USE OF SELF-ION BOMBARDMENT TO STUDY VOID SWELLING IN...

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

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USE OF SELF-ION BOMBARDMENT TO STUDY VOID SWELLING IN ADVANCED

RADIATION-RESISTANT ALLOYS

F. A. Garner1,2*, L. Shao2, M. B. Toloczko3, S. A. Maloy4, V. N. Voyevodin5

1Radiation Effects Consulting, Richland WA USA

2Texas A&M University, College Station TX USA

3Pacific Northwest National Laboratory, Richland WA USA 4Los Alamos National Laboratory, Los Alamos NM USA

5Kharkov Institute of Physics and Technology, Kharkov, Ukraine

ABSTRACT

Ion bombardment has been used successfully in earlier fast breeder and fusion materials programs to study the swelling behavior of simple metals and austenitic alloys, and more recently used on ferritic and

ferritic martensitic alloys for fusion, fast and spallation reactor systems. Currently there is interest to

apply this surrogate irradiation technique to help develop and test improved alloys for light water reactor service, especially with respect to development of accident-tolerant fuel cladding materials. A review is

presented of those features of ion bombardment that simulate neutron-induced behavior and those that do

not simulate as well. Some results of ongoing studies on ferritic martensitic alloys and their ODS variants

are shown to illustrate the value of the charged particle simulation technique.

1.0 Introduction

Since the Fukushima catastrophe there is strong interest in development of structural alloys with enhanced

resistance both to irradiation and to post-irradiation accident scenarios such as hydrogen release. Such

development efforts are hampered, however, by the current unavailability of high-flux reactors for irradiation testing. One approach to overcome this problem is to use ion irradiation as a surrogate for

neutrons, especially using self-ions (major elements of the alloy such as Fe, Cr, Ni) rather than light

particles such as protons, to assess the microstructural and dimensional stability of advanced alloys. Self-

ions are in general preferable to lighter ions because they produce damage on the atomic level that is closer to that produced by fission neutrons. They also cause much less heating per dpa during irradiation,

allowing ion irradiation to proceed at much higher displacement rates than are possible with energetic

protons or other light ions that deposit large amounts of energy and significantly raise the target temperature. Both of these issues are discussed in detail in refs. [1, 2].

The use of self-ion irradiation as a surrogate for neutron irradiation was very popular in the late 1970s

and most of 1980s after the discovery of void swelling when it was realized that this phenomenon would

have a very strong impact on the lifetime, performance and perhaps safety of fast reactors and also fusion reactors. Ion irradiation was primarily used to study the parametric sensitivities of void swelling, first to

identify those alloys that might be more swelling-resistant and later to guide the selection of materials to

test in fast reactors. In the years while waiting for specimens to become available from fast reactors there was large government financial support for ion irradiation, but once neutron-irradiated specimens became

easily available, interest and funding subsided and most further swelling-related studies involving ion

irradiation moved to academic centers primarily. As interest in nuclear power declined worldwide following the Chernobyl and Three-Mile Island events, the use of ion irradiation to study void swelling

also declined, although the use of ion irradiation as a tool grew strongly in other areas, especially in

electronic materials, surface modification and various exotic material applications.


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More recently, however, there has been a renewed interest or renaissance in nuclear power, especially

involving various advanced reactor concepts designed to run to very high fuel burn-up and therefore high neutron exposure. For such concepts to be economically viable, however, it is necessary to overcome a

major limitation encountered in fast reactor applications that arise from void swelling.

All of the various national fast reactor programs eventually focused on the use of titanium-modified

austenitic steels with optimized compositions and fabrication for use as fuel cladding and structural components. It had been earlier determined that these and all austenitic steels would eventually swell at

~1%/dpa after some incubation dose at all reactor-relevant temperatures and that alloy optimization

would only delay somewhat, but not preclude, the eventual onset of this higher swelling regime [3-5].

Additionally, it was learned that at ~10% void swelling, austenitic stainless steels developed a new form

of severe void-induced embrittlement that became the life-limiting condition for its use as fuel cladding

[6,7]. A consequence of this was a limitation on fuel burn-up to 10-12%, far short of the ~30% burn-up limitation imposed by fission product accumulation, thereby strongly impacting the economics of power

generation in fast reactors. In effect, the economics of power production were dictated by the cladding

and not by the much more expensive fuel.

As interest and financial support for fast reactors and nuclear power were declining, some progress was being made to explore the use of ferritic and ferritic-martensitic steels as substitutes for austenitic steels.

Iron-based alloys with a bcc structure were found to swell less that iron-based alloys with fcc structure,

possessing not only longer incubation regimes before transitioning to higher swelling rate, but also a much lower steady-state swelling rate, ~0.2%/dpa rather than ~1%/dpa for austenitic alloys [8-9]. A

successful demonstration of the superior swelling resistance and overall performance was completed in

the FFTF fast reactor just as the US fast reactor program was being terminated [10, 11]. In this campaign a series of subassemblies utilizing a 12Cr ferritic-martensitic alloy designated HT9 reached ~155 dpa with

very limited swelling of ~0.3%, and only then in isolated areas of the specimen [12,13].

In the last decade there has been an increasing interest to reach very high damage levels for advanced

reactor concepts, first to ~250 dpa and then to increasingly higher levels of 500 to 600 dpa. The primary focus of this effort is to use ferritic and ferritic-martensitic alloys to resist swelling and allow higher burn-

ups. Additionally, there is a focus on oxide-dispersion-strengthened (ODS) variants of these alloys in

order to overcome a major weakness of this class of steels, the tendency to lose strength at higher temperatures. It is also thought that that high dispersoid densities will contribute to swelling suppression.

However, all of the high-flux fast reactors have been shut down in the USA, UK, France and effectively

Japan, leaving only BOR-60, BN-600 and eventually BN-800 in Russia and PFBR in India, with one new

reactor just coming on line in China and another in India in a few years. Even for the currently operating fast reactors there are significant issues concerning availability and cost, however, that effectively

preclude their use for efficient alloy development and testing. Even when available, however, these

reactors can deliver neutron dose at a relatively low rate compared to the need. In BOR-60 center core region, for instance, the accumulated dose will be less than 20 dpa per year, requiring a minimum of 20

years to reach 400 dpa. Therefore, once again, the reactor and materials communities are turning to ion

irradiation at accelerated damage rates as a tool to explore the swelling behavior of new alloys.

The use of surrogate ion testing is not without controversy, however, and reactions of scepticism are

often met when first hearing that ion irradiation will proceed at an accelerated pace of 3-4 orders of

magnitude over a depth often less than the width of a single grain, and within a micron or less of the

specimen surface. Therefore, the first objective of this paper will be to provide a small taste of previous success of ion bombardment to study void swelling, focusing especially on insights gained from ion

bombardment that were later observed to be correct in neutron irradiation studies. Afterwards we will

discuss the features of ion irradiation that are atypical of neutron irradiation and then discuss how the impact of these differences can be minimized. Finally, some examples of recent studies will be presented.


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2.0 Examples of ion-derived insights that were later confirmed by neutron irradiation studies

Some of the most enlightening studies conducted on swelling using either ions or neutrons involve the use of simple model alloys, where in absence of solute elements, the compositional distribution of the alloy

matrix evolves very little, allowing observation of the basic compositional sensitivity of swelling. One of

the earliest insights by Johnston and coworkers, using 5 MeV nickel ions, was that the swelling of fcc Fe-

Ni-Cr alloys decreased very strongly with increasing nickel content, reaching minimum levels in the 35-60% Ni range before reversing thereafter, but climbing more slowly with increasing nickel content

[14,15]. Fig. 1 shows this behavior was confirmed in neutron studies published five years later [3,16,17].

As shown in Fig. 2 Johnston also showed that the dependence on nickel arose primarily from changes in the duration of the incubation period with a much smaller effect on the post-incubation swelling rate and

that this post-incubation swelling rate was on the order of ~1%/dpa, a rate that appeared to decline slowly

at higher nickel levels. Johnston also showed that the dependence on nickel content was preserved in more complex solute-added alloys, almost always producing a reduction in swelling arising from an

increased duration of the incubation dose preceding the onset of high rate swelling, as also shown in Fig.

2 for the case of silicon.

Later neutron studies showed that the steady-state rate was indeed ~1%/dpa, but this rate did not decline significantly with nickel content, foreshadowing one of the neutron-atypical aspects of ion bombardment.

Fig. 1 Comparison of ion and neutron irradiation data on the nickel dependence of void swelling of simple ternary Fe-Cr-Ni alloys [15,16]. .


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Fig. 3 shows a fraction of the extensive neutron data base on the same simple model alloys examined by

Johnston. These demonstrate that ion irradiation got some of the most important features of swelling correct but missed others, such as swelling at lower temperatures being invariant with Ni and Cr

composition, as well as temperature under some low temperature conditions. Of particular importance is

that ion irradiation came close to predicting the steady-state swelling rate, but leads to the misleading

conclusion that it was dependent on composition, a conclusion incompatible with the neutron results.

Fig. 3 Compilation of data from the EBR-II fast reactor showing that at low temperatures and low nickel

contents swelling is relatively independent of these variables, but as the nickel level or temperature

increases, the incubation period of void swelling begins to increase, but the steady state swelling rate

remains essentially constant [3,16]. Note on the lower left figure that the tendency of swelling to increase with increasing chromium, as observed by Johnston in Fig. 1 (left), is also seen during neutron irradiation,

but is again manifested only in the duration of the incubation regime of swelling.

The temperature dependence of void swelling, as observed during ion irradiation, is also rather distorted and spans a smaller range of temperatures compared with that of neutron irradiation. In particular, ion

irradiation does not clearly show that the action of temperature is only to determine the duration of the

incubation period, as shown in Fig. 4. Other studies conducted in the FFTF reactor show that the role of


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increasing dpa rate is increase the incubation regime in model ternary alloys, but not the steady-state

swelling rate [18]. On the basis of this observation alone, one might conclude that ion irradiation at accelerated dpa rates might lead to longer incubation periods than observed in neutron irradiations,

Fig. 4 Compilation of data from the EBR-II fast reactor showing that the incubation period of swelling in

any given Fe-Cr-Ni ternary alloys is sensitive to temperature, increasing with temperature [16].

Fig. 5 Comparative ion irradiation of ferritic 85Fe-15Cr with several austenitic ternary alloys, showing that ferrite swells less than austenite at 625°C, not only because it is inherently more swelling-resistant,

but also because it inhabits a lower temperature regime of swelling [19]. (Note the different right and left

scales for measured swelling). When duplex alloy Uranus 50 with ferrite and austenite grains was ion-

irradiated to 625°C Johnston showed the ion-incident surface is raised by the higher-swelling of austenite relative to lower-swelling ferrite grains [20], foreshadowing the relative swelling-resistance of ferrite.


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Focusing on several more examples of Johnston's work in Figs. 5 and 6, it is obvious that he forecast the

large inherent difference in swelling of fcc and bcc iron-base alloys, but he also showed that the lower temperature regime of ferrite swelling would place much of the swelling regime below the inlet

temperature of most fast reactors, making such alloys appear to be even more swelling resistant. Fig. 6

shows that the steady-state swelling rate approaches ~0.2%/dpa [19], in good agreement with the value

found in EBR-II and FFTF fast reactors [8].

Fig. 6 Comparison of swelling produced by ion [19] and neutron [8] irradiation of Fe-Cr alloys. The

longer incubation period in FFTF compared to that in EBR-II reflects the higher dpa rate and much lower helium generation rates in FFTF, both of which tend to delay the onset of steady-state swelling.

Finally, to demonstrate the utility of ion irradiation as a tool for assessing the relative swelling resistance

of various alloys, a previously unpublished, decades-old, data set from the doctoral thesis of one of the

authors (Voyevodin) is presented in Fig. 7. The five US and Russian alloys appear to swell with progressively longer incubation periods, in exactly the same order as observed in various fast reactor

irradiation campaigns. Note also that the swelling rate is ~1%/dpa for the earliest-swelling alloys but the

swelling rate declines somewhat with increasing incubation period, as was observed in Johnston's studies.

Fig. 7 Void swelling observed at 650°C in comparative irradiation with 3 MeV Cr ions at 10-2

dpa/s. Data were taken 200 – 300 nm from the surface. There are three Russian alloys; an analog of AISI 321, EI-847

(higher nickel, Nb-stabilized) and CW ChS-68 (early Ti, Nb stabilized, D9-like alloy) and two US alloys,

316 and D9, the latter being the most swelling resistant alloy developed in the US program. The

irradiation details are similar to those in refs. 44-47.


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3. Neutron-atypical features of ion irradiation and their consequences.

Only a few of many possible examples of ion-neutron comparative irradiations were presented above, but these are sufficient to show that ion irradiation, if conducted to minimize various atypical factors, can

reproduce the general swelling law observed in neutron irradiation for both fcc and bcc alloys, i.e. steady-

state after a transient incubation regime. It was also demonstrated that ion irradiation can come close to

producing the steady-state swelling rate for both fcc and bcc alloys, but gives the misleading impression that the rate falls as the transient regime increases.

Actually, the examples presented were selected because they avoided, inadvertently perhaps, some of the

worst consequences of neutron-atypical aspects of ion irradiation. For instance, many early publications on ion studies of austenitic steels, including those involving the first author of this paper, reported much

lower steady-state swelling rates arising from the action of neutron-atypical factors in ion irradiation.

To demonstrate the origin of many of these atypical factors, it is best to illustrate the scale and range of damage distribution. Fig. 8 shows the distributions of damage and injected ions for irradiation of pure

iron with 3.5 MeV Fe ions, using the SRIM code and a displacement threshold of 40 eV and the Kinchin-

Pease option [21]. Note that all damage occurs within ~1.4 microns of the surface.

Fig. 8 Damage and injected ion rest distributions calculated for 3.5 MeV Fe ions, for a peak dose of 800 dpa. Note the strong gradients in dose with depth, the short distance to the surface, and the strong overlap

of the higher damage rate region and the implanted ion distribution.

Neutron-atypical processes in self-ion irradiation involve a variety of phenomena, all operating in competition with each other over short distances which are typically on the order of only one micron.

Primary effects arise firstly from the strong and often underestimated influence of the ion-incident

surface, producing atom loss by sputtering, loss of radiation-produced defects and dislocations through

the surface acting as a sink, and compositional alteration as the surface serves as a sink for point defects, creating vacancy and interstitial gradients that alter the composition by processes such as the Inverse

Kirkendall and solute drag effects.

It is not generally recognized that at a given dpa rate the upper temperature limit of void swelling is determined primarily by the effect of the surface to reduce the vacancy supersaturation and thereby

strongly reducing void nucleation and growth [22,23]. Thus, the upper temperature limit reached in ion

irradiation is more an artifact of the irradiation technique than of the material properties. Bullough and Haynes calculate that for 4.2 MeV ions in stainless steel the void swelling rate is reduced by surface sink

effects as deep as 1 micron at 525°C and as much as 5 microns at 700°C [23]. Blamires and Worth

compared ion-induced swelling of M316 steel using 4.2 and deeper-penetrating 46.5 MeV ions, and saw

that while the results agreed well at lower temperatures, above 600°C the two ions produced very divergent behavior as the lower energy irradiation produced much less swelling [22], as shown in Fig. 9.


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Fig. 9 Comparison of ion-induced swelling in annealed M316 stainless steel using 4.2 MeV Fe ions and 46.5 MeV Ni ions [22]. At 525°C (left) the swelling is very similar, but diverges above ~600°C. At 35-40

dpa the chemical alteration by either Fe or Ni ions is not very large and hopefully we can ignore the

difference in bombarding ion.

A second class of atypical phenomena are associated with vacancy-interstitial defect imbalances arising

first from a slight forward scattering of interstitials in collisions of incident ion and matrix atoms to

produce a slightly larger separation between the vacancy and interstitial Frenkel pair, and second , especially, from the injected ion itself. The combined effects of these two aspects on swelling

distribution can be very strong [24]. Where the ion comes to rest, the implanted atom is effectively an

interstitial without an accompanying vacancy. In this capacity the injected interstitial strongly competes

with the otherwise balanced distribution of vacancy and interstitial fluxes and strongly suppresses void nucleation, and also suppresses somewhat the steady-state swelling rate once sufficient voids nucleate

[25-29]. Plumton and Wolfer demonstrated how potent the suppressive effect of injected interstitials

could be in austenitic steels and also showed that the effect increased strongly with dpa rate [28].

In ferritic alloys the accumulating evidence is that the injected interstitial suppression is even stronger.

One example is given in Fig. 10 to show the strength of such suppression even in pure iron, where

swelling is almost completely suppressed at depths where the injected interstitial comes to rest.

Fig. 10 Depth distributions of void swelling in pure Fe irradiated with 3.5 MeV self-ions to peak values of

35, 70 and 105 dpa, with normalized dpa and rest distribution curves shown for comparison [24]. Note the

near-total suppression of void formation in the zone where the ions come to rest.


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In the early days of ion irradition studies directed toward void swelling it was the general practice to take

swelling data at the peak damage region which lies in the distribution of injected ions, as shown in Fig.8, thereby inadvertently producing strong underestimates of the steady-state swelling rate.

Additionally, the injected interstitial also functions as a chemical species where it comes to rest. Johnston

estimated that a peak of ~2% nickel ions were deposited per each ~100 dpa at the peak damage region

[19]. Since the nickel level is a strong determinant of the duration of the transient regime, it tends to depress the onset of swelling and extend the transient, and the effect becomes stronger for alloys which

resist swelling the longest. Johnston found that adding 2% Ni indeed measurably decreased the swelling.

This compositional modification by injected ions may be the major reason that the apparent swelling rate under ion irradiation tends to fall with increasing transient duration. A similar effect may be occurring in

Fig. 7 where chromium ions come to rest. When going to many hundreds of dpa the injected element can

reach double digit values in added concentration, significantly altering the local composition even if radiation-assisted diffusion spreads out the injected element.

A third major category of effects arises from factors related to displacement rate. First, as seen in Fig. 8

there are internal variations in displacement rate that can affect swelling directly [30] or indirectly via

gradient-induced modifications of the composition [31]. Second, void swelling in typical reactor structural alloys is often proceeded by radiation-induced formation or modification of precipitates that

remove from the matrix swelling-suppressive elements such Si, P and Ni [3-5]. These precipitation

sequences are known to be sensitive to displacement rate, but probably have different rate dependencies compared to those of dislocation and void structures.

The most often cited rate effect is that of the much higher displacement rate in production of point defects

and production of equivalent dislocation and void microstructure, as best described by Mansur [32]. In effect, to produce equivalent microstructures at higher dpa rates requires a movement of the swelling

regime to higher temperatures. This phenomenon is described as the "temperature shift". However, the

description of this shift envisions an infinite medium and does not include the strong effect of the

specimen surface which is very sensitive to the dpa rate. Additionally and more importantly, it does not include the very strong rate dependence of the injected interstitial, acting to decrease void nucleation at

lower temperatures and higher dpa rates [28].

These two factors mean that whereas swelling in neutron irradiation is limited by point defect recombination at low temperatures and vacancy reemission at higher temperatures, the situation under ion

bombardment is more complicated, such that the lower temperature limit of swelling is dominated by the

injected interstitial effect and the higher temperature limit by the effect of the specimen surface. Both of

these processes are strongly sensitive to dpa rate and can produce a temperature shift all by themselves, although they probably operate in concert with the microstructural equivalence mechanisms proposed by

Mansur.

As mentioned earlier, the effect of higher dpa rate as observed in neutron irradiations is to extend the duration of the transient regime [5,18]. A similar effect has been observed in these same model alloys

using ion irradiation at three different dpa rates [33].

The net effect of the various dpa rate-sensitive processes acting in addition to surface effect , compositional modifications, etc. is that the temperature dependence of swelling produced by self-ions is

very distorted with respect to that produced in neutron irradiation. It is therefore unrealistic to expect that

self-ion irradiation at the energies available to most researchers (≤5 MeV) can closely match the swelling

dependence observed in a neutron environment.

Another neutron-atypical factor is interesting but of lesser importance. The thin film undergoing swelling

is constrained by the non-swelling underlying bulk, producing a planar compressive stress state of

relatively low magnitude with a shear vector perpendicular to the ion-incident surface [34,35]. The consequence of this stress state is that all mass moves anisotropically toward the surface rather than


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distributing strains isotropically as typical of unconstrained neutron-irradiated specimens. In effect, ion

irradiation is as much an irradiation creep experiment as it is a swelling experiment, with two-thirds of the mass flow directed to the third dimension. The actual movement of mass must follow crystallographic

restraints involving dislocation and loop orientation with respect to the stress state, but the net movement

of mass is to raise the surface, with differently oriented grains responding to produce visible surface

topography [34-37].This atypical factor is thought to be transient in its influence with no significant long-term or high dose consequences.

Another neutron-atypical aspect can be easily avoided. An ASTM specification states that the ion beam

should not be swept or rastered across the specimen surface, as it will produce a strong suppressive effect on void nucleation, primarily by evaporation of small vacancy clusters during the beam-off phase of the

irradiation [38]. Recent studies confirm that rastering does indeed suppress void nucleation [39,40].

However, the theory used to describe such suppression was originally used to describe pulsed irradition characteristic of some inertial fusion reactor concepts [41, 42] but does not take into account the strong

action of injected interstitials on void nucleation and growth.

As an example, let's assume a typically used rastering procedure that produces at one location a beam-on

time that is only 2.5% of the total. The instantaneous dpa rate for a 97.5% beam-off time is therefore 40 times greater in the beam-on period than the quoted average dpa rate. Defect recombination is sensitive to

the square of the dpa rate and the strong dpa rate dependence of the injected interstitial phenomenon

strongly suppresses void nucleation during the beam-on period, acting in addition to the suppression by evaporation of vacancy clusters in the beam-off period. Thus, rastering during ion bombardment involves

much greater suppression than pulsing in the neutron environment.

The nucleation-suppressive effect of the various atypical variables can be partially mitigated by pre-injection of helium [43], but this procedure in itself is a neutron-atypical process. A better way, not only

to overcome such suppression effects, but to also more closely simulate the effect of helium and or

hydrogen is to co-inject these two gases during ion irradiation. An excellent example of this technique

was recently published that involves the use of a single-beam, three-ion facility rather than using two or three accelerators operating concurrently to introduce both damaging ions and gas ions [44].

In summary, ion irradiation can reproduce some aspects of neutron irradiation, but the various neutron-

atypical factors do not allow a faithful reproduction of the temperature dependence or the duration of the transient regime of swelling. The best procedure to minimize the impact of these atypical factors is to

conduct examination of the swelling at a distance somewhere between the surface and the region of

injected interstitials, with the latter being more important to avoid. The best uses of ion bombardment as a

simulation tool are 1) comparison of alloys under identical irradiation conditions to assess their relative swelling resistance and 2) fundamental studies concerning the stability or role of particular

microstructural components such as radiation-produced precipitates or deliberately added dispersoids.

4.0 Some recent examples of ion irradiation to study void swelling and microstructural stability

The ion studies conducted by the authors of this paper are conducted in two major facilities. To

illustrate the utility of ion bombardment to study void swelling we will concentrate in this paper

on several irradiations series performed in the Ukraine at the Kharkov Institute of Physics and

Technology. Other studies conducted at Texas A&M university are covered in references 24 and

40, with a number of other papers on various commercial or developmental alloys in varying

stages of publication.

Of all ferritic-martensitic steels the most published data is for the 12Cr alloy dual phase alloy

EP-450, having a ~50-50 mixture of ferrite and tempered martensite (the latter called sorbite in

Russian) and has been irradiated in all Russian and Kazakh fast reactors. It is routinely used for


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the hexagonal fuel assembly wrappers in BN-600. This alloy was irradiated with 1.8 MeV Cr

ions to very high doses and the swelling studied at a depth far from the influence of the injected

interstitial [44,45]. Figure 11 shows the preirradiation microstructure and Figure 12 shows the

surface typography of the ion-irradiated surface. Note that the ferrite grains rise above the

original surface while the sorbite grains are not visibly elevated. Figure 13 shows void

microstructures observed after irradiation in two specimens at high doses. Note that the ferrite

grains swell earlier and much more than the adjacent sorbite grains. Figure 14 shows that the

ferrite grains have already reached at 150 dpa the ferrite-characteristic ~0.2% swelling per dpa

value while the swelling rate of sorbite grains is still increasing at 300 dpa.

From this study it was learned that ferrite should be avoided if swelling is to be minimized .

Fig. 11 Typical preirradiation microstructures of EP-450, showing 50-50 duplex structure.

Russian units of HM correspond to nm in English. Larger grains are ferrite and smaller grains are

primarily tempered martensite or sorbite.

Fig. 12 Post-irradiation surface morphology showing that ferrite grains rise above the level of

surrounding sorbite grains at 300 dpa and 480°C.


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Fig. 13 (top) Void structure in an area composed primarily of ferrite grains after 225 dpa at

480°C, showing rather uniform swelling. (bottom) Void microstructure in area with three ferrite

(F) grains and two sorbite (S) grains, showing very different swelling levels in the two types of

grains.


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Fig. 14 Swelling behavior observed in the ferrite and sorbite grains of EP-450 at 480°C [44,45].

It is not always possible to completely avoid ferrite formation during tempering of ferritic-

martensitic alloys. As shown in Fig. 15 where the INL US program heat of HT9 retained ~5%

ferrite. Fig. 16 shows that the ferrite grains indeed swell more than adjacent tempered martensite

grains. The swelling of this HT9 heat is relatively low compared to that of EP-450, partly as a

result of different composition and preparation, but also because the ferrite fraction is smaller.

Note, however, that the two ion irradiations were conducted at different temperatures.

When a 33% cold-worked fusion heat of HT9 without any ferrite was irradiated at 480°C it was

shown that the tempered martensite grains eventually reached the ~0.2%/dpa value known to be

steady state swelling rate of iron-base ferritic alloys. Note in Fig. 17 that ~35% swelling was

reached at 600 dpa, indicating that hopes of forever resisting swelling are too optimistic.

Fig. 15 Photos showing small amount of ferrite remaining in the INL heat of tempered HT9

supplied by the USDOE though Los Alamos National Laboratory. There is ~5% ferrite in this

heat, seen as larger white features in the left hand micrograph.


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Fig. 16 Selected micrographs extracted from areas of HT9 irradiated at 425°C where ferrite is

surrounded by tempered martensite grains. The unirradiated microstructure was shown in Fig 15.

Fig. 17 Another older "fusion" heat of HT9 in the 760°C/0.5hr + 33% CW condition waited

until approximately 350 dpa before accelerating to 0.2%/dpa. All carbides that were originally in

the grain interior have disappeared, and it is assumed that the carbon from these vanished

precipitates are now concentrated in the grain boundary precipitates shown in the micrograph on

the right.

The most important question is whether the attainment of the ~0.2%/dpa swelling rate can be

delayed in any alloy to sufficiently high enough dose that the alloy can serve out its life without

exceeding the maximum swelling defined for its specified objective. Two recent studies show

that such dose extension may be reached in oxide dispersion hardened alloys. Two recently


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published papers show that nano-featured ODS alloys can attain very high doses before

swelling accelerates to the 0.2% rate [46,47].

Fig. 18 Voids formed in thin elongated grains of MA957 at 450°C and 500 dpa [46].

Fig. 19 Preirradiation microstructure of 14YWT with more equiaxed nanostructured grains,

producing relatively low levels of swelling after 500 dpa at 450°C [47]. Also shown is a

comparison of the various alloys discussed in this section.


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CONCLUSIONS

We have demonstrated in this paper that ion irradiation can simulate some, but not all aspects of

neutron-induced void swelling. Successful use of the technique requires an understanding of

those factors inherent in ion irradiation that are atypical of the neutron environment. When

conducted to minimize the impact of atypical factors ion irradiation can be used to compare the

relative swelling response of various alloys. It can also be used to identify those factors that can

be optimized to further resist swelling. In the examples presented it was shown that ferrite should

be avoided and that nano-structuring of grains held pinned by dispersoids is a good way to

increase the swelling resistance of ferritic-martensitic alloys.

ACKNOWLEDGEMENTS

Texas A&M University acknowledges the support by US Department of Energy, NEUP

program, through grant no. DE-NE0008297. Los Alamos National Laboratory acknowledges the

support of DOE-NE Fuel Cycle R&D program. This research was funded at Pacific Northwest

National Laboratory by the Fuel Cycle R&D Program Core Materials research area sponsored by

the U.S. Department of Energy, Office of Nuclear Energy. Pacific Northwest National

Laboratory is operated for the U.S. Department of Energy by Battelle Memorial Institute under

Contract DE-AC06-76RLO 1830. Collaboration with Kharkov Institute of Physics and

Technology in the Ukraine was made possible by various sources, primarily by grants or

contracts placed via the Civilian Research and Development Foundation and via Radiation

Effects Consulting LLC.

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17

7. V. S. Neustroev and F. A. Garner, “Very High Swelling and Embrittlement Observed in a

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18

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19

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influence of High Levels of Helium and Hydrogen on Void Swelling of 8-12%Cr

Ferritic-Martensitic Steels", J. Nucl. Mater. , accepted for publication 2015.

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at High Doses and Gas Concentrations", Proceedings of Accelerator Applications 2013,

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47. M. B. Toloczko, F. A. Garner, V. N. Voyevodin, V. V. Bryk, O. V. Borodin, V. V.

Mel'nychenko, A. S. Kalchenko, S. A. Maloy and D. T. Hoelzer, "Ion Irradiation of the

ODS Ferritic Alloy 14YWT to 100-500 dpa", accepted for publication in J. Nucl. Mater.

2015.

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