3FTFBSDI

Grain Boundary Segregation in Commercial Nuclear Reactor
Pressure Vessel Steels
Grain boundary (GB) segregation of impurity/soluteatoms in surveillance test specimens of a commercial
western-type pressurized water reactor pressure vessel(RPV) steel has been observed by using state-of-the-art 3D
atom probe (3DAP) field ion microscope (laser-assisted
local electrode-type atom probe: laser-LEAP). The GB
segregation is considered to be an origin of the embrittlementof the RPV steels without hardening. After the low dose
irradiation, P, C and Mo atoms are segregated on GB. Afterthe high dose irradiation, the segregation of Si, Mn and As
atoms, in addition to P, C and Mo, is observed.
The irradiation-induced embrittlement of the nuclear
reactor pressure vessel (RPV) steels is currently a vital issue
for ensuring safe operation of nuclear power plants in theworld since the reactors of the first generation are going to
exceed their initially designed operation lifetimes. Thus, it isof great importance to reveal the nanostructural change,
attributed to the origins of the embrittlement, during long-term
operation of the commercial nuclear reactors [1,2].
Here we show the grain boundary (GB) segregation of
impurity/solute atoms in surveillance test specimens of a RPV
steel with high Cu impurity (0.30 wt.%); pressurized water
reactor (PWR), Doel-2 (in Belgium), by using laser-assisted
local electrode- type 3D atom probe (laser-LEAP). The GB
segregation is a possible candidate of the embrittlement of RPV
steels without hardening. Since the embrittlement by the GB
segregation is considered to be noticeable after high dose
irradiation , the detailed observation of the GB is very important
for aged RPVs.
The key for the successful observations is to prepare fineneedle specimens with GB at few ten nm below the top of theneedles. We successfully made the specimens by combining
focused ion beam (FIB), electropolishing and transmission
electron microscope (TEM). Another key is to use new 3DAP,
laser-LEAP. It enables us to analyze much larger volume
than conventional 3DAP and to reduce the possibility of thefracture of the irradiated needle specimens.
Figure 1 shows 3DAP atom maps of the solute distributions
in Doel-2 RPV surveillance test specimens, around GB, of
low dose ((a) 0.83×1019 n/cm2) and high dose ((b) 5.1×1019
n/cm2), corresponding to the in-service irradiation periods forabout 3 and 20 years, respectively. After the low dose
irradiation, the segregation of P, C and Mo atoms on GB isclearly observed. Precipitation of Cu on the GB is also
observed. After the high dose irradiation, Si, Mn and As
atoms are segregated in addition to P, C and Mo.
We also observed low angle GB in the high dose
specimen. As shown in Fig.2(a), P and C atoms aresegregated and Cu atoms are precipitated on the GB. (Mo isalso segregated, although it is not plotted in the figure.)
However, in this case, the segregation is not uniform on the
GB surface. Figure 2(b) shows the lines of P and C atoms inthe same spacing. These lines correspond to the array of
dislocation lines, and P and C are segregated along thedislocations. The Cu precipitates are also on the dislocationlines [3].

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Advanced Nuclear Energy Materials

(a)

(b)

Fig. 1. 3DAP atom maps of RPV surveillance test specimensof Doel-2 with GB. (a) 0.83×1019 n/cm2, (b) 5.1×1019 n/cm2.

Fig. 2. 3DAP atom maps for low angle GB.
References[1] Y.Nagai, T.Toyama, Y.Nishiyama, M.Suzuki, Z.Tang and
M.Hasegawa, Appl. Phys. Lett. 87, 261920 (2005).
[2] T.Toyama, Y.Nagai, Z.Tang, M.Hasegawa, A.Al Mazouzi,
E.vav Walle, R. Gerard, Acta Materialia 55, 6852 (2007).
[3] T. Toyama, Y.Nagai and M.Hasegawa, MTERE 46, 812(2007) (in Japanese).
Contact toYasuyoshi Nagai (Irradiation Effects in Nuclear and Their
Related Materials Division)e-mail: nagai@imr.tohoku.ac.jp

IMR KINKEN Research Highlights 2008

3FTFBSDI

Effects of Electron and Neutron Irradiations on the Propertiesof Functional Materials
for Fusion Applications:Role of Hydrogen in Radiation Effects on Oxide Ceramics
Effects of hydrogen (proton) in the electrical conductivityof ceramics were studied, with protonic conductive
perovskite-oxides. Large electrical conductivity wasobserved at elevated temperatures under fission reactor
irradiation, which could be explained by the contribution of
protonic conduction. At elevated temperatures, existence ofhigh concentration hydrogen (proton) will increase the ionic
conductivity. Reduction of ions constituting the oxides wasobserved in electron irradiation as well as 14 MeV neutron
irradiation at room temperature, which will result in radiationinduced phase changes. The results imply that role ofhydrogen in radiation effects in oxide ceramics is important,which may evoke technological problems in fusion
machines.
The behavior of electrical conductivity of perovskiteoxides was evaluated, under 1.8–2.0 MeV electrons, 14 MeV
neutrons and fission reactor irradiations. Effects of hydrogen(protons) on the electrical conductivity were specially studied.
All of the perovskite-oxides studied had relatively large baseelectrical conductivity and normal RIC by the excited electrons
was marginal in the irradiation. The electrical conductivity of
SrCe0.95Yb0.05O3-δ increased by 8X10-9 S/m from the baseelectrical conductivity under the 1.8 MeV electron irradiation[1] with a 10 Gy/s dose rate, which is within the range ofscatter of accumulated data of the radiation induced electricalconductivity (RIC) of several ceramic insulators shown in Fig.
1. In the meantime, the base electrical conductivity decreased
from 4.6X10-8 S/m to about 5X10-10 S/m after the 2.85 MGy
irradiation. A similar decrease of the base electricalconductivity was observed in the 14 MeV neutron irradiations
at 293 and 373 K [2]. The observed decrease of the baseelectrical conductivity is thought to be related to a radiationinduced reduction, namely change of a charge state of ionsin the oxides. In the case of SrCe0.95Yb0.05O3-δ, Ce4+, thismay being reduced to Ce3+.
With decrease of the base electrical conductivity from
2X10-8 S/m to about (2-5)X10-10 S/m, a radiation induced(enhanced) electrical conductivity of about 2X10-9 S/mFig. 2. Electrical conductivity of BaCe0.9Y0.1O3-δduring JMTR
startup operation.
surfaced even in 14 MeV neutron irradiation at 293 K [2],after the irradiation up to about 2X108 n/m2. The electronic
excitation dose rate was estimated to be in the range of 10-2
Gy/s and the observed increase of the electrical conductivity
was higher than the accumulated RIC data as shown in Fig.
1. Hydrogen atoms picked up from the atmosphericenvironment may play a role in the observed radiationinduced (enhanced) electrical conductivity.
The electrical conductivities of BaCe0.9Y0.1O3-δ with onehydrogen implanted in its anode zirconium film and anotherwithout hydrogen implantation are shown in Fig. 2 as afunction of the electronic excitation dose rate at the startup of
JMTR [3]. It is clearly shown that the specimen with hydrogen
had larger electrical conductivity than that without hydrogen.
Two other oxides, SrCe0.95Yb0.05O3-δ and CaZr0.9In0.1O3-vδ
showed similar results. The results clearly showed that thehydrogen implantation into an anode increased the electricalconductivity. The observed increase of electrical conductivity,
both in the specimens with hydrogen and without hydrogen,with the increase of the reactor power (thus the increase of
the neutron flux and the electronic excitation dose rate) does
not directly correspond to the increase of the normal RIC. In
normal RIC, the electronic conductivity is dominant and the
RIC has very weak temperature dependence, in the present
case, the electrical conductivity observed during the JMTR
irradiation had a large temperature dependence.
References[1] B. Tsuchiya, A. Moroño, E.R. Hodgson, T. Yamamura, S.
Nagata, K. Toh, T. Shikama, Phys. Stat. Sol. (C) 2, (1), 204(2005).
[2] B. Tsuchiya, S. Nagata, K. Toh, T. Shikama, M. Yamauchi,
T. Nishitani, Fus. Sci. Technol. 47, (4), 891 (2005).
[3] T. Shikama, B. Tsuchiya, E. R. Hodgson, J. Nucl. Mater.
367-370, 995 (2007).

Fig. 1. Radiation induced electrical conductivity (RIC) of
several ceramic insulators.

KINKEN Research Highlights 2008 IMRContact to
Tatsuo Shikama (Nuclear Materials Science Division)e-mail: shikama@imr.tohoku.ac.jp

Advanced Nuclear Energy Materials

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3FTFBSDI

Effects of Impurities on One-dimensional Migration of Interstitial
Clusters in Iron under Electron Irradiation
One-dimensional (1D) migration of small interstitial-type
dislocation loops was studied for iron specimens of different
purities under electron irradiation using a high-voltage
electron microscope. The distance and frequency of 1D
jumps were greatly reduced by impurity atoms. We
proposed a model describing the experimental 1D jump
behavior with a support by molecular dynamics
simulations.
Recent molecular dynamics (MD) simulations of displacement
collision cascades in copper and iron have suggested thatinterstitial clusters are directly formed from cascade damage andthat the clusters escape from the cascade damage zone throughone-dimensional (1D) migration with an activation energy as low
as 0.03-0.05 eV. Migration of single vacancies, on the other hand,is through a three-dimensional (3D) random walk. The differencein the migration mechanism between interstitial atoms (including
their clusters) and vacancies has been shown to affect total defectstructural development under irradiation with high-energy particles.
Therefore, the study of 1D migration is important for materials

Fig. 2 Distribution of 1D jump distances observed in five iron specimens

development for future nuclear applications. In addition, recent

from high (Fe1) to low (Fe5) purities under irradiation with 1250 kVyears have seen a renewal of interest in the nature of self-interstitial

electrons at room temperature.

atoms and their clusters: the 1D migration observed in experiments
and MD simulations raises a question about our understanding of
the configuration and migration of this basic unit of crystal latticemigration of liberated clusters at low activation energy, and (4)defects in both fcc and bcc metals.

another trapping by some impurities.

We investigated the experimental 1D migration of interstitial

Figure 1 schematicallyshows the present model.

clusters in iron specimens of different purities under electron

Based on the proposed model, we consider a certain volume

irradiation using a JEMARM-1250 high-voltage electron microscopearound individual interstitial clusters, and assume that a certain impurity

at Tohoku University. Most 1D migration appeared as discrete

atom in this volume prevents the free migration of an interstitial cluster

jumps (stepwise positional changes) at irregular intervals, anddue to their binding. Let the cross-section of the interaction volume

sometimes involved back and forth motion between certain points.

projected along the direction of 1D migration be σ. We also assume a

The distribution of jump distances extended to over 100 nm in

random distribution of impurity atoms at the concentration Ci (<<1) andhigh-purity specimens; it was less than 30 nm in low-purity

neglect the migration of impurity atoms during the fast 1D migration of

specimens. Jump frequency was almost proportional to electroninterstitial clusters for simplicity. Then a 1D jump of length n, namely,

beam intensity, and was on the same order as the rate of atomic

an interstitial cluster detrapped from an impurity atom and trapped

displacement by electron irradiation.

again by another impurity atom after 1D migration for n atomic distances,

Molecular dynamics simulation suggested the suppression ofis approximated to have a probability of P(n) = (1−σCi )n−1 ・σCi.

1D migration of an interstitial cluster (7i) by an oversized solute Cu

The distribution of 1D jump distance in iron specimens from highatom located in the dilatational strain field of the cluster. We(Fe1) to low (Fe5) purities is plotted in Fig. 2. When we neglect the

proposed that the 1D jump process occurs in the following

region of jump distances less than 10 nm for experimental difficulties

sequence: (1) interstitial clusters in a stationary state being trapped

and also that of long distances for poor statistics, the middle range forby impurity atoms, (2) detrapping from impurity atoms caused byeach specimen is well described by the straight line. The σCi values

displacement of impurity atoms by incident electrons, (3) fast 1D

obtained from the gradient of the line were greater for the low-purity
specimens, and corresponded to the results of chemical impurity

cross-sectionof interaction σ

1250 keV electrons〜0.02 dpa/s

Emax 〜100 eV

fast

analysis of the specimens. The experimental 1D migration behavior
was successfully explained by the simple model.
Reference

ion
igrat
1D m

[1] Y. Satoh, H. Matsui, T. Hamaoka, Phys. Rev., B77, 94135(2008).
impurity atom
c =10-3〜10-5

volume ofinteraction

Contact toYuhki Satoh (Nuclear Materials Engineering Division)e-mail: ysatoh@imr.tohoku.ac.jp

Fig. 1 Schematic illustration of the present model of 1D jumps ofinterstitial clusters observed under HVEM irradiation conditions.

40

Advanced Nuclear Energy Materials

IMR KINKEN Research Highlights 2008

3FTFBSDI

Development of Ultra-fine Grained W-(0.25-0.8)wt%TiC and Its
Superior Resistance to Neutron and 3MeV He-ion Irradiations
Poor ductility and radiation induced embrittlement due
mainly to weak grain boundaries in tungsten have beenmajor concern with its application. This summary document
describes the current status on the developmental work
performed to mitigate the issue [1].
Tungsten (W), a refractory transition metal in group VIA,
is the most promising for use as divertors and structural
materials exposed to high heat loading and irradiation
environments because of its high melting point, good thermal
conductivity, low thermal expansion coefficients, reduced
radio activation, etc.
However, it undergoes serious
embrittlement in several regimes, i.e., low temperatureembrittlement, recrystallization embrittlement and radiation
embrittlement.
For the mitigation of the issue, we have developed ultrafine grained (UFG) W-TiC with average grain sizes of 50 ~200 nm and a relative density of approximately 99% byutilizing mechanical alloying (MA) and hot isostatic pressing(HIP) processes. Such UFG W-TiC consolidates exhibited amaximum fracture strength of approximately 2GPa at room
temperature,however no ductility is assured prior to fracture.
The assurance of room-temperature ductility hence remains
to be accomplished for UFG W-TiC.
A noted finding regarding this respect is that plastic
working after consolidation significantly increases the fracture
strength accompanied with appreciable room-temperatureductility and the ductility increase becomes prominent withdecreasing grain size from 2 to 0.6µm [2]. Therefore, theabove main problem is expected to be overcome by applying
a sufficient degree of plastic working without a significant
change in the UFG structures (grain size: 0.05-0.2µm) topoorly workable W-TiC. UFG materials are known to exhibit
superplasticity. Superplastic deformation can be utilized as aplastic working process for poorly workable W-TiC and thus
it is important to reveal high temperature deformation of UFG
W-TiC in connection with superplasticity.
Regarding radiation resistance of W-TiC, on the other
hand, it has been shown that fine-grained W-0.3TiC (grain
size: 0.9µm) exhibits higher resistance to neutron irradiation
at 563K to 9 x 1023 n/m2 (E > 1MeV) than commerciallyavailable pure W [3]. However there are no reports on theeffects of neutron and He ion irradiations for UFG W-TiC.
The objective of this study is to show the features of
microstructures and mechanical properties at room and hightemperatures in the unirradiated state. Effects of irradiation
with neutrons or 3MeV helium-ions on microstructures and room
temperature mechanical properties are also presented.
Neutron irradiation was performed in a helium atmosphereto a fluence of 2×1024 n/m2 (about 0.15 dpa) at 873K in theJapan Materials Testing Reactor (JMTR). For comparison,
TEM disks of commercially available pure W with a grain sizeof approximately 10µm in the hot-rolled, stress-relieved
condition were also prepared and examined.
He irradiation was conducted for the TEM disk specimens
by a dynamitron type accelerator at Tohoku University with 3
MeV He4+ at 823 K up to a fluence of 1023 He/m2.
The main results and conclusions are as follows [1]:1. The microstructural features of UFG W-(0.25-0.8)%TiC
are equiaxed grains with average diameters of 50 to
190nm and nearly full density of approximately 99%.

KINKEN Research Highlights 2008 IMR

2.

3.

4.

5.

6.

Significant grain refinement occurs by TiC additions anduse of Ar in the MA atmosphere which gives rise to alarge number of nano-size Ar bubbles in the as-HIPed
W-TiC consolidates.
Vickers microhardness (HV) values of UFG W-TiC are ina range of 900 to 1160 and determined by grain size
strengthening.
The three-point bend fracture strength at room temperaturedepends strongly on the TiC addition and MA atmosphere.
The strength takes a maximum of approximately 2GPafor 0.25~0.5%TiC additions and increases in the order of
H2, Ar and N2 as the MA atmosphere, which is mainly
attributable to decrease in detrimental effects of pores
and/or bubbles due to easier removal of H2 than Ar and N2
from the W-TiC consolidates.
W-0.5TiC/H2 exhibits superplastic deformation at 16731973K, with a larger strain rate sensitivity of flow stress,
m, of 0.5~0.6. On the other hand, W-0.5TiC/Ar does notexhibit superplastic deformation, with a smaller m value ofapproximately 0.2 [4]. This suggests that the Ar bubbles
cause an adverse effect on superplastic deformation.
For fast neutron irradiation to a fluence of 2×1024 n/m2 at
873K, W-0.5TiC/H2 and W-0.5TiC/Ar exhibit no radiation
hardening as measured with Vickres microhardness,whereas commercially available pure W shows aconsiderable hardening of ∆HV = 98. Radiation induced
interstitial loops and voids are observed with much lessincrease in void density for UFG W-0.5TiC than that forcommercially available pure W. No significant difference
in radiation hardening between the consolidates
fabricated via the different MA atmospheres is recognizedunder this irradiation condition.
The critical fluence for exfoliation and surface cracking
along grain boundaries to occur by 3MeV He irradiation isabout 2×1022 He/m2 for commercially available W
materials (pure W/R, K-doped W/R and K-doped W/SR.
R: Recrystallized, SR: Stress Relieved) and above
2~3×1023 He/m2 for UFG W-0.3TiC/H2. This indicatesthat surface damage resistance to 3 MeV He irradiations
for UFG W-0.3TiC/H2 is more than ten times as high asthat for the commercially available W materials.

References[1] H. Kurishita, S. Kobayashi, K. Nakai, T. Ogawa, A. Hasegawa,
K. Abe, H. Arakawa, S. Matsuo, T. Takida, K. Takebe, M. Kawaiand N. Yoshida, J. Nucl. Mater. 377, 34 (2008).
[2] Y. Ishijima, H. Kurishita, H. Arakawa, M. Hasegawa, Y.
Hiraoka, T. Takida and K. Takebe, Mater. Trans. 46, 568(2005).
[3] H. Kurishita, H. Arakawa, Y. Amano, S. Kobayashi, K.
Nakai, H. Matsui, Y. Hiraoka, T. Takida and K. Takebe, J.
Nucl. Mater. 367-370, 1453 (2007).
[4] H. Kurishita, S. Matsuo, H. Arakawa, S. Kobayashi, K.
Nakai, T. Takida, K. Takebe and M. Kawai, Mater. Sci. Eng.
A477, 162 (2008).
Contact toHiroaki Kurishita (International Research Center for Nuclear
Materials Science)e-mail: kurishi@imr.tohoku.ac.jp

Advanced Nuclear Energy Materials

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