Radiation Tolerance of Ultrafine Grained Materials Fabricated by Severe Plastic Deformation

Radiation Tolerance of Ultra

fi

ne-Grained Materials Fabricated by Severe Plastic

Deformation

Nariman A. Enikeev

, Valentin K. Shamardin

and Bertrand Radiguet

2_{Laboratory of Mechanics of Advanced Bulk Nanomaterials for Innovative Engineering Applications, Saint Petersburg State University,}

Peterhof, St. Petersburg, 198504, Russia

3_{Research Institute of Atomic Reactors JSC“SSC RIAR”, Dimitrovgrad, Ulyanovsk Region, 433510, Russia}

4_{Normandie Univ, UNIROUEN, INSA Rouen, CNRS, Groupe de Physique des Matériaux, 76000 Rouen, France}

This paper presents a historical overview of studies on the irradiation behaviour of ultrafine-grained materials produced by severe plastic deformation (SPD), which allows fabricating bulk materials with submicron grain size and nanostructural features, essentially enhancing their strength and functional properties. The dramatically increased interface fraction also provides substantially improved microstructure tolerance and ultrafine-grained material properties for electron, proton, ion or neutron irradiation. Issues related to considering SPD metals and alloys as advanced radiation-resistant materials are discussed. [doi:10.2320/matertrans.MF201931]

(Received February 25, 2019; Accepted June 14, 2019; Published July 26, 2019)

Keywords: radiation resistance, severe plastic deformation, ultrafine-grained materials, nanostructured materials, grain boundaries, radiation-induced defects, ion irradiation, neutron irradiation, mechanical performance, microstructure

1. Introduction

Industrial progress in the twenty-first century significantly increases the need for electric power, produced in safe and eco-friendly ways. In parallel with the development of sustainable energy production technologies, the power industry gives keen attention to radically improving the durability and reliability of existing nuclear reactors and aims to switch to fusion energy harvesting in the near future. This ambitious goal generates the task of developing new materials with markedly enhanced functional properties both in terms of mechanical performance and irradiation tolerance.1,2) The internal structures of nuclear systems function under extreme conditions of high temperature and severe irradiation by fast neutrons and high-energy particles. As a result, the microstructure of irradiated materials is damaged by radiation-induced defects. The atomic displace-ment cascades caused by the interaction of the particles with atoms lead to the generation of self-interstitials and vacancies with concentrations significantly exceeding the equilibrium values. With increasing damage dose, these defects agglomerate to form clusters, voids and dislocation loops. They also facilitate solute precipitation, segregation, phase transformation and formation of gas bubbles. All these factors result in a noticeable degradation of the functional performance of irradiated materials. In particular, irradiated materials are subjected to void swelling, irradiation hardening and creep, neutron-induced embrittlement, irradiation as-sisted stress corrosion cracking and so on.2­7) These troublesome effects significantly reduce the lifetime of components and can destabilize the reliable operation of nuclear systems, especially, fusion reactors.

Thus, modern nuclear engineering vitally needs to increase the radiation resistance of materials to provide long-term and safe functioning of reactors. Progress in this filed requires fundamental studies, which demand intensive

multidiscipli-nary research combining materials science, solid state and nuclear physics, computer simulation and mechanical engineering.

Traditionally, researchers and metallurgists try to enhance the irradiation tolerance of metallic materials by alloying with specific chemical elements. However, alloying could affect the other parameters of the material’s performance.2­6) _For

instance, austenitic stainless Cr­Ni steels have excellent corrosion resistance and ductility but they are apt to swelling and are characterized by relatively low yield stress and high redundant radioactivity. Ferritic/martensitic Cr-based steels have high strength, reduced redundant radioactivity and high resistance to thermal and swelling effects, but they are prone to radiation embrittlement and high-temperature creep.

Another approach to this problem is based on the microstructural design of a material with a given chemical composition. The idea behind this approach consists of increasing the density of internal sinks for radiation-induced point defects. One of the recent breakthroughs in nuclear materials science relates to introducing an ultrahigh density of dispersion-strengthening nanofeatures, effectively sup-pressing radiation damage (see Refs. 5, 6), for example).

Neutral sinks, such as grain boundaries also have a great capability to adsorb both vacancies and self-interstitials. Moreover, as shown in Ref. 8) interstitials that are radiation-driven into the boundary, could then be emitted to the bulk, annihilating with vacancies there. This can additionally reduce damage induced by irradiation. Hence, increasing the fraction of interfaces in a material could significantly improve its radiation resistance. To realize this idea, one should dramatically refine the grain size. In Ref. 7) the author mentioned that for grain sizes less than 10 µm, absorption of point defects by grain boundaries becomes comparable to the effects of other sinks (voids, dislocations and so on) in metallic materials. One would expect a dramatic improve-ment in radiation stability with further reduction in grain size, especially, down to the nanoscale.

+_{Corresponding author, E-mail: nariman.enikeev@ugatu.su}

2. Early Studies on the Radiation Resistance of Nano-materials

After the introduction of nanocrystalline materials to the scientific community,9)it was natural to expect experimental testing of their irradiation tolerance. Pioneering studies on the effect of ion and electron model irradiation by on nanocrystalline materials produced by inert gas condensation demonstrated their much higher ability to digest radiation-induced damage compared to materials with coarser grains using the example of nano-Pd and ZrO2 specimens with a

grain size in the range of 10­300 nm bombarded with Kr+ ions.10,11) _{Both materials in the nanocrystalline state}

demonstrated a notably reduced tendency for accumulating radiation-induced point defects compared to the conventional samples.10) _{Moreover, as shown in Ref. 11), a decrease in}

grain size from 100 nm to 40 nm reduced the accumulation of radiation-induced defects by three times for ZrO2. In Pd,

a similar reduction of grain size provided a four times difference in the defect agglomeration rate (Fig. 1). For grain sizes below 15­30 nm, defect formation was not observed at all in both materials.11) Chimi et al.12) reported consistent results where they estimated the defect accumulation rate in irradiated nano-Au with the help of electric resistivity measurements.

Neutron irradiation experiments with nanocrystalline 316 steel specimens consolidated by hot isostatic pressure using ball-milled steel and TiC powders13)_{and nano-Ni and N-W}

produced by electrical deposition13,14) _{also proved that}

materials in a nanocrystalline state possess much higher resistance to radiation hardening and microstructure damage. The threshold value of grain size when the size effect becomes considerable was found to be about 300 nm.14)

Consistent results were reported for the case of nano-structured V-Y alloys prepared by mechanical alloying and subjected to neutron irradiation to a dose of less than 1 dpa.15) Molecular dynamics simulations carried out in the beginning of this century for the irradiation behaviour of nano-Ni with a grain size of about 10 nm16­19)revealed that self-interstitials effectively sink and annihilate with free volumes in triple lines and grain boundaries, which does not entail significant structural changes. Similar numerical

studies reported in Ref. 20) confirmed that interstitial atoms were accommodated by interfaces while vacancies evolved to vacancy clusters.

Thus, the results of the early experimental and theoretical investigations showed that a higher number of interfaces can significantly affect defect annihilation during irradiation. After achieving these preliminary promising results, research activity in thisfield has been rapidly growing. Studies have also covered the cases of thin films, multi-layers, nano-composites, coatings and amorphous materials. Thefindings reported in the numerous subsequent publications have been summarized in many papers, thoroughly reviewed in Refs. 21­25). These works unambiguously confirmed that reducing grain size could be an effective way to develop new radiation-resistant materials.

However, numerous reviewed results relate to research on materials obtained by powder solidification after gas consolidation or ball milling, etc. These materials are characterized by redundant porosity, contamination and limited dimensions, the latter problem relates to thin films or coatings as well. This occasion complicates both precise systematic studies of such materials (especially from the viewpoint of mechanical performance) and evaluation of their potential application.

In recent decades, materials scientists focused attention on a technique that takes advantage of large strains or severe plastic deformation (SPD)26) _{to refine the microstructure of}

metallic materials. SPD materials typically have submicron grain size; for some alloys, the grain size can be reduced to a value less than 100 nm. The SPD-induced changes in a material’s properties are primarily defined by the ultrafi ne-grained (UFG) structure and, consequently, by a dramatically pronounced grain boundary network. Depending on the processing conditions, grain boundaries in SPD materials can be characterized by different misorientation spectra, misorientation character distribution, specific defect structure, excess volume, long-range stressfields, enhanced diffusivity and so on.27)Significantly enhanced atomic mobility can also lead to untypical precipitation, phase transformation and local redistribution of solutes in SPD alloys,26,28)including strain-induced non-equilibrium grain boundary segregation29)quite similar to radiation-induced segregation. On the one hand, the diversity of these phenomena complicates the investigation and analysis of SPD materials; on the other hand, purposeful manipulation of the output nanostructural parameters can provide extra-capabilities to tailor and tune improved multifunctional properties for the produced UFG materials, including irradiation tolerance.28) _{In addition, SPD allows}

producing bulk and fully dense UFG specimens without contamination and oxidation. SPD techniques such as equal-channel angular pressing (ECAP), its modification designed to produce long-sized billets (continuous ECAP) and numerous alternative recent SPD developments30­32) are capable of fabricating UFG billets with excellent mechanical properties and dimensions attractive for specific advanced applications. Taking all these considerations into account, bulk SPD materials could be quite promising for nuclear engineering as radiation resistant materials with enhanced mechanical performance; however, literature studies available in this field are not numerous and comprehensive. This Fig. 1 Dependence of radiation-induced defect density on grain size for

overview aims at summing up the progress in the research on the radiation tolerance of UFG materials produced by SPD.

3. Tolerance of SPD Materials to Model Irradiation

3.1 The effect of model irradiation on the micro-structure of SPD materials

Model experiments using ion or protonfluxes represent a fast and convenient way to study the effect of irradiation on the microstructure of metals and alloys. To the best of our knowledge, thefirst data on the irradiation behaviour of SPD materials were reported in Refs. 33, 34). Nanostructured Ni and Cu­0.5%Al2O3specimens produced by ECAP followed

by high-pressure torsion (HPT), as well as electrodeposited nano-Ni specimens, were bombarded with 590 MeV protons and 840 keV self-ions with a maximum damage of 5 dpa. Nita

et al.33,34) _{showed that the density of radiation-induced}

stacking fault tetrahedra in SPD materials was an order of magnitude less than in conventional materials subjected to both proton and ion irradiation.

The next studies confirmed these encouraging results. Heavy 1.5 MeV Ar+ irradiation of nanocrystalline Ti­Ni shape memory alloy produced by HPT showed the improved abilities of SPD materials to sustain irradiation damage.35,36) As shown in Fig. 2, X-ray studies of a TiNi alloy with a grain size of 23 nm (Fig. 2(a)) showed improved stability of the crystal structure. In the coarse-grained alloy (Fig. 2(b)), B2 superlattice disordering was observed after irradiation up to 0.4 dpa followed by increasing amorphization, which becomes much more pronounced after 2.5 dpa.35) _{In TiNi}

after HPT, the superlattice was found to be stable after irradiation to 1.8 dpa and amorphization was not observed even at a 5.6 dpa damage dose. Similar results have been obtained for ion irradiated SPD-produced nanocrystalline intermetallic Fe­Al alloy with a 35 nm grain size studied by X-ray diffraction and Mössbauer spectroscopy.36)

Nuclear engineering researchers draw special attention to enhancing the radiation stability of commercial materials used to fabricate the internal structures of nuclear reactors. The authors of Refs. 37­39) reported results related to the

microstructure stability of an austenitic stainless 316 steel nanostructured by HPT and irradiated with 160 keV Fe+ions to doses of 5 and 10 dpa. They pointed out radiation-assisted grain growth (average grain size increased from 40 to 60 nm) in the nanostructured steel.37) Atom probe tomography revealed that radiation-induced grain boundary segrega-tions37­39) in nano-steel were somewhat different from those typical for irradiated coarse-grained materials. Moreover, the nanostructured 316 steel demonstrated no tendency toward intragranular segregation, clustering, precipitation, or Frank loop formation;37) however, these phenomena are peculiar to irradiated coarse-grain austenite stainless steels. The radiation-induced segregation at grain boundaries (depletion in Cr and enrichment with Ni and Si) was found to be smaller in the nanostructured 316 steel compared to the conventional material7­39)_{. These effects were suggested to be due to}

enhanced annihilation of point defects in boundaries, which could lead to potential suppression of irradiation-assisted stress corrosion cracking and void swelling37) _{as supported}

by dynamic modelling results.39)

It is worth mentioning that SPD itself is capable of forming strain-induced grain boundary segregations in stainless steels depending on the processing and post-annealing conditions, as shown in Refs. 40, 41). Researchers pointed out a similar background for strain and radiation-driven segregation kinetics;42) the excess density of externally induced point defects provides the concentration gradient between the grain interior and interfaces providing a corresponding driving force for recombination of the defects.

The studies analysed above37­41) _{and further}

investiga-tions43,44)_{revealed an interesting difference in grain boundary}

segregation driven by strain and radiation. Cr, Si and Mo atoms tend to segregate at grain boundaries in the steel nanostructured by HPT at elevated temperature (starting from 400°C) or nanostructured at room temperature and then annealed at 450°C and higher.40,41,43,44) However, Ni does not segregate at boundaries, while it is well known4) that irradiation of Cr­Ni steels entails enrichment of boundaries with Ni and depletion in Cr. Figure 3 shows distributions of alloying elements, Si, Cr and Ni, across grain boundaries for

(a) (b)

316 steel nanostructured at 450°C, before and after irradiation by self-ions.43)

Unlike SPD, radiation led to notable segregation of Ni in most boundaries. Alongside that, some boundaries became depleted in Cr while the others kept the Cr concentration at a slightly decreased level compared to steel processed by HPT and annealed at 450°C. Different behaviours of solutes under severe straining and irradiation could be linked to a different production rate of the point defects. Displacement cascades produced by ions or neutrons generate comparable amounts of interstitials and vacancies, while deformation under high pressure generates vacancies at a higher concen-tration than interstitials, as discussed in Ref. 27). Competi-tion between effects induced by strain and irradiation could be an additional way to tune a material’s behaviour under external influence.

3.2 The effect of model irradiation on the properties of SPD materials

The evolution of Cr in boundaries driven by

nano-structuring and irradiation was studied in detail in Ref. 44) with special attention to corrosion properties defined by the ability of Cr to form a passivation layer on the stainless-steel surface. Irradiation-induced grain boundary depletion in Cr could significantly affect the corrosion resistance of stainless steels, promoting unwanted irradiation-assisted stress corro-sion cracking.45)

Un-irradiated nanostructured steel demonstrated slightly weaker corrosion resistance in a Na­Cl solution than coarse-grained steel, while in general, the corrosion property was excellent for both materials.44) Figure 4 reveals differences in the corrosion behaviour of coarse-grained 316 steel and the UFG steel produced by HPT in as-received and ion-irradiated conditions.

The experiments showed that the passivation state of conventional steel had significantly degraded after irradi-ation, while the effect of irradiation on corrosion resistance was moderate for the HPT steel.44)_{Nanostructured steel also}

demonstrated better resistance to corrosion in terms of pitting and weight loss in the irradiated state. Revathy Rajanet al.44)

(a) (b)

Fig. 3 Atom probe tomography images of (a) annealed UFG sample and (b) irradiated UFG sample showing the distribution of Si. Line profiles of elemental distribution across grain boundaries as indicated for Si, Cr and Ni. Reproduced from Ref. 43).

explained the reduction of negative irradiation effect on the corrosion properties of the HPT steel by the state of grain boundaries, less depleted in Cr after irradiation.

Application of model irradiation significantly limits the capability to evaluate mechanical performance because the representative damaged area is limited to a thin layer under the surface of an irradiated specimen. However, nano-indentation tests revealed that grain refinement notably improved the materials’ resistance to irradiation harden-ing.34,44) This can be explained both from the viewpoint of the decreased defect accumulation rate and high hardening of UFG materials in an un-irradiated state provided by grain refinement and the high density of crystallographic defects induced by SPD.

3.3 Does the grain size refinement always improve radiation tolerance?

The publications analysed above gave quite an optimistic view on the radiation tolerance of UFG materials produced by SPD. However, some studies claimed a careful approach to the grain size effect. In Ref. 46), Shen discussed the effect of grain size on the irradiation resistance of metallic materials in terms of an energetic analysis of grain boundaries. The balance between the free energy of interfaces and point defects can change a positive grain refinement effect to negative.46) To resist long-term exposure to high-energy irradiation, the nanostructures must be thoughtfully tuned in any specific material.47)

Moreover, nanostructured materials, including those produced by SPD,28) _{are known to possess limited thermal}

stability. Assuming the potential application implies oper-ation of materials simultaneously at elevated temperatures and high irradiation fluxes, one should take the possible microstructure recovery into account. Indeed, most of the studies cited above (starting with Ref. 11)) reported radiation-assisted grain growth in nanostructured materials, depending on the irradiation experiment temperature.

Recent experiments on steels exposed to huge doses of ion irradiation provided somewhat contradictory results.48,49) A specially designed low activation steel, RUSFER EK-181,50) in coarse-grained and nanostructured states was subjected to irradiation with Fe ions to as much as 400 dpa at 400­ 500°C.48)_{The study revealed that nanostructuring accelerated}

the onset of swelling accompanied by significant thermally activated grain growth at high doses. Radiation-induced voids in HPT steel were three times smaller but their concentration was noticeably higher than for the coarse-grained material. Further, despite the positive results reported for ECAP steel T91 irradiated with self-ions to 150 dpa, where the swelling rate was reduced by three times thanks to microstructure refinement,51)Gigaxet al.49)showed that ion irradiation to 1000 dpa provided a considerably higher level of void swelling in the SPD-produced alloy than in the conventional one.

It is important to note that the steels tested in Refs. 48, 49) belong to the ferritic-martensitic type, which are known to be highly resistant to swelling. In addition, the results of heavy ion irradiation must be analysed carefully with respect to the surface effect and do not directly correspond to the effect of neutron experiments for the reasons discussed in the next

section. Finally, the experiments48,49)_{were carried out at the}

temperature where the UFG material recrystallised, so two processes of irradiation and thermal grain growth could interfere and accelerate each other. Gigax et al. pointed out that grain growth in the un-irradiated area of the SPD material was far less pronounced than in irradiated regions,49) which leads to the assumption that irradiation can facilitate degradation of the UFG microstructure when the material is thermally unstable. Hence, grain boundary stability needs careful analysis with respect to the specific material in the given irradiation and thermal conditions.

As reported in Refs. 8, 22), the ability of grain boundaries to recover irradiation damage can be increased with temperature, which is schematically illustrated in Fig. 5. After a collision, the produced interstitials are absorbed by boundaries, while excess vacancies remain in the grain interior (thefirst three frames). At elevated temperatures (the next two frames), mechanisms with low activation energy, such as interstitial emission, can be activated and some amount of the damage will also be recovered by annihilating the emitted interstitials with the extra-vacancies in the grain interior. With further increases in temperature (the last two frames), the vacancy mobility increases; they drift to boundaries due to a vacancy concentration gradient and maximize the self-healing effect.22) Hence, increasing the fraction of interfaces in a material could significantly improve its radiation resistance, especially in a high-temperature environment.

Keeping these findings in mind, researchers should thoroughly design a particular nanostructured material aimed at enhanced performance in the given environment assuring the optimal grain size and grain boundary stability at the operational temperature and considering the diffusion proper-ties of atoms and vacancies in an alloy with a specific chemical composition.

Currently, the amount of research in model irradiation of UFG and nanostructured materials demonstrates a sky-rocketing growth. The scientific community is obtaining new evidence of the great potential of UFG metals and alloys as high-performance radiation-resistant materials produced by various SPD techniques, accumulative roll-bonding,52)

ECAP53) _{and HPT,}54,55) _{etc. At the same time, the amount}

of radiation experiments on UFG materials involving fast neutron fluxes is relatively negligible.

4. Behaviour of UFG Materials Produced by SPD in the Conditions of a Nuclear Reactor Core

As shown in the previous section, experiments with the help of model irradiation give a clear indication that, in general, SPD materials can effectively sustain irradiation-induced damage compared to their coarse-grained counter-parts. The strategy to enhance radiation tolerance by tailoring the optimal microstructure with an increased fraction of interfaces has proven to be very fruitful and promising.

depth. That distance significantly complicates the correct and consistent investigation of ion irradiation-induced damage and makes it impossible to evaluate the mechanical performance of irradiated samples by standard mechanical tests. Neutrons produce atomic displacement cascades through the bulk of the materials, leading to homogeneous damage in the whole specimen. High-energy neutronfluxes could also promote forming transmutation elements, addi-tionally altering the irradiation effect on the microstructure. As a result, Songet al. discuss a different effect of ion and neutron irradiation on swelling in steels, depending on dpa rate in particular.51)

Hence, systematic neutron irradiation experiments are vitally needed to reliably evaluate the behaviour of SPD materials in the internal environment of commercial reactors. Currently, the results of such experiments are few in the literature. Few laboratories in the world have access to research reactors and testing facilities to study high-dose neutron irradiation effects, keeping in mind the high redun-dant radioactivity of neutron-irradiated metallic samples.

Nevertheless, the first published data on the behaviour of UFG steels, produced by ECAP and subjected to neutron irradiation, confirmed a positive effect of nanostructuring on their radiation tolerance. The authors of Refs. 56, 57) carried out investigations on 321 type stainless-steel subjected to 4 passes of ECAP and irradiated to 5.3 dpa at 350°C in the conditions of research atomic reactor BOR-60.58) _Tensile

tests at 20 to 650°C were performed to characterize the

impact of neutron irradiation on the mechanical properties of UFG steel in comparison with its reference state obtained by cold-working.56,57)The first results did not reveal a striking difference in mechanical properties between the two steels after irradiation (Fig. 6(a)). However, the study demonstrated that the UFG steel exhibited higher strength and higher resistance to irradiation hardening in the entire testing temperature range.56,57) Interestingly, the UFG steel, having notably less ductility than a conventional material in the un-irradiated stated, showed similar values of elongation to failure at testing temperatures exceeding 500­550°C after irradiation (Fig. 6(b)).

Thesefirst results did not establish a marked advantage of grain size refinement to increase the ability of this material to sustain a considerable damaging dose. However, it should be noted that the microstructure of the investigated ECAP steel was not completely refined, grain size varied from several hundreds of nanometres to tens of microns, grain boundaries had preferably low misorientations. These studies showed the importance of optimizing SPD processing routes to fabricate UFG specimens with specified parameters describing grain size and grain boundaries. It is important to notice that in Ref. 56), the researchers found that nanostructuring a 321 steel, even non-optimally, slightly enhanced its corrosion resistance in a sodium chloride environment after irradiation, which is consistent with the data reported in Ref. 44).

Later, Alsabbagh et al.59,60) _{investigated a UFG}

low-carbon steel-10 produced by ECAP-C and subjected to Fig. 5 Illustration of a transition in the recombination of radiation-induced interstitials (blue spheres) and vacancies (red spirals) with

temperature increasing.22)

(a) (b)

neutron irradiation up to 1.37 dpa. Mechanical tests and hardness measurements showed that the radiation-hardening effect was effectively suppressed in the UFG steel. They also demonstrated that a slightly increased strength for the irradiated UFG steel is related to precipitation of nano Mn­Si particles rather than to dislocation loop generation.

Maksimkin et al.61) featured neutron irradiation experi-ments carried out on 321 steel nanostructured by HPT. HPT was used to achieve UFG microstructure with a grain size in the range of 50­300 nm and mostly high angle grain boundaries. This UFG structure remained stable after irradiation with afluence of 2©1020_n/cm2_{. The study also}

showed that nanostructured steel had a high resistance to irradiation hardening while the corrosion property notably degraded. The authors linked the observed decrease in corrosion resistance to carbide precipitation.61) Strain-induced martensite, often found in HPT-produced 321 steel strongly depending on the SPD conditions,62) could also significantly affect the corrosion resistance.

A recent publication63) reported on changes in the microstructure and mechanical properties of conventional and ECAP 321 steels irradiated in a research nuclear reactor at 350 and 430°C to doses of 12 and 15 dpa, respectively. Annealing experiments before irradiation revealed that UFG steel with a 300 nm average grain size and high angle grain boundary fraction exceeding 50% was thermally stable up to 600°C. This result assured that no thermal grain growth was going to occur at the temperature of the irradiation experiment. Irradiated UFG steel showed a lower concen-tration of Frank loops compared to conventional 321 steel neutron irradiated at similar conditions.64) In addition, Shamardin et al.63)observed no void nucleation in the UFG steel, while swelling in coarse-grained steels can take place after irradiation with similarfluence at 450°C, as was shown for thefirst time in Ref. 65). Later studies reported the onset of void nucleation at a fluence of 5©1021n/cm2 and 370­ 380°C,66) _{which is considerably less than achieved in}

Ref. 63)®1.9©1022_n/cm2_.

Post-irradiation tensile tests at room temperature using standard samples confirmed UFG steel’s better resistance to irradiation hardening (Fig. 7(a)). Mechanical examination at

elevated temperatures (up to 650°C) confirmed the con-clusion reported in Ref. 57)®the strength of the irradiated SPD steel is higher compared to conventional steel in the entire testing temperature range. However, the finer grains and bigger fraction of high angle grain boundaries achieved in the UFG steel studied in Ref. 63) made this difference more substantial.

Apart from that, the UFG steel demonstrated surprisingly higher plasticity at a testing temperature of 500­650°C (Fig. 7(b)). Together with increased strength, this result shows the superior mechanical performance of the SPD steel in the environment of a nuclear reactor. This phenomenon could also indicate an ability of the UFG steel to resist high-temperature radiation embrittlement typical to conventional Cr­Ni steels.67)

Further investigations in the field are vitally needed to clarify uncovered gaps in exploring radiation tolerance issues concerning UFG and nanostructured materials, the effect of irradiation parameters (type, dose, temperature); the sink efficiency of different boundaries; the effect of defect distribution, precipitation and solute segregation in alloys with different chemical composition; different corrosion behaviour; and so on. On this basis, many additional ways to improve radiation tolerance of UFG materials could be found. Kurishita et al.68) _{managed to significantly enhance}

the mitigation of recrystallization and irradiation-induced embrittlement using grain boundary modification in a UFG material via precipitation and segregation of TiC constituents. Du et al.69) _{show an example of improving strength and}

radiation tolerance in combination with outstanding thermal stability for a nanocrystalline steel thanks to alloying. Additional capacity for increasing the density of sinks for radiation-induced defects in metallic materials could be provided by introducing nanotwins70) or via interface engineering with the help of SPD-producing nanolayered composites,53) and so on. These opportunities demand systematic and multidisciplinary studies involving nuclear engineering, materials science, computer simulation, solid state and radiation physics, and the development of SPD technologies.

(a) (b)

5. Conclusion

Grain refinement proved to be a powerful pathway to design high-performance radiation-tolerant materials. Bulk UFG metals and alloys produced by SPD and characterized by small grain size, fine nanostructural features, full density and macroscopic dimensions demonstrate excellent mechani-cal performance and improved resistance to model and neutron irradiation.

There are still many open questions covering numerous issues related to their behaviour at extra-high damaging doses, such as the corrosion property, the role of grain boundary parameters in interaction with point defects, a combination of alloying elements, etc. The physical back-ground of SPD materials’ performance in the conditions of a nuclear reactor core accounting for thermal, stress and irradiation effects must be clarified to purposefully enhance the radiation tolerance of bulk UFG billets to be used in advanced energy applications.

Acknowledgements

This work was partly supported by Saint Petersburg State University via the grant“Lot 3 Applied”(id:26130576). The authors would like to acknowledge the fruitful collaboration with the colleagues contributed to the joint research and discussions­X. Sauvage, A. Etienne, E. Hug, T. Bulanova, A. Korsakov, S. Dobatkin, O. Rybalchenko, B. Kalin, M. Ganchenkova, O. Emelyanova, P. Dzhumaev, F. Garner, L. Shao, M. Short, A. Kilmametov, Y. Ivanisenko, M. Abramova, I. Alexandrov, G. Raab and R.Z. Valiev.

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