrity. The results obtained in the HIDOBE and future irradiation programs will be used for update and consolidation of the existing materials’ properties database in the form of Material Database (MDBR) and Material Assessment (MAR) Reports31 to be used for various stages of the TBMs design qualification. Many design issues such as, e.g., tritium accumulation and release, are intimately related to the development of the radiation induced microstructure and cannot be resolved without a deep understanding of the basic mechanisms of beryllium behavior under irradiation. Although many questions could be answered by performing dedicated experiments the obtained results should be reliably extrapolated to the fusion blanket operational conditions. Moreover, reliable modeling tools for prediction of tritium inventory and material behavior under normal and accidental conditions need to be developed. Recent advances in the technology of fabrication of titanium beryllides pebbles should draw more attention to this promising neutron multiplier material. New neutron irradiation tests are necessary for proving if its advantageous properties withstand irradiation. The use of titanium beryllides would allow increasing blanket operating temperature range and, hence, the overall efficiency of energy production in fusion reactor. After successful test in fission reactor, these materials can be verified in the ITER TBMs at the later stages of ITER operation.

ACKNOWLEDGMENTS Characterization of alternative routes of Be pebbles production and HIDOBE-01 PIE, was supported by the European Joint Undertaking for ITER and the Development of Fusion Energy (F4E) under the grant contracts F4E-2009GRT-030-A2 and -A3. Modeling of beryllium was partly

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performed within the EFDA IREMEV program. The views and opinions expressed herein reflect the authors’ views and do not necessary reflect those of the European Commission. F4E is not liable for any use that may be made of the information contained therein.

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