22D15
Materials Science of PFM-W -- Current Status, Issues and Prospect -「PFM-Wの材料科学 -現状、課題、そして展望-」
Akihiko Kimura, Hiroaki Kurishita, Akira Hasegawa, Yoshio Ueda,
Koichiro Ezato, Takashi Masuzaki
木村晃彦、栗下裕明、長谷川晃、上田良夫、江里幸一郎、増崎 貴
Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, Japan京都大学エネルギー理工学研究所 〒611-0011 宇治市五ヶ庄
Institute for Materials Science, Tohoku University, 2145-2 Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1313, Japan
東北大学金属材料研究所 〒311-1313 茨城県東茨城郡大洗町成田町2145-2
Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita 565-0871, Japan
大阪大学工学研究科 〒565-0871 吹田市山田丘2-1
Naka Fusion Institute, Japan Atomic Energy Agency, 811-1 Mukoyama, Naka-shi, Ibaraki-ken 311-0193, Japan
日本原子力研究開発機構那珂核融合研究所 〒311-0193 茨城県那珂市向山801-1
National Institute for Fusion Science, National Institutes of Natural Sciences, 322-6, Oroshi, Toki, Gifu 509-5292, Japan

自然科学研究機構核融合科学研究所 〒509-5292 岐阜県土岐市下石町322-6
Tungsten (W) is not a common structural material. The application of W as an armor material of fusion
divertor is a big challenge. Many efforts of W-R&D have been made to overcome a number ofrequirements, among which fracture toughness is a critical issue. Materials science often played an
important role in technology innovation as well as in understanding fundamental aspects. W-materials
science certainly provides most suitable pass way to fabricate a reasonable divertor component with anenough ductility and resistance to irradiation and corrosion. Joining and irradiation performance will be theother issues for practical application.

1. Introduction
Materials are essential for any objects, andTungsten (W) is not common as structural
material mostly because of its brittleness, while
W has high resistance against sputtering and low
tritium retention [1].
Fusion plasma imposes so serious requirements
to armor materials of divertor that there is nomaterial to meet full requirements. W is one of
the candidates, and W R&D for PFM has been
performed with a lot of efforts to overcome
critical issue to improve fracture toughness of W.
W has a body centered cubic (BCC) latticesimilar to Fe, V, Nb, Ta and Mo, which have acharacteristic feature of ductile-brittle transition(DBT) behavior in a temperature region. W isductile at above DBT temperature (DBTT) that is
as high as 400 C [2], above which W is ductile.
If the temperature of not only service but
processing were always above the DBTT of theW, the critical issue of brittleness is solved
irrespective of irradiation.
Generally, irradiation embrittlement is crucial
for fusion structural materials. However, it may
not be true for PFM-W because irradiation
embrittlement usually appears at relatively lower
irradiation
temperatureregionand
thetemperature of the PFM can be kept above theDBTT by engineering design. Since irradiation
embrittlement as well as transmutation helium
embrittlement of W has not been fully understood,
there are still uncertainties about the mechanismof irradiation and helium embrittlement [3,4].
In this presentation, current status, issues andfuture prospect are discussed to show a pass way
to the fabrication of W-armored fusion divertor
based on both materials science and engineering
design technology.
2. Ductility improvement
Ultra-fine grained (UFG) W-(0.25-1.5)%TiC
alloys with recrystallized, equiaxed grain sizes of50-200nm and fine TiC dispersoids at grainboundaries were produced by powder metallurgy
methods utilizing mechanical alloying processing.
TiC was selected because of its high melting point(~3200°C) and a self-adjustment capability of thelattice constant by forming a solid solution with W
and non-stoichiometric TiCx. UFG W-0.5TiC
alloys exhibit negligibly small hardening by
neutron irradiation, no blistering and flaking by
3MeV-He irradiation, and superplastic deformation
by grain boundary sliding above 1400°C [5].
The new idea consists in synthesizing aW-laminate made of several layers of W-foils.
During fabrication, the foil was exposed to a highdegree of deformation and therefore, it is ductile. It
can be bend plastically even at RT. The tensile
properties at RT also show ductile behavior;
however, they are anisotropic and strongly dependon the rolling direction [6].
A novel toughening concept based on W-wirereinforcement and the potential benefit ofcorresponding W-composites were demonstrated.
This toughening method utilizes both, inherent
toughness of W-wires and pseudo-toughness byenergy absorption at the engineered wire/matrix
interfaces [7].
3. Joining technology R&D
Oxide dispersion strengthened (ODS) steels andtungsten (W) are considered to be a desired coupleof promising candidate as structural and plasma
facing materials, respectively, of the first wall anddivertor components in DEMO fusion reactor[8-10]. However, ODS steels and W show asignificant difference in their physical properties,
particularly the mismatch of coefficients of thermal
expansion (CTE), W=4.3 × 10-6 K-1, α-steel = 10.5
× 10-6 K-1. Diffusion bonding is an attractive
technique to obtain W-layered ODS steel
component because of their availability of thesufficient thickness of the component and possible
excellent interfacial strength. Two types of joints of
ODS steel and W, liquid phase diffusion bonding(LPDB) and solid state diffusion bonding (SSDB),
were produced and their joint microstructure andstrength were investigated.
Nano-hardness distribution in the cross section ofboth joint regions, (a) LPDB and (b) SSDB (notethe different hardness scale) are shown in Fig. 1. Inorder to evaluate the shear strength at joint region,
the miniaturized torsion tests were performed atroom temperature and the results are shown in Fig.
2. Both joint regions have little plastic strain.

Fig. 1. Hardness distributions on the crosssectional area of W and ODS steel joint [11].

However, the shear strength of LPDB joint region isas high as 300 MPa and higher than that of SSDB
joint region.
4. Materials science and engineering design
technology
Material behavior of W is not very wellunderstood based on currently available materials
science and knowledge. It can be said, however,that there are several pass ways to reach the fusion
divertor production. One is through the essence ofmaterials science and the other is sophisticatedengineering design technology. Both materials
science and engineering design technology also
include several methods which have been left for
definite consideration. The best selection andcombination of both the science and engineeringcertainly offer the most attractive divertor in near
future.

Acknowledgments
A part of this work is supported by The Japan
Society of Plasma Science and Nuclear Fusion Research
as an activity of one of special committees. Authors
thank Dr. Sanghoon Noh, JAEA, for providing his Ph.D
work results on joining of W and ODS steel.
References[1] Y. Ueda et al, J. Nucl. Mater., 313-316(2003)32.
[2] S. Wurster et al, J. Nucl. Mater., submitted.
[3] A. Hasegawa et al, J. Nucl. Mater., 191-194(1992) 910.
[4] A. Kimura et al, J. Nucl. Mater. 283-287(2000) 827.
[5] H. Kurishita, et al., J. Nucl. Mater. 386-388(2009) 579.
[6] M. Rieth et al, Int. J. Ref. Met. Hard Mater. 28(2010) 679.
[7] J. Du, et al., Comp. Sci. and Tech. 70(2010)1482.
[8] R.J. Kurtz et al., J. Nucl. Mater., 386-388(2009) 411[9] J. Reiser et al., Fusion Eng. Design, 83 (2008)1126[10] A. Kimura et al, J. of Nucl. Mater., in press.
[11] S. Noh, Ph.D thesis, Kyoto University.

Fig. 2 The results of miniaturized torsion tests ofW-ODS steel joints. Three tests were carried out
for each joint [11].