Vanadium Alloys for Fusion Blanket Applications

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1 Materials Transactions, Vol. 46, No. 3 (2005) pp. 405 to 411 Special Issue on Fusion Blanket Structural Materials R&D in Japan #2005 The Japan Institute of Metals OVERVIEW Vanadium Alloys for Fusion Blanket Applications Takeo Muroga National Institute for Fusion Science, Toki , Japan Vanadium alloys are attractive candidate structural materials for breeding blanket of fusion reactors because of their low activation properties and high temperature strength. Studies on various blanket designs are being carried out using vanadium alloys as structural materials and liquid lithium or molten salt Flibe as breeding and coolant materials. Recently significant progress in fabrication technology has been made for vanadium alloys through a program of producing large ingots of high purity V-4Cr-4Ti (NIFS-HEATs). Fundamental understandings on the effects of interstitial impurities (C, N, O) on mechanical properties, radiation effects on microstructure and mechanical properties, corrosion and compatibility in various environments were enhanced. New promising candidates were identified for the MHD insulator coating. Among the remaining critical issues are the effects of transmutant helium on mechanical properties and development of long life MHD coating. Development of vanadium alloys will be carried out in coordination with IFMIF and ITER-TBM schedules as essential tools for verification of their performance in fusion blanket environments. (Received September 17, 2004; Accepted December 16, 2004) Keywords: vanadium alloy, breeding blanket, fusion reactor 1. Introduction There are enough rationales to claim that vanadium alloys are promising blanket structural materials for fusion power systems. 1,2) They have low induced activation characteristics, high temperature strength and high thermal stress factors. Recent researches have successfully resolved many of the critical issues and enhanced the feasibility of vanadium alloys as fusion blanket structural materials. The research emphasis of vanadium alloys for fusion reactors has been made on V-4Cr-4Ti alloy as a reference vanadium alloy. Large and medium heats of V-4Cr-4Ti have been made in the US, 3) Japan 2,4) and Russia. 5) Especially high purity V-4Cr-4Ti ingot made by collaboration of NIFS and Japanese Universities (NIFS-HEAT-1 and 2) showed superior properties for manufacturing due to the reduced level of oxygen impurities. 2,6) Figure 1 compares radioactivity after use in the first wall of a fusion commercial reactor for four reference alloys. The full-remote and full-hands-on recycle limits are shown to indicate the guideline for recycling and reuse. 7) SS316LN-IG Contact dose rate (Sv/h) SS316LN IG F82H NIFS HEAT 2 SiC (no impurity) FFHR Li Blanket First wall 1.5MW/m 2 30 years operation Decay time after shutdown (year) Full-remote recycling Full-hands-on recycling Fig. 1 Radioactivity after use in first wall of a fusion commercial reactor for four reference alloys. SS316LN-IG: the reference ITER structural material 8) F82H: reference Reduced Activation Ferritic/Martensitic Steel 9) NIFS-HEAT-2: Reference V-4Cr-4Ti alloy 2,4) SiC/SiC: assumed to be impurity-free. (the reference ITER structural material) 8) will not reach the remote-recycling limit after cooling and thus the recycling is not feasible. F82H (candidate low activation ferritic/martensitic steel) 9) and NIFS-HEAT-2 behave similarly but NIFS-HEAT-2 shows significantly lower radioactivity before 100 year cooling. The activity of F82H and NIFS-HEAT-2 reached the level almost one order higher than the hands-on recycle limit by cooling for 100 and 50 years, respectively. SiC/SiC composite (assumed to be free from impurities because of lack of reference composition) behaves very differently, with much lower activity at <1 year cooling but slightly higher activity at >100 year cooling relative to F82H and NIFS-HEAT-2. This paper highlights the recent progress in vanadium alloys for use in fusion blanket systems with emphases on (1) development of concepts and key technologies for fusion blankets with vanadium alloys, (2) fabrication technology, (3) impurity effects, (4) corrosion, compatibility and hydrogen effects, (5) radiation effects, (6) development of advanced alloys and (7) Magneto-Hydrodynamic (MHD) insulator coating. Remaining critical issues and development strategy for vanadium alloys are also discussed. 2. Design for Fusion Blanket System Using Vanadium Alloys Tritium breeding fusion blankets with vanadium alloys as structural materials and liquid lithium as breeding and cooling materials (self-cooled Li/V blankets) have been designed as advanced concepts for DEMO and commercial fusion reactors. 10,11) One of the attractive features of this blanket system is that the neutron multiplying beryllium is in most cases not necessary for obtaining the required Tritium Breeding Ratio (TBR). This is confirmed recently by neutronics calculation both for Tokamak and Helical reactor systems. 12,13) Without beryllium, the system is released from the issues concerning beryllium such as handling safety and natural resource limit. Moreover, the replacements frequency of the blanket will be reduced once long life structural materials are developed, because the blanket system is free from periodic replacement due to the life time of beryllium.

2 406 T. Muroga (a) Vanadium duct Magnetic Field (b) Lorentz-Force Liquid Metal Flow Insulating Coating Fig. 3 Schematic illustration of Magneto-Hydrodynamic (MHD) pressure drop and the role of insulator coating. (a) Without coating, and (b) with insulator coating. Fig. 2 Illustration of self-cooled Li blanket with V-4Cr-4Ti structural material. Without beryllium, the blanket can be designed with much more simple structures. Figure 2 shows the illustration of self-cooled lithium blanket. The blanket is composed of Li cooling channels, reflectors and a shielding area, which is in contrast to solid breeder blankets composed of a solid breeder zone, a neutron multiplier beryllium zone, cooling (gas or water) channels and tritium recovery gas flow channels, in addition to reflectors and shieldings. The self-cooled Li/V blanket has also advantages of high heat transfer capability due to physical property of vanadium alloys and high heat transfer characteristics of liquid lithium, and low tritium leakage because of high solubility of tritium in liquid lithium. In addition the system has advantages which are common to liquid blanket systems such as Li-Pb and molten salt systems, relative to the solid breeder/water blanket systems, i.e. high temperature operation with potential high energy conversion efficiency, continuous compositional adjustments of breeding materials including lithium isotopic adjustment, impurity trapping and tritium recovery. Major key issues of the self-cooled Li/V blanket are tritium recovery from liquid lithium, MHD pressure drop and development of vanadium alloys. Self-cooled Li blanket using neutron multiplier beryllium was also designed. 14) This concept enables even higher TBR but involves structural complexity and new issues such as Li/ Be compatibility. Requirements to vanadium alloys for the use in Li/V blankets include dimensional stability, high temperature strength and low temperature ductility during irradiation, as well as fundamental requirements for industrial materials such as large scale manufacturing technologies. Figure 3 illustrates the MHD pressure drop when electrically conductive coolants such as liquid lithium is flowing in a magnetic field. Since coating with insulating ceramics on the pipe interior is considered as a promising method to eliminate the MHD pressure drop, the coating technology on vanadium alloys is one of the key technologies. 15) Vanadium alloys could be a candidate structural material for molten salt Flibe (LiF-BeF 2 ) blanket. One of the critical feasibility issues of Flibe/V blanket is high tritium inventory in vanadium alloys because of low and high tritium solubility in Flibe and vanadium alloys, respectively. However, a concept was proposed to dissolve WF 6 or MoF 6 into Flibe for corrosion protection of the wall surfaces by precipitation of W or Mo and reduction of the tritium inventory in vanadium alloys by enhancing reaction from T 2 to TF which is highly soluble in Flibe. 16) With this concept, vanadium alloys are potential candidate materials. Neutronics investigations have shown that Flibe/V blankets may obtain marginal TBR without using neutron multiplying beryllium but have superior neutron shielding capability for the superconductor magnet systems. 12,17) One of the critical issues of this concept is the removal of W or Mo precipitates from flowing Flibe. Table 1 summarizes the blanket concept using vanadium alloys with its advantages and critical issues. 3. Recent Progress in the Development of Vanadium Alloys 3.1 Fabrication technology The examination on microstructural changes during the processing of NIFS-HEAT ingots into various products showed that optimization of size and distribution of Ti-(C, O, Table 1 Breeding blanket concepts using vanadium alloys. Concept V/Li V/Be/Li V/Flibe Breeder and Coolant materials Liquid Li Liquid Li Molten salt Flibe Use of neutron multiplier Be no yes no Advantages Simple structure High TBR Small MHD pressure drop Critical issues MHD coating MHD coating REDOX control T recovery from Li Li/Be compatibility Recovery of W or Mo T recovery from Li Increase in TBR

3 Vanadium Alloys for Fusion Blanket Applications 407 N) precipitates is crucial for good mechanical properties of the V-4Cr-4Ti products. 18) Since clustered structures of the precipitates result in low impact properties, rolling to high reduction ratio is necessary for making thin band structure or homogenized distribution of the precipitates. Plates, sheets, rods and wires were fabricated minimizing the impurity pickup and maintaining grain and precipitate sizes. Thin pipes, including those for pressurized creep tube specimens, were also successfully fabricated maintaining the impurity level, fine grain size and straight band precipitate distribution by maintaining a constant reduction ratio between the intermediate heat treatments. 19) Through the efforts for fabricating the creep tubes including plugging of end caps, the fine-scale EB welding technology was enhanced. 20) Joining of V-4Cr-4Ti by Gus Tungsten Arc (GTA) 21) and laser welding 22) methods was demonstrated. Optimization of the Post Weld Heat Treatment (PWHT) was made by optimizing precipitation in weld metals. Only limited data on irradiation effects on the weld joint are available at present. They showed enhanced defect cluster density at the weld metals, 23) and potential elimination of radiation-induced degradation by applying appropriate conditions of PWHT. 24) A low pressure plasma-spraying method was applied for coating W on V-4Cr-4Ti for the use on the plasma-facing surfaces. Hardening of substrate V-4Cr-4Ti by the coating was shown to be acceptable. 25) Figure 4 is a collection of the products from NIFS-HEAT Impurity effects The effect of interstitial impurities (C,O,N) on mechanical properties is a concern because impurity transportation is thought to occur between vanadium alloys and the environments during operation. Although oxygen is introduced into the alloys in oxidizing environments such as water and helium, oxygen is removed and nitrogen is introduced into the alloys in liquid lithium environments. Since both the precipitates (size and distribution) and solid solution impurities influence strongly the mechanical property, the impurity effects are thought to be a function of the heat treatment conditions. One of the concerns by reducing interstitial impurities in vanadium alloys is loss of strength at high temperature. Since thermal creep properties of NIFS-HEAT-2, containing C, O and N impurities of 100 massppm each, were shown to be similar to those of other V-4Cr-4Ti alloys having higher levels of the impurities, 26) reduction of the impurities to 100 massppm is thought not to induce loss of strength. However, significant increase in creep strain rate was reported recently in V-4Cr-4Ti with oxygen level of 10 massppm. 27) 3.3 Corrosion, compatibility and hydrogen effects Corrosion of vanadium alloys in liquid Li might not be a concern if all inner wall is covered with insulating ceramic coating. However, since an idea to cover the insulator ceramic coating again with a thin vanadium or vanadium alloy layer was presented, for the purpose of preventing liquid lithium from intruding into the cracks in the ceramics coating, 15) the corrosion of vanadium alloys in liquid lithium is attracting attention. It is known that the corrosion of vanadium alloys in liquid lithium is highly dependent on the alloy composition and lithium chemistry. 28,29) The corrosion of vanadium alloys in oxidizing environ- (mm) 26t 6.6t 4.0t 1.9t 1.0t 0.5t 0.25t 2d 8d Plates, Sheets, Wires and Rods Thin pipes Creep tubes W coating by plasma spraying Laser weld joint Fig. 4 Collection of the products from NIFS-HEAT-2 (high purity V-4Cr-4Ti).

4 408 T. Muroga Elongation (%) Total elongation Uniform elongation Total elongation SWIP Heat NIFS-HEAT Hydrogen Concentration (massppm) Fig. 5 Tensile properties as a function of hydrogen level for NIFS-HEAT- 2 (O level of 148 massppm) and SWIP-Heat (O level of 900 massppm). 34) ments is a concern for the performance of the pipe exterior out of the breeding blanket. The data should also be useful for the database of vanadium alloys for potential non-li coolant systems. Addition of Si, Al and Y and increase in Cr level were shown to be effective in suppressing the corrosion in air and water environment, respectively. 30,31) Vanadium alloys are known to lose their ductility when hydrogen is charged. It was reported that the loss of ductility by hydrogen was enhanced by impurity oxygen. 32) Figure 5 compares elongation vs. hydrogen concentration for two types of V-4Cr-4Ti alloys, NIFS-HEAT-2 with O level of 148 massppm and SWIP-heat 33) with O level of 900 massppm. The figure indicates advantage of reducing oxygen level in suppressing susceptibility to hydrogen embrittlement. 34) The effect of hydrogen is also a concern for the use of vanadium alloys for Plasma Facing Component (PFC). Deuterium ion implantation followed by thermal desorption showed that deuterium retention of V-4Cr-4Ti is much higher than and comparable to those of other PFC candidate materials (graphite and tungsten) at 380 K and 773 K, respectively. 35) A hydrogen absorption study showed that the rate of absorption is highly influenced by prior heat treatment inducing Ti surface segregation. The formation of Ti oxide on the surface suppressed significantly the absorption rate. 36) 3.4 Radiation effects Neutron irradiation can significantly influence the performance of vanadium alloys during operation in fusion reactors. Void swelling is known to be small if alloyed with Ti. Among the feasibility issues of radiation effects of vanadium alloys are loss of ductility at lower temperature, embrittlement enhanced by transmutant helium at high temperature, and irradiation creep at intermediate to high temperature. The loss of uniform elongation of vanadium alloys at relatively low temperature (<673 K) is accompanied by dislocation channel microstructure, which implies flow localization during the deformation. 37) Although the mechanism of the flow localization is not well understood, it is inferred that interaction of dislocations with radiationinduced defect clusters, precipitates or complexes of the two species is responsible. The increase in uniform elongation by addition of Al, Si and Y, which getter O, N and C in the matrix, may suggest that reduction of interstitial impurities in solution would enhance the radiation resistance. 38) Helium embrittlement is a critical issue which may determine the upper temperature limit for vanadium alloys. The past experimental evaluation of the helium effects varied from weak to very strong. 1) Since the technique to generate helium and displacement damage simultaneously is limited, recent progress in experimental evaluation of the helium effect is also limited. Clearly a 14 MeV neutron source is essential to evaluate the helium effects in fusion conditions. Dynamic Helium Charging Experiment (DHCE) 39) using fission reactors can extend our understanding on the helium effects before the 14 MeV neutron source becomes available. The irradiation creep data are also very sparse. The data are limited to relatively low temperature and low dose. However, irradiation creep tests are making progress partly because of the progress in fabricating high quality pressurized creep tube specimens with reduced impurity levels. The irradiations of creep tubes are being carried out in HFIR and JOYO. The progress in creep tube fabrication also enhanced thermal creep tests including those in liquid lithium environments. 40) The effects of helium pre-implantation on the creep property of NIFS-HEAT-2 were investigated using helium ion irradiation. 41) The results showed reduced creep deformation by reduced minimum creep rate for the case of helium pre-implantation at 973 K. The contributions of radiationinduced defects, helium clusters and other impurities introduced during the pre-implantation were potential reasons of the increased creep resistance. 3.5 Development of advanced alloys The reference composition of V-4Cr-4Ti was decided as a compromise of ductility at lower temperature and strength at high temperature. 1,2) However, the requirements to the vanadium alloys depend on the blanket design. Generally larger windows for operation temperature of structural materials allow us to design compact blanket systems with higher efficiency. For increasing the plant efficiency, increase in the coolant temperature is essential. For this application, high temperature strength is of greater importance than low temperature ductility. Therefore, efforts have been made to develop advanced vanadium alloys which have larger operation temperature windows, especially with potential use at higher temperature. Increase in Cr level in V-Cr-Ti is known to increase high temperature strength, bartering with loss of ductility at low temperature. Recent detailed survey in V-xCr-4Ti alloys showed that the strength at high temperature increases with small change in the DBTT with the Cr level to 7%. 42) A new effort has been made to examine V-W or V-W-Ti alloys. V-6W-1Ti clearly showed better resistance to hydrogen embrittlement. High temperature strength including creep performance is under investigation. 43)

5 Vanadium Alloys for Fusion Blanket Applications 409 Mechanically alloyed V-Y alloys were fabricated and their irradiation response was examined. Fine grain and oxide dispersion increased high temperature strength and inhibited formation of interstitial loops in the matrix by neutron irradiation, because of the enhanced defect sink. 44) Thus mechanically alloyed vanadium alloys have potentiality to extend both low and high temperature operation limits. 3.6 MHD coating development Years ago, developments of MHD coating were carried out in limited research groups focusing on CaO or AlN as candidate materials. Recent reexamination of those materials showed that they have problems in the stability in liquid lithium at high temperature. 45) Thus efforts in recent years were focused on identifying new candidate materials which withstand lithium corrosion at high temperature and developing coating technology of the new candidates. By the recent efforts, promising candidate ceramics of Er 2 O 3 and Y 2 O 3, which were stable to 1073 K in liquid lithium, were identified. 15) Feasibility of the coating with Er 2 O 3 and Y 2 O 3 on V-4Cr-4Ti was demonstrated by EB- PVD, 15) Arc Source Plasma Deposition 46) and RF sputtering. 47) In addition to the physical deposition methods, in-situ coating with Er 2 O 3 on V-4Cr-4Ti is being developed. 48) In this process, a Er 2 O 3 thin insulating layer is formed on V- 4Cr-4Ti during its exposure to liquid lithium by reaction of pre-charged oxygen in the vanadium alloy and pre-doped Er in Li. The in-situ coating method is a quite attractive technology because it will enable coating on complex surfaces after fabrication of components and have the potential to heal cracks in the coating without disassembling the component. All the above progresses have been made by static immersion tests. The static tests are valuable for identifying potentiality of the candidates but not useful for verifying their function in the blanket condition. For example, erosion rate of the candidate coating ceramics in liquid lithium could be significantly different between in static and flowing conditions. Thus loop experiments are essential for validating the performance of MHD coating in fusion blanket conditions. 4. Remaining Critical Issues As a result of recent significant progress in developing vanadium alloys, critical issues for the future researches have been focused into limited numbers. As to the available data, thermal and irradiation creep, helium effects on high temperature mechanical properties and radiation effects on fracture properties are insufficient. Conclusive evaluation of the irradiation properties is possible only with the use of 14 MeV neutrons, motivating the construction of the 14 MeV neutron source. Maintaining impurity (C, O, N) levels and optimizing precipitate (Ti-CON) size and distribution are crucial for the mechanical properties of the alloy products and weldments. Systematic studies to optimize the microstructure and mechanical properties are necessary for enhancing the performance of vanadium alloys. MHD insulator coating is a critical feasibility issue for Li/ V blanket. Although recent progress in MHD coating is large, further intensive efforts are necessary including tests in flowing lithium conditions. Concept exploration efforts are being made to consider variety of blankets using vanadium alloys, which include potential use of molten salt for the coolants. General common necessary technology would be coating for MHD insulation, corrosion protection or tritium diffusion barrier. Close coupling of the materials development and blanket design activities is essential. 5. Strategy of Vanadium Alloy Development for Fusion Blanket In Japanese fusion materials development strategy, the candidate structural materials are categorized into reference and advanced materials. 49) As the reference materials, Reduced Activation Ferritic/Martensitic Steels (RAFMs) were selected because they have the most matured industrial infrastructure. Development of the reference materials is crucial for realization of DEMO in timely manner. On the other hand, vanadium alloys and SiC/SiC were nominated as the advanced materials, which will contribute to increasing attractiveness of the fusion system in terms of cost of electricity and environmental benignity. It is recognized that the development of the advanced materials must also be enhanced now, due to the long lead time necessary for their development. This strategy is common to that of EU. 49,50) For the qualification of materials up to the full lifetime of DEMO and Power Plant Reactors, irradiation testing with IFMIF (International Fusion Materials Irradiation Test Facility, a 14 MeV neutron source) is recognized to be essential. 51) The Test Blanket Module (TBM) to be installed in ITER is also considered to be an important milestone for technological integration. 52) Figure 6 shows the roadmap of vanadium alloy development. This is arranged from the road map shown before 49) for showing exclusively for vanadium alloys. The materials development is planned to proceed with IFMIF and ITER in a coordinated way. For vanadium alloys, small scale testing will be carried out in initial operation period of IFMIF and ITER-TBM followed by full size testing, in contrast to the full size testing planned for the reference materials (RAFM) from the initial period of IFMIF and ITER-TBM. 6. Summary Vanadium alloys are promising candidate structural materials for breeding blanket of fusion reactors because of their low induced activation characteristics, high temperature strength and high thermal stress factors. The self-cooled liquid lithium blanket with structural materials of vanadium alloys is an attractive concept because of its high heat transfer capability, high temperature operation, simple structure, high tritium breeding capability and low tritium leakage. Among the major issues of the blanket are tritium recovery, MHD insulator coating and development of vanadium alloys. The efforts of developing vanadium alloys have been focused on V-4Cr-4Ti alloys as reference materials. Recent researches have successfully resolved many of the critical issues and enhanced the feasibility of the alloys as fusion

6 410 T. Muroga Calendar year Advanced Powerplant Design (Full blanket) Power Generation Plant Design Construction Operation (Sector blanket) ITER Blanket Module Test V Alloy Development Alloy optimization (composition, processing) Qualification by fission neutrons (creep, fracture, He effect) V/Li Blanket Technology MHD coating (static and dynamic corrosion test) Tritium recovery Blanket module design/r&d (Small unit test module) (Miniaturized specimen test) (Full size test module) (Engineering database) (Design data) (Design data) IFMIF Fig. 6 Irradiation Test, Materials Qualification Roadmap of vanadium alloy development. blanket structural materials. Major remaining issues of vanadium alloys are thermal and irradiation creep, helium effects on high temperature mechanical properties and radiation effects on fracture properties. For conclusive characterization of these properties, the use of IFMIF is essential. In addition, development of advanced vanadium alloys for the purpose of enhancing further high temperature strength, low temperature ductility or radiation resistance is in progress. New promising candidates were identified for MHD insulator coating on vanadium alloys for use in self-cooled lithium blanket. The coating technology is in progress for the new candidate materials. However, further intensive efforts are necessary including tests in flowing lithium conditions for verifying the performance of the coatings in the blanket environment. Vanadium alloys are categorized into advanced materials among the candidate low activation structural materials. The development of vanadium alloys is planned to proceed with IFMIF and ITER-TBM in a coordinated way but in long-term schedule relative to the reference candidate materials of Reduced Activation Ferritic/Martensitic Steels. REFERENCES 1) R. J. Kurtz, K. Abe, V. M. Chernov, D. T. Hoelzer, H. Matsui, T. Muroga and G. R. Odette: J. Nucl. Mater (2004) ) T. Muroga, T. Nagasaka, K. Abe, V. M. Chernov, H. Matsui, D. L. Smith, Z.-Y. Xu and S. J. Zinkle: J. Nucl. Mater (2002) ) W. R. Johnson and J. P. Smith: J. Nucl. Mater (1998) ) T. Muroga, T. Nagasaka, A. Iiyoshi, A. Kawabata, S. Sakurai and M. Sakata: J. Nucl. Mater (2000) ) M. M. Potapenko, V. M. Chernov, V. A. Drobyshev, G. N. Ermolaev, I. V. Golikov, I. A. Gubkin, M. I. Solonin, A. K. Shikov, A. V. Vatulin, G. P. Vedernikov and V. S. Zurabov: Proc. IEA/JUPITER-II Workshop on Critical Issues of Vanadium Alloy Development for Fusion Reactor Applications, Dec 15 16, 2003, NIFS, Japan. 6) T. Nagasaka, T. Muroga, M. Imamura, S. Tomiyama and M. Sakata: Fusion Technol. 39 (2001) ) T. J. Dolan and G. J. Butterworth: Fusion Technol. 26 (1994) ) G. Kalinin, W. Gauster, R. Matera, A. F. Tavassoli, A. Rowcliffe, S. Fabritsiev and H. Kawamura: J. Nucl. Mater (1996) ) A. Hishinuma, A. Kohyama, R. L. Klueh, D. S. Gelles, W. Dietz and K. Ehrlich: J. Nucl. Mater (1998) ) F. Najmabadi and The ARIES Team: Fusion Eng. Design 38 (1997) ) Y. A. Sokolov, et al.: Fusion Eng. Design 41 (1998) ) T. Tanaka, T. Muroga and A. Sagara: Fusion Science and Technology, to be published. 13) T. Muroga and T. Tanaka: Fusion Science and Technology, to be published. 14) RF DEMO Team, Development of DEMO-C fusion reactor, Technical Report of Efremov Institute, St. Petersburg, 1999 (in Russian). 15) B. Pint, P. F. Tortorelli, A. Jankowski, J. Hayes, T. Muroga, A. Suzuki, O. I. Yeliseyeva and V. M. Chernov: J. Nucl. Mater (2004) ) D. K. Sze: Fusion Technol. 10 (1986) ) D. K. Sze, K. A. McKarthy, M. E. Sawan, M. S. Tillack, A. Y. Ying and S. J. Zinkle: Fusion Technol. 39 (2001) ) T. Nagasaka, N. J. Heo, T. Muroga and M. Imamura: Fusion Eng. Design (2002) ) T. Nagasaka, T. Muroga and T. Iikubo: Fusion Sci. Technol. 44 (2003) ) K. Fukumoto, H. Matsui, M. Narui, T. Nagasaka and T. Muroga: J. Nucl. Mater. 335 (2004) ) T. Nagasaka, M. L. Grossbeck, T. Muroga and J. F. King: Fusion

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