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IFMIF specifications from the users point of view
Fusion Engineering and Design 86 (2011) 611–614
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
IFMIF specifications from the users point of view P. Garin a , E. Diegele b , R. Heidinger c , A. Ibarra d,∗ , S. Jitsukawa e,f , H. Kimura e,f , A. Möslang g , T. Muroga h , T. Nishitani i , Y. Poitevin b , M. Sugimoto i , M. Zmitko b a
CEA, IFMIF/EVEDA Project Team, Rokkasho, Japan F4E, Barcelona, Spain F4E, Garching, Germany d CIEMAT, Avda. Complutense 22, Madrid 28040, Spain e JAEA, Tokai, Japan f JAEA, Rokkasho, Japan g KIT, Karlsruhe, Germany h NIFS, Toki, Japan i JAEA, IFMIF/EVEDA Project Team, Rokkasho, Japan b c
a r t i c l e
i n f o
Article history: Available online 16 February 2011 Keywords: IFMIF Radiation effects Fusion materials
a b s t r a c t This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the facility. Special attention is given to the different roadmaps of fusion pathway towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing. © 2011 Published by Elsevier B.V.
1. Introduction IFMIF (international fusion materials irradiation facility) is a neutron source based on a stripping reaction between two deuteron beams and a lithium target aimed to generate a fusion reactor relevant radiation environment with a high neutron flux. It is fusion specific and representative for fusion in-vessel components in terms of the n-spectrum, the damage as well as He and H transmutations generated, and, moreover, offers flexibility in loading conditions and instrumentation allowing simultaneous irradiation of typically thousands of specimens in sufficient volume to perform post-irradiation experiments for generation and validation of nuclear code qualified materials. The EVEDA phase of IFMIF is part of the Japan-EU Broader Approach (BA) agreement. In 2009, a Specification Working Group was established in order to update the IFMIF users requirements for the facility. The main terms of reference given to the group were: (i) to update the role and definition of IFMIF in the fusion (materials) roadmap, (ii) to define the main requirements to IFMIF, both for the irradiation programme and the post irradiation examination including a proper scheduling, (iii) to define the irradiation
∗ Corresponding author. E-mail address: [email protected] (A. Ibarra). 0920-3796/$ – see front matter © 2011 Published by Elsevier B.V. doi:10.1016/j.fusengdes.2011.01.109
characteristics of IFMIF, (iv) to target objectives for the irradiation modules related to any materials used for breeding blankets in terms of flux, temperature and type of tests, and (v) to propose guidelines for usage of the low flux and very low flux volumes, (vi) to identify other missing elements, for example in terms of other modules of interest or reference irradiation cycles. This paper summarizes the output of the work [1]. Specific attention will be paid to the roadmap of IFMIF, its role within the overall fusion roadmap and its relationship to other facilities, materials R&D activities for any of the materials, structural, (blanket) functional or non-metallic ones. 2. The role of IFMIF in the fusion roadmap The roadmaps towards fusion power developed in different countries commonly foresee the construction of two fusion machines before the industrial prototypes: ITER and DEMO [2–4]. It has also been recognized since more than two decades that material development and validation under irradiation are not only of highest importance for the economical success but are on the critical path for early use of fusion power. As a consequence, IFMIF is also considered an indispensable element of international roadmaps to fusion powers [4]. Taking a decade as a typical time constant in fusion roadmaps, then ITER will start operation in hydrogen in approximately a
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decade from today. DEMO engineering design phase and final design phase/construction are anticipated to last a decade, each. A key issue that determines the relative schedule of IFMIF operation and DEMO is the availability of a comprehensive database for irradiated structural materials. The timely availability of validated data is essential for the final design, licensing, safety assessment and the basis for reliable operation and lifetime evaluation of DEMO components. Data generated from IFMIF are at least needed and expected within the same time frame as results from ITER operation. This statement holds also for machines based on different physics, in particular stellarators and other facilities that can be categorized as fusion nuclear technology test devices. In the development strategy for fusion materials each of the following elements is essential and has its benefit and added value: irradiation in (fission) reactors, multi-ion beam irradiations, modeling of radiation defects. However, any arbitrary method to simulate He and H transmutation, is not sufficient because of various reasons like insufficient amount of transmutations or/and incorrect production rates, limited irradiation volume or homogeneity in irradiation conditions. Consequently, the final validation in terms of engineering data or data for licensing will be only provided by a facility that combines in due time (i) high flux of fusion specific neutrons in a sufficiently large volume (ii) allowing the simultaneous irradiation of about 1000 specimens in a wide temperature range (iii) with a reliable and affordable technology. As a result of high ranking advisory panels the consensus was reached that the concept of an accelerator based source which utilizes the deuterium–lithium stripping reaction (D–Li-source, later called IFMIF) for neutron production was the only choice with the potential to fulfill the demanded requirements within a realistic time scale [5,6]. The generic fusion roadmap and the basic relations between the different pillars should not change by a relative shift in time, e.g. caused by the delay of ITER operation. However, depending on the relative start of ITER (D–T) phase compared to IFMIF operation, IFMIF could prepare or contribute to address unresolved issues and/or uncertainties that impose potential risk on ITER test blanket modules (TBM) operation. Main IFMIF expected contributions can be summarized in the following: • To provide data for the engineering design for DEMO (main role in the initial period of operation), • To provide information to define performance limits of materials and materials systems for DEMO and beyond, • To contribute to the completion and validation of (existing) databases to gather and confirm data required for licensing and safety assessment, • To contribute to the selection or optimization of different alternative fusion materials, • To validate the fundamental understanding of radiation response of materials including benchmarking of irradiation effects modeling at length-scale and time-scale relevant for engineering application, and • To tests blankets and functional materials prior to or complementary to ITER test blanket modules. The assessment of the role of IFMIF and its expected contribution to the fusion roadmap described above are valid independently of the detailed definition of DEMO and the time scale. In particular, the analyses presented in the following sections are based on present standard operational scenarios (neutron wall loading, regular exchange of plasma facing components), anticipated loading conditions and current design values (e.g. cooling temperatures). The values might be subject to change, however, the proposals and specification detailed below should hold.
3. IFMIF specifications for structural materials One of the main objectives of IFMIF is to evaluate irradiation performance of the structural materials under fusion (first wall/blanket) typical conditions for DEMO engineering design. Although a number of innovative concept DEMO designs with advanced materials have been proposed, the current primary candidate materials for the DEMO in, both, the EU and Japan are reduced activation ferritic/martensitic (RAF/M) steels, together with water or helium cooling, including martensitic oxide dispersion strengthened (ODS) steel. Ferritic oxide dispersion strengthened steels, vanadium alloys and SiC/SiC composite materials are candidate structural materials for advanced blanket concepts. Tungsten alloys are an option for gas-cooled divertor concepts [7]. A key issue of materials development for structures exposed to intense fusion neutrons, is to generate data on irradiation effects on microstructure, mechanical and physical properties, as well as on compatibility including also various types of joints and welds – under various conditions and for different reasons, for example for conceptual and engineering design, lifetime evaluation and safety analyses. It is also important to have these data as early as possible as complete as possible and at lowest uncertainties. The expected maximum neutron fluence for the materials of the first wall for DEMO is ∼4 MWa/m2 (early Demo) up to ∼10 MWa/m2 with a even higher development goal for a power plant. This corresponds to a displacement damage level higher than (40–) 100 dpa with He levels of (500–) 1400 appm and H levels of (1500–) 4000 appm for iron base alloys. For IFMIF, such damage levels and dose rata should be to achieved in around the expected lifetime of the blanket (3–5 years) within a volume that is sufficient to allow simultaneous irradiation of more than 1000 samples to provide the key pillar of a fusion materials database. Presently, the irradiation volume available at the high flux test region is expected to be of about 500 cm3 and several liters in the medium flux test region. Typical service conditions the temperature window to be tested ranges from ∼0.3 to ∼0.5 of the homologous temperature, i.e. 300–550 ◦ C for the RAF/M steels (∼400–750 ◦ C for ODS steels), 400–800 ◦ C for vanadium, 650–1200 ◦ C for tungsten and 600–1100 ◦ C, for SiC/SiC, There are severe restrictions on the homogeneity and stability of the irradiation temperature: with a temperature stability of ±5% in the 80% of the samples volume. Evolution of microstructure under irradiation governs the degradation of properties and is known to be strongly dependent from the damage level and from the level of transmutations and also to some extend from the respective damage rates. Therefore, it is desirable to minimize neutron flux changes during the irradiation. A requirement on time structure of IFMIF neutron beam has been made to be quasi continuous. Irradiation effects on mechanical properties of the materials may be categorized into the properties exhibiting cumulative property changes with damage levels and those exhibiting changes depending on damage rate (i.e. exposure time). Tensile property (strength and ductility), fracture toughness (including DBTT), fatigue crack growth and creep deformation measurement are typical items examined after irradiation and are of high importance to determine allowable values in the terms of temperature-applied load (stress, strain, strain range, stress rate, strain rate). These specimens may be also used for the tests to evaluated high temperature He embrittlement. Tests on mechanical properties using current international standards would require large specimens occupying (too) large irradiation volume. This drives the development of SSTT (small scale test technique). Specimens of IFMIF-SSTT often are categorized to be “macro-miniaturized” specimens, similar in shape to standard specimen, smaller in size ranging typically from one tenth
P. Garin et al. / Fusion Engineering and Design 86 (2011) 611–614
to one third, depending on the tests. The limit is given for example by a minimum number of grains through thickness or a suitable length of flaws in pre-cracked specimen. The methodology still has to be demonstrated and has to be implemented in international standards and norms. Deformation by pure fatigue loading, pure constant (creep) load and damage caused by high energy neutrons introduce microstructural changes in different manners indicating interactions between different damage mechanisms. Therefore, tests on fatigue, creepfatigue interaction, fatigue crack growth and crack growth by creep-fatigue interaction are desirable to be conducted by controlled in situ testing during irradiation. Except for the creep tests under irradiation with pressurized tubes, in situ mechanical tests tend to occupy relatively large irradiation volumes. Therefore, these experiments cannot be performed in the small area of high n-flux. Compatibility may be evaluated by post irradiation examination except for the effect of radiolysis occurred during irradiation.
4. IFMIF specifications for blanket functional materials systems DEMO blanket has to withstand high neutron flux typically in the order of 15–30 dpa/year under continuous operation. So far the blanket is planned to be replaced every 2–3 years, mainly due to the irradiation damage of the structural material. In the helium cooled pebble bed (HCPB) and water cooled ceramic (WCC) blankets, Li-based ceramic pebble bed (Li2 TiO3 or Li4 SiO4 ) will be used as tritium breeder and metal beryllium or beryllium alloy pebble bed as neutron multiplier. These so-called functional materials have to withstand up to about 50 dpa, which is about 80% of expected damage on the first wall. Therefore, IFMIF irradiation capability is essential for the qualification of blanket functional materials, which could also be completed with dedicated fission reactor irradiations. In IFMIF irradiation tests in the medium and low flux regions are desired, as well as post irradiation examination (PIE) of irradiated samples/specimens. The tritium breeder material of HCPB and WCCB blanket concepts is a Li-based ceramic in the form of pebbles bed. Presently, two options are envisaged: the lithium orthosilicate (‘Si’, Li4 SiO4 ) and the lithium metatitanate (‘MTi’, Li2 TiO3 ). In a DEMO blanket, the temperature limit of the tritium breeder pebbles is set up to 1000 ◦ C. Main irradiation issues are related to the change in the tritium release and recovery characteristics (in order to assure a minimum tritium inventory), thermal properties (mainly thermal conductivity), mechanical stability and the compatibility with the structural materials. As in the previous section, some of these tests, mainly those related to tritium release, must be made in situ. Irradiation temperatures will range between 300 and 1100 ◦ C up with a typical dose rate around 15–30 dpa/year. Solid breeder blankets with lithium ceramics as a breeder and steel as a structural material require beryllium as a neutron multiplier to increase the tritium breeding ratio (TBR) performance. Beryllium or beryllium alloys is used in the form of pebbles. Main irradiation issues are related to changes in the tritium release characteristics (in order to assure a minimum tritium inventory), thermal properties (mainly thermal conductivity), mechanical stability and the compatibility with the structural materials. Irradiation temperatures will range between 300 and 900 ◦ C up with an expected maximum neutron fluence up to ∼50–90 dpa and ∼15,000–20,000 appm He. In liquid breeder blankets, Li–Pb, Li or molten salt Flibe will be used as tritium breeder. The tritium breeder materials of the liquid breeder alternatives are free from radiation damage but compositional change by transmutation and compatibility with structural materials under irradiation can be issues. These properties must
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be tested in the temperature range from 300 to 1000 ◦ C and under relevant temperature gradient conditions. On the other side, development of ceramics coatings is a key technology for many of the blanket concepts. Coatings and/or flow channel insert (FCI) are envisaged for tritium permeation reduction in blankets using water, gas, Li–Pb or Flibe as coolants, and for electrical insulation to mitigate MHD pressure drop for selfcooled Li–Pb and Li blankets. Coatings can also be envisaged, when needed, as anti-corrosion barrier between the structural material and the liquid breeder. The temperature of the coating will be near that of the coolant. The neutron fluence is expected to be up to about 20 dpa. Main irradiation issues are related to the permeation reduction factor during irradiation and/or the electrical resistivity although the thermal conductivity as well as the mechanical stability and compatibility aspects are also important. The first ones must be measured in situ in the temperature range from 300 to 900 ◦ C. Besides this, it has been also identified that IFMIF can be used to define TBM fluence relevant experiments that could be beneficial to the qualification and licensing of TBM in ITER, provided it is available on time, as well as reduced size DEMO design-oriented experiments for testing or qualification of different breeder alternatives. Typically the first ones will require irradiation temperatures between 500 and 850 ◦ C up to doses lower than 2 dpa and onsite measurement of tritium release. On the other side, the second ones will require the irradiation of an assembly of a breeder and a structural material up to 50 dpa at temperatures between 300 and 1100 ◦ C as well as on-site measurement of tritium release.
5. IFMIF specifications for “non-metallic materials” The fusion reactor will require not only metallic structural materials, but also non-metallic materials, mainly ceramics, under various irradiation conditions. Generally speaking, these materials are mainly used in the different diagnostics systems (e.g. bolometers, magnetic pick-up coils, Thompson scattering systems) as well as in the heating and current drive systems (for example, the RF windows in the ECR system) and in the magnetic system (superconductor and insulation resin between the superconductors and layers). Most of these materials show higher sensitivity to radiation than structural metals, mainly because the critical properties in the case of their application for fusion reactors are physical properties much more sensitive to radiation effects. In case of non-metallic materials, irradiation effects on those characteristics are affected by both atomic displacement and electric excitation. Furthermore, dynamic irradiation effects are important in those irradiation effects. So the in situ measurement techniques are essential in those irradiation tests of the non-metallic materials. In some of the latter cases, the radiation effects can show a strong influence on the atmosphere around the material (vacuum, gas, etc.), which needs to be also taken into account. Insulating materials will be used in feedthroughs, connectors, mechanical supports, and mineral insulated (MI) cables for the diagnostics, control and heating devices. In these materials not only electrical properties (mainly radiation induced conductivity and radiation induced electrical degradation), but also thermal conductivity and mechanical properties such as dimensional stability are important functions and are irradiation issues. Some ceramics insulator will be used near the first wall, such as in-vessel diagnostic sensors or breeding blanket instruments so the irradiation condition will be up to more than ∼10 dpa. Irradiation temperatures should range between 20 and 500 ◦ C. Generally speaking, measurement of electrical properties must be made during irradiation and, in some specific cases, atmosphere control is required. Non metallic materials required in diagnostic and heating and current drive systems for the advanced fusion devices are typically
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window (transmission) materials, in which the main functional requirement is the maintenance of the high optical transmission in the first case and the maintenance of low dielectric loss and high thermal conductivity in the second. Due to the high radiation sensitivity of most of these materials, usually they are located quite far away from the first wall and so the irradiation condition will be lower than ∼1 dpa (with the exception of plasma facing reflectors in which neutron fluences can be much higher). Irradiation temperatures should range also between 20 and 500 ◦ C and the measurement of optical properties as well as impurities diffusion and electrical properties (as indicator for dielectric loss increase) must be made during irradiation. More detailed dielectric properties characterization at specific frequency bands in the 1 MHz and 200 GHz range, optical properties – transmission and reflection – thermal conductivity and mechanical properties characterization – dimensional stability and ultimate bending strength – can be made off-line. In the superconducting magnets, superconducting materials and organic insulation materials such as glass fiber reinforced plastics are used. The neutron flux at the superconducting magnet is not so high (around 0.01 dpa), but the lifetime of the fusion reactor is limited by the irradiation damage of the superconducting materials and the organic insulator of the superconducting magnet. In this last case the main irradiation issues are the electrical insulation as well as the chemical stability and mechanical properties. The irradiation must be made at low temperatures (around 100 K). 6. Conclusions This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the plant. Obtained
main conclusions related to materials characterization are quite similar to those previously discussed in [8], although the essential role of material technologies testing (joints and welds) as well as materials systems is greatly emphasized. New identified findings points out the role of IFMIF for testing small scale design-oriented mockups and the possible role of IFMIF for validation of ITER oriented TBM licensing aspects. A significant effort is required in the future to confirm the potential of IFMIF for all the different types of experiments and experimental conditions identified. The interaction between the IFMIF design team and IFMIF users will thus continue, to regularly check that the design proposed is well fulfilling their expectations. Acknowledgments The IFMIF/EVEDA Project is one of the three projects of the Broader Approach agreement between Europe and Japan. References [1] P. Garin et al, IFMIF Users Specifications and Proposals, BA D 224ERJ Report, March 2010. [2] K. Lackner, R. Andreani, D. Campbell, M. Gasparotto, D. Maisonnier, M.A. Pick, J. Nucl. Mater. 307–311 (2002) 10. [3] T. Muroga, M. Gasparotto, S.J. Zinkle, Fusion Eng. Des. 61–62 (2002) 13. [4] D. Maisonnier, J. Hayward, Technological and engineering challenges of fusion, in: 2nd IAEA Technical Meeting on First Generation of Fusion Power Plant, Vienna, 20–22 June, 2007. [5] A. Cottrell et al., Report of the Panel on Fusion Materials Research and Testing, IEA Implementing Agreement R&D Fusion Materials, 1983. [6] J.E. Leis et al., Report on the International Fusion Irradiation Facility, IEA Workshop San Diego, USA, February 14–17 1989, vol. 1: Evaluation Panel Report, vol. 2: Technical Presentations. [7] N. Baluc, et al., Nucl. Fusion 47 (2007) S696–S717. [8] IFMIF International Team, IFMIF Comprehensive Design Report, January 2004.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
IFMIF specifications from the users point of view P. Garin a , E. Diegele b , R. Heidinger c , A. Ibarra d,∗ , S. Jitsukawa e,f , H. Kimura e,f , A. Möslang g , T. Muroga h , T. Nishitani i , Y. Poitevin b , M. Sugimoto i , M. Zmitko b a
CEA, IFMIF/EVEDA Project Team, Rokkasho, Japan F4E, Barcelona, Spain F4E, Garching, Germany d CIEMAT, Avda. Complutense 22, Madrid 28040, Spain e JAEA, Tokai, Japan f JAEA, Rokkasho, Japan g KIT, Karlsruhe, Germany h NIFS, Toki, Japan i JAEA, IFMIF/EVEDA Project Team, Rokkasho, Japan b c
a r t i c l e
i n f o
Article history: Available online 16 February 2011 Keywords: IFMIF Radiation effects Fusion materials
a b s t r a c t This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the facility. Special attention is given to the different roadmaps of fusion pathway towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing. © 2011 Published by Elsevier B.V.
1. Introduction IFMIF (international fusion materials irradiation facility) is a neutron source based on a stripping reaction between two deuteron beams and a lithium target aimed to generate a fusion reactor relevant radiation environment with a high neutron flux. It is fusion specific and representative for fusion in-vessel components in terms of the n-spectrum, the damage as well as He and H transmutations generated, and, moreover, offers flexibility in loading conditions and instrumentation allowing simultaneous irradiation of typically thousands of specimens in sufficient volume to perform post-irradiation experiments for generation and validation of nuclear code qualified materials. The EVEDA phase of IFMIF is part of the Japan-EU Broader Approach (BA) agreement. In 2009, a Specification Working Group was established in order to update the IFMIF users requirements for the facility. The main terms of reference given to the group were: (i) to update the role and definition of IFMIF in the fusion (materials) roadmap, (ii) to define the main requirements to IFMIF, both for the irradiation programme and the post irradiation examination including a proper scheduling, (iii) to define the irradiation
∗ Corresponding author. E-mail address: [email protected] (A. Ibarra). 0920-3796/$ – see front matter © 2011 Published by Elsevier B.V. doi:10.1016/j.fusengdes.2011.01.109
characteristics of IFMIF, (iv) to target objectives for the irradiation modules related to any materials used for breeding blankets in terms of flux, temperature and type of tests, and (v) to propose guidelines for usage of the low flux and very low flux volumes, (vi) to identify other missing elements, for example in terms of other modules of interest or reference irradiation cycles. This paper summarizes the output of the work [1]. Specific attention will be paid to the roadmap of IFMIF, its role within the overall fusion roadmap and its relationship to other facilities, materials R&D activities for any of the materials, structural, (blanket) functional or non-metallic ones. 2. The role of IFMIF in the fusion roadmap The roadmaps towards fusion power developed in different countries commonly foresee the construction of two fusion machines before the industrial prototypes: ITER and DEMO [2–4]. It has also been recognized since more than two decades that material development and validation under irradiation are not only of highest importance for the economical success but are on the critical path for early use of fusion power. As a consequence, IFMIF is also considered an indispensable element of international roadmaps to fusion powers [4]. Taking a decade as a typical time constant in fusion roadmaps, then ITER will start operation in hydrogen in approximately a
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decade from today. DEMO engineering design phase and final design phase/construction are anticipated to last a decade, each. A key issue that determines the relative schedule of IFMIF operation and DEMO is the availability of a comprehensive database for irradiated structural materials. The timely availability of validated data is essential for the final design, licensing, safety assessment and the basis for reliable operation and lifetime evaluation of DEMO components. Data generated from IFMIF are at least needed and expected within the same time frame as results from ITER operation. This statement holds also for machines based on different physics, in particular stellarators and other facilities that can be categorized as fusion nuclear technology test devices. In the development strategy for fusion materials each of the following elements is essential and has its benefit and added value: irradiation in (fission) reactors, multi-ion beam irradiations, modeling of radiation defects. However, any arbitrary method to simulate He and H transmutation, is not sufficient because of various reasons like insufficient amount of transmutations or/and incorrect production rates, limited irradiation volume or homogeneity in irradiation conditions. Consequently, the final validation in terms of engineering data or data for licensing will be only provided by a facility that combines in due time (i) high flux of fusion specific neutrons in a sufficiently large volume (ii) allowing the simultaneous irradiation of about 1000 specimens in a wide temperature range (iii) with a reliable and affordable technology. As a result of high ranking advisory panels the consensus was reached that the concept of an accelerator based source which utilizes the deuterium–lithium stripping reaction (D–Li-source, later called IFMIF) for neutron production was the only choice with the potential to fulfill the demanded requirements within a realistic time scale [5,6]. The generic fusion roadmap and the basic relations between the different pillars should not change by a relative shift in time, e.g. caused by the delay of ITER operation. However, depending on the relative start of ITER (D–T) phase compared to IFMIF operation, IFMIF could prepare or contribute to address unresolved issues and/or uncertainties that impose potential risk on ITER test blanket modules (TBM) operation. Main IFMIF expected contributions can be summarized in the following: • To provide data for the engineering design for DEMO (main role in the initial period of operation), • To provide information to define performance limits of materials and materials systems for DEMO and beyond, • To contribute to the completion and validation of (existing) databases to gather and confirm data required for licensing and safety assessment, • To contribute to the selection or optimization of different alternative fusion materials, • To validate the fundamental understanding of radiation response of materials including benchmarking of irradiation effects modeling at length-scale and time-scale relevant for engineering application, and • To tests blankets and functional materials prior to or complementary to ITER test blanket modules. The assessment of the role of IFMIF and its expected contribution to the fusion roadmap described above are valid independently of the detailed definition of DEMO and the time scale. In particular, the analyses presented in the following sections are based on present standard operational scenarios (neutron wall loading, regular exchange of plasma facing components), anticipated loading conditions and current design values (e.g. cooling temperatures). The values might be subject to change, however, the proposals and specification detailed below should hold.
3. IFMIF specifications for structural materials One of the main objectives of IFMIF is to evaluate irradiation performance of the structural materials under fusion (first wall/blanket) typical conditions for DEMO engineering design. Although a number of innovative concept DEMO designs with advanced materials have been proposed, the current primary candidate materials for the DEMO in, both, the EU and Japan are reduced activation ferritic/martensitic (RAF/M) steels, together with water or helium cooling, including martensitic oxide dispersion strengthened (ODS) steel. Ferritic oxide dispersion strengthened steels, vanadium alloys and SiC/SiC composite materials are candidate structural materials for advanced blanket concepts. Tungsten alloys are an option for gas-cooled divertor concepts [7]. A key issue of materials development for structures exposed to intense fusion neutrons, is to generate data on irradiation effects on microstructure, mechanical and physical properties, as well as on compatibility including also various types of joints and welds – under various conditions and for different reasons, for example for conceptual and engineering design, lifetime evaluation and safety analyses. It is also important to have these data as early as possible as complete as possible and at lowest uncertainties. The expected maximum neutron fluence for the materials of the first wall for DEMO is ∼4 MWa/m2 (early Demo) up to ∼10 MWa/m2 with a even higher development goal for a power plant. This corresponds to a displacement damage level higher than (40–) 100 dpa with He levels of (500–) 1400 appm and H levels of (1500–) 4000 appm for iron base alloys. For IFMIF, such damage levels and dose rata should be to achieved in around the expected lifetime of the blanket (3–5 years) within a volume that is sufficient to allow simultaneous irradiation of more than 1000 samples to provide the key pillar of a fusion materials database. Presently, the irradiation volume available at the high flux test region is expected to be of about 500 cm3 and several liters in the medium flux test region. Typical service conditions the temperature window to be tested ranges from ∼0.3 to ∼0.5 of the homologous temperature, i.e. 300–550 ◦ C for the RAF/M steels (∼400–750 ◦ C for ODS steels), 400–800 ◦ C for vanadium, 650–1200 ◦ C for tungsten and 600–1100 ◦ C, for SiC/SiC, There are severe restrictions on the homogeneity and stability of the irradiation temperature: with a temperature stability of ±5% in the 80% of the samples volume. Evolution of microstructure under irradiation governs the degradation of properties and is known to be strongly dependent from the damage level and from the level of transmutations and also to some extend from the respective damage rates. Therefore, it is desirable to minimize neutron flux changes during the irradiation. A requirement on time structure of IFMIF neutron beam has been made to be quasi continuous. Irradiation effects on mechanical properties of the materials may be categorized into the properties exhibiting cumulative property changes with damage levels and those exhibiting changes depending on damage rate (i.e. exposure time). Tensile property (strength and ductility), fracture toughness (including DBTT), fatigue crack growth and creep deformation measurement are typical items examined after irradiation and are of high importance to determine allowable values in the terms of temperature-applied load (stress, strain, strain range, stress rate, strain rate). These specimens may be also used for the tests to evaluated high temperature He embrittlement. Tests on mechanical properties using current international standards would require large specimens occupying (too) large irradiation volume. This drives the development of SSTT (small scale test technique). Specimens of IFMIF-SSTT often are categorized to be “macro-miniaturized” specimens, similar in shape to standard specimen, smaller in size ranging typically from one tenth
P. Garin et al. / Fusion Engineering and Design 86 (2011) 611–614
to one third, depending on the tests. The limit is given for example by a minimum number of grains through thickness or a suitable length of flaws in pre-cracked specimen. The methodology still has to be demonstrated and has to be implemented in international standards and norms. Deformation by pure fatigue loading, pure constant (creep) load and damage caused by high energy neutrons introduce microstructural changes in different manners indicating interactions between different damage mechanisms. Therefore, tests on fatigue, creepfatigue interaction, fatigue crack growth and crack growth by creep-fatigue interaction are desirable to be conducted by controlled in situ testing during irradiation. Except for the creep tests under irradiation with pressurized tubes, in situ mechanical tests tend to occupy relatively large irradiation volumes. Therefore, these experiments cannot be performed in the small area of high n-flux. Compatibility may be evaluated by post irradiation examination except for the effect of radiolysis occurred during irradiation.
4. IFMIF specifications for blanket functional materials systems DEMO blanket has to withstand high neutron flux typically in the order of 15–30 dpa/year under continuous operation. So far the blanket is planned to be replaced every 2–3 years, mainly due to the irradiation damage of the structural material. In the helium cooled pebble bed (HCPB) and water cooled ceramic (WCC) blankets, Li-based ceramic pebble bed (Li2 TiO3 or Li4 SiO4 ) will be used as tritium breeder and metal beryllium or beryllium alloy pebble bed as neutron multiplier. These so-called functional materials have to withstand up to about 50 dpa, which is about 80% of expected damage on the first wall. Therefore, IFMIF irradiation capability is essential for the qualification of blanket functional materials, which could also be completed with dedicated fission reactor irradiations. In IFMIF irradiation tests in the medium and low flux regions are desired, as well as post irradiation examination (PIE) of irradiated samples/specimens. The tritium breeder material of HCPB and WCCB blanket concepts is a Li-based ceramic in the form of pebbles bed. Presently, two options are envisaged: the lithium orthosilicate (‘Si’, Li4 SiO4 ) and the lithium metatitanate (‘MTi’, Li2 TiO3 ). In a DEMO blanket, the temperature limit of the tritium breeder pebbles is set up to 1000 ◦ C. Main irradiation issues are related to the change in the tritium release and recovery characteristics (in order to assure a minimum tritium inventory), thermal properties (mainly thermal conductivity), mechanical stability and the compatibility with the structural materials. As in the previous section, some of these tests, mainly those related to tritium release, must be made in situ. Irradiation temperatures will range between 300 and 1100 ◦ C up with a typical dose rate around 15–30 dpa/year. Solid breeder blankets with lithium ceramics as a breeder and steel as a structural material require beryllium as a neutron multiplier to increase the tritium breeding ratio (TBR) performance. Beryllium or beryllium alloys is used in the form of pebbles. Main irradiation issues are related to changes in the tritium release characteristics (in order to assure a minimum tritium inventory), thermal properties (mainly thermal conductivity), mechanical stability and the compatibility with the structural materials. Irradiation temperatures will range between 300 and 900 ◦ C up with an expected maximum neutron fluence up to ∼50–90 dpa and ∼15,000–20,000 appm He. In liquid breeder blankets, Li–Pb, Li or molten salt Flibe will be used as tritium breeder. The tritium breeder materials of the liquid breeder alternatives are free from radiation damage but compositional change by transmutation and compatibility with structural materials under irradiation can be issues. These properties must
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be tested in the temperature range from 300 to 1000 ◦ C and under relevant temperature gradient conditions. On the other side, development of ceramics coatings is a key technology for many of the blanket concepts. Coatings and/or flow channel insert (FCI) are envisaged for tritium permeation reduction in blankets using water, gas, Li–Pb or Flibe as coolants, and for electrical insulation to mitigate MHD pressure drop for selfcooled Li–Pb and Li blankets. Coatings can also be envisaged, when needed, as anti-corrosion barrier between the structural material and the liquid breeder. The temperature of the coating will be near that of the coolant. The neutron fluence is expected to be up to about 20 dpa. Main irradiation issues are related to the permeation reduction factor during irradiation and/or the electrical resistivity although the thermal conductivity as well as the mechanical stability and compatibility aspects are also important. The first ones must be measured in situ in the temperature range from 300 to 900 ◦ C. Besides this, it has been also identified that IFMIF can be used to define TBM fluence relevant experiments that could be beneficial to the qualification and licensing of TBM in ITER, provided it is available on time, as well as reduced size DEMO design-oriented experiments for testing or qualification of different breeder alternatives. Typically the first ones will require irradiation temperatures between 500 and 850 ◦ C up to doses lower than 2 dpa and onsite measurement of tritium release. On the other side, the second ones will require the irradiation of an assembly of a breeder and a structural material up to 50 dpa at temperatures between 300 and 1100 ◦ C as well as on-site measurement of tritium release.
5. IFMIF specifications for “non-metallic materials” The fusion reactor will require not only metallic structural materials, but also non-metallic materials, mainly ceramics, under various irradiation conditions. Generally speaking, these materials are mainly used in the different diagnostics systems (e.g. bolometers, magnetic pick-up coils, Thompson scattering systems) as well as in the heating and current drive systems (for example, the RF windows in the ECR system) and in the magnetic system (superconductor and insulation resin between the superconductors and layers). Most of these materials show higher sensitivity to radiation than structural metals, mainly because the critical properties in the case of their application for fusion reactors are physical properties much more sensitive to radiation effects. In case of non-metallic materials, irradiation effects on those characteristics are affected by both atomic displacement and electric excitation. Furthermore, dynamic irradiation effects are important in those irradiation effects. So the in situ measurement techniques are essential in those irradiation tests of the non-metallic materials. In some of the latter cases, the radiation effects can show a strong influence on the atmosphere around the material (vacuum, gas, etc.), which needs to be also taken into account. Insulating materials will be used in feedthroughs, connectors, mechanical supports, and mineral insulated (MI) cables for the diagnostics, control and heating devices. In these materials not only electrical properties (mainly radiation induced conductivity and radiation induced electrical degradation), but also thermal conductivity and mechanical properties such as dimensional stability are important functions and are irradiation issues. Some ceramics insulator will be used near the first wall, such as in-vessel diagnostic sensors or breeding blanket instruments so the irradiation condition will be up to more than ∼10 dpa. Irradiation temperatures should range between 20 and 500 ◦ C. Generally speaking, measurement of electrical properties must be made during irradiation and, in some specific cases, atmosphere control is required. Non metallic materials required in diagnostic and heating and current drive systems for the advanced fusion devices are typically
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window (transmission) materials, in which the main functional requirement is the maintenance of the high optical transmission in the first case and the maintenance of low dielectric loss and high thermal conductivity in the second. Due to the high radiation sensitivity of most of these materials, usually they are located quite far away from the first wall and so the irradiation condition will be lower than ∼1 dpa (with the exception of plasma facing reflectors in which neutron fluences can be much higher). Irradiation temperatures should range also between 20 and 500 ◦ C and the measurement of optical properties as well as impurities diffusion and electrical properties (as indicator for dielectric loss increase) must be made during irradiation. More detailed dielectric properties characterization at specific frequency bands in the 1 MHz and 200 GHz range, optical properties – transmission and reflection – thermal conductivity and mechanical properties characterization – dimensional stability and ultimate bending strength – can be made off-line. In the superconducting magnets, superconducting materials and organic insulation materials such as glass fiber reinforced plastics are used. The neutron flux at the superconducting magnet is not so high (around 0.01 dpa), but the lifetime of the fusion reactor is limited by the irradiation damage of the superconducting materials and the organic insulator of the superconducting magnet. In this last case the main irradiation issues are the electrical insulation as well as the chemical stability and mechanical properties. The irradiation must be made at low temperatures (around 100 K). 6. Conclusions This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the plant. Obtained
main conclusions related to materials characterization are quite similar to those previously discussed in [8], although the essential role of material technologies testing (joints and welds) as well as materials systems is greatly emphasized. New identified findings points out the role of IFMIF for testing small scale design-oriented mockups and the possible role of IFMIF for validation of ITER oriented TBM licensing aspects. A significant effort is required in the future to confirm the potential of IFMIF for all the different types of experiments and experimental conditions identified. The interaction between the IFMIF design team and IFMIF users will thus continue, to regularly check that the design proposed is well fulfilling their expectations. Acknowledgments The IFMIF/EVEDA Project is one of the three projects of the Broader Approach agreement between Europe and Japan. References [1] P. Garin et al, IFMIF Users Specifications and Proposals, BA D 224ERJ Report, March 2010. [2] K. Lackner, R. Andreani, D. Campbell, M. Gasparotto, D. Maisonnier, M.A. Pick, J. Nucl. Mater. 307–311 (2002) 10. [3] T. Muroga, M. Gasparotto, S.J. Zinkle, Fusion Eng. Des. 61–62 (2002) 13. [4] D. Maisonnier, J. Hayward, Technological and engineering challenges of fusion, in: 2nd IAEA Technical Meeting on First Generation of Fusion Power Plant, Vienna, 20–22 June, 2007. [5] A. Cottrell et al., Report of the Panel on Fusion Materials Research and Testing, IEA Implementing Agreement R&D Fusion Materials, 1983. [6] J.E. Leis et al., Report on the International Fusion Irradiation Facility, IEA Workshop San Diego, USA, February 14–17 1989, vol. 1: Evaluation Panel Report, vol. 2: Technical Presentations. [7] N. Baluc, et al., Nucl. Fusion 47 (2007) S696–S717. [8] IFMIF International Team, IFMIF Comprehensive Design Report, January 2004.