Document Sample
					                                                                                      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


                                      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 [1]. 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 [14]. 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 [5].                 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

                                                                     Total neutron
              Place of specimen          Distance from core                                 Dose,         irradiation       Dose rate,
                   cutting                 midplane, mm                                      dpa         temperature,        dpa/s
                                                                      1026 n/m2

           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 [24].
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 [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.

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

                  0.1 m

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.
     +500              6.03            1.87             365                            Fontevraud 5, Contribution of Materials Investigation to
                                                                                       the Resolution of Problems Encountered in Pressurized
                                                                                       Water Reactors, 23-27 September, 2002, paper #19, on
Finally, it should be noted that most previous perceptions                             CD format, no page numbers.
concerning the lower boundary of void swelling and the flux-                      2.   Bryne, S. T., Wilson I. and Shogan, R., “Microstructural
dependence of the lower temperature limit of swelling were                             Characterization on Baffle Bolts” Proc. Fontevraud 5,
established using reactors with relatively high inlet temperatures,                    Contribution of Materials Investigation to the
such as 350ºC in BR-10 and 365-370ºC in FFTF and EBR-II. As                            Resolution of Problems Encountered in Pressurized
shown in Figure 8, when swelling data on the same steel are                            Water Reactors, 23-27 September, 2002, paper #96, on
compiled from reactors with different inlet temperatures and from                      CD format, no page numbers.
data derived from both fueled and unfueled zones, then the                        3.   Troyanov, V.M., Likhachev, Yu.I., Khmelevsky, M.Ya.
apparent lower limit of swelling moves toward the lowest inlet                         et al.., “Evaluation and Analysis of Thermo-mechanical
temperature. Thus the previously published BR-10 in-core data                          Behavior of Internals of VVERs Taking into Account
upon extrapolation imply that swelling ceases somewhere                                Irradiation Effects”, in Russian, Proceedings of 5th
between 400 to 430ºC [14], but this is a misperception arising                         Russian       conference    on     Reactor    Materials
from the inlet temperature and the strong flux gradient near the                       Science,Dimitrovgrad, Russia, 8-12 September 1997,
bottom of the core. Swelling actually develops down to                                 Vol.2, Part I, pp. 3-18.
significantly lower temperatures, as seen in both the vessel                      4.   Neustroev, V.S., Golovanov, V.N., Shamardin, V.K.,
specimen and especially in specimens taken from the reflector                          Ostrovsky, Z.E., and Pecherin,         . ., “Irradiation
region of BN-350 with its lower inlet temperature of 280ºC [15].                       Phenomena in 18 10 Steel Irradiated in Different
                                                                                       Reactors under Conditions Relevant to Operation
                             Conclusions                                               Conditions of Internals of VVERs”, Proceedings of 6th
                                                                                       Russian conference on Reactor Materials Science,
In agreement with many previous studies it appears that                                Dimitrovgrad, Russia, 11-15 September 2000, Vol..3,
decreasing rates of displacement rate lead to an earlier onset of                      Part I, pp. 3-23.

                             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
[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

5.    Simonen, E. P., Garner, F. A. and Klymyshyn, N. A.                     of Materials in Nuclear Power Systems – Water
      and M. B. Toloczko, “Response of PWR Baffle Bolt                       Reactors, 2003, pp. 647-656.
      Loading to Swelling, Irradiation Creep and Bolt                  15.   Porollo, S. I., Konobeev, Yu. V., Dvoriashin, A. M.,
      Replacement as Revealed Using Finite Element                           Krigan, V. M. and Garner, F.A., “Determination of the
      Modeling”, in this symposium.                                          Lower      Temperature Limit of Void Swelling of
6.    Garner, F. A., Chapter 6: "Irradiation Performance of                  Stainless Steels at PWR-relevant Displacement Rates”,
      Cladding and Structural Steels in Liquid Metal                         10th Int. Conf. on Environmental Degradation of
      Reactors," Vol. 10A of Materials Science and                           Materials in Nuclear Power Systems – Water Reactors,
      Technology: A Comprehensive Treatment, VCH                             August 5-9, 2001, issued in CD format, no page
      Publishers, 1994, pp. 419-543.                                         numbers.
7.    Porter, D.L., and Garner, F.A ,“Swelling of AISI Type            16.   Neustroev, V. S., Shamardin, V. K., Ostrovsky, Z. E.,
      304L Stainless Steel in Response to Simultaneous                       Pecherin, A. M. and Garner, F. A., “Temperature-Shift
      Variation in Stress and Displacement Rate”, Effects of                 of Void Swelling Observed at PWR-Relevant
      Radiation on Materials: Twelfth International                          Temperatures in Annealed Fe-18Cr-10Ni-Ti Stainless
      Symposium, ASTM STP 870, F.A Garner and J.S.                           Steel Irradiated at High and Low Dpa Rates in BOR-
      Perrin, Eds., American Society for Testing and                         60", International Symposium on "Contribution of
      Materials, Philadelphia, 1985, pp.212-220.                             Materials Investigation to the Resolution of Problems
8.    Seran, J.L., and Dupouy, J.M., “The Swelling of                        Encountered in Pressurized Water Reactors, 14-18 Sept.
      Solution Annealed 316 Cladding in RAPSODIE and                         1998, Fontevraud, France, pp. 261-269.
      PHENIX,” Effects of Radiation on Materials: Eleventh             17.   Neustroev, V. S., Shamardin, V. K., Ostrovsky, Z. E.,
      Symposium, ASTM STP 782, H.R. Brager and J.R.                          Pecherin, A. M., and Garner, F. A., "Temperature-Shift
      Perrin, Eds., American Society for Testing and                         of Void Swelling Observed at PWR-Relevant
      Materials, 1982, pp.5-16.                                              Temperatures in Annealed Fe-18Cr-10Ni-Ti Stainless
9.    Garner, F. A., Greenwood L. R., and Harrod, D. L.,                     Steel Irradiated in the Reflector Region of BOR-60",
      "Potential High Fluence Response of Pressure Vessel                    Effects of Radiation on Materials: 19th International
      Internals Constructed from Austenitic Stainless Steels",               Symposium, ASTM STP 1366, M. L. Hamilton, A. S.
      Proc. Sixth Intern. Symp. on Environmental                             Kumar, S. T. Rosinski and M. L. Grossbeck, Eds.,
      Degradation of Materials in Nuclear Power Systems -                    American Society for Testing and Materials, pp. 792-
      Water Reactors, San Diego, CA. August 1-5, 1993, pp.                   800 (2000).
      783-790.                                                         18.   Okita, T., Sato, T., Sekimura, N., Garner, F. A. and
10.   Garner, F. A., "Materials Issues Involving Austenitic                  Greenwood, L. R., “The Primary Origin of Dose Rate
      Pressure Vessel Internals Arising From Void Swelling                   effects on Microstructural Evolution of Austenitic
      and Irradiation Creep", Trans. Am. Nucl. Soc., 71                      Alloys during Neutron Irradiation”, Journal of Nuclear
      (1994) 190.                                                            Materials 207-211 (2002) 322-326.
11.    Garner, F. A. and Toloczko, M. B., "Irradiation Creep           19.   Okita, T., Sato, T., Sekimura N., Garner, F. A., and
      and Void Swelling of Austenitic Stainless Steels at Low                Wolfer, W. G. “ Combined Effect of Temperature,
      Displacement Rates in Light Water Energy Systems", J.                  Displacement Rate and Composition on the Neutron-
      of Nucl. Mater. 251 (1997) 252-261.                                    Induced Swelling of Austenitic Alloys”, 11th
12.   Garner, F. A., Edwards, D. J., Bruemmer, S. M.,                        International Conference on Environmental Degradation
      Porollo, S. I., Konobeev, Yu. V., Neustroev V. S.,                     of Materials in Nuclear Power Systems – Water
      Shamardin, V. K. and Kozlov, A. V., "Recent                            Reactors, August 2003, issued on CD format, no page
      Developments Concerning Potential Void Swelling of                     numbers.
      PWR Internals Constructed from Austenitic Stainless              20.   Edwards, D. J., Simonen, E. P., Garner, F. A., Oliver, B.
      Steels", Proc. Fontevraud 5, Contribution of Materials                 A., and Bruemmer, S. M., “Sensitivity of
      Investigation to the Resolution of Problems                            Microstructural Evolution Due to Temperature and
      Encountered in Pressurized Water Reactors, 23-27                       Dose Gradients in Neutron-Irradiated 316SS”, Journal
      September, 2002, paper #22, on CD format, no page                      of Nuclear Materials 317 (2003) 32-45.
      numbers.                                                         21.   Budylkin, N. I., Bulanova, T. M., Mironova, E. G., N.
13.   Garner, F. A., Porollo, S. I., Vorobjev, A. N.,                        M. Mitrofanova, N. M., Porollo, S. I., Chernov, V. M.,
      Konobeev, Yu. V., Dvoriashin, A. M., Krigan, V. M.,                    Shamardin, V. K. and Garner, F. A., “The Strong
      Budylkin, N. I. and Mironova, E. G., "Void-Induced                     Influence of Displacement Rate on Void Swelling in
      Swelling and Embrittlement in the Russian Equivalent                   Variants of Fe-16Cr-15Ni-3Mo Austenitic Stainless
      of AISI 316 Stainless Steel at PWR-Relevant End-of-                    Steel Irradiated in BN-350 and BOR-60”, Journal of
      Life Conditions", International Symposium on                           Nuclear Materials 329-333 (2004) 621-624.
      "Contribution of Materials Investigation to the                  22.   Maksimkin, O. P., Tsai, K. V., Turubarova, L. G.,
      Resolution of Problems Encountered in Pressurized                      Doronina, T., and Garner, F. A., “Characterization of
      Water Reactors, 14-18 Sept. 1998, Fontevraud, France,                  08Cr16Ni11Mo3 Stainless Steel Irradiated in the BN-
      pp. 249-260.                                                           350 Reactor,” Journal of Nuclear Materials 329-333
14.   Garner, F. A., Budylkin, N. I., Konobeev, Yu. V.,                      (2004) 625-629.
      Porollo, S. I., Neustroev, V. S., Shamardin, V. K., and          23.   Maksimkin, O. P., Tsai, K. V., Turubarova, L. G.,
      Kozlov, A. V. “The Influence of DPA rate on Void                       Doronina, T., and Garner, F. A., “Void Swelling of AISI
      Swelling of Russian Austenitic Stainless Steels,” 11th                 321 Analog Stainless Steel Irradiated at Low dpa Rates
      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.


Shared By: