Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture
Ціль запропонованої роботи – отримати результати для того, щоб зробити питомий внесок у розуміння механізму радіаційного обумовленного міжзеренного корозійного розтріскування аустенітних нержавіючих сталей. Цель предложенной работи – получение новых результатов для того, чтобы внести конструктивный...
Saved in:
| Published in: | Вопросы атомной науки и техники |
|---|---|
| Date: | 2002 |
| Main Authors: | , , , , , |
| Format: | Article |
| Language: | English |
| Published: |
Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2002
|
| Subjects: | |
| Online Access: | https://nasplib.isofts.kiev.ua/handle/123456789/79290 |
| Tags: |
Add Tag
No Tags, Be the first to tag this record!
|
| Journal Title: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Cite this: | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture / Yu.D. Goncharenko, V.A. Kazakov, V.K. Shamardin, A.M. Pechyorin, G.V. Filyakin, Z.Ye. Ostrovsky // Вопросы атомной науки и техники. — 2002. — № 6. — С. 69-75. — Бібліогр.: 7 назв. — англ. |
Institution
Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1859824646060244992 |
|---|---|
| author | Goncharenko, Yu.D. Kazakov, V.A. Shamardin, V.K. Pechyorin, A.M. Filyakin, G.V. Ostrovsky, Z.Ye. |
| author_facet | Goncharenko, Yu.D. Kazakov, V.A. Shamardin, V.K. Pechyorin, A.M. Filyakin, G.V. Ostrovsky, Z.Ye. |
| citation_txt | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture / Yu.D. Goncharenko, V.A. Kazakov, V.K. Shamardin, A.M. Pechyorin, G.V. Filyakin, Z.Ye. Ostrovsky // Вопросы атомной науки и техники. — 2002. — № 6. — С. 69-75. — Бібліогр.: 7 назв. — англ. |
| collection | DSpace DC |
| container_title | Вопросы атомной науки и техники |
| description | Ціль запропонованої роботи – отримати результати для того, щоб зробити питомий внесок у розуміння механізму
радіаційного обумовленного міжзеренного корозійного розтріскування аустенітних нержавіючих сталей.
Цель предложенной работи – получение новых результатов для того, чтобы внести конструктивный вклад в
понимание механизма радиационно-обусловленного межзеренного коррозионного растрескивания аустенитных сталей.
The objective is to obtain the new results in order to make a constructive contribution into understanding of the mechanism of
austenitic stainless steels (ASS) irradiation-assisted intergranular stress corrosion cracking (IAISCC).
|
| first_indexed | 2025-12-07T15:28:18Z |
| format | Article |
| fulltext |
РАЗДЕЛ ТРЕТИЙ
МАТЕРИАЛЫ РЕАКТОРОВ НА ТЕПЛОВЫХ НЕЙТРОНАХ
УДК 669.018.2
IRRADIATION-ASSISTED INTERGRANULAR STRESS CORROSION
CRACKING OF AUSTENITIC STAINLESS STEEL
IN STEAM-WATER MIXTURE
Yu.D.Goncharenko, V.A.Kazakov, V.K.Shamardin, A.M.Pechyorin,
G.V.Filyakin, Z.Ye.Ostrovsky
Federal State Unitary Enterprise
“State Scientific Center of Russian Federation Research Institute of Atomic Reactors”
Dimitrovgrad, Ulyanovsk region, Russia
Ціль запропонованої роботи – отримати результати для того, щоб зробити питомий внесок у розуміння механізму
радіаційного обумовленного міжзеренного корозійного розтріскування аустенітних нержавіючих сталей.
Цель предложенной работи – получение новых результатов для того, чтобы внести конструктивный вклад в
понимание механизма радиационно-обусловленного межзеренного коррозионного растрескивания аустенитных сталей.
The objective is to obtain the new results in order to make a constructive contribution into understanding of the mechanism of
austenitic stainless steels (ASS) irradiation-assisted intergranular stress corrosion cracking (IAISCC).
INTRODUCTION
Susceptibility of austenitic stainless steels (ASS) to
stress corrosion cracking (SCC), which is of
intergranular type in most cases (ISCC), is one of the
important problems of their usage as in-vessel devices
components (IVDC) of nuclear and fusion reactors in
water-steam environment [1-4]. This fact is caused by
chromium depletion of the boundaries and their
adjoining zones in the sensitive areas near the welded
joints as a result of precipitation of chromium carbides
such as Ме23С6 and/or Ме6С [1,2].
Neutron irradiation facilitates the SCC propagation
due to the radiation-induced segregation and depletion
processes at the interphase surfaces. Moreover, the
specific mechanisms of sulfur release from MnS
sulfides due to nuclear transformation of 54Mn into 56Fe,
cascade intermixing on the sulfide surface and advanced
penetration of manganese deep into the metal because of
inverse Kirkendall effect [5] have been proposed for the
irradiation conditions. The processes are followed by
etching of grain boundaries with sulfur, fluorine and
chlorine that are dangerous elements from the corrosion
point of view.
In-vessel devices components of 35 BWR-type
German reactors made of austenitic stainless steel [6]
were subjected to examination not long ago. Cracks
caused by ISCC were discovered in IVDC of 24
reactors. They were 390mm in length as a maximum
and 25…30mm in depth. All cracks were discovered in
the areas of residual stress caused by cold surface
hardening in machining and welding.
The real IVDC that have been operated for a long
time are the most informative material for SCC
investigation under neutron irradiation. The external
0X18H10T ASS tube of VK-50 control assembly was
chosen as the subject of examination in this case. It was
damaged in the welded joint area after 30 years of
operation during the scheduled replacement.
The objective– to obtain new results in order to
make a constructive contribution into understanding of
the mechanism of ASS irradiation-assisted intergranular
stress corrosion cracking (IAISCC).
1. MATERIAL, SPECIMENS, OPERATING
CONDITIONS
The chemical composition of the tube (∅152×3 mm)
made of 0Х18Н10Т is as follows: C ≤ 0.08; Si ≤ 0.8;
Mn ≤ 2.0; Cr = 17.0…19.0; Ni = 9.0…11.0; Ti = 0.4…
0.7; S ≤ 0.02; P ≤ 0.035; Cu ≤ 0.3 mass%. The tube was
operated during 271 500 hours at a temperature of 285,
264 and 250ОС. The tube was damaged in the welded
joint area at a distance of 10-20mm from the melting
zone of the welded joint. The neutron fluence in the
place of fracture was 2.2⋅ 1021 n/cm2 (E > 0.5 MeV) or
3.0⋅ 1021 n/cm2 (E > 0.1 MeV). The chemical
composition of the steam-water mixture was the
following: water hardness–1.2 µmole/dm3; pH – 6.0…
6.2; Fe3+ – 0.012 mg/dm3; Cl- – ≤ 0.05 mg/dm3; Cu2+ –
0.009…0.03 mg/dm3; nitric oxide NO – 0.011…0.03
ml/dm3; Zn2+ – 0.01…0.04 mg/dm3; specific electric
conductivity – 0.18…0.2 µS/cm.
A wrapper was operated under alternate cooling with
subboiling water and steam-water mixture at the
saturation temperature in the place of fracture. The
visual inspection of the fracture place reveled cracks
____________________________________________________
ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2002. № 6. 69
Серия: Физика радиационных повреждений и радиационное материаловедение (82), с.69-75.
both in transverse and longitudinal directions. Some
cracks were 120…140mm in length.
2. RESULTS OF EXPERIMENT
2.1. METALLOGRAPHY, FRACTOGRAPHY,
MECHANICAL PROPERTIES
Fig.1 demonstrates the typical view of the fracture
surface and cracks. It was noticed that cracks grow
along the grain boundary only and they grow in both
transverse and longitudinal directions. The fracture
exhibits brittle, grain-boundary character although there
are traces of fresh cleavages caused by transcrystalline
fracture in the fine areas that adjoin the grain
boundaries. Flat tensile specimens were cut out form the
tube section adjacent to the place of fracture using the
electro-erosion technique. A test portion of specimens
was 15×3×3 mm in size. The yield stress of irradiated
specimen increased by a factor of two at 20ОC when
compared to the initial state and reached 850…950
MPa, the uniform relative elongation δr came to 12…
25% and the total one δt comprised 26…38%. δr fell to
0.8…4% and δt – to 14…15% at a test temperature of
280ОC. The material damaged with the formation of
developed neck in both cases and the fracture exhibited
transcrystalline character.
2.2. TRANSMISSION ELECTRON MICROSCOPE
INVESTIGATION
As illustrated in Fig.2, segregation of secondary
phases that range in size from several tens to several
hundreds of nanometers and align themselves in the
main along the grain boundaries and in the near-
boundaries areas (Fig.2a, b) takes place during the
0Х18Н10Т steel irradiation. Their chemical
composition is as follows: 37…46 Fe, 11…15Cr, 32…
45Ti, 5…6Ni, atomic%. Based on the structure, these
precipitates are likely to be Me23C6 or Me6C type
carbides. The large precipitate in Fig.2c is the titanium
carbide TiC. Fine-dyspersated precipitates of a round
shape are discovered in the grain body. They are
assumed to be the G-phase. Their medium size is about
8 nanometers and density constitutes 5⋅ 1015 cm−3. The
average diameter of dislocation loops was 9 nanometers,
their density comprised 1.4⋅ 1016 cm−3.
70
50µm
a
50µm
b
25µm
c
Fig.1. Fracture surface (a,c) and typical view of cracks (b).
Optical metallography (a, b); scanning electron microscope investigation (c)
Fig.2. The tube material microstructure after irradiation.
Transmission electron microscope investigation
500nмa
500nм
b
200nm
c
2.3. HYDROGEN ANALYSIS
The hydrogen content was analyzed using a etalon-
free spectral-isotopic method. Hydrogen constitutes
0.003 mass% in the Х18Н10Т steel in the initial state.
The hydrogen content increased insignificantly and
made up 0.0045 mass% after operation. For comparison,
the hydrogen content in Х18Н10Т increased nine times
in the temperature range 300-395ОC and at a neutron
fluence of 1⋅ 1022 ÷ 1.7⋅ 1023 n/cm2 after operation in the
BOR-60 reactor.
2.4. AUGER SPECTROSCOPY
A fresh fragment of the fracture surface to be 2×2
×0,3 mm in size was used for examination. The primary
electron beam with a diameter of 3µm was focused into
a raster to light a fracture section to be ~ 200×200 µm in
size in order to obtain a surface- average values for
concentrations of different elements. A multiple ion
etching of the surface under investigation was carried
out after the Auger spectra recording (10…12 atomic
layers) from the fracture surface to analyze the
elemental composition in depth. The thickness of the
remote layer was calculated based on the etching rate
that is approximately equal to 20 angstorm per second.
As illustrated in Fig.3a, chromium and nickel are
fully absent but titanium is present in large quantities.
Its content is 60 times greater than in the matrix. The
changed layer depth varies between 0.2 and 1.2 µm. The
carbon content falls to the level of initial state at ~1.2µ
m, whereas the high level of oxygen is kept constant
and makes up about 40 atomic% at a depth of 1 µm
(Fig.3b). The depth of sulfur and chlorine layers makes
up about 0.2 and 0.4 µm respectively. It is about 1 µm
for copper. The calcium content decreases sharply in 0.2
µm thick layer but it disappears completely at a depth of
1 µm (Fig.3c).
0
20
40
60
80
0 0,2 0,4 0,6 0,8 1 1,2
Расстояние от поверхности разрушения,
мкм
Ко
нц
ен
тр
ац
ия
, а
т%
Ti
Ni
C
Fe
0
20
40
60
0 0,2 0,4 0,6 0,8 1 1,2
Расстояние от поверхности разрушения,
мкм
Ко
нц
ен
тр
ац
ия
, а
т%
С
О
0
1
2
3
4
0 0,2 0,4 0,6 0,8 1 1,2
Расстояние от поверхности разрушения,
мкм
Ко
нц
ен
тр
ац
ия
, а
т%
C
Cl Ca
S
а d c
Fig.3. Content of the primary alloying elements ( a, except for light and impurity ones), oxygen, carbon (b), as
well as sulfur, chlorine, copper and calcium (c) against the fracture surface distance in the irradiated steel
In order to obtain data on uniform or non-uniform
distribution of different elements on the fracture surface
we recorded series of elemental images at an
accelerating voltage of 3 and 9kV. Fig.4 shows the
images of the fracture surface for one of the grains that
is arranged in the center of each picture. As shown in
Fig.4a, chlorine is distributed uniformly. Contrary to
chlorine sulfur is found in aggregations to be 5…6 µm
in size (Fig.4b). This fact is well visible at lesser
magnifications and accelerating voltage of 3 kV
(Fig.4c). As a rule, the sulfur aggregations are
characterized by higher concentration of iron (Fig.4e),
chromium (Fig.4h) and nickel as well. The sulfur
aggregations appear to be the complex sulfide
compounds, which incorporate iron, chromium and
nickel.
Copper is distributed by separate aggregations along
the boundaries of grain under examination (Fig.4d).
Higher concentration of copper has been noted in
several places around the periphery of a facet but it is
absent on its surface. A map of carbon distribution
(Fig.4g) those Auger peak is the most intensive one is
much the same as the image of the fracture surface in
absorbed or secondary electrons (Fig.4i) according to
the topographic tinctures.
Any specific carbon aggregations are absent but there
are regions of its lower concentration where the oxygen
content is higher (Fig.4f, g). Higher oxygen content in
the right lower part of the facet (Fig.4f) is followed by
higher nickel concentration and slightly increased
chromium concentration (Fig.4h).
3. DISCUSSION OF RESULTS
The following conclusions can be drawn according
to the obtained results:
• Fracture occurs mainly along the grain boundaries.
• Ti segregates on the grain boundaries, but Cr and
Ni deplete them.
71
• Chlorine, sulfur and copper that are dangerous
elements from the corrosion point of view, are
available on the fracture surface.
The thickness of the segregation layer on the
fracture surface is about 1…1.5 µm; in any case after
ion etching at a depth of ~ 3 µm segregation phenomena
disappear practically (Fig.5).
Table 1
Local elemental composition
of the fracture surface
Ele-
ment
Content, atomic %
«A» «B» «C» «D» Raster
Fe 4.0 12.2 1.7 7.4 2.4
Cr 0.5 2.8 0.8 1.6 0.6
Ni 3.8 5.1 1.4 6.5 0.3
Ti 1.0 0.6 0.6 1.2 0
C 75.9 61.7 82.5 47.0 91.5
O 13.0 13.1 6.3 30.8 3.9
Cl 0.5 0.4 0.6 0.3 0.3
S 0.4 2.7 0.1 0 0.4
Cu 0.8 1.4 6.2 3.0 0.5
Ca 0 0 0 2.3 0
3.1.
DISTRIBUTION OF ELEMENTS ON THE
FRACTURE SURFACE
Images provide the qualitative representation of
element distribution. But the quantitative information
can be gained from the local elemental composition
analysis at different points when the diameter of the
electron probe is 0.1…0.3 µm (in our case it is
performed at points “A – D», Fig.4, Table 1). For
comparison purposes the Table includes the results of
integral analysis of the elemental composition in a raster
of 200×200 µm. The facet under examination entered its
area. The results of local measurements agree well with
the quantitative information obtained from the maps. As
the Table indicates chlorine is the only element that is
distributed on the fracture surface rather uniformly. The
72
b
B
Sulfur
a
Chlorine
e
A
c
Sulfur
d
C
Copper Iron
e f
Oxygen
Carbon
g h
Chromiu
m
D
i
Absorbed electrons
Fig 4. Images for distribution of elements on the irradiated steel fracture surface
(a,b,d-i – 1400Ч; c – 400Ч; A,B,C,D – places of local elemental composition analysis)
a Sulfur b
Iron
c
Absorbed electrons
Fig.5. Images of sulfur and iron distributions on the fracture surface of the irradiated steel
at a depth of 3 µm after ion etching. 1500×
concentration of other elements varies within rather
wide limits.
3.2. SUMMARY OF FACTORS IMPORTANT
FOR IAISCC OF ASS
Literature data [1-7] and obtained results enable us
to define three groups of factors: 1) principal, 2)
determining and 3) contiguous.
Principal factors: 1) Tensile stress. Cracks were ob-
served only in the welded joint area in the upper part of
the tube. In accordance with data [4] tensile stress in
this area may reach 300MPa. In the base metal in the
lower part of the tube, where neutron fluence was high-
er, cracks were not observed. Therefore, it is logically to
assume that tensile stresses in IAISCC dominate over
radiation-induced segregation/depletion processes. 2)
Sensitivity. Heating of the area near welded joint up to
550…800ОС during welding leads to precipitation of
large Me23C6 and/or Me6C type carbides along the grain
boundaries and in the near-boundaries areas. The size of
these carbides is several hundred nanometers (Fig.2),
their chemical composition is as follows: Cr-13%, Fe-
42%, Ti-39% and Ni-6%.
This is one of the sources of grain boundaries depletion
in chromium. The next depletion source is pure thermal
low-temperature sensitivity in the temperature range
250…350ОС. As the evalution shows, the time neces-
sary for this process is 12 years at 285ОС [4].
This time period is in two and a half lesser than the
tube operation time. The third and probably the most
powerful source of grain boundaries depletion in
chromium is the radiation-induced one. It seems to com-
bine 3 processes: a) growth of the existing carbides; b)
radiation depletion of grain boundaries in chromium; c)
precipitation of fine-dispersed phases containing
chromium.
Table 2
Radiation effect on the change of chemical composition of SS 304 grain body
and near-boundaries areas at 300оC [7]
Elemen
t
Energy dispersion analysis, TEM
investigations, mass%
Auger-electron microscopy investigation,
atomic%
Initial state
Grain
body
Grain
boundaries
5⋅ 1021 n/cm2,
grain boundaries
Initial state 5⋅ 1021 н/см2
Grain
body
Grain
boundaries
Grain
body
Grain
boundaries
Fe score 62-64 61 69.39 74.9 68.4 60.6
Cr 18.54 24-25 16 19.58 16.4 18.7 14.2
Ni 8.28 9 15-16 7.75 7.8 8.43 14.1
Mo 0.32 1.7-1.9 1 0.18 Not analyzed 1.53 0.2
Mn 1.52 Not analyzed Not analyzed Not
analyzed
Not analyzed Not
analyzed
Not
analyzed
Si 0.55 1 4.2-4.5 1.08 Not analyzed 1.25 8.3
C 0.069 Not analyzed Not analyzed Not ana-
lyzed
Not analyzed Not ana-
lyzed
Not ana-
lyzed
P 0.023 0.8 1.1 0.04 0.7 0.06 1.9
S 0.021 Not analyzed Not analyzed Not ana-
lyzed
Not analyzed Not ana-
lyzed
Not ana-
lyzed
Paper [7] presents the analysis of segregation
processes at the grain boundaries of steel 304 before and
after irradiation at 300ОC (Table 2). The Table indicates
that chromium and silicon segregate noticeably but
molybdenum and phosphorus segregate especially
strong at the grain boundaries in the initial state
according to the results of energy dispersion analysis
(EDA). The results of Auger-electron microscopy
investigation didn’t prove chromium segregation. The
depletion of grain boundaries in chromium,
molybdenum and their enrichment in nickel, phosphorus
and especially strong in silicon was revealed after
irradiation according to the EDA results. As for
irradiated state the results of EDA and Auger-
spectroscopy agree qualitatively but they correlate better
for molybdenum, silicon and phosphorous that are
subjected to Auger-electron microscopy investigations.
The results of these examinations with regard to
chromium agree with the results of the given paper
principally but the effect is much stronger. But they are
in direct opposition to nickel: as for nickel the depletion
of grain boundaries up to zero is revealed in our case,
and in work [7] the enrichment of grain boundaries in
nickel is observed.
Therefore, the conclusion may be drawn that
radiation-induced segregation processes at the grain
boundaries present the second important factor in
IAISCC phenomenon.
Determining factors: 1) High oxygen concentration
in the water-steam mixture, which may reach 0.2…0.5
mg/l in water and up to 20-50 mg/l in steam [1,2] as a
result of radiolysis, facilitates the increase of metal dis-
solution rate at the intergranular channels tops. 2) Chlo-
rine precipitation at the boundaries. Coolant is probably
the major chlorine source. The capability of chlorine to
be accumulated at the steam/water interface facilitates
the process of chloride cracking even if the chlorine-ion
content is <0.1 mg/l. The additional, although not so
strong chloride source, is transmutation of sulfur into
chlorine and possible chlorine release from decaying
manganese sulphides [5]. 3) Sulfur poisoning of grains
boundaries. Sulfur, acting as poison during electrochem-
ical processes and decreasing the rate of repassivation
after oxide film rupture, facilitates the development of
73
local anode processes. Coolant and sulfides unbound
sulfur are probably the main sulfur source. The addi-
tional source of sulfur can be radiation-induced MnS
sulfide decay as shown in work [5]. In some places of
the fracture surface sulfur content reaches 2.7%. 4) Cop-
per effect. It is known [1,2] that bivalent copper ions fa-
cilitate chloride corrosion cracking. Besides ~0.3% cop-
per present in steel it was also present in coolant (up to
0.03 mg/l). It was enough for copper concentration reach-
ing 6.2% on the fracture surface and in the near boundary
areas.
Contiguous factors: 1) enrichment of grain bound-
aries with carbon and oxygen, the content of which
reaches 50 atomic %. 2) Effect of steam/water separa-
tion boundary. The alternative cooling with water un-
derheated by 5-7oC up to a boiling temperature and with
water-steam mixture was observed at the investigated
tube section during the reactor operation. Therefore, the
section was located at the steam/water interface that is
characterized by the greatest danger for corrosion crack-
ing resulting from chlorides accumulation at multiple
dryouts and facilitation of oxygen access.
3.3 ASS IAISCC MECHANISM
The analysis of literature and obtained results with-
out any claim to originality enables us to assume the
ASS IAISCC mechanism in the following way.
At the primary stage, when cracks, pores and other
defects are present on the metal oxide film the rate of
general corrosion process is relatively high. The gener-
ated slightly insoluble corrosion products gradually heal
the defects of oxide film on the grains and decrease the
area and number of anode sections. The effect of anodes
gradually propagates to the sensitive grain boundaries.
This period is considered as incubation period [1,2].
Then the process proceeds at a low rate suitable for
the anode dissolution rate in passive state. The maxi-
mum rate is observed in the areas with minimum
chromium concentration and as a result of it the corro-
sion has a nodular character. As the intergranular cracks
propagate, the corrosion products obstruct the access of
fresh portions of water/steam to the anode sections and
removal of metal ions in the opposite direction. More-
over the access of cathode depolarizer (oxygen) is ob-
structed. As a result, the cathode process is transferred
onto the walls of intergranular channels close to the sur-
face [1,2].
Gradually the poisoning process of grain boundaries
with sulfur and chloride released from coolant and from
radioactive decay of MnS manganese sulfides becomes
stronger. Copper precipitation begins on the surfaces of
the generated intergranular cracks. Complex aggrega-
tions on the basis of sulfur and copper are being gener-
ated. They consist of iron, chromium, nickel, carbon and
oxygen also.
The effect of radiation-induced processes of grain
boundaries depletion in chromium and nickel with their
simultaneous enrichment in titanium, phosphorus and
silicon increases. The chemical composition of bound-
ary areas differs more and more from the matrix chemi-
cal composition. Under the effect of the residual tensile
welded stresses the intergranular crack nuclei grow in
width and length. At the final stage they reach the
length of some tens of millimeters and form one main
cross crack. The process of intergranular corrosion
cracking results into the transverse fracture of the tube.
4. CONCLUSIONS
Х18Н10Т ASS tube was subjected to elementary
and structural analysis after 30 years of operation in
VK-50 at a temperature of 285…250ОC in the
steam/water mixture. The tube was fractured in the
welded joint area where the neutron flux was 3.0⋅ 1021
n/cm2 (E>0.1 MeV). The following conclusions can be
drawn according to the obtained results:
1. The tube was fractured along the main cross crack
as a result of intergranular corrosion cracking in the
welded joint area where the generation of crack network
to be 140 mm in length was revealed.
2. The mechanical properties of the tube in the base
metal adjacent zone remained high but the
transcrystalline corrosion took place at high necking of
specimens. The transmission electron microscope
investigations revealed the precipitation of Ме23С6
and/or Ме6С carbides, TiC titanium carbides to be
several hundreds nanometers in size, G-phase of the
medium size 8 µm as well as generation of dislocation
loops of the average diameter 9µm.
3. The principal factors of the intergranular
corrosion cracking are: 1) residual tensile welded
stresses; 2) sensitization of austenite in the welded joint
area during welding; 3) radiation-induced low-
temperature sensitization in irradiation; 4) radiation-
induced segregation and depletion processes at the grain
boundaries.
4. Chemical composition of grains and their adjacent
areas to be 1µm in width after irradiation had nothing in
common with the initial chemical composition of
matrix. They depleted in chromium and nickel greatly
and enriched in titanium. As a rule the width of zones is
different for different elements and varies between 0.2
and 1 µm.
5. The wide application of Auger-spectroscopy
followed by construction of the elemental maps of the
fracture surface, raster and point elemental spectra
allowed us to link all data together with the help of
quantitative analysis of the elemental and structural
peculiarities at the grains boundaries.
REFERENCES
1. I.I. Vasilenko & al. Corrosion cracking of steels К.:,
«Naukova Dumka», 1977.
2. V.P. Pogodin & al. Intergranular corrosion and cor-
rosion cracking of stainless steels in water environment
М.:, Atomizdat, 1970.
3. G.G. Ulig, R.U. Revi. Corrosion and corrosion con-
trol. Introduction into corrosion science and technolo-
gy. L.: Chemistry, 1989.
4. A.A. Nazarov. Susceptibility of steel to intergranular
corrosion and current methods of its evaluation.// Re-
view. «Prometey». 1991.
5. F.A.Garner, L.R.Greenwood, H.M.Chung. Irradia-
tion-induced instability of MnS precipitates and its pos-
sible contribution to IASCC in light water reactors.
//Proc. of 8-th Internat. Sympos. on Environm. Degra-
74
dation of Materials in Nuclear Power Systems–Water
Reactors. August 10…14, p.857-860.
6. O.Wachten, U.Wesseling, J.Bruns, R.Kilian, A.Roth.
Crack initiation in the Nb-stabilized austenitic steel
(A347) in the core shroud and top and core guide of a
German boiling water reactor – description of the extent
of the damage and explanation of its causes. //Proc. of
8-th Internat. Sympos. on Environm. Degradation of
Materials in Nuclear Power Systems–Water Reactors.
August 10…14, 1997, USA, Florida, p. 725-733.
7. J.F.Williams, P.Spellward, J.Walmsley, T.R.Mager,
M.Koyama, H.Mimaki, I.Suzuki. Microstructural effects
in austenitic stainless steel materials irradiated in a
pressurized water reactor. //Proc. of 8-th Internat.
Sympos. on Environm. Degradation of Materials in
Nuclear Power Systems–Water Reactors. August 10…
14, 1997, USA, Florida, Refer, p. 812-822
75
УДК 669.018.2
Introduction
Raster
Fe
|
| id | nasplib_isofts_kiev_ua-123456789-79290 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 1562-6016 |
| language | English |
| last_indexed | 2025-12-07T15:28:18Z |
| publishDate | 2002 |
| publisher | Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| record_format | dspace |
| spelling | Goncharenko, Yu.D. Kazakov, V.A. Shamardin, V.K. Pechyorin, A.M. Filyakin, G.V. Ostrovsky, Z.Ye. 2015-03-30T09:53:06Z 2015-03-30T09:53:06Z 2002 Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture / Yu.D. Goncharenko, V.A. Kazakov, V.K. Shamardin, A.M. Pechyorin, G.V. Filyakin, Z.Ye. Ostrovsky // Вопросы атомной науки и техники. — 2002. — № 6. — С. 69-75. — Бібліогр.: 7 назв. — англ. 1562-6016 https://nasplib.isofts.kiev.ua/handle/123456789/79290 669.018.2 Ціль запропонованої роботи – отримати результати для того, щоб зробити питомий внесок у розуміння механізму радіаційного обумовленного міжзеренного корозійного розтріскування аустенітних нержавіючих сталей. Цель предложенной работи – получение новых результатов для того, чтобы внести конструктивный вклад в понимание механизма радиационно-обусловленного межзеренного коррозионного растрескивания аустенитных сталей. The objective is to obtain the new results in order to make a constructive contribution into understanding of the mechanism of austenitic stainless steels (ASS) irradiation-assisted intergranular stress corrosion cracking (IAISCC). en Національний науковий центр «Харківський фізико-технічний інститут» НАН України Вопросы атомной науки и техники Материалы реакторов на тепловых нейтронах Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture Article published earlier |
| spellingShingle | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture Goncharenko, Yu.D. Kazakov, V.A. Shamardin, V.K. Pechyorin, A.M. Filyakin, G.V. Ostrovsky, Z.Ye. Материалы реакторов на тепловых нейтронах |
| title | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| title_full | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| title_fullStr | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| title_full_unstemmed | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| title_short | Irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| title_sort | irradiation-assisted intergranular stress corrosion cracking of austenitic stainless steel in steam-water mixture |
| topic | Материалы реакторов на тепловых нейтронах |
| topic_facet | Материалы реакторов на тепловых нейтронах |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/79290 |
| work_keys_str_mv | AT goncharenkoyud irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture AT kazakovva irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture AT shamardinvk irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture AT pechyorinam irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture AT filyakingv irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture AT ostrovskyzye irradiationassistedintergranularstresscorrosioncrackingofausteniticstainlesssteelinsteamwatermixture |