ODS steel as structral material for high temperature nuclear reactors
Oxide dispersed strengthened (ODS) ferritic-martensitic steels are investigated as possible structural material for the future generation of High Temperature Gas Cooled Nuclear Reactors. The Ni based austenitic ODS superalloys are not considered, because of the Ni presence, which is unfavorable unde...
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| Опубліковано в: : | Вопросы атомной науки и техники |
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| Дата: | 2005 |
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Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2005
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| Цитувати: | ODS steel as structral material for high temperature nuclear reactors / M.A. Pouchon, M. Dobeli, R. Schelldorfer, J. Chen, W. Hoffelner, C. Degueldre // Вопросы атомной науки и техники. — 2005. — № 3. — С. 122-127. — Бібліогр.: 9 назв. — англ. |
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Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1860259976837070848 |
|---|---|
| author | Pouchon, M.A. Dobeli, M. Schelldorfer, R. Chen, J. Hoffelner, W. Degueldre, C. |
| author_facet | Pouchon, M.A. Dobeli, M. Schelldorfer, R. Chen, J. Hoffelner, W. Degueldre, C. |
| citation_txt | ODS steel as structral material for high temperature nuclear reactors / M.A. Pouchon, M. Dobeli, R. Schelldorfer, J. Chen, W. Hoffelner, C. Degueldre // Вопросы атомной науки и техники. — 2005. — № 3. — С. 122-127. — Бібліогр.: 9 назв. — англ. |
| collection | DSpace DC |
| container_title | Вопросы атомной науки и техники |
| description | Oxide dispersed strengthened (ODS) ferritic-martensitic steels are investigated as possible structural material for the future generation of High Temperature Gas Cooled Nuclear Reactors. The Ni based austenitic ODS superalloys are not considered, because of the Ni presence, which is unfavorable under neutron irradiation. ODS-steels are considered to replace other high temperature materials for tubing or structural parts. Interestingly, ODS is also considered as material being used in future fusion applications. The oxide particles serve for interfacial pinning of moving dislocations. Therefore the creep resistance is improved. In case of the usage of these materials in reactor, the behavior under irradiation must be further clarified. In this paper the effects induced by He implantation are investigated. The induced swelling is measured and the mechanical behavior of the irradiated surface is investigated. These first tests are performed at room temperature, where a clear swelling and hardening could be observed.
Окисно дисперговані зміцнені (ОДЗ) феритно-мартенситні сталі досліджуються як можливі конструкційні матеріали для нового покоління високотемпературних ядерних реакторів з газовим охолодженням. Аустенітні ОДЗ-сплави на основі никелю не вивчаються завдяки никелю, присутність якого під дією опромінення небажана. ОДЗ-сталі разглядаються як можливі кандидати на заміну інших високотемпературних матеріалів для вигoтовлення трубопроводів або інших конструкційних вузлів. Цікаво, що ОДЗ-матеріали розглядаються також з точки зору їх можливого використання для майбутнього застосування в термоядерних пристроях. Окисні частки служать як міжфазні пастки для закріплення дислокацій, що рухаються. У разі використання ціх материалів в реакторі їх поведінка під опроміненням повинна вивчатись більш ретельно. В роботі досліджуються ефекти, зумовлені проникненням Не. Вимірюється зумовлене розпухання та механічні характеристики опроміненої поверхні. Ці перші дослідження були виконані при кімнатній температурі, коли можно чітко спостерігати розпухання та зміцнення.
Окисно диспергированные упрочненные (ОДУ) ферритно-мартенситные стали изучались как возможные конструкционные материалы для нового поколения высокотемпературных ядерных реакторов с газовым охлаждением. Аустенитные ОДУ-суперсплавы на базе никеля не рассматривались из-за присутствия никеля, который нежелателен при нейтронном облучении. ОДУ-стали рассматривались как возможные заменители других высокотемпературных материалов для изготовления трубопроводов или других композиционных узлов. Интересно, что ОДУ рассматривается так же, как возможный кандидат для использования в термоядерных устройствах. Окисные частицы служат как межфазные ловушки для закрепления движущихся дислокаций. Поэтому сопротивление ползучести увеличивается. В случае использования этих материалов в реакторе их поведение под облучением должно изучаться более тщательно. В предлагаемой работе исследуются эффекты, обусловленные внедрением Не. Измеряется обусловленное распухание и механические характеристики облученной поверхности. Эти первые испытания были выполнены при комнатной температуре, когда можно явно наблюдать распухание и упрочнение.
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| first_indexed | 2025-12-07T18:53:57Z |
| format | Article |
| fulltext |
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ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2005. № 3.
Серия: Физика радиационных повреждений и радиационное материаловедение (86), с. 122-127. 122
РАЗДЕЛ ТРЕТИЙ
МАТЕРИАЛЫ ПЕРСПЕКТИВНЫХ ЯДЕРНЫХ РЕАКТОРОВ
УДК 669.15-194
ODS STEEL AS STRUCTURAL MATERIAL FOR
HIGH TEMPERATURE NUCLEAR REACTORS
M.A. Pouchon, M. Döbeli, R. Schelldorfer, J. Chen, W. Hoffelner, C. Degueldre
Paul Scherrer Institute, 5232 Villigen PSI, Switzerland
Oxide dispersed strengthened (ODS) ferritic-martensitic steels are investigated as possible structural material for
the future generation of High Temperature Gas Cooled Nuclear Reactors. The Ni based austenitic ODS superalloys
are not considered, because of the Ni presence, which is unfavorable under neutron irradiation. ODS-steels are
considered to replace other high temperature materials for tubing or structural parts. Interestingly, ODS is also
considered as material being used in future fusion applications. The oxide particles serve for interfacial pinning of
moving dislocations. Therefore the creep resistance is improved. In case of the usage of these materials in reactor,
the behavior under irradiation must be further clarified. In this paper the effects induced by He implantation are
investigated. The induced swelling is measured and the mechanical behavior of the irradiated surface is investigated.
These first tests are performed at room temperature, where a clear swelling and hardening could be observed.
1. INTRODUCTION
1.1. THE HIGH TEMPERATURE PROJECT
The temperature of current light water reactors
(LWR) does not exceed 350 ºC. Energy conversion is
done by steam generation and steam turbines. Materials
for these applications exist. Research focuses mainly on
safety and life assessment under fuel high burnup
conditions and reactor ageing. In contrast to the LWRs,
concepts for future (generation IV) nuclear fission
plants focus on much higher temperatures necessary to
increase the thermal cycle efficiency and to include
process heat applications (e.g. hydrogen production).
Fast neutron spectra are also considered. Concepts for
nuclear plants are studied world wide within the
framework of the Generation IV international forum
(GIF), of which Switzerland is also full member.
1.2. PM2000 AS POTENTIAL MATERIAL
The role of ODS-materials for fusion applications
was reviewed by Ukai and Fujiwara [2] and also by
Hoelzer [3]. Both conclude that ODS materials are
interesting for future nuclear applications. It is worth
mentioning that ODS materials have not been
considered for HTRs so far. They have, however,
proven their capabilities with respect to high
temperature gas cooled reactor applications in the
meanwhile and they should be considered as serious
candidates for future nuclear high temperature
applications according to the opinion of the members of
the GIF VHTR and GFR Steering Committees and
researchers from FZ Jülich [4]and Petten [5].
PM2000 is an oxide dispersion strengthened (ODS)
alloy of the composition 20 wt% Cr, 5.5 wt% Al,
0.5 wt% Ti, 0.5 wt% Y2O3, balance Fe, manufactured
by mechanically alloying in a high energy mill to
produce a solid solution which contains a uniform
dispersion of yttria. The powder is consolidated using
hot isostatic pressing followed by a hot and cold rolling
procedure. Then a thermal treatment finalizes the
production [6,7].The alloy is supplied by Plansee
GmbH.
2. EXPERIMENTAL
2.1. SAMPLE PREPARATION
PM2000 ODS samples are cut and polished in the
following way. Several samples of the dimension 6 x 6
x 1 mm are cut in a long transverse direction. One
surface is then ground with SiC papers down to a P-
Grading of 4000. A polishing with a 6 and 3 μm
diamond suspension is performed, and the polishing is
then finalized with OPS for 2’.
2.2. IRRADIATION
The samples described in Section 2.1. are irradiated
with a 4He++ beam at the Swiss Federal Institute of
Technology in Zürich, using a Tandem Accelerator. The
main difference from a thermal neutron irradiation is the
high He to dpa ratio, however, for this preliminary study
this fact was accepted and the data gained from this
experiment also represents an extreme situation, which
might be faced in a fast spectrum. For further
information about the simulation of n irradiation by
charged particles, see [7]. An even distribution is
desirable; therefore the irradiation is performed under 4
different incident angles (ranging from 0° to 66°) and an
energy of 1.5 MeV. The damage as a function of depth
is then already quite even, being about 0.7 to 1.3 dpa
from 1 to 2.5 µm. Figure 1,a) shows a TRIM simulation
of the induced damage profile. TRIM tends to
underestimate the width of the implantation-profile and
the peaks appear sharper than in reality.
123
The irradiated fluences are 1.4·1016, 2.8·1016,
5.6·1016 and 1.12·1017 cm-2. The irradiation is performed
through a Mesh 400 TEM grid covering the sample (see
Fig. 1,b), the bar periodicity is 63.5 μm with a bar width
of 30 μm.
Therefore a bar pattern occurs on the sample. This
bar pattern is used to investigate the sample by AFM
and derive the swelling behavior. The influence of
irradiation is studied consequently with a nanoindenter.
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
0.0 0.5 1.0 1.5 2.0 2.5 3.0
depth [µm]
di
pl
. d
am
ag
e
[d
pa
]
0˚
37˚ 53˚ 66˚
a b
Fig. 1. a – TIRM simulation of the He implantation using 4 angles (0°, 37°, 53° and 66°) with a total He fluence of
5.6·1016 cm-2 and an energy of 1.5 MeV. The single fluences for each angle (α) are: cos(α)·2.0·1016 cm-2; b – ODS
sample with Ni parallel bars (400 lines/inch)
2.3. AFM INVESTIGATIONS
The samples are investigated by atomic force
microscopy, using an apparatus from Digital
Instruments “Dimension 3100”. The maximum scan
region with the used AFM head is 124 x 124 μm with a
maximum scanning depth of 6.6 μm. The depth
resolution is mainly limited by noise. The scanning is
performed in tapping mode, in order to reach higher
precision and to avoid friction related artifacts of the
irradiated surface. The sample preparation for the
PM2000 is rather difficult, because the dispersed yttria
nano-particles are pulled out during ceramographic
preparation and scratch the surface. Therefore the AFM
sometimes shows some artifacts on the scan. These
peaks are removed with a computer program, and the
resulting AFM images are well analyzable. In all cases
the swelling is large enough to be well distinguishable
from any noise and the remaining artifacts on the scan.
An example of such an analysis is shown in Fig. 4,a.
The x-direction of the scans is then averaged and a third
order fit of two constantly separated curves is performed
through the original and the swelled region (see Fig.
4,b).
2.4. NANO-INDENTATION
In order to test the mechanical behavior a nano-
indenter from CSM Instruments SA in Neuchâtel
(Switzerland) is applied, the maximum load of the
apparatus being 500 mN with a load resolution of 0.04
μN and a maximum depth of 20 μm with a resolution of
0.04 nm. Two measurements series are performed using
a Berkovich tip (a tetrahedron that comes to a sharp
point). In both cases the indent depth is fixed. In the
first series it is set to 500 nm, and in the second one to
1 μm. The targeted indent positions are localized with
an optical microscope being connected to the indenter.
The positioning is performed with a precision of about 1
μm. Two series of indents are performed on a sample
irradiated with the dose and damage profile described in
Fig. 1. In each series the indents are positioned such,
that both, non irradiated and irradiated parts of the
sample are tested. Figure 2 shows the first indentation
series, where the two first indents are placed in the non-
irradiated part, the following three in the irradiated part
and the last one in a non-irradiated section on the other
side of the irradiated stripe.
Fig. 2. First indent series with 500 nm penetration
depth. The lighter surface displays the non irradiated
part and the darker one the irradiated part
of the sample
The reproducibility of the results is very good, as
illustrated in Fig. 3. All load/penetration curves comply
with each other in both, the irradiated and the non-
irradiated area respectively.
-5
0
5
10
15
20
25
30
35
0 100 200 300 400 500
Penetration [nm]
Fo
rc
e
[m
N
]
indent on non-irradiated surface
indent on irradiated surface
Fig. 3. Force to penetration depth characteristic of the
500 nm indents performed on the irradiated and the
non-irradiated surfaces
3. RESULTS
3.1. SWELLING BEHAVIOR
The AFM shows the swelling pattern of the
irradiated ODS bars. Figure 4,a shows a sample
irradiated with He ions under the conditions described
in Fig. 1,a. Figure 4,b is a third order fit through both,
the irradiated and the non-irradiated parts of the profile.
The swelling is deduced from the distance of the two
fitted curves.
a b
Fig. 4. Surface of the ODS sample, investigated by AFM, original data after peak removal on the left side and the
averaged profile on the right (a). Third order fitting through the averaged profile (b). Separate curves are derived
for the irradiated and the non-irradiated locations on the profile, a step height of 24.1±2.8 nm is found
for this example
Four different irradiation doses are applied (see 2.2).
For each of the irradiated samples, the swelling is
determined by AFM, resulting in a displacement
dependency as a function of fluence. The displacements
are shown in Fig. 5,a and they are represented as strain
data in Fig. 5,b.
0
10
20
30
40
0 5 10 15
fluence [1016 cm-2]
di
sp
la
ce
m
en
t [
nm
]…
calculated
experimental
0.0
0.5
1.0
1.5
2.0
0.0 0.5 1.0 1.5
displ. dose [dpa]
st
ra
in
[%
]..
...
trend line
corr. trend line
experimental
a b
Fig. 5. Swelling data as measured. The calculated data is based on the corrected strain trend line in (b) and the
implanted profile with the strain profile in Fig. 6 (a). Strain data as function of the averaged displacements of the
depth region from 0.5 to 2.5 μm with the corresponding trend line (b). The corrected trend line is based on the
strain profile shown in Fig. 6 and shows the strain as a function of the exact (non-averaged) dpa value
The strain data are used to determine a bi-linear
trend line. Because of the small amount of data points
and for simplicity, this form of the trend line is chosen
rather than a more sophisticated function. The x-axis for
strain data and the trend line represents the average
displacement in the depth from 0.5 to 2.5 μm. Because
of the non-linearity of strain as a function of
displacements and because of the displacement damage
not being constant in the relevant depth region, the data
points can not be assigned to the precise displacement
value, but only to the average value for the actual
displacement distribution. This distribution is depicted
124
in Fig. 6,a (solid thin line), the profile is directly
obtained by overlying TRIM simulations for each
incident angle.
TRIM tends to underestimate the profile width for
the single simulations; therefore the initial profile is
fitted with a polynomial of seventh degree. This
polynomial is also presented in Fig. 6,a (doted line) and
is used for further calculations. Fig. 6,a also shows the
strain profile, which is calculated, using the corrected
strain as a function of displacements shown in 5,b and
the displacement polynomial. The corrected strain to
displacement function is found on an iterative way, by
introducing correction factors to the parameters of the
trend line in Fig 5,b. By calculating the swelling of the
known displacement distribution with the corrected
strain to displacement function (integration), a direct
comparison with the measured data can be performed
(see Fig 5,a. Subsequently the parameters of the
corrected trend line can be adjusted to reach a good
matching of the experimental data points and the
calculated ones.
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
0.00 1.00 2.00
depth [µm]
di
sp
l.
do
se
[d
pa
]
0
0.2
0.4
0.6
0.8
1
1.2
1.4
st
ra
in
[%
]…
.
TRIM prof ile
fitted profile
strain
0.0
0.5
1.0
1.5
2.0
0.0 1.0 2.0
depth [µm]
st
ra
in
[%
]…
..
a b
Fig. 6. Displacement profile for a fluence of 5.6·1016 cm-2 calculated with TRIM and an averaged profile used for
calculations (a). With the averaged profile and the corrected trend line of relative strain (see Fig. 5,b), the strain
profile is calculated. All strain profiles for the different implantation fluence (1.4·1016, 2.8·1016, 5.6·1016 and
1.12·1017cm-2). Because of the non linearity of the strain to displacement function, the shape varies with the total
fluence (b)
MECHANICAL PROPERTIES
In this paper the hardness is investigated using a
nanoindenter, see Section 2.4. Results are presented for
the non-irradiated and the irradiated surface of the
sample. Figure 8 contains the results for both
indentation series, the one with an indentation depth of
500 nm and the one with 1000 nm. For both series a
consistent increase of the irradiation induced increase in
hardness, can be observed. For the 1000 nm indent
series, two indents are placed on the border line between
the irradiated and the non-irradiated surface, accounting
for mixed regions. Figure 7 shows such an indent,
where about 40% of the indent is performed on an
irradiated surface. The axis of abscissae in Fig. 8 does
therefore not only contain the distinctive cases of non-
irradiated and irradiated, but represents the percentage
of the indent accounting for an irradiated region.
The intermediate data points properly fit between the
distinctive cases. The hardness is compared to values
reported in [8], in this thesis the hardness is measured
by a Vickers indenter with a mass of 30 kg and a 2/3"
objective. For untreated ODS samples a Vickers
hardness of 326±6 HV30 is reported. Transforming this
into Si units results in a hardness of 3197±59 MPa
which agrees well with the result of 3207 MPa found for
the 1000 nm indents on the non-irradiated surface (see
Fig. 8).
a
b
125
Fig 7. Optical (a) and SEM (b) picture of a 1 μm indent at the border between the irradiated and the non irradiated
surface of the sample
In Fig. 8 the hardness decreases with increasing the
indentation depth. The depth-dependent hardness has
been observed in various materials such as metals,
diamond-like carbon, polymers, ceramics, etc before.
The depth-dependent hardness has been explained by
the theory of strain-gradient plasticity and the surface
effect [9].
2500
3000
3500
4000
4500
-20 0 20 40 60 80 100 120
Irradiated surface [%]
H
ar
dn
es
s [
M
Pa
]
1000 nm indents
0500 nm indents
Fig. 8. Hardness as a function of the irradiated surface
to non-irradiated surface fraction of the indentation
region
4. DISCUSSION
The increased hardness of the irradiated surfaces
well agrees with results for ferritic steels found in
literature. In [10], low activation ferritic steel is
investigated by micro-indentation technique in order to
find the ion induced hardening. At an irradiation
temperature of 673 K, irradiation-produced dislocation
loops are observed by TEM, the number and density of
these loops is clearly increased and their mean size is
decreased with increasing He concentration. The
irradiation induced micro hardness changes were no
more than 10 at.%673 K. In this paper the irradiation
induced hardness changes at room temperature are 29%
for the 1000 nm indents and 26% for the 500 nm
indents. This difference might be attributed to a
superimposed effect of the compressive residual stresses
in the samples being investigated in this work.
However, the differences might only come from the
different radiation conditions.
In the present study only the transverse grain
direction is studied. As the material is anisotropic, the
other grain orientations will also be investigated. In [8]
the hardness was distinguished for the longitudinal and
the perpendicular extrusion direction. Differences were
mainly found for thermally treated samples, where the
hardness generally decreased by up to 23% and the
transversal extrusion direction was 10% harder than the
longitudinal one. For the untreated samples the hardness
was almost isotropic.
5. CONCLUSIONS
At room temperature and displacement dose ranges
around 0.7 dpa, PM2000 already shows a major increase
of its hardness by 30% and also a strain of around 1%.
Annealing effects at higher temperatures will potentially
decrease the hardening and swelling effects. Future
experimental series will address this question. A first
hint for a temperature decreased hardness can be found
in [8,10], where the thermal treatment significantly
decreased the hardness, and the irradiation induced
hardening effect was reduced at higher temperatures.
The present work demonstrates that the nano/micro
scale approach to characterize the material is valid. In
case of the hardness a good agreement with literature
data exists, and the swelling induced displacements are
also within expectations.
6. ACKNOWLEDGEMENTS
The authors wish to thank Tomislav Rebac at the
Paul Scherrer Instiute for his preparation of the ODS
plates. The nano-indents were performed at the CMS
Company in Neuchâtel by Philippe Kempe. The
financial support of this project by the department
Nuclear Energy and Safety of PSI is greatly
acknowledged.
REFERENCES
1. S. Ukai and M. Fujiwara. Perspective of ODS alloys
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Pint, P.J. Mazias. Advanced Alloy Systems //Fusion
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University of California-Santa Barbara,
http://www.fusionmaterials.pnl.gov/peerreview/hoelzer
_advanced.pdf
3. Prof. L. Singheiser, private communication, January
2003.
4. M.A. Fütterer //JRC Petten, private communication,
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5. A. Czyrska-Filemonowicz and B. Dubiel.
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of Materials Processing Technology. 1997, v. 64(1-3),
p. 53–64.
6. Dispersion-Strengthened High-Temperature
Materials /Material properties and applications,
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DE.04.03(1000)RWF.
7. ASTM Designation: E 521 – 96 /Standard Practice
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8. V.A. Yardley. Magnetic Detection of
Microstructural Change in Power Plant Steels
/Dissertation at the University of Cambridge, May 2003,
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10. Y.Katoh et al. The influence of helium co-
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OДУ СТАЛИ КАК КОНСТРУКЦИОННЫЕ МАТЕРИАЛЫ ДЛЯ
ВЫСОКОТЕМПЕРАТУРНЫХ ЯДЕРНЫХ РЕАКТОРОВ
М.А. Пушон, М. Дюбели, Р. Шелдорф, Ж. Шен, В. Хёфельнер, С. Дегельдр
Окисно диспергированные упрочненные (ОДУ) ферритно-мартенситные стали изучались как возможные
конструкционные материалы для нового поколения высокотемпературных ядерных реакторов с газовым охлаждением.
Аустенитные ОДУ-суперсплавы на базе никеля не рассматривались из-за присутствия никеля, который нежелателен при
нейтронном облучении. ОДУ-стали рассматривались как возможные заменители других высокотемпературных
материалов для изготовления трубопроводов или других композиционных узлов. Интересно, что ОДУ рассматривается
так же, как возможный кандидат для использования в термоядерных устройствах. Окисные частицы служат как
межфазные ловушки для закрепления движущихся дислокаций. Поэтому сопротивление ползучести увеличивается. В
случае использования этих материалов в реакторе их поведение под облучением должно изучаться более тщательно. В
предлагаемой работе исследуются эффекты, обусловленные внедрением Не. Измеряется обусловленное распухание и
механические характеристики облученной поверхности. Эти первые испытания были выполнены при комнатной
температуре, когда можно явно наблюдать распухание и упрочнение.
ОДЗ СТАЛІ ЯК МОЖЛИВІ КОНСТРУКЦІЙНІ МАТЕРІАЛИ ДЛЯ
ВИСОКОТЕМПЕРАТУРНИХ ЯДЕРНИХ РЕАКТОРІВ
М.А. Пушон, М. Дюбелі, Р. Шелдорф, Ж. Шен, Ф. Хофельнер, С. Дегельдр
Окисно дисперговані зміцнені (ОДЗ) феритно-мартенситні сталі досліджуються як можливі конструкційні матеріали
для нового покоління високотемпературних ядерних реакторів з газовим охолодженням. Аустенітні ОДЗ-сплави на
основі никелю не вивчаються завдяки никелю, присутність якого під дією опромінення небажана. ОДЗ-сталі
разглядаються як можливі кандидати на заміну інших високотемпературних матеріалів для вигoтовлення трубопроводів
або інших конструкційних вузлів. Цікаво, що ОДЗ-матеріали розглядаються також з точки зору їх можливого
використання для майбутнього застосування в термоядерних пристроях. Окисні частки служать як міжфазні пастки для
закріплення дислокацій, що рухаються. У разі використання ціх материалів в реакторі їх поведінка під опроміненням
повинна вивчатись більш ретельно. В роботі досліджуються ефекти, зумовлені проникненням Не. Вимірюється
зумовлене розпухання та механічні характеристики опроміненої поверхні. Ці перші дослідження були виконані при
кімнатній температурі, коли можно чітко спостерігати розпухання та зміцнення.
1.
1. INTRODUCTION
1.1. THE HIGH TEMPERATURE PROJECT
1.2. PM2000 AS POTENTIAL MATERIAL
2. EXPERIMENTAL
2.1. SAMPLE PREPARATION
2.2. IRRADIATION
2.3. AFM INVESTIGATIONS
2.4. NANO-INDENTATION
3. RESULTS
3.1. SWELLING BEHAVIOR
4. DISCUSSION
5. CONCLUSIONS
6. ACKNOWLEDGEMENTS
REFERENCES
ОДЗ СТАЛІ ЯК МОЖЛИВІ КОНСТРУКЦІЙНІ МАТЕРІАЛИ ДЛЯ ВИСОКОТЕМПЕРАТУРНИХ ЯДЕРНИХ РЕАКТОРІВ
М.А. Пушон, М. Дюбелі, Р. Шелдорф, Ж. Шен, Ф. Хофельнер, С. Дегельдр
|
| id | nasplib_isofts_kiev_ua-123456789-80401 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 1562-6016 |
| language | English |
| last_indexed | 2025-12-07T18:53:57Z |
| publishDate | 2005 |
| publisher | Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| record_format | dspace |
| spelling | Pouchon, M.A. Dobeli, M. Schelldorfer, R. Chen, J. Hoffelner, W. Degueldre, C. 2015-04-17T16:40:04Z 2015-04-17T16:40:04Z 2005 ODS steel as structral material for high temperature nuclear reactors / M.A. Pouchon, M. Dobeli, R. Schelldorfer, J. Chen, W. Hoffelner, C. Degueldre // Вопросы атомной науки и техники. — 2005. — № 3. — С. 122-127. — Бібліогр.: 9 назв. — англ. 1562-6016 https://nasplib.isofts.kiev.ua/handle/123456789/80401 669.15-194 Oxide dispersed strengthened (ODS) ferritic-martensitic steels are investigated as possible structural material for the future generation of High Temperature Gas Cooled Nuclear Reactors. The Ni based austenitic ODS superalloys are not considered, because of the Ni presence, which is unfavorable under neutron irradiation. ODS-steels are considered to replace other high temperature materials for tubing or structural parts. Interestingly, ODS is also considered as material being used in future fusion applications. The oxide particles serve for interfacial pinning of moving dislocations. Therefore the creep resistance is improved. In case of the usage of these materials in reactor, the behavior under irradiation must be further clarified. In this paper the effects induced by He implantation are investigated. The induced swelling is measured and the mechanical behavior of the irradiated surface is investigated. These first tests are performed at room temperature, where a clear swelling and hardening could be observed. Окисно дисперговані зміцнені (ОДЗ) феритно-мартенситні сталі досліджуються як можливі конструкційні матеріали для нового покоління високотемпературних ядерних реакторів з газовим охолодженням. Аустенітні ОДЗ-сплави на основі никелю не вивчаються завдяки никелю, присутність якого під дією опромінення небажана. ОДЗ-сталі разглядаються як можливі кандидати на заміну інших високотемпературних матеріалів для вигoтовлення трубопроводів або інших конструкційних вузлів. Цікаво, що ОДЗ-матеріали розглядаються також з точки зору їх можливого використання для майбутнього застосування в термоядерних пристроях. Окисні частки служать як міжфазні пастки для закріплення дислокацій, що рухаються. У разі використання ціх материалів в реакторі їх поведінка під опроміненням повинна вивчатись більш ретельно. В роботі досліджуються ефекти, зумовлені проникненням Не. Вимірюється зумовлене розпухання та механічні характеристики опроміненої поверхні. Ці перші дослідження були виконані при кімнатній температурі, коли можно чітко спостерігати розпухання та зміцнення. Окисно диспергированные упрочненные (ОДУ) ферритно-мартенситные стали изучались как возможные конструкционные материалы для нового поколения высокотемпературных ядерных реакторов с газовым охлаждением. Аустенитные ОДУ-суперсплавы на базе никеля не рассматривались из-за присутствия никеля, который нежелателен при нейтронном облучении. ОДУ-стали рассматривались как возможные заменители других высокотемпературных материалов для изготовления трубопроводов или других композиционных узлов. Интересно, что ОДУ рассматривается так же, как возможный кандидат для использования в термоядерных устройствах. Окисные частицы служат как межфазные ловушки для закрепления движущихся дислокаций. Поэтому сопротивление ползучести увеличивается. В случае использования этих материалов в реакторе их поведение под облучением должно изучаться более тщательно. В предлагаемой работе исследуются эффекты, обусловленные внедрением Не. Измеряется обусловленное распухание и механические характеристики облученной поверхности. Эти первые испытания были выполнены при комнатной температуре, когда можно явно наблюдать распухание и упрочнение. The authors wish to thank Tomislav Rebac at the Paul Scherrer Instiute for his preparation of the ODS plates. The nano-indents were performed at the CMS Company in Neuchâtel by Philippe Kempe. The financial support of this project by the department Nuclear Energy and Safety of PSI is greatly acknowledged. en Національний науковий центр «Харківський фізико-технічний інститут» НАН України Вопросы атомной науки и техники Материалы перспективных ядерных реакторов ODS steel as structral material for high temperature nuclear reactors ОДЗ сталі як можливі конструкційні матеріали для високотемпературних ядерних реакторів OДУ стали как конструкционные материалы для высокотемпературных ядерных реакторов Article published earlier |
| spellingShingle | ODS steel as structral material for high temperature nuclear reactors Pouchon, M.A. Dobeli, M. Schelldorfer, R. Chen, J. Hoffelner, W. Degueldre, C. Материалы перспективных ядерных реакторов |
| title | ODS steel as structral material for high temperature nuclear reactors |
| title_alt | ОДЗ сталі як можливі конструкційні матеріали для високотемпературних ядерних реакторів OДУ стали как конструкционные материалы для высокотемпературных ядерных реакторов |
| title_full | ODS steel as structral material for high temperature nuclear reactors |
| title_fullStr | ODS steel as structral material for high temperature nuclear reactors |
| title_full_unstemmed | ODS steel as structral material for high temperature nuclear reactors |
| title_short | ODS steel as structral material for high temperature nuclear reactors |
| title_sort | ods steel as structral material for high temperature nuclear reactors |
| topic | Материалы перспективных ядерных реакторов |
| topic_facet | Материалы перспективных ядерных реакторов |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/80401 |
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