Plasma and Fusion Research: Overview Articles

Volume 11, 2505090 (2016)

Oxide Dispersion Strengthened Steels for Advanced Blanket
Systems∗)
Akihiko KIMURA, Wentuo HAN, Hwanil JE, Kiyohiro YABUUCHI and Ryuta KASADA
Institute of Advanced Energy, Kyoto University, Uji 611-0011, Japan(Received 13 December 2015 / Accepted 19 April 2016)
Oxide dispersion strengthened (ODS) steels with nano-scaled oxide particles in high density in ferrite/martensite matrix and on grain boundaries of ultra-fine grains have excellent properties as structural material
for fusion blankets, while some issues of fabrication processes are still remaining to be solved. In this overview,material properties of recently developed ODS steels are introduced in comparison to ferritic/martensitic (F/M)
steels: 1) mechanical properties, 2) corrosion behavior and 3) radiation tolerance in ODS steels, which are outstandingly superior to the F/M steels. Coupling use of the ODS and F/M steels is eﬀective to expand the design
margin of fusion blankets. R&D of dissimilar joining of FM/ODS steels is indispensable.
c 2016 The Japan Society of Plasma Science and Nuclear Fusion ResearchKeywords: structural material, strength, corrosion, radiation tolerance, high thermal eﬃciency, long lifetime
DOI: 10.1585/pfr.11.2505090

1. Introduction
Reduced activation ferritic/martensitic (RAFM) steels
have been considered to be a candidate of structural material for fusion blanket because of their maturity in manufacturing products and balanced materials properties [1–5].
The sound properties of ferritic/martensitic (F/M) steels isdue to the complex martensitic structure including dislocations in high density, lath boundaries, packet boundaries
and fine carbides which provide a number of trapping sites
for point defects and their clusters [6–8].
International Thermonuclear Experimental Reactor(ITER) blanket module testing is planned to qualify the
integrity of candidate blanket concepts and structures. Together with the blanket development, material development is being performed on a prime candidate material
that is RAFM steel, and on silicon carbide composite(SiCf/SiC) [9–11], vanadium alloy [12–14] and ODS ferritic steel [15–17] as next stage structural materials for advanced blankets. The RAFM steels were chosen as the
structural material of test blanket modules (TBM) of the
ITER by all the participating parties. The most of the reasons of the selection of the material other than the othercandidates is the maturity of the steel production including
manufacturing technology and infrastructure as well as thebalanced materials performance.
However, the RAFM steels have still limits of operation conditions as structural materials for the advanced
blanket systems with high thermal eﬃciency and long lifetime in DEMO reactors. Because of the demands of thesociety to accelerate the development of fusion energy, aauthor’s e-mail: kimura@iae.kyoto-u.ac.jp, wt-han@iae.kyoto-u.ac.jp,
hi-je@iae.kyoto-u.ac.jp∗) This article is based on the presentation at the 25th International Toki
Conference (ITC25).

target of operation condition of the Japanese solid breeder
water cooling DEMO blanket is set for a conservative design using the RAFM steels.
Oxide dispersion strengthened (ODS) steels have beendeveloped for application to fuel claddings of fast reactors [18–25] where high performance materials propertiesare required to meet the severer conditions of in-core environment such as high doses of neutron irradiation at hightemperatures. The ODS steel consists of a number of nanoscaled fine oxide particles in the matrix of the RAFM steel,and resultantly, the trapping sites for radiation defects increase with the number of the particles in comparison to
the RAFM steel since the oxide particles/matrix (P/M) interfaces also play a role of trapping defects induced by irradiations [26]. The oxide particles also work as obstacles
to dislocation motion providing strain field at the P/M interfaces, which results in the strengthening of the steel atelevated temperatures. As for compatibility issue, high-Cr
ODS steels have much better corrosion resistance in a supercritical pressurized water than the RAFM steels [27].
Although the ODS steels show excellent performanceas structural material, the steels are produced by means of
powder metallurgy processing that gives a limitation of thefabrication of large products. It is considered that the coupling utilization of RAFM and ODS steels is eﬀective toexpand the design margin of blankets where the ODS steels
are placed in severer environment and the RAFM steels areused for large components in mild environment.
In this overview, the ODS steels R&D for both fusion as well as fission applications is introduced to show
their excellent materials performance. The impacts of theachievement of the materials innovations of the ODS steels
can give an insight into the advancement of the current con-2505090-1

c 2016 The Japan Society of Plasma
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Plasma and Fusion Research: Overview Articles

Volume 11, 2505090 (2016)servative blanket design to a certain higher level meetingthe anticipated demands of the society.

2. ODS Steels
The high strength at elevated temperatures of the
ODS steels is due to the fine microstructure that consistsof a high number density of nano-scaled oxide particles(∼5 nm) and very fine grains (∼ 0.5 µm) as well as dislocations in high density.
The chromium (Cr) content in the previous ODS steels
is in the range of 9 ∼ 19 wt.%, while the content of the otheralloying elements (Al, W, Mo, Nb, Ti, Zr etc.) are at a levelof a few percent or less, while some of them are excluded
from alloy elements for fusion application because of radio activation. The most popular oxide particle of the ODS
steels is Y2 O3 with a size smaller than 10 nm of diameter
mostly added with up to 0.5 wt.% at most. Along with yttrium oxides, the alumina [28] and titania [29] were alsoused.
At present there is a great variety of ODS steel with
diﬀerent grade and type, as grouped according to developer countries (Table 1). Also, it can be categorized into
3 groups as 9Cr-martensitic, 12Cr-ferritic/martensitic and14(15)Cr-ferritic ODS steels.

in comparison with the other F/M steels (closed symbols).
It can be found that the ODS steels show much higher tensile stresses at all the test temperatures than the non-ODS
steels such as F82 H. According to the type of the ODS
steels, the tensile properties are ranging over a wide variety of strength level, which is due to the diﬀerence in themorphologies of oxide particles, such as, size distribution
and the number density as well as the grain size and grain
shape among these ODS steels. Until now, many related
works have been done to investigate the factors controlling
tensile properties. From Fig. 1, it can be said that the upper
limit of operation temperature of 15Cr-ODS steels, which
is defined as the tensile stress equal to the Mises stress of
359 MPa, is 800 ◦ C, while that is 550 ◦ C for F82 H.

3.2 Creep properties
Figure 2 shows Larson-Miller diagram of the ODS
steels [36] with a value of 25 as the coeﬃcient parameter to gather creep data measured between 600 and 900 ◦ C.
The data of the F/M steel, F82H, and SUS316L are also

3. Mechanical Properties
Tensile stressFig. 1 F/M steels (•  ), 9Cr-ODS steels ( ), 15Cr-ODS
steels (  ), RAFM steel (F82H: ) [32–40].
High strength at elevated temperatures of the ODS
steels was obtained through refinement of microstructures
including a high number density of small Y2 O3 and TiO2
particles dispersed in a complex ferrite/martensite matrix[30, 31]. The ODS steels for structural components in fusion reactor would allow an increase in the operation temperature to approximately 650 ◦ C or more [32]. Therefore,
the ODS steels are widely considered to be promising candidates for structural materials of fusion reactors as well as
the cladding of the next generation fast reactors.
In order to meet the diﬀerent requirements in the nuclear systems, various ODS steels have been fabricated inJapan, Europe and the United States. Figure 1 shows thetensile results of some of the ODS steels (open symbols)

3.1

Table 1 ODS steels developed in each country.

Fig. 2 Larson-Miller diagram of the ODS steels [36].

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plotted in the figure for comparison. Although the data of
the ODS steels are widely distributed in a band because
the data was obtained from various sorts of ODS steels,
it is clear that the creep stress of the ODS steels is muchhigher than those of F82H. Figure 2 indicates that the creep
strength of 200 MPa is available for ODS steels at 780 ◦ Cwith 10,000 h of rapture time, while the test temperatureand rapture time of F82H is 550 ◦ C and 5000 h, respectively.
The extrusion process leads to extremely elongatedfine grains parallel to the extruding direction, which results in the anisotropy in the strength. This is particularly detrimental in the hoop direction of the claddings
as a primary stress mode in fission gas-pressurized fuelpins [39]. In the case of ODS martensitic steels, the ferrite to austenite transformation caused by austenitization
heat treatment makes the elongated grains be adequately
equiaxed. However, in the case of ODS ferritic steels,the matrix has no this kind of transformation. Therefore,
the recrystallization treatment is necessary to reduce theanisotropy and to relieve its internal stress. Thus, the creation of equiaxed grains has been considered to be critical
for the development of ODS ferritic steel claddings. The
recrystallization processing has been considered to be aneﬀective way to overcome the anisotropy. Related report[39] showed that after a suitable recrystallization, the creep
rupture strength of 12Cr-ODS ferritic steels in the hoop direction was significantly improved, resulting from breaking of the elongated grains and generating more equiaxed
grains by means of recrystallization processes.
Sakasegawa et al. investigated the feature of the creep
rupture of an ODS steel which raptured along pre-powder
boundaries where so called void coalescence happened toform large voids that played a role as nucleation site ofductile rupture. The important is that the ductile rupture
occurs in steady creep region without appearance of thethird stage of tertiary creep [41], although the rapture time
is much longer than the conventional ferritic steels.

3.3

Fe-9Cr-2W-0.4Ti-0.2V-0.05C-0.3Y2 O3 , respectively [44].
In their previous research on fracture mechanism using
the NFA-14YWT, it was found that the low-energy grainboundary decohesion in fracture process at a high temperature (>200 ◦ C) resulted in low fracture toughness [45,46].
Lately, eﬀorts have been devoted to explore an integrated
process to enhance grain boundary strength. Isothermal
annealing (IA) and controlled rolling (CR) treatments in
two-phase region were used to enhance diﬀusion acrossthe interfaces and boundaries. The PM2 alloy after CR
treatments showed high fracture toughness (KJQ ) at repre√
sented temperatures: 240 - 280 MPa m at room tempera√ture and 160 - 220 MPa m at 500 ◦ C. This indicates that√the target fracture toughness of over 100 MPa m has been
well achieved in the temperature range of nuclear application. It was also confirmed that the CR treatment on
9YWTV-PM2 resulted in its high fracture toughness similar to or even higher than those of the conventional F/Msteels such as HT9 and NF616.

4. Corrosion in Supercritical Pressurized Water
Corrosion is another critical issue for blanket structural material. Corrosion tests were carried out for several
materials in a supercritical pressurized water (SCPW) condition to compare the corrosion behavior. The tests werecarried out at 500 ◦ C for 600 h at a pressure of 25 MPa in
SCPW dissolved with 8 ppm of oxygen, and the weightgain was measured. The results were shown in Fig. 3, indicating that the weight gain considerably depended on Cr
content. As expected, F82H that contains 8 wt.% Cr shows
the maximum weight gain among the materials. The 9CrODS steel without addition of Al also shows a large weightgain similar to F82H, while the 15 wt.% Cr ODS steel
shows the smallest value in comparison to the other candidate alloys, such as nickel or iron based stainless steels

Fatigue

The research on low cycle fatigue properties of ODS
steels is very limited. Ukai et al. [42] investigated the low
cycle fatigue properties of 9Cr ODS martensitic and 12Cr
ODS ferritic steel at high temperatures. They summarizedthat the 9Cr-ODS and 12Cr-ODS steels have similar fatigue properties, and the ODS steels show superior fatigue
life under the high cycle conditions compared to the conventional ferritic steels such as Mod. 9Cr–1Mo steel. They
also confirmed a high strength after fatigue tests of ODS
steels, that is, cyclic stability, i.e. no cyclic softening [43].

3.4

Fracture toughness

Byun et al. produced two ODS steels (NFAs) designated as 9YWTV-PM1 and 9YWT-PM2, with nominal
components: Fe-9Cr-2W-0.4Ti-0.2V-0.12C-0.3Y2 O3 andFig. 3 Weight gains of each material after corrosion test in
SCPW [15].

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Fig. 4 The dependence of corrosion layer thickness on theamount of Al addition and test temperature. The values
are estimated from the experimental data for 1000 h.

with Cr contents higher than 15 wt.%. Figure 3 indicatesthat the corrosion rate of the 15Cr-ODS steel is two ordersof magnitude smaller than F82H. Considering the application to water cooled advanced blanket, the elevation ofwater temperature above the critical point is required toincrease the thermal eﬃciency. However, F82H is not adequate for the application to SCPW cooled blanket withkeeping long lifetime.
As for the eﬀects of the other alloying elements thanCr, it was also shown that an addition of Al is eﬀective to
reduce the weight gain of the ODS steels and the eﬀect becomes significant with decreasing Cr content from 19 to
15 wt.% [47, 48]. The estimated corrosion layer thickness
as a function of temperature and Al contents is shown in
Fig. 4 for ODS ferritic steels with diﬀerent amount of Al
addition after exposure to SCPW for 1 year. It was foundthat the corrosion layer thickness of the ODS steel significantly depended on the temperature. At 400 ◦ C, the Al
addition to ODS steel increases slightly their corrosion resistance. However, the contribution of Al to the corrosion
resistance is increased at higher temperatures. Especiallyat 600 ◦ C, the eﬀect of Al is remarkable. After exposure
to SCPW at 600 ◦ C for 1 year, the oxide thickness of Aladded ODS steel was about 5 µm, indicating that the addition of 3.5 wt.%Al to 15Cr-ODS steel improves its corrosion resistance at higher temperatures, although Al is
not adequate for fusion application from the view of reduced activation. The previous research on the chemical
analysis and XRD measurements of the corrosion layer revealed that the layer of Al-free 15Cr-ODS steels consistsof three parts, Fe2 O3 outer layer, Fe3 O4 middle one and(Fe, Cr)3 O4 inner one, after the test for 300 h at 510 ◦ C in
SCPW dissolved with 8 ppm oxygen [49]. In the Al-added
ODS steels, a thin alumina layer was formed at the mostinner region of the multi-corrosion layers [47–49].

Fig. 5 The stress-strain curves of 9Cr steels with and without
oxide dispersion tested at room temperature after irradiation at 300 ◦ C to 2.8 dpa. (solid:irr., dotted:unirr.)

5. Radiation Tolerance
5.1 Tensile properties
Nano-sized oxide particles dispersion resulted in aunique behavior under irradiation. Generally, irradiationinduces hardening and embrittlement in metallic materials.
Figure 5 shows the stress-strain curves of 9Cr F/M steel
and 9Cr-ODS steel before and after neutron irradiation at
300 ◦ C up to 2.8 dpa in the HFIR in Oak Ridge National
Laboratory. The 9Cr F/M steel shows a large irradiation
hardening of about 200 MPa and the hardening is accompanied by a drastic reduction of tensile elongation showing
no work hardening or homogeneous deformation. The lossof elongation is interpreted in terms of the irradiation enhancement of localized deformation. The localized deformation is well known as channel deformation in neutron irradiated metals, where once the dislocations move over theirradiation-induced defects, the following moving dislocations can move easily because the defects are absorbed by
the leading dislocations. Since localized plastic deformation accelerates necking or ductile fracture, the total elongation is reduced. In the case of the ODS steel, however,
irradiation-induced hardening is not accompanied by the
reduction of total elongation. This could be due to the suppression of localized deformation caused by the activationof many dislocation sources by oxide particles. Since thefine dispersion of oxide particles play a role as obstacles todislocation motion, the leading dislocation can be trappedat the particles and a back stress is applied to suppress thecontinuous dislocation source activation. Therefore, in the
ODS steel, a number of dislocation sources are activatedin the whole tensile specimen, and homogeneous deformation occurs.

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Fig. 6 Void swelling in F/M steels and MA957 [50].

5.2

Fig. 7 Charpy impact absorbed energy vs. temperature curves[25].

Swelling

Void swelling behavior up to several hundreds dpa
was investigated by Toroczko et al. with use of a van deGraf electrostatic accelerator of heavy ions with an external injector of 1.8 MeV Cr3+ ions [50]. Three diﬀerent
steels were irradiated: HT-9 and MA957 at 450 ◦ C, and480 ◦ C for EP-450 (Fig. 6). The EP-450, which is a ferritic steel, shows a linear increase in the swelling beyond
150 dpa at the swelling rate of 0.2%/dpa. The HT-9 tempered martensitic steel shows good resistance to swellingup to 400 dpa, then shows a rapid increase in swelling at
the similar rate with EP-450. The swelling of MA957,
however, is still small even after the irradiation up to 500
dpa, which is considered to be due to large trapping capacity of defects in the microstructures of the ODS steel,such as fine oxide particles, grain boundaries and dislocations [50].

5.3

Helium embrittlement

Transmutation helium eﬀects on fracture behavior willbe critical for integrity of fusion structural materials. Implantation helium eﬀect on the impact fracture behavior was investigated for F82 H, 9Cr-ODS and 14Cr-ODS
steels. Results of the impact tests with using sub-sized
specimens (1.5 × 1.5 × 20 mm3 ) of these steels are shownin Fig. 7 [25]. Materials were implanted with helium ionsof 40 MeV with a penetration depth of 0.4 mm at the Vnotch region. The Charpy impact energy vs. temperature curves of F82H before and after He implantation revealed that the DBTT was shifted by 70 ◦ C by 1000 appm
He implantation. The Vickers hardness test results indicate a slight hardening (ΔHv = 40) in the He-implantedarea of these specimens. A SEM micrograph of the Heimplanted specimens revealed that intergranular fracturesurface was observed only from the bottom of the notch
to a depth of about 400 µm which corresponded to the Heimplanted region. No grain boundary fracture was observed in the range beyond the He-implantation. Not somany researches had dealt with the microstructural characterization of He-implanted specimens, except for thatby Takahashi et al. They examined the microstructureof F82H after 1000 appm He implantation at 550 ◦ C, andshowed that large He bubbles were formed both in thegrains and at grain boundaries [51]. Thus, intergranularfracture was presumably initiated by crack formation at He
bubbles on grain boundaries.
The fracture mode in both the He-implanted and unimplanted regions of the 9Cr-ODS steel was cleavage, andno intergranular fracture was observed. The hardness ofthe He-implanted area appeared to be decreased.
The DBTT of both the as-received and He-implanted
14Cr-ODS steel was −100 ◦ C. No fracture mode change
due to He implantation was observed. The DBTT of 14CrODS steel did not change upon the He implantation but
a slight decrease of hardness (ΔHv = −15) was observedin the He-implanted area. From the view point of microstructure, the 9Cr-ODS steel has martensitic structure
and the 14Cr-ODS steel has ferritic structure. Both specimens have fine and dense oxide particles [27, 52]. Yutani
et al., showed that a number of small He bubbles were observed at the fine oxide particles after ion irradiation [53].
These results revealed that the fine and dense oxide particles in the ODS steels had enough trapping capacity of He
up to 1000 appm at around 600 ◦ C.

6. Coupling RAFM and ODS Steels
Technological advancement of structure materials is
indispensable for the realization of the high eﬃciency blanket beyond the DEMO reactor. The maximum acceptable
operation temperature for the RAFM steel has been predicted as 550 ◦ C based on the creep property. Using finite element method, the stress distribution in the first wallof water cooled TBM was evaluated by two-dimensional
thermo-mechanical analysis [54], which indicated that thehighest TRESCA stress of 359 MPa appeared at the same

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place as the highest temperature. This stress value was
evaluated to satisfy the 3Sm value for F82H. According
to this TRESCA criteria, the highest temperature for the
ODS steel can be evaluated to be 780 ◦ C from Figs. 1 and2. Furthermore, the corrosion behavior should be taken
into account in consideration of the lifetime of blankets. At
first, it should be recognized that the weight gain of highCr steel in SCPW remarkably depends on Cr content. The
corrosion rate of the ODS steel is almost similar to that of
RAFM steel when Cr content is 9 wt.% [55–57]. According to the previous research [58], the corrosion layer thickness could be estimated to be 600 and 5 µm for RAFM steel
and 15Cr-ODS steel, respectively, at 500 ◦ C for 1 year.
As for helium embrittlement, it was reported that there
was a critical amount of helium content at grain boundaries to induce grain boundary embrittlement [59]. In
F/M steels, no helium eﬀects were observed up to 600
appm of helium-implantation. However, after implantation
to 1000 appm of He, F82H showed a definite intergranular cracking, indicating that 1000 appm He is an enoughamount to cause intergranular cracking of F82H, while the
ODS steel never showed susceptibility to the embrittlement
up to 1000 appmHe. As for the eﬀect of irradiation on
the tensile elongation, the RAFM steel showed a significant loss of total elongation but the ODS steel did show noloss of elongation. The above mentioned is summarized inTable 2.
Thus, ODS steels possess rather high materials performance than RAFM steels. However, it is diﬃcult tofabricate massive products of ODS steel because of thelimitation of the mechanical alloying process. It can besaid that both F/M steels and ODS steels are used as a coupling component where ODS steels are used to face severer
environment and RAFM steels are for production of huge

Table 2 Comparison of materials performance between RAFM
and ODS steels.

structural container.
It is recognized that early realization of the nuclear
fusion energy is essential for acceleration of fusion reactor
development. However, thermal eﬃciency of power plantis also important for the reduction of future energy source
consumption. For the progress in fusion science and technology, aiming high level of target may push people forward to realization of fusion reactors.

7. Dissimilar Joint (FSW)
Dissimilar joining of ODS ferritic steel and RAFM
steel is necessary to expand the flexible design margin andget benefits from each material in a functional way. Numerous researches have shown that conventional melting
welding techniques are inadequate for ODS alloys, because the melting process results in the agglomeration of
nano-sized oxide particles, which causes degradation of
performance of ODS alloys [60–62]. In addition, melting
weld process of F82H, a martensitic steel, is accompaniedby the phase transformation that causes significant changes
in material performance [63–65]. The solid state joining
technique, friction stir welding (FSW), has been shown tobe an appropriate method for welding both ODS steels andF82H steels [66, 67].
In FSW process, the stir zone (SZ) of the joint suffers drastic thermomechanical eﬀects imposed by the stirring tool, and undergoes complicated microstructural evolutions. Chung [67] reported that defect-free F82H joints
could be successfully fabricated by FSW, without formation of neither the δ ferrite nor inter-metallic compounds.
However, complicated phase transformations and the grain
structure changes were found in SZ. Several researches[64–66, 68] on microstructural evolution during FSW process of ODS ferritic steels reached an agreement that therecrystallization occurred in the ODS SZ.
Figure 8 presents grain morphologies of BM and SZ in
the ODS part of the dissimilar joint. Comparing the elongated and preferentially orientated grains in the ODS BM,
the SZ grains exhibit an equiaxed shape and random orientations. As the welding temperature is about 980 ◦ C during

Fig. 8 Grain characteristics of ODS side of the dissimilar joint:(a) IPF of BM, (b) IPF of SZ [72].

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Fig. 9 Grain characteristics ofF82H side of the dissimilar joint:(a) IPF of BM, (b) IPF of SZ [72].

the FSW process, the 15Cr-ODS steel stays in a single αphase, and no phase transformation could occur. The remarkable changes in the grain morphology in ODS SZ arestrong evidence that the SZ grains have experienced recrystallization. However, the 15Cr-ODS ferritic steel usually needs heat treatment at about 1300 ◦ C for more than
1 h for recrystallization [69]. Contrarily, the recrystallization during the FSW process happens quite faster even atmuch lower temperature. It indicates that the recrystallization in FSW should be achieved by a specific mechanism.
Fonda et al., [70, 71] and Han et al. [65, 69, 72] investigated the recrystallization behavior during FSW of ODS
steel. The common findings are that recrystallized grainsin the SZ are formed by the continuous dynamic recrystallization, in which subgrains are recrystallized not by the
nucleation/growth mechanism but by plastic flow mechanism.
Significant changes in grain morphologies of BM andSZ can also be found on F82H side of the joint, as shownin Fig. 9. However, the mechanism of microstructural evolutions in the F82H part is quite diﬀerent from that in the
ODS part. The evolution mechanism in F82H SZ is identified as phase transformations, since the welding temperature exceeds the AC1 (840 ◦ C) of F82H steel. During theheating process of FSW, the martensite grains in F82H SZ
can transform to the austenite grains. After the stir tool
was removed from the welded area, γ–α transformation
happened in the sequential cooling process. As the coolingrate is quite high, during the γ–α transformation, martensitic grains with fine grain size and high hardness value(about 475 Hv in F82H SZ) can be generated, while F82H
BM comprises much coarser grains and processes a lower
hardness value (235 Hv).

8. Summary
RAFM steels have been considered as candidates forstructure material of ITER-TBM. On the other hand, theirmaterials performances are insuﬃcient in considering theapplication to advanced blankets bearing high eﬃciency
and long lifetime for commercial reactors. It is thoughtthat the application of ODS steels is eﬀective in advancing blanket technology, particularly in the case that highthermal eﬃciency and long lifetime are demanded. ODS
steels surpass RAFM steels in several materials performance, and the ODS steels raise the operation temperature of the blanket in comparison to the RAFM steels. On
the other hand, a diﬃculty is the limit in the productivityof ODS steels by means of powder metallurgy processing,and the ODS steels might be unfavorable in the manufacturing huge components where a large quantity of materials
are necessary.
Therefore, in consideration of merits and demerits ofboth the RAFM steel and ODS steel, it is vital to use bothsteels according to a way of thinking the right material in
the right place. The supporting structure body of a blanket needs a large quantity of steel, and F82H meets therequirement. The design of a blanket using both RAFM
steel and ODS steel may impose high thermal eﬃciency
with the extension of lifetime by using rather small amountof ODS steel. In the coupling use of RAFM steel and ODS
steel, joining technology development becomes indispensable. The research on the eﬀect of irradiation on the properties of the joints is another issue.

Acknowledgements
The authors appreciate many thanks to collaborators
including staﬀs of Institute of Advanced Energy, Kyoto
University, students of Graduate School of Energy Science,
Kyoto University and research group members on SuperODS steels R&D.
Present study includes the result of “R&D of corrosion
resistant super ODS steel for highly eﬃcient nuclear systems” entrusted to Kyoto University by the Ministry of Education, Culture, Sports, Science and Technology of Japan(MEXT) and “R&D of high-Cr ODS steels for practicalapplication to LWR” by the Ministry of Economy, Tradeand Industry (METI).
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