Dpa Irradiation Research Articles

On the origin of superior stability of coherent nanoparticles under ion irradiation

Recently, a new anti-irradiation mechanism relying on reversible disorder-ordering transition of coherent nanoparticles was discovered, which significantly improves the microstructural stability and void swelling resistance of metallic materials. However, the factors that govern the outstanding stability and superb radiation tolerance are still not clear. Here, two kinds of FeCrAl alloys were designed, each strengthened by a different type of coherent phase: L21-ordered Fe2AlV and B2-ordered Ni (Al, Fe). It was observed that the two alloys exhibited disparate responses to high-dose ion irradiations at elevated temperatures. The L21-Fe2AlV precipitates were found to be completely dissolved after 50 dpa of ion irradiation at 500–600 °C, whereas the B2-NiAl precipitates remained stability even after 200 dpa irradiation. This research challenges the conventional wisdom that the stability of nanoparticles is governed by the balance between radiation-enhanced coarsening and radiation dissolution. Instead, it demonstrates the re-nucleation and subsequent interface-controlled solute reshuffling processes govern the stability of coherent chemically-ordered nanoparticles under radiation at high temperatures. We demonstrate that the low interfacial energy, which is related to both the simply chemically-ordered lattice structure and low lattice misfit interface, is crucial for enabling such short-range elemental reshuffling process to repeatedly form nanoprecipitates that cannot be suppressed by radiation.

Effect of initial microstructure on irradiation induced microstructure evolution in δ-ferrite in an austenitic stainless steel weld

The irradiation induced microstructure evolution of δ-ferrite in an austenitic stainless steel weld was investigated. To simulate long-term service in light water reactor (LWR) at operating temperature, accelerated thermal aging at 400 °C up to 30,000 h was performed. Then, proton irradiation up to 10 displacement per atom (dpa) was subsequently conducted for unaged and aged samples. For the unaged samples, formation of Cr-rich (α’) phase by spinodal decomposition increased with irradiation dose. Meanwhile, formation of G-phase was promoted after 1 dpa, but decreased after 10 dpa. For the aged samples, after 1 dpa irradiation, Cr-rich (α’) phase formed by spinodal decomposition during thermal aging was destroyed while population of G-phase increased. After further irradiation up to 10 dpa, α’ phase was re-precipitated and coarsened significantly, but the degree of which was significantly less compared to the unaged samples. Meanwhile, with increase of irradiation to 10 dpa, G-phase population increased for the aged samples. The difference in the evolution of microstructural features during irradiation were discussed in view of the initial microstructure of δ-ferrite.

In-situ investigation of helium effect on diverse precipitate/matrix interface areas in irradiated 5052 aluminum alloy

Due to its low neutron cross-section and high irradiation swelling resistance, 5052 aluminum alloy is extensively utilized in the core structure and irradiation shielding of research nuclear reactors. Indeed, irradiation-induced defects may accumulate preferentially in the precipitates of aluminum alloys during service, leading to prior impairment of these precipitates, thereby inducing the degradation of mechanical properties. The interaction between irradiation defects and precipitates under irradiation, however, remains inadequately understood. In this work, the microstructure of 5052 aluminum alloy irradiated in-situ with He+ at different temperatures and subsequent annealing was investigated by transmission electron microscopy. After 1.0 dpa irradiation at different temperatures, the evolution of He bubbles within the precipitate was greater than in the matrix, and manifested a pronounced growth threshold between 373 K and 473 K. During post-irradiation annealing, compared to the matrix, also the dynamic evolution of bubbles was observed on the inner side of the precipitate interface, where bubbles coalesced and rapidly dissolved. The dynamic behavior of He bubbles within different precipitates manifested a certain discrepancy, and sink strength of two typical Al-containing precipitates was estimated. The difference in bubble evolution can be attributed to the different sink efficiency and vacancy concentration gradient around the precipitate.

Redistribution of Nb and other alloying elements in Nb-doped Zr alloy under high dose ion irradiation

ABSTRACT To understand the microstructural changes in Nb-doped zirconium alloy cladding at high irradiation doses, the Mitsubishi Developed Alloy (MDA) cladding was ion- irradiated up to 40 dpa at 400 ◦ C and the distribution of alloying elements was in- vestigated using atom probe tomography (APT) and scanning transmission electron microscopy and energy dispersive X-ray spectroscopy (STEM-EDS). APT analysis revealed two types of Nb nanoclusters with different Nb contents after the ion irradia- tion. The Nb concentration in the matrix decreased significantly with ion irradiation up to 10 dpa and then decreased slowly to 40 dpa. STEM-EDS analysis estimated that the secondary phase precipitates (SPPs) of Zr(Fe,Cr,Nb)2 and Zr(Fe,Cr)2 were formed before the ion irradiation. Although Fe, Cr, and Nb were dissolved from the SPPs during the ion irradiation, these alloying elements still remained in the SPPs after 40 dpa irradiation. Other Fe-Cr nanoclusters, Nb-rich phase, and Fe-rich phase were also formed during the ion irradiation. These radiation-induced precip- itates were observed until after 40 dpa irradiation and were expected to be stable under irradiation conditions.

Dislocation loops and hardening of silicon ion irradiated W and W-3Re alloy at 400 °C and 550 °C

Si2+ irradiation with the dose of 33.8–100 dpa was used to study the dislocation loops and hardening in W and W-3Re alloy at 400 °C and 550 °C using transmission electron microscope (TEM) and nano-indentation tests. When irradiated at 550 °C, larger dislocation loops and numerous dislocation network formed with the dose increased from 33.8 dpa to 100 dpa for both materials. The density of dislocation loops might be saturated between 33.8–100 dpa for W and saturated at ∼53.8 dpa for W-3Re alloy. Compared with pure W, the mean size of dislocation loop was smaller in W-3Re alloy when irradiated to 33.8 dpa for both temperatures. Meanwhile, it was found that the irradiation hardening rate of W was larger than that of W-3Re alloy under the same irradiation conditions, indicating that the addition of 3 wt.% Re element could indeed decrease the irradiation hardening rate of W. The hardening rate might be saturated after ∼53.8 dpa irradiation for W-3Re alloy, which might be related to the saturated loop densities.

Short communication: Complete dissolution of MX-phase nanoprecipitates in fusion steels during irradiation by heavy-ions

New RAFM steels, with intragranular VN precipitates for enhanced creep strength, were irradiated with 2MeV Fe+ ions at 600 °C to explore MX-type precipitate response to extreme end-of-life conditions in breeder blanket components. Utilizing TEM, Analytical-STEM, and APT, the steels were characterized in the unirradiated and irradiated states. Unirradiated grains contained needle-like VN precipitates (∼1 × 1022/m2). However, throughout the 40–100 dpa irradiation damage layer, VN precipitates were completely annihilated. APT confirmed no fine-scale VN clustering, indicating re-solution with the matrix, suggesting instability. Further investigations via reactor irradiations are crucial to assess this instability under breeder blanket conditions.

Ten-Year Recycling of Vanadium Alloy in Fusion Reactors

Low-activation characteristics and dominant radioactive isotopes limiting materials recycling were analyzed for a fusion-reactor-grade vanadium alloy, NIFS-HEAT-2. In order to reduce contact dose rate after use in fusion reactors, purification of the base metal vanadium was examined by chemical aqueous separation, electron-beam vacuum melting and zone refining, focusing on the removal of the dominant high-activation impurities, such as Co, Cu, Fe, Nb, Ni and Mo. Based on the measured impurity levels, remote recycling of vanadium alloy is possible within ten years after use in fusion reactors under operation condition with 100 dpa irradiation. Early quasi-hands-on recycling requires further purification and re-design of alloy composition especially with low Ti and high Cr content. The present paper discusses status of material R&Ds for the ten-year recycling and impact on operation of fusion reactors.

Effect of the free surface on near-surface void swelling in self-ion irradiated single crystal pure iron considering the carbon effect

The influence of ion-incident free surfaces on void swelling in near-surface regions was systematically studied using self-ion irradiation of single crystal pure iron. The key interest was to evaluate the effect of free surfaces not only on the formation of void-denuded zones but also on the swelling in the region immediately adjacent to the void-denuded zone. The irradiation matrix included beam energies varying from 1.0 to 5.0 MeV, peak displacements-per-atom levels of 50 and 100 dpa, and irradiation temperatures at 425 °C (at 6 × 10−3 peak dpa/sec, 475 °C (at 1.2 × 10−3 peak dpa/sec) and 525 °C (at 6 × 10−3 peak dpa/sec). The observed denuded depth Δx obtained from transmission electron microscopy was modified to incorporate the sputtering effect. The derived activation energy governing the denuding process is Ex4=1.65±0.03eV, which is significantly higher than the vacancy migration energy EVm known for Fe to be 0.67 eV. This difference is speculated to be due to the strong effect of relatively small amounts of dissolved carbon at 103 appm, which reduces the effective vacancy mobility and thereby increases the effective migration energy. Based on the effective vacancy migration energy extrapolated from a previous first-principle calculation for Fe containing various carbon levels, rate theory calculation was used to model the surface influence and the results support the speculation of carbon influence. For the region immediately adjacent to the void-denuded zone, swelling is locally suppressed, rising to reach the bulk swelling at a depth increment essentially equal to the denuded width. This suppression effect causes shifting of swelling vs. local dpa curves obtained from different peak dpa irradiations. Such shifting is proposed in the present study as a method to define the boundary of the surface-induced suppressed swelling zone beyond which the swelling can be considered to be fully free of surface influence.

Heavy ion irradiation effects on CrFeMnNi and AlCrFeMnNi high entropy alloys

Co-free but Al-included medium/high entropy alloys (M/HEAs) have gained increasing interests due to their lower cost and the potential to tune the multi-phase microstructure. The irradiation response of two Co-free HEAs, face-centered cubic (FCC) CrFeMnNi with limited Cr enriched α′ phase and body-centered cubic (BCC) AlCrFeMnNi with B2 s phase and nanoprecipitates were explored. Ion irradiations using 5 MeV Fe2+ ions were performed at 500 °C to a peak fluence of 50 and/or 100 displacements per atom (dpa). In dual-phase AlCrFeMnNi, there was no significant radiation induced segregation or chemical intermixing at the coherent matrix (FeCrMn-rich)/second phase (AlNi-rich) boundaries. In CrFeMnNi, limited voids were only detected at the peak damage location of ∼ 50 dpa. On the other hand, voids were widely distributed in AlCrFeMnNi: under 50 and 100 dpa irradiation conditions, voids were found with larger dimension and denser distribution in the FeCrMn-rich matrix, smaller and slightly lower density in an AlNi-rich second phase. In addition, the diameter of the FeCMn-rich nanoprecipitates didn't reveal any tendency of dissolution or growth. This is correlated with their superior structural stability against irradiation. Significant radiation-induced hardening (increases from 3.8 ± 0.2 GPa to 4.7 ± 0.6 GPa) was measured in CrFeMnNi, but only ∼ 4% hardness increase (from 7.4 ± 0.8 GPa to 7.7 ± 0.4 GPa) was noted in AlCrFeMnNi. In addition to the radiation-induced defects, such as voids, dislocation loops and point defects, other factors, such as chemical short-range ordering may play an important role.

In-situ microstructure observation of oxidized SiC layer in surrogate TRISO fuel particles under krypton ion irradiation

In accidental scenarios of high temperature gas-cooled reactors, both oxidation of and irradiation to the SiC layer in tri-structural-isotropic (TRISO) fuel particles can change the microstructure and integrity of the fuel elements. In the present study, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under 1.2 MeV krypton ion irradiation was observed by in-situ transmission electron microscopy. The SiC layers oxidized in water vapor at 1200 °C were irradiated azt room temperature and 800 °C and at damage levels of 0.28–11.2 dpa, respectively. SiC and SiO2 were found to still be in their crystal structures at the damage level of 11.2 dpa at 800 °C, while SiC was observed to have been amorphized at only 0.56 dpa irradiation at room temperature. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference.

Structural stability of β -Ga2O3 under ion irradiation

β-Ga2O3 was suggested to have excellent irradiation hardness which makes β-Ga2O3-based devices extremely attractive for nuclear and space applications. To discern the fundamental nano-scale structural changes with irradiation, an in situ irradiation experiment in a transmission electron microscope was carried out using 400 keV Ar ions of fluences up to 8 × 1015 cm−2 (equivalent to four displacements per atom). Contrary to previous works, which indicate a phase transition of β-Ga2O3 into the κ polymorph, the β-Ga2O3 structure was found to remain intact throughout except for (i) anisotropic lattice distortions, which are most significant at low levels of irradiation, and (ii) the appearance of additional weak reflections above 2 dpa irradiation. The origin of the extra reflections is discussed.

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe2+ up to 600 peak displacement-per-atom (dpa) at 450 ℃. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. The comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

Effect of purity on the vacancy defects induced in self–irradiated tungsten: A combination of PAS and TEM

Vacancy defects in tungsten induced by self-ion irradiation were characterized by positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Taking advantage of their complementarity, defects ranging from single vacancies up to vacancy clusters were detected, and the effects of sample purity (99.95 wt.% and 99.9999 wt.%) on their formation and evolution under low damage dose (0.01 to 0.02 dpa) irradiation with 20 and 1.2 MeV-W ions were studied at room temperature (RT), 500 and 700 °C respectively. When irradiation was performed at RT, PAS probed mainly single vacancies. At high irradiation temperatures, larger vacancy clusters were detected by both techniques. Both PAS and TEM revealed larger vacancy clusters in the purest samples after irradiation at 500 and 700 °C. A significant difference in the evolution of vacancy-type defects related to sample purity was observed. In view of the properties (migration, trapping) of some light-element impurities (LEs), present at high concentrations, determined by first-principles calculation, it is reasonable to assign the effect of purity during the irradiation response to the formation of vacancy-impurity complexes, which is critical for the understanding of the behavior of tungsten in the future fusion reactor.

MoS2 Nanocomposite Films with High Irradiation Tolerance and Self-Adaptive Lubrication.

Although nanostructures and oxide dispersion can reduce radiation-induced damage in materials and enhance radiation tolerance, previous studies prove that MoS2 nanocomposite films subjected to several dpa heavy ion irradiation show significant degradation of tribological properties. Even in YSZ-doped MoS2 nanocomposite films, irradiation leads to obvious disordering and damage such as vacancy accumulation to form lamellar voids in the amorphous matrix, which accelerates the failure of lubrication. However, after thermal annealing in vacuum, YSZ-doped MoS2 nanocomposite films exhibit high irradiation tolerance, and their wear duration remains unchanged and the wear rate was nearly three orders of magnitude lower than that of the as-deposited films after 7 dpa irradiation. This successful combination of anti-irradiation and self-adaptive lubrication mainly results from the manipulation of the nanosize and the change of composition by annealing. Compared with the smaller nanograins in as-deposited MoS2/YSZ nanocomposite films, the thermally annealed MoS2 nanocrystals (7-15 nm) with fewer intrinsic defects exhibited remarkable stabilization upon irradiation. Abundant amorphous nanocrystal phases in ion-irradiated thermally annealed films, where each has advantages of their own, greatly inhibit accumulation of voids and crack growth in irradiation; meanwhile, they can be easily self-assembled under induction of friction and achieve self-adaptive lubrication.

Effects of nanosized precipitates on irradiation behavior of CoCrFeNi high entropy alloys

Evolution of nanosized precipitates in typical (CoCrFeNi)94Ti2Al4 high entropy alloys under irradiation and its effect on the irradiation resistance were investigated in detail. The as-solutionized samples and the aged samples with massive nanosized precipitates were irradiated at ambient temperatures with a dose ranging from 10 to 49 dpa of 4 MeV Au ions. It was found that the ordered precipitates became disordered quickly at the early stage under 10 dpa irradiation, accompanying with the dissolution process which became more and more severer at higher doses. After irradiation, the hardness slightly changed in the aged specimens containing massive precipitates, but increased by> 30% in the as-solutionized ones with no precipitates. The obvious hardness increment in the as-solutionized samples is due to the creation of defects while the small variations of the hardness in the aged specimens can be ascribed to the disordering and dissolution of precipitates. In addition, the size of dislocation loops induced by the irradiation in the aged specimens is much smaller than that in the as-solutionized samples. We confirmed that presence of the nanosized precipitates delayed defect evolution and reduced the sizes of dislocation loops.

Oxide dispersoid coherency of a ferritic-martensitic 12Cr oxide-dispersion-strengthened alloy under self-ion irradiation

Ferritic-martensitic oxide-dispersion-strengthened (ODS) alloy has shown excellent mechanical property and high radiation tolerance. However, the stability of dispersoids during displacive irradiation in the individual ferritic and martensitic phases is unclear. Here, the correlation among dispersoid coherency, size, density, and matrix phase are studied in dual-phase 12Cr ODS after 100 peak displacements-per-atom (dpa) irradiation at 475°C, using 3.5 MeV Fe2+ self-ions. The size and density changes of coherent and incoherent dispersoids were analyzed as a function of irradiation depth. The average dispersoid size decreased after irradiation in both phases, and the large incoherent dispersoids in the tempered martensite phase underwent a more dramatic change than those in the ferrite phase. The dispersoid density significantly increased in the ferrite phase within the irradiated region, mostly resulting in an increase in coherent dispersoid density. On the other hand, only a small change in density was observed in the tempered martensite phase for both coherent and incoherent dispersoids. This study shows that dispersoid evolution is dramatically different in ferrite and tempered martensite phases.

In Situ Micro-Pillar Compression to Examine Radiation-Induced Hardening Mechanisms of FeCrAl Alloys

The effects of 5 MeV Fe2+ ion irradiation at 300°C on the microstructure evolution and deformation behavior of a FeCrAl C26M alloy are presented. It has been found that dislocation loop density increases an order of magnitude from 1 dpa to 16 dpa irradiations, whereas, the dislocation loop size saturates with increasing damage. Micropillars, 600 nm in diameter and 1.3 µm in height, were fabricated and compressed inside grains with <001>, <011> and <111> crystallographic orientations, respectively. {112} <111> has been identified as the primary slip system in both unirradiated and irradiated alloys. The increase in yield stress after irradiation is observed with measurable variation along <001> and <011> vs. along <111>. By applying the Orowan dispersed barrier model, the increase of yield stress is found mainly due to the slip resistance of radiation generated defect loops. Detailed transmission electron microscopy (TEM) studies were performed to quantify the Burgers vector and the distribution of irradiation induced dislocations at elevated strains. It is revealed that localized shear instability is caused by avalanche slip events of ½<111> dislocations gliding out of tested pillars. Simultaneously, a large number of sessile/immobile <100> dislocations formed in the vicinity of slip band, leading to the hardening at elevated strains.

In-situ observation of nano-oxide and defect evolution in 14YWT alloys

Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450 °C up to 10 dpa. It has been found that irradiations at 450 °C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of <111> loops compared to the literature values. High resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y2Ti2O7 particles with pyrochlore crystal structure after both RT and 450 °C irradiations, similar to those observed before irradiation.

Improvement in irradiation resistance of FeCu alloy by pre-deformation through introduction of dense point defect sinks

The irradiation resistance of pre-deformed FeCu alloy was studied using a 3 MeV Fe ion irradiation experiment at room temperature in comparison with that of the as-received sample. Nanoindentation and atom probe tomography (APT) were used to characterize the mechanical properties and solute distribution. The stress–strain curve obtained by nanoindentation shows that the yield strength (σ0.2) of the pre-deformed sample is unexpectedly reduced with an increase in the irradiation dose to five displacements per atom (dpa). We suggest that it results both from the decrease in the dislocation density and the suppression of defects during irradiation. APT shows that the nucleation of the Cu cluster is suppressed; however, its growth is promoted in the pre-deformed sample, resulting in the formation of sparse and coarse clusters at 1 dpa irradiation. These coarse Cu clusters were then unexpectedly refined to finer grains with an increase in the irradiation dose to 5 dpa. Theoretically, the improvement in the resistance to irradiation in the pre-deformed sample is attributed to the dense point-defect sinks, that is, the dislocations and grain boundaries introduced by pre-deformation. In addition, the contributions of the dislocations and grain boundaries to the sink strength are estimated for both the as-received and pre-deformed samples. The results indicate that dislocations, rather than grain boundaries, play a major role after deformation.

Development of long-term irradiation testing technology at HANARO

As the High Flux Advanced Neutron Application Reactor (HANARO) has been recently required to support new R&D relevant to future nuclear systems requiring a much higher neutron fluence, the development of irradiation capsule technology for long-term irradiation testing was performed in three steps (3, 5, 10 dpa). At first, several design improvements of a standard capsule were suggested based on a failure analysis of the capsule and successfully applied for irradiation testing at HANARO at up to eight reactor operation cycles equivalent to 3 dpa. Based on a schematic stress analysis of the vulnerable parts of the previous capsule, an optimized design of the capsule was made for 5 dpa irradiation. The newly designed capsule was safely out-pile tested up to 450 days, which was equivalent to 5 dpa irradiation in the reactor. The test results were submitted to the Reactor Safety Review Committee of HANARO and irradiation testing for 5 dpa was approved. The capsule was also successfully out-pile tested to evaluate the possibility of irradiation testing for 10 dpa. For a higher neutron fluence exceeding 10 dpa, new capsule technologies, including a new capsule that has a different bottom design and neutron flux boosting capsule, were also suggested.