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
F. A. Garner1, S. I. Porollo2, Yu. V. Konobeev2 and O. P. Maksimkin3
Pacific Northwest National Laboratory, Richland, WA, USA
Institute of Physics and Power Engineering, Obninsk, Russian Federation
Institute of Nuclear Physics, Almaty, Kazakhstan Republic
Keywords: void swelling, stainless steels, dpa rate, temperature, LWRs, fast reactors
Abstract VVER-1000, BN-600) are made of type X18H9 or X18H9T
(18Cr-9Ni or 18Cr-10Ni- i, analogs of AISI 304 and AISI 321,
In the last decade the PWR community has become aware of the respectively) austenitic stainless steels. In Western PWRs and
potential for void swelling in austenitic internal components, BWRS the steel AISI 304 (with composition similar to 18Cr-9Ni)
especially those constructed from AISI 304 and AISI 316 stainless is used for such purposes. Soviet-design fast reactors also use
steels. Predictive equations for swelling of these steels were 18Cr-10Ni- i as a pressure vessel material, whereas Western-
usually developed from data derived in irradiations conducted in design reactors use low-alloy ferritic steels. For purposes where
high flux fast reactors. A number of recent studies have shown, higher strength is required (bolts, springs) cold-worked steels such
however, that swelling increases at a given dose and temperature as AISI 316 are employed in the West while cold-worked
as the dose rate decreases to levels characteristic of PWR X16H11Mo3 and similar steels are used in Soviet reactors.
internals. Most importantly, the lower temperature limit of
swelling appears to be on the order of 300ºC, i.e. at temperatures Decreasing displacement rates have been shown not only to cause
relevant to a large portion of most PWR and BWR internals. an earlier onset of swelling with dose at all reactor-relevant
temperatures, but also to induce swelling that extends to much
Several new studies are presented here that provide additional lower than previously expected temperatures. Based on some
insight on void swelling at lower dpa rates, focusing on the earlier studies [6-8] Garner and coworkers predicted that
Russian analogs of AISI 321 and 316 stainless steel that were austenitic steels serving as internal components in PWRs might
irradiated in two different fast reactors. In the 321 and 316 analog exhibit high levels of void swelling and the potential for severe
steels, voids were observed at irradiation temperatures as low as void-induced embrittlement [9-11]. Even more importantly, it was
281oC and doses of 0.65 and 1.3 dpa, respectively, when concluded that high dose data derived from in-core regions of
irradiated in BN-350 at 3.9x10-9 dpa/sec. In the 321 analog, high flux fast reactors would strongly under-predict the swelling
swelling of ~0.1% was observed at 0.6 dpa when the irradiation that would arise at lower dpa rates characteristic of PWR baffle-
proceeded in BR-10 at 350-430ºC and only 1.9 10-9 dpa/sec. former assemblies, BWR shrouds, out-of-core regions of fast
reactors and some components of proposed fusion devices.
A number of recent studies by Garner and various Japanese, Russian
Void swelling of austenitic internals has been identified as an and Kazakh workers have shown that void swelling in austenitic
issue with potential to influence license extension for pressurized stainless steels strongly increases at lower dpa rates [12-24], often
water reactors, both Western PWRs [1, 2] and Russian VVERs [3, allowing the observation of the lower swelling temperature limit (<300ºC)
4]. Due to the very low dose (2-3 dpa maximum) expected in the at very low dpa levels. This increased swelling arises from a decrease in
shrouds of BWRs, void swelling per se is not expected to be the duration of the transient regime of swelling at lower dpa rates. As the
license extension issue for BWRs. dpa rate goes below ~10-8 dpa/sec the transient regime approaches 0 dpa.
Recently, other researchers have also reported “cavities” produced at low
There is a growing body of evidence that shows that a decrease in dpa rates and low temperatures in PWR-irradiated austenitic components
atomic displacement rate to levels characteristic of PWR internals [25-28]. Most significantly, measurable swelling has been observed in
leads to larger swelling levels than would be predicted using data PWR cold-worked 316 baffle bolts [20, 26], even though it is expected that
generated at much higher displacement levels characteristic of fast such bolts should swell less than the 304 plates in which they are
reactors. To date the potential for significant amounts of swelling embedded.
( 5%) appears to be concentrated in small volumes of the
reentrant corners of PWR baffle-former assemblies constructed In a previous paper a series of data and micrographs derived from
from AISI 304 stainless steel . Even at lower swelling levels, irradiations conducted in Soviet fast or thermal reactors demonstrated the
however, differential swelling of annealed 304 baffle-former acceleration of void swelling as the displacement rate decreased . In this
plates and cold-worked 316 baffle bolts is being considered as a paper are presented new data that further addresses the two issues, the
possible contributor to corrosion and cracking of bolts . continued flux-dependent reduction in the incubation period and the
movement of the lower boundary of the swelling regime to lower
Near-core internals of Russian power reactors (VVER-440,
TABLE 1 Irradiation conditions for sections cut from the BR-10 reactor vessel
Place of specimen Distance from core Dose, irradiation Dose rate,
cutting midplane, mm dpa temperature, dpa/s
Level of basket bottom. 425 0.35 0.64 350 1.9 10-9
Level of upper flange 1890 … … 80 ...
for 20 years at ~80ºC.
Swelling of 12X18 9T pressure vessel of BR-10 fast reactor
The 12X18 9T austenitic stainless steel is sometimes used as a Using a remote milling machine, strips 10 mm 2 mm or 7 mm 2
containment vessel for Soviet reactors that do not involve high mm were cut from the original sections in an axial direction. Then
pressurization. Examples are the BR-10, BOR-60 and BN-600 fast from these strips TEM specimens and flat specimens for
reactors. In these applications the displacement rates experienced measurements of short-term mechanical properties were prepared.
by the steel are very low, even compared to the rates experienced The mechanical measurements are reported elsewhere .
by PWR internals.
TEM specimens in the form of disks of 3 mm in diameter with a
Samples for investigation of microstructure were recently cut perforated central hole were prepared using a standard technique
from the first vessel of the BR-10 fast reactor. This vessel was employing the two-jet-polishing “STRUERS” device.
replaced by a new vessel in 1979. The first vessel was variable in Microstructural investigations were performed at an accelerating
width with a maximum outside diameter of 535 mm and a total voltage of 100 kV using a JEM-100CX electron microscope
length just over 4 m. At the location of fuel assemblies the vessel equipped with a lateral goniometer.
has the outside diameter of 366 mm and wall thickness of 7 mm.
The vessel material is 12X18H9T austenitic stainless steel in the The microstructure of the unirradiated steel at the upper flange
solution treated condition. The nominal chemical composition of 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
the steel is (wt. %): 0.12; Si 0.8; Mn 2.0; Cr at 17-20; Ni
microns. Austenitic grains, in turn, are divided into subgrains by
at 8-11; Ti 0.8.
dislocation walls with sizes ranging from ~1 to 5 microns (Figure
1). The average dislocation density is equal to (4-5) 1013 m-2. In
The first vessel was in operation for 20 years (July 1959 till
addition, twins, large TiC precipitates with mean diameter of 0.5 to
October 1979) during which the core was composed of three fuel
1 microns, and much smaller precipitates distributed uniformly and
phases, the first two with PuO2 fuel and the third with UC fuel.
at much higher density within the grains (Figure 2) were observed.
Each phase lasted ~7 years. The total reactor operation during this
The diameter of the small precipitates ranges from 50 to 60 nm,
period was 3930 days or 2562.6 effective full power days. The
total neutron fluence accumulated by the vessel at the core with their concentration at ~3 1019 m-3. An analysis of micro-
diffraction patterns obtained from these precipitates showed that
midplane was 8.44 1026 n/m2 corresponding to an exposure dose
these precipitates have the fcc-structure with the lattice parameter
of 33.1 dpa (NRT). On its inner side the vessel was in contact
of 0.43 nm, identifying them also to be TiC carbides.
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
The microstructure of the irradiated steel at the elevation of the
vessel and a safety vessel. In the first and last fuel phases the inlet
basket bottom is shown in Figures 3 and 4. Even at the low dose
temperature of the vessel was 350ºC, but during the second cycle
of 0.64 dpa the microstructure has changed significantly,
it was higher at 430 . producing a non-uniform spatial distribution of dislocation loops
(Figure 3) and voids (Figure 4). Frank dislocation loops with a
To study the microstructure cross-sectional specimens were cut
mean diameter of 33 nm and mean concentration of 3 1021 m-3 are
from two elevations of the vessel. Irradiation conditions for these
seen in extended cluster arrays (Figure 3). The size of such arrays
cross sections are shown in Table 1.
coincides with the size of sub-grains observed in the unirradiated
steel and thus it can be assumed that the dislocation loops formed
One specimen was cut from the elevation corresponding to the
preferentially on the dislocation walls separating the sub-grains.
bottom level of the fuel basket, in which the lower ends of fuel
assemblies were located. Another specimen was cut at the
The spatial distribution of voids is also rather non-uniform. Large
elevation of the upper flange of the primary coolant circuit. This
voids are located mainly in zones having high loop concentration,
second specimen was effectively unirradiated but had been aged
i.e. in the former dislocation walls (Figure 4). Smaller voids,
however, are distributed nearly uniformly throughout the grain.
As measured by microscopy the swelling of the steel is 0.1 %,
with a mean void diameter of 11 nm and concentration of 6 1020
m-3. 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
higher dpa rates on the order 1-3 x 10-7 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 . 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.
Figure 1 Microstructure of unirradiated 12 18 9 from the Figure 3 Dislocation loops in 12 18 9 steel irradiated to 0.6
upper flange of the BR-10 vessel, aged at ~80ºC for 20 years. 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.
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
produced using the thermal-mechanical treatment mentioned above.
The measured temperature at the bottom of each assembly was 280oC 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
a) 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.
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
b) working liquid. Disks of 3 mm diameter for microscopy studies
Figure 4 Voids in 12 18 9 steel irradiated to 0.6 dpa at were prepared from 300 m sections cut from the mid-section of
350/430/350ºC: ) large voids on sub-grain boundaries, b) spatial the duct face. Mechanical grinding and polishing with subsequent
distribution of smaller voids. electrochemical polishing were used for final preparation of TEM
disks. The irradiation conditions for specimens examined are
shown in Table 2.
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
Figure 5 Dependence of steel 12 18 9 swelling in BR-10 on shown in Figure7 and Table 4. Note that once again void swelling
dose and dose rate. Light circles are wrappers of fuel assemblies was observed at temperatures as low as 281ºC at 1.3 dpa. Only at
and fuel pin claddings at 430ºC, black circle is the reactor vessel 280ºC and 0.25 dpa were voids not observed.
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
Void swelling of 321 and 316 analog steels following irradiation agreement with expectations, especially with respect to the
in the BN-350 fast reactor different nickel content of the alloys. It is known that the swelling
of 300 series stainless steels is a strong function of the nickel
Two hexagonal blanket assemblies with faces 50 mm wide and 2 mm content, decreasing with increasing nickel content [6, 29, 30].
thick were irradiated in the reflector region of the BN-350 reactor,
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 BN-350
365ºC, 6.03 dpa
The specimens examined in this study were all irradiated in fast swelling in austenitic steels, as demonstrated by the BR-10 data at
reactors where both helium generation and hydrogen generation 350-430 ºC and dpa rates on the order of 10-9 dpa/sec. When the
and retention are much lower than in light water reactors [31, 32]. dpa rate lies is in the range of 10-8 to10-9 dpa/sec it becomes easier
Therefore the formation of voids is occurring in BN-350 and BR- to see that void swelling in austenitic steels exists over a
10 under conditions that are less conducive to void nucleation and temperature range that reaches lower temperatures (~280ºC) than
stabilization than found in both PWRs and BWRs. Therefore the previously expected, and occurs at doses that are very low, often
results of these studies allow us to speculate that it might be even at < 1 dpa.
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
Table 2a-- Dose and temperatures of specimens over the height of assemblies. While the swelling in PWRs may become high
the hexagonal assembly for 12Cr18Ni10Ti enough to be a potential license extension issue, the total swelling
expected in a BWR should be small enough such that it will most
Distance 12Cr18Ni10Ti likely not be designated as a life extension issue.
from midplane, Dose Damage rate Temperature
mm (dpa) (x10-8 dpa/sec) (°C) Acknowledgements
-900 0.65 0.12 281
-375 7.3 1.36 294 The Russian portion of this work was supported by the Russian
0 12.28 2.3 313 Foundation for Basic Research under the Project # 04-02-17278.
+75 12.6 2.34 318 The Kazakh portion of this work was supported by the Ministry of
+375 7.25 1.35 333 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
Table 2b-- Dose and temperatures of specimens over the height of Basic Energy Sciences, and the Office of Fusion Energy, US
the hexagonal assembly for 08Cr16Ni11Mo3 Department of Energy.
Distance 08Cr16Ni11Mo3 The authors are indebted to Natalia A. Brikotnina of Interpreter
from midplane, Dose Damage rate Temperature and Translation Services for her assistance in the conduct and
mm (dpa) (x10-8 dpa/sec) (°C) interpretation of these experiments, and for translation of original
Russian texts into English.
-1200 0.25 0.08 280
-900 1.27 0.39 281
-500 7.08 2.2 309 1. Tang, H. T. and Gilreath, J. D., “Aging Research and
0 15.6 4.85 337 Management of PWR Vessel Internals”, Proc.
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Finally, it should be noted that most previous perceptions CD format, no page numbers.
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dependence of the lower temperature limit of swelling were Characterization on Baffle Bolts” Proc. Fontevraud 5,
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upon extrapolation imply that swelling ceases somewhere Irradiation Effects”, in Russian, Proceedings of 5th
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from the inlet temperature and the strong flux gradient near the Science,Dimitrovgrad, Russia, 8-12 September 1997,
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Table 3 Microstructural data on cavities for irradiated stainless steel 12Cr18Ni10Ti
Distance from Range for void Mean void Peak void diameter, Void density,
midplane, sizes, nm diameter, nm 1015cm-3 Swelling,
mm 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 Range for void Mean void Peak void diameter, Void density,
midplane, sizes, nm diameter, nm 1015cm-3 Swelling,
Mm nm %
-1200 - - - no voids -
-900 <7 - - Some -
-500 10 -15 10.0 8.0 0.61 0.04
0 4 -15 8.6 10.0 2.57 0.13
+500 10 -35 14.0 10.0 0.78 0.16
Figure 8 Comparison of swelling data on annealed austenitic steel 18Cr-10Ni- i derived from three separate sources in two fast reactors
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International Conference on Environmental Degradation in the BN-350 Reactor, submitted to Fusion Reactor
Materials Semiannual Progress Report, July 2005,: also
to be submitted to ICFRM-12.
24. Porollo, S. I., Dvoriashin, A. M., Konobeev, Yu. V.,
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
25. 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.
26. 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.
27. 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.
28. 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
29. F. A. Garner, "Recent Insights on the Swelling and
Creep of Irradiated Austenitic Alloys," Invited Paper, J.
of Nucl. Mater., 122 and 123, (1984), pp. 459-471.
30. 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.
31. 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.
32. 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.