Proceedings of the 12th International Conference on
Environmental Degradation of Materials in Nuclear Power System – Water Reactors –
Edited by T.R. Allen, P.J. King, and L. Nelson TMS (The Minerals, Metals & Materials Society), 2005
VOID SWELLING OF AUSTENITIC STEELS IRRADIATED WITH NEUTRONS AT LOW
TEMPERATURES AND VERY LOW DPA RATES
1,
2,
2
F. A. Garner S. I. Porollo Yu. V. Konobeev and O. P. Maksimkin
3
1
2
Pacific Northwest National Laboratory, Richland, WA, USA
Institute of Physics and Power Engineering, Obninsk, Russian Federation
3
Institute of Nuclear Physics, Almaty, Kazakhstan Republic
Keywords: void swelling, stainless steels, dpa rate, temperature, LWRs, fast reactors
VVER-1000, BN-600) are made of type X18H9 or X18H9T
(18Cr-9Ni or 18Cr-10Ni- i, analogs of AISI 304 and AISI 321,
respectively) austenitic stainless steels. In Western PWRs and
BWRS the steel AISI 304 (with composition similar to 18Cr-9Ni)
is used for such purposes. Soviet-design fast reactors also use
18Cr-10Ni- i as a pressure vessel material, whereas Westerndesign reactors use low-alloy ferritic steels. For purposes where
higher strength is required (bolts, springs) cold-worked steels such
as AISI 316 are employed in the West while cold-worked
X16H11Mo3 and similar steels are used in Soviet reactors.
Abstract
In the last decade the PWR community has become aware of the
potential for void swelling in austenitic internal components,
especially those constructed from AISI 304 and AISI 316 stainless
steels. Predictive equations for swelling of these steels were
usually developed from data derived in irradiations conducted in
high flux fast reactors. A number of recent studies have shown,
however, that swelling increases at a given dose and temperature
as the dose rate decreases to levels characteristic of PWR
internals. Most importantly, the lower temperature limit of
swelling appears to be on the order of 300ºC, i.e. at temperatures
relevant to a large portion of most PWR and BWR internals.
Decreasing displacement rates have been shown not only to cause
an earlier onset of swelling with dose at all reactor-relevant
temperatures, but also to induce swelling that extends to much
lower than previously expected temperatures. Based on some
earlier studies [6-8] Garner and coworkers predicted that
austenitic steels serving as internal components in PWRs might
exhibit high levels of void swelling and the potential for severe
void-induced embrittlement [9-11]. Even more importantly, it was
concluded that high dose data derived from in-core regions of
high flux fast reactors would strongly under-predict the swelling
that would arise at lower dpa rates characteristic of PWR baffleformer assemblies, BWR shrouds, out-of-core regions of fast
reactors and some components of proposed fusion devices.
Several new studies are presented here that provide additional
insight on void swelling at lower dpa rates, focusing on the
Russian analogs of AISI 321 and 316 stainless steel that were
irradiated in two different fast reactors. In the 321 and 316 analog
steels, voids were observed at irradiation temperatures as low as
o
281 C and doses of 0.65 and 1.3 dpa, respectively, when
-9
irradiated in BN-350 at 3.9x10 dpa/sec. In the 321 analog,
swelling of ~0.1% was observed at 0.6 dpa when the irradiation
-9
proceeded in BR-10 at 350-430ºC and only 1.9 10 dpa/sec.
Introduction
A number of recent studies by Garner and various Japanese, Russian
and Kazakh workers have shown that void swelling in austenitic
stainless steels strongly increases at lower dpa rates [12-24], often
allowing the observation of the lower swelling temperature limit
(<300ºC) at very low dpa levels. This increased swelling arises from a
decrease in the duration of the transient regime of swelling at lower
dpa rates. As the dpa rate goes below ~10 -8 dpa/sec the transient
regime approaches 0 dpa. Recently, other researchers have also
reported “cavities” produced at low dpa rates and low temperatures in
PWR-irradiated austenitic components [25-28]. Most significantly,
measurable swelling has been observed in PWR cold-worked 316
baffle bolts [20, 26], even though it is expected that such bolts should
swell less than the 304 plates in which they are embedded.
Void swelling of austenitic internals has been identified as an
issue with potential to influence license extension for pressurized
water reactors, both Western PWRs [1, 2] and Russian VVERs [3,
4]. Due to the very low dose (2-3 dpa maximum) expected in the
shrouds of BWRs, void swelling per se is not expected to be
license extension issue for BWRs.
There is a growing body of evidence that shows that a decrease in
atomic displacement rate to levels characteristic of PWR internals
leads to larger swelling levels than would be predicted using data
generated at much higher displacement levels characteristic of fast
reactors. To date the potential for significant amounts of swelling
( 5%) appears to be concentrated in small volumes of the reentrant
corners of PWR baffle-former assemblies constructed from AISI
304 stainless steel [1]. Even at lower swelling levels, however,
differential swelling of annealed 304 baffle-former plates and
cold-worked 316 baffle bolts is being considered as a possible
contributor to corrosion and cracking of bolts [5].
In a previous paper a series of data and micrographs derived from
irradiations conducted in Soviet fast or thermal reactors
demonstrated the acceleration of void swelling as the
displacement rate decreased [14]. In this paper are presented new
data that further addresses the two issues, the continued fluxdependent reduction in the incubation period and the movement of
the lower boundary of the swelling regime to lower temperatures.
Near-core internals of Russian power reactors (VVER-440,
439
TABLE 1 Irradiation conditions for sections cut from the BR-10 reactor vessel
Place of specimen
cutting
Distance from core
midplane, mm
Total neutron
fluence,
10
26
n/m
2
Dose,
dpa
Average
irradiation
temperature,
°
Dose rate,
dpa/s
Level of basket bottom.
425
0.35
0.64
350
1.9 10
Level of upper flange
1890
…
…
80
...
-9
for 20 years at ~80ºC.
Swelling of 12X18 9T pressure vessel of BR-10 fast reactor
Using a remote milling machine, strips 10 mm 2 mm or 7 mm 2
mm were cut from the original sections in an axial direction. Then
from these strips TEM specimens and flat specimens for
measurements of short-term mechanical properties were prepared.
The mechanical measurements are reported elsewhere [24].
The 12X18 9T austenitic stainless steel is sometimes used as a
containment vessel for Soviet reactors that do not involve high
pressurization. Examples are the BR-10, BOR-60 and BN-600 fast
reactors. In these applications the displacement rates experienced
by the steel are very low, even compared to the rates experienced
by PWR internals.
TEM specimens in the form of disks of 3 mm in diameter with a
perforated central hole were prepared using a standard technique
employing
the
two-jet-polishing
“STRUERS”
device.
Microstructural investigations were performed at an accelerating
voltage of 100 kV using a JEM-100CX electron microscope
equipped with a lateral goniometer.
Samples for investigation of microstructure were recently cut
from the first vessel of the BR-10 fast reactor. This vessel was
replaced by a new vessel in 1979. The first vessel was variable in
width with a maximum outside diameter of 535 mm and a total
length just over 4 m. At the location of fuel assemblies the vessel
has the outside diameter of 366 mm and wall thickness of 7 mm.
The vessel material is 12X18H9T austenitic stainless steel in the
solution treated condition. The nominal chemical composition of
the steel is (wt. %): 0.12; Si 0.8; Mn 2.0; Cr at 17-20; Ni at 8-11;
Ti 0.8.
The microstructure of the unirradiated steel at the upper flange
elevation is shown in Figures 1 and 2. It is observed that the steel
had the anticipated austenitic structure with a grain size of ~10-20
microns. Austenitic grains, in turn, are divided into subgrains by
dislocation walls with sizes ranging from ~1 to 5 microns (Figure
13 -2
1). The average dislocation density is equal to (4-5) 10 m . In
addition, twins, large TiC precipitates with mean diameter of 0.5 to
1 microns, and much smaller precipitates distributed uniformly and
at much higher density within the grains (Figure 2) were observed.
The diameter of the small precipitates ranges from 50 to 60 nm,
19 -3
with their concentration at ~3 10 m . An analysis of microdiffraction patterns obtained from these precipitates showed that
these precipitates have the fcc-structure with the lattice parameter
of 0.43 nm, identifying them also to be TiC carbides.
The first vessel was in operation for 20 years (July 1959 till
October 1979) during which the core was composed of three fuel
phases, the first two with PuO2 fuel and the third with UC fuel.
Each phase lasted ~7 years. The total reactor operation during this
period was 3930 days or 2562.6 effective full power days. The
total neutron fluence accumulated by the vessel at the core
26
2
midplane was 8.44 10 n/m corresponding to an exposure dose
of 33.1 dpa (NRT). On its inner side the vessel was in contact
with sodium coolant flowing from bottom to top, but on its outer
side it was in contact with air contained in the gap between the
vessel and a safety vessel. In the first and last fuel phases the inlet
temperature of the vessel was 350ºC, but during the second cycle
it was higher at 430 .
The microstructure of the irradiated steel at the elevation of the basket
bottom is shown in Figures 3 and 4. Even at the low dose of 0.64 dpa
the microstructure has changed significantly, producing a nonuniform spatial distribution of dislocation loops (Figure 3) and voids
(Figure 4). Frank dislocation loops with a mean diameter of 33 nm
and mean concentration of 3 1021 m-3 are seen in extended cluster
arrays (Figure 3). The size of such arrays coincides with the size of
sub-grains observed in the unirradiated steel and thus it can be
assumed that the dislocation loops formed preferentially on the
dislocation walls separating the sub-grains.
To study the microstructure cross-sectional specimens were cut
from two elevations of the vessel. Irradiation conditions for these
cross sections are shown in Table 1.
One specimen was cut from the elevation corresponding to the
bottom level of the fuel basket, in which the lower ends of fuel
assemblies were located. Another specimen was cut at the
elevation of the upper flange of the primary coolant circuit. This
second specimen was effectively unirradiated but had been aged
The spatial distribution of voids is also rather non-uniform. Large
voids are located mainly in zones having high loop concentration,
i.e. in the former dislocation walls (Figure 4). Smaller voids,
however, are distributed nearly uniformly throughout the grain.
440
As measured by microscopy the swelling of the steel is 0.1 %,
20
with a mean void diameter of 11 nm and concentration of 6 10
-3
m . Precipitates observed in the irradiated steel were essentially
identical to those in the unirradiated steel.
One can compare in Figure 5 the swelling observed for the
pressure vessel with that of wrappers and pin cladding of BR-10
fuel assemblies made from the same steel but irradiated in-core at
-7
higher dpa rates on the order 1-3 x 10 dpa/sec. The data base on
swelling of the steel was obtained from examination of wrappers
and fuel pins of the BR-10 reactor when where the inlet sodium
temperature was equal to 430 [14]. For this comparison only
swelling data derived from bottom of the wrappers and claddings
were selected to insure an essentially isothermal data set. The
larger BR-10 in-core data base including these data will be
presented in Figure 8.
a)
b)
Figure 1 Microstructure of unirradiated 12 18 9 from the upper
flange of the BR-10 vessel, aged at ~80ºC for 20 years.
Figure 3 Dislocation loops in 12 18 9 steel irradiated to 0.6 dpa at
350/430/350ºC: ) general view, b) dislocation loop cluster along
preexisting sub-grain boundaries.
The 430ºC data for cladding and wrappers are shown as a function
of dose in Figure 5 together with the single datum for the vessel. It
is seen from Figure 5 that after the incubation dose of 4-7 dpa the
swelling of the wrapper and cladding at ~430 is an approximately
linear function of dose with the swelling rate of 0.08 to 0.13
%/dpa. In general, one would expect that the vessel specimen,
which spent two-thirds of its life at 350ºC and only one-third at
430ºC, would swell less because of its average lower temperature,
but the swelling of the vessel steel is higher than expected (~0.1
% at only 0.64 dpa) than one would anticipate based on the
extrapolation of the 430ºC curve to 0.6 dpa.
On the basis of this one non-isothermal comparison alone, a clear
effect of lower dpa rate to accelerate the onset of swelling can not
be conclusively demonstrated. When combined with larger data
base on flux-affected swelling cited earlier, however, this
comparison is consistent with the previously observed strong
effect of dpa rate on void swelling in 300 series stainless steels.
Figure 2 Dislocations and TiC-precipitates from the unirradiated
upper flange specimen, aged at ~80ºC for 20 years.
441
reaching a maximum of 12.6 and 15.6 dpa at average maximum dpa rates
of 3.8 to 4.9 10-8 dpa/sec averaged over their lifetime in reactor. The first
duct was constructed from 12Cr18Ni10Ti stainless steel, a Soviet analog
of AISI 321 steel, and was produced with the final thermal-mechanical
treatment of the duct being 15-20% cold deformation followed by
annealing at 800oC for 1 hour. The second duct was constructed from
08Cr16Ni11Mo3, a Soviet analog of AISI 316 stainless steel. It was also
producedusing the thermal-mechanical treatmentmentioned above.
The measured temperature at the bottom of each assembly was 280 oC and
the calculated temperature at the top of the 321 analog assembly was
430oC and was 420ºC for the 316 analog assembly. Specimens were
chosen for examination between elevations having LWR-relevant
temperatures between 280 and 333ºC for the 321 analog. For the highest
elevation location chosen for the 316 analog the calculated temperature
was 365ºC. Due to the thinness of the duct wall, the internal temperature
was not raised significantly by gamma heating. Thus, the temperature of
the steel is expected to be within 1-2 ºC of the local coolant temperature.
The temperatures were relatively constant through the irradiation with the
calculational uncertainty very small at the lower end of the duct at 280 ºC,
rising to perhaps ±5ºC at the highest temperature elevation examined,
which was 365 ºC.
a)
At the BN-350 site specimens with 10 mm height and 50 mm width were
cut from the duct walls at various locations. Subsequent reduction of these
specimens was conducted in a hot cell at INP-Almaty for microstructural
analysis and microhardness measurements. Plate-shape specimens with
sizes of 5 6 mm were prepared for metallography investigations,
microhardness measurements and hydrostatic weighing. To date only the
microscopy examination has been completed.
The examination technique involved transmission electron
microscopy (TEM) using a JEM-100CX electron microscope
operating at 100 keV. The density was measured using an
immersion density technique employing a CEPN-770 electronic
balance with methyl alcohol as the working liquid. Disks of 3 mm
diameter for microscopy studies were prepared from 300 m
sections cut from the mid-section of the duct face. Mechanical
grinding and polishing with subsequent electrochemical polishing
were used for final preparation of TEM disks. The irradiation
conditions for specimens examined are shown in Table 2.
b)
Figure 4 Voids in 12 18 9 steel irradiated to 0.6 dpa at
350/430/350ºC: ) large voids on sub-grain boundaries, b) spatial
distribution of smaller voids.
Microscopy examination confirms the presence of void swelling
of the 321 analog in the range 281 to 333ºC, as shown in Table 3
and Figure 6. Most significantly, it is seen that even at 0.65 dpa
and 281ºC voids are clearly visible, adding additional support to
the growing body of evidence that swelling extends down to
unexpectedly low temperatures and low doses if the displacement
rate is low enough. Note that there is essentially no uncertainty in
this temperature, being defined by the inlet temperature.
Similar swelling indications were observed in the 316 analog, as
shown in Figure7 and Table 4. Note that once again void swelling
was observed at temperatures as low as 281ºC at 1.3 dpa. Only at
280ºC and 0.25 dpa were voids not observed.
Figure 5 Dependence of steel 12 18 9 swelling in BR-10 on dose
and dose rate. Light circles are wrappers of fuel assemblies and
fuel pin claddings at 430ºC, black circle is the reactor vessel at
350/430/350ºC. Displacement rates are shown for each data set.
In general the swelling at near-comparable conditions was larger
in the 321 analog compared to that in the 316 analog, is in general
agreement with expectations, especially with respect to the
different nickel content of the alloys. It is known that the swelling
of 300 series stainless steels is a strong function of the nickel
content, decreasing with increasing nickel content [6, 29, 30].
Void swelling of 321 and 316 analog steels following irradiation
in the BN-350 fast reactor
Two hexagonal blanket assemblies with faces 50 mm wide and 2 mm
thick were irradiated in the reflector region of the BN-350 reactor,
442
443
0.1 m
65 nm
280ºC, 0.25 dpa
281ºC, 1.27 dpa
0.1 m
0.1 m
309ºC, 7.08 dpa
337ºC, 15.6 dpa
Fig. 7 Microstructure of AISI 316
analog steel irradiated in the BN350 reactor.
0.1 m
365ºC, 6.03 dpa
444
swelling in austenitic steels, as demonstrated by the BR-10 data at
-9
350-430 ºC and dpa rates on the order of 10 dpa/sec. When the
-8
-9
dpa rate lies is in the range of 10 to10 dpa/sec it becomes
easier to see that void swelling in austenitic steels exists over a
temperature range that reaches lower temperatures (~280ºC) than
previously expected, and occurs at doses that are very low, often
at < 1 dpa.
The specimens examined in this study were all irradiated in fast
reactors where both helium generation and hydrogen generation
and retention are much lower than in light water reactors [31, 32].
Therefore the formation of voids is occurring in BN-350 and BR10 under conditions that are less conducive to void nucleation and
stabilization than found in both PWRs and BWRs. Therefore the
results of these studies allow us to speculate that it might be even
easier to form voids in LWR austenitic components, especially for
very low dose rate conditions found far from the reactor core.
Since the inlet temperatures of both BWRs and PWRS are in this
range this implies that a low level of voids will probably be
observed in BWR shrouds as well as in PWR baffle-former
assemblies. While the swelling in PWRs may become high
enough to be a potential license extension issue, the total swelling
expected in a BWR should be small enough such that it will most
likely not be designated as a life extension issue.
Table 2a-- Dose and temperatures of specimens over the height of
the hexagonal assembly for 12Cr18Ni10Ti
Distance
from midplane,
mm
-900
-375
0
+75
+375
12Cr18Ni10Ti
Damage rate
-8
(x10 dpa/sec)
0.12
1.36
2.3
2.34
1.35
Dose
(dpa)
0.65
7.3
12.28
12.6
7.25
Temperature
(°C)
281
294
313
318
333
Acknowledgements
The Russian portion of this work was supported by the Russian
Foundation for Basic Research under the Project # 04-02-17278.
The Kazakh portion of this work was supported by the Ministry of
Energy and Mineral Resources of the Republic of Kazakhstan,
and under ISTC project number K-437. The US portion was
jointly sponsored by the Materials Science Branch, Office of
Basic Energy Sciences, and the Office of Fusion Energy, US
Department of Energy.
Table 2b-- Dose and temperatures of specimens over the height of
the hexagonal assembly for 08Cr16Ni11Mo3
Distance
from midplane,
mm
Dose
(dpa)
-1200
-900
-500
0
+500
0.25
1.27
7.08
15.6
6.03
The authors are indebted to Natalia A. Brikotnina of Interpreter
and Translation Services for her assistance in the conduct and
interpretation of these experiments, and for translation of original
Russian texts into English.
08Cr16Ni11Mo3
Damage rate
Temperature
-8
(x10 dpa/sec)
(°C)
0.08
0.39
2.2
4.85
1.87
280
281
309
337
365
References
1.
Finally, it should be noted that most previous perceptions
concerning the lower boundary of void swelling and the fluxdependence of the lower temperature limit of swelling were
established using reactors with relatively high inlet temperatures,
such as 350ºC in BR-10 and 365-370ºC in FFTF and EBR-II. As
shown in Figure 8, when swelling data on the same steel are
compiled from reactors with different inlet temperatures and from
data derived from both fueled and unfueled zones, then the
apparent lower limit of swelling moves toward the lowest inlet
temperature. Thus the previously published BR-10 in-core data
upon extrapolation imply that swelling ceases somewhere
between 400 to 430ºC [14], but this is a misperception arising
from the inlet temperature and the strong flux gradient near the
bottom of the core. Swelling actually develops down to
significantly lower temperatures, as seen in both the vessel
specimen and especially in specimens taken from the reflector
region of BN-350 with its lower inlet temperature of 280ºC [15].
2.
3.
4.
Conclusions
In agreement with many previous studies it appears that
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445
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mm
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Range for void
Mean void
Peak void diameter, Void density,
15 -3
sizes, nm
diameter,
nm
10 cm
Swelling,
nm
%
-900
<5-12
7.7
<5nm / 5-10
0.84
0.03
-375
<10 - 15
11.6
10
0.47
0.05
0
<10 - 20
11.2
10
2.9
0.25
+75
<8 -18
9.0
8
8.2
0.33
+375
<10 -35
15.3
15
1.0
0.23
Table 4. Microstructural data on cavities for irradiated stainless steel 08Cr16Ni11Mo3
Distance from
midplane,
Mm
-1200
-900
Range for void
sizes, nm
<7
Mean void
diameter,
nm
-
-500
0
+500
10 -15
4 -15
10 -35
10.0
8.6
14.0
Peak void diameter, Void density,
15 -3
nm
10 cm
-
8.0
10.0
10.0
no voids
Some
scattered
voids
0.61
2.57
0.78
Swelling,
%
-
0.04
0.13
0.16
BR-10
in-core
data
BN-350
reflector
BR-10
vessel
Figure 8 Comparison of swelling data on annealed austenitic steel 18Cr-10Ni- i derived from three separate sources in two fast reactors
[14-15]. The BR-10 data shown at 430ºC was presented earlier in Fig. 5. The added datum of the BR-10 vessel is shown at 350ºC only for
convenience.
446
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Temperatures in Annealed Fe-18Cr-10Ni-Ti Stainless
Steel Irradiated at High and Low Dpa Rates in BOR60", International Symposium on "Contribution of
Materials Investigation to the Resolution of Problems
Encountered in Pressurized Water Reactors, 14-18 Sept.
1998, Fontevraud, France, pp. 261-269.
17. Neustroev, V. S., Shamardin, V. K., Ostrovsky, Z. E.,
Pecherin, A. M., and Garner, F. A., "Temperature-Shift
of Void Swelling Observed at PWR-Relevant
Temperatures in Annealed Fe-18Cr-10Ni-Ti Stainless
Steel Irradiated in the Reflector Region of BOR-60",
Effects of Radiation on Materials: 19th International
Symposium, ASTM STP 1366, M. L. Hamilton, A. S.
Kumar, S. T. Rosinski and M. L. Grossbeck, Eds.,
American Society for Testing and Materials, pp. 792800 (2000).
18. Okita, T., Sato, T., Sekimura, N., Garner, F. A. and
Greenwood, L. R., “The Primary Origin of Dose Rate
effects on Microstructural Evolution of Austenitic
Alloys during Neutron Irradiation”, Journal of Nuclear
Materials 207-211 (2002) 322-326.
19. Okita, T., Sato, T., Sekimura N., Garner, F. A., and
Wolfer, W. G. “ Combined Effect of Temperature,
Displacement Rate and Composition on the NeutronInduced Swelling of Austenitic Alloys”, 11th
International Conference on Environmental Degradation
of Materials in Nuclear Power Systems – Water
Reactors, August 2003, issued on CD format, no page
numbers.
20. Edwards, D. J., Simonen, E. P., Garner, F. A., Oliver, B.
A., and Bruemmer, S. M., “Sensitivity of
Microstructural Evolution Due to Temperature and
Dose Gradients in Neutron-Irradiated 316SS”, Journal
of Nuclear Materials 317 (2003) 32-45.
21. Budylkin, N. I., Bulanova, T. M., Mironova, E. G., N.
M. Mitrofanova, N. M., Porollo, S. I., Chernov, V. M.,
Shamardin, V. K. and Garner, F. A., “The Strong
Influence of Displacement Rate on Void Swelling in
Variants of Fe-16Cr-15Ni-3Mo Austenitic Stainless
Steel Irradiated in BN-350 and BOR-60”, Journal of
Nuclear Materials 329-333 (2004) 621-624.
22. Maksimkin, O. P., Tsai, K. V., Turubarova, L. G.,
Doronina, T., and Garner, F. A., “Characterization of
08Cr16Ni11Mo3 Stainless Steel Irradiated in the BN350 Reactor,” Journal of Nuclear Materials 329-333
(2004) 625-629.
23. Maksimkin, O. P., Tsai, K. V., Turubarova, L. G.,
Doronina, T., and Garner, F. A., “Void Swelling of AISI
321 Analog Stainless Steel Irradiated at Low dpa Rates
in the BN-350 Reactor, submitted to Fusion Reactor
447
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Ivanov, A. A., Shulepin, S. V. and Garner, F. A.,
“Microstructure and Mechanical Properties of
Austenitic Stainless Steel 12X18H9T Irradiated in the
Pressure Vessel of BR-10 at Very Low Displacement
Rates”,submitted to J. Nuclear Materals: also submitted
to Fusion Materials Semiannual Progress Report, July
2005.
Fujii, K., Fukuya, K., Furutani, G., Torimaru, T.,
Kohyama A. and Katoh, Y., “Swelling in 316 Stainless
Steel Irradiated in a PWR”, 10th Int. Conf. on
Environmental Degradation of Materials in Nuclear
Power Systems – Water Reactors, issued in CD format,
no page numbers.
Thomas, L. E. and Bruemmer, S. M., “Analytical
Transmission Electron Microscopy Characterization of
Stress Corrosion Cracks in an Irradiated Type 316
Stainless Steel Core Component”, Proc. Fontevraud 5,
Contribution of Materials Investigation to the
Resolution of Problems Encountered in Pressurized
Water Reactors, 23-27 September, 2002, paper #60, on
CD format, no page numbers.
Edwards, D. J., Simonen, E. P., Bruemmer S. M, and
Efsing, P.,”Microstructural Evolution in Neutron
Irradiated Steel: Comparison of LWR and Fast-Reactor
Irradiations,” in these proceedings.
Fujimoto, K., Yonezawa , T., Wachi , E., Yamaguchi ,
Y., Nakano, M., , Shogan , R. P., Massoud , J. P. and
Mager, T. R., “Effect of the Accelerated Irradiation and
Nuclear Transmuted Gas on IASCC Characteristics for
Highly Irradiated Austenitic Stainless Steels,” in these
proceedings.
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Creep of Irradiated Austenitic Alloys," Invited Paper, J.
of Nucl. Mater., 122 and 123, (1984), pp. 459-471.
F. A. Garner and H. R. Brager, "Swelling of Austenitic
Fe-Cr-Ni Ternary Alloys During Fast Neutron
Irradiation," Effects of Radiation on Materials: Twelfth
International Symposium, ASTM STP 870, F. A.
Garner and J. S. Perrin, Eds., ASTM, Philadelphia, PA,
1985, pp. 187-201.
F. A. Garner, B. M. Oliver, L. R. Greenwood, D. J.
Edwards, S. M. Bruemmer and M. L. Grossbeck,
“Generation and Retention of Helium and Hydrogen in
Austenitic Steels Irradiated in a Variety of LWR and
Test Reactor Spectral Environments”, 10th International
Conference on Environmental Degradation of Materials
in Nuclear Power Systems – Water Reactors, 2001,
issued on CD format, no page numbers.
F. A. Garner and L.R. Greenwood, “Survey of Recent
Developments Concerning the Understanding of
Radiation Effects on Stainless Steels Used in the LWR
Power Industry,” 10th International Conference on
Environmental Degradation of Materials in Nuclear
Power Systems – Water Reactors, 2003, pp. 887-909.
448