Peak Damage Level Research Articles

Influence of suppressed blistering by heavy ion pre-damage on deuterium retention in tungsten

Surface damage and fuel retention are one of the major threats to the performance of plasma-facing materials (PFMs) in ITER and future fusion reactors. This work aims to investigate the influence of suppressed blistering by heavy ion pre-damage on deuterium (D) retention in tungsten (W). Recrystallized W samples were irradiated with 3.5 MeV iron (Fe13+) ions at room temperature to create displacement damage with a peak damage level of 0.1 dpa. Afterwards, a series of low-energy (38 eV) D plasma exposures were performed at 500 K. Three exposure fluences below and above the blistering threshold (5 × 1024 ∼ 3 × 1025 D m−2) of the pre-damaged W are selected to decouple and couple the damage-induced defects and blistering-induced defects, respectively. Surface observations show that no blisters are formed in un-damaged W after low-fluence D exposure (5 × 1024 D m−2), whereas severe blistering (surface coverage ratio: 34.2 %) occurs in the high-fluence case (3 × 1025 D m−2). In contrast, only a small number of blisters (4.2 %) are formed in Fe-damaged W when D fluence reaches 3 × 1025 D m−2. Moreover, Fe pre-damage increases D retention by a factor of 3.33 and 1.20 at the low-fluence (5 × 1024 D m−2) and medium-fluence (1 × 1025 D m−2) D exposure, respectively. While in the high-fluence case (3 × 1025 D m−2), the enhancement effect of D retention in Fe-damaged W is significantly weakened, such that retention is lower for damaged W, probably due to the significant suppression effect on surface blistering and its accompanying defect formation. This work highlights the suppressed blistering-induced D retention by pre-existing damage in W.

A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals

We propose and demonstrate a microstructurally-based experimental method to quantitively determine the depth regions in self-ion-irradiated metals that are affected by the injected interstitial effect and various surface effects, focusing on the choice of safe analysis zones to minimize the impact of these phenomena. The goal is to define the depth ranges where extracted data can be confidently applied to ion-neutron correlations for reactor application. Since ion energies in the range of 1–5 MeV are most frequently employed by the radiation effects community, irradiations were conducted at four energies in this range, all proceeding at 475 °C. The experiment was conducted on relatively pure single crystal iron to focus only on physical phenomena, avoiding the influence of possibly confounding chemical or segregation processes. Care was also taken to minimize the influence of other physical factors such as crystalline orientation. It was shown that, at 475 °C, ion energies of ≤1 MeV were too shallow in penetration and did not yield a safe depth range, but irradiations at 2.5 MeV and above yielded useful safe zones with predicted swelling behavior becoming independent of ion energy. The surface-affected zone width was found to be roughly twice that of the void-denuded zone width and to be independent of accumulated displacement dose. The largest injected-interstitial effect arises from the injected interstitial depression of void nucleation and growth. The interstitial-affected region starts at about one half of the projected range and does not show any “spreading” of its influence in depth as the peak damage level increases from 50 to 100 dpa. This study provides some confidence that enhances the credibility of ion simulation when applied to prediction of void swelling in neutron environments.

Dynamic substrate reactions during room temperature heavy ion irradiation of CoCrCuFeNi high entropy alloy thin films

High entropy alloys (HEAs) are promising materials for various applications including nuclear reactor environments. Thus, understanding their behavior under irradiation and exposure to different environments is important. Here, two sets of near-equiatomic CoCrCuFeNi thin films grown on either SiO2/Si or Si substrates were irradiated at room temperature with 11.5 MeV Au ions, providing similar behavior to exposure to inert versus corrosion environments. The film grown on SiO2 had relatively minimal change up to peak damage levels above 500 dpa, while the film grown on Si began intermixing at the substrate–film interface at peak doses of 0.1 dpa before transforming into a multi-silicide film at higher doses, all at room temperature with minimal thermal diffusion. The primary mechanism is radiation-enhanced diffusion via the inverse Kirkendall and solute drag effects. The results highlight how composition and environmental exposure affect the stability of HEAs under radiation and give insights into controlling these behaviors.

Dislocation structure of tungsten irradiated by medium to high-mass ions

Single crystalline tungsten was irradiated by the medium-mass ion Si with 7.5 MeV and high mass-ion W with 20.3 MeV up to a calculated peak damage level of 0.04 dpa and 0.5 dpa. The obtained dislocation structure of the damage zone was investigated by transmission electron microscopy and systematically compared with each other. Bright-field kinematical images were taken under four different two-beam diffraction conditions g = −200, 020, −110, 110 close to the [100] zone axis. The observed damage depth and damage peak position is in good agreement with the SRIM calculated damage depth profiles. The dislocation structures were investigated at the region of the damage peak because there the damage levels are comparable. In both irradiations (Si and W), the dislocation structures were similar. At the low damage level of 0.04 dpa dislocation loops and dislocation-loop clusters were found. The size of the dislocation loops in the W-irradiated tungsten sample was up to 20% higher than for the Si-irradiated sample. At the high damage level of 0.5 dpa a dislocation network consisting of dislocation-loop chains and dislocation lines was found for both irradiations. The dislocation line density was about 12% higher for the W-irradiated sample. Through comparison of the damage zone to SRIM damage depth profiles it was found that the transition from dislocation loops and dislocation-loop clusters to an ordered dislocation network takes place at about 0.08–0.1 dpa. Despite the large differences in ion mass and irradiation energy the dislocation structures were very similar.

Dependence of blistering and deuterium retention on damage depth in damaged tungsten exposed to deuterium plasma

The effect of different damage depth on blistering and deuterium (D) retention has been investigated in heavy-ion-damaged tungsten (W) with exposure to D plasma (40 eV, 1 × 1022 ions m−2 s−1) at 550 K. Different damage depths are realized via copper (Cu) ion irradiation with energies of 1, 3, and 6 MeV on W samples with the same calculated peak damage level of 0.5 dpa. The plasma-induced blister density reduces with increasing damage depth, which is explained based on the recently proposed dislocation nucleation mechanism of blistering. Comparison of D retention measured by nuclear reaction analysis (NRA) and thermal desorption spectroscopy (TDS) reveals that retention at depths larger than 7.4 μm—which is far beyond the ion damage depth—increases with damage depth. Such a phenomenon indicates a gradual increase of diffusion flux inside the damaged sample with the increasing damage depth. It is suggested that it originates from the observed difference in blister density. Besides the widely acknowledged enhanced D retention due to ion damage, this work shows a strong impact of the damage depth on blistering such as the blister density, and by which the D diffusion flux inside W and total D retention are further affected.

Deuterium retention in W and binary W alloys irradiated with high energy Fe ions

To investigate systematically the effects of Re and other elements on deuterium (D) retention, W and binary W-alloys (Re, Mo and Ta) were irradiated with 6.4 MeV Fe ions at high temperatures (1073–1273 K) and then exposed to D2 gas at 673 K. Depth profiles of D were measured by nuclear reaction analysis (NRA), and D retention was determined by thermal desorption spectrometry (TDS) and NRA. The addition of 5 at.% Re into W reduced the content of D trapped at radiation-induced defects created by irradiation with the Fe ions at 1273 K to the peak damage level of 5 displacements per atom (dpa). At Re fractions of 1, 3 and 5 at.%, comparable effects on D retention were observed after irradiation to 0.5 dpa at 1073 K. The addition of Mo and Ta to W resulted in no visible effects in D retention.

Effects of temperature on helium bubble behaviour in Fe–9Cr alloy

In the present study, Fe–9Cr model alloys were irradiated by 275 keV He+ ions at 500 to 800 °C to a peak damage level and implanted helium content of 0.5 dpa and 0.75 at% He, respectively. Effects of temperature on He bubble behaviour were studied by transmission electron microscopy (TEM) to reveal the temperature effects. The bubble spatial distribution changes from homogenous at 500 and 600 °C to heterogeneous at 700 and 800 °C, due to preferential bubble precipitation on dislocations and grain boundaries at high temperature. Faceted bubbles with low energy faces along the {001} planes were present at all four temperatures and are attributed to surface energy minimization. Bubble precipitation on the GBs became more pronounced with increasing temperature. Meanwhile, the cavity swelling increased with increasing temperature, from 0.59% at 500 °C to 8.21% at 800 °C with a sharp increase above 700 °C. By analysing the temperature dependence of bubble size and density, a transition temperature between low and high temperature regimes was obtained in terms of different activation energies which are related to different bubble nucleation and formation mechanisms.

Micro-compression testing of gold ion irradiated zirconolite glass-ceramics as nuclear waste forms

The mechanical properties of zirconolite glass-ceramics designed for immobilising actinide-rich nuclear wastes, have been investigated through micro-compression tests pre and post 12 MeV gold ion irradiation at room temperature to a peak damage level of 100 displacements per atom (dpa). Whilst zirconolite crystals have become amorphized because of irradiation, transmission electron microscope analysis found no voids within the irradiated zirconolite. Nevertheless, voids of ∼25 nm in diameter were observed within the glass phase and appear to be homogeneously distributed. Surprisingly, micro-compression testing found the zirconolite glass-ceramic to exhibit a lower modulus, lower compression yield and greater toughness post irradiation.

Numerical simulation of reinforced concrete structures under impact loading

AbstractThis study presents the performance of a combined finite‐discrete element method for prediction of the structural response of reinforced concrete beams under impact loading. A combination of finite and discrete element methods enables the modelling of the concrete and the reinforcement before the concrete cracking, as well as a discontinuous nature of the concrete caused by fracture and fragmentation under high impact loading. Discretization of the concrete with triangular finite elements is coupled with one‐dimensional reinforcing bars embedded inside the concrete finite elements. The cracking in the concrete activates the joint elements used to simulate the non‐linear behavior of both concrete and reinforcement. Numerical analysis based on experimental test data has been carried out to simulate the main features of the reinforced concrete beams impacted by free‐falling drop‐weights. A high level of accuracy was demonstrated in various comparisons between the experimental tests and the analysis results, including peak displacement, crack pattern, damage level and failure modes of reinforced concrete beams.

Understanding ion irradiation resistance of a silicon diffused nanostructured ferritic alloy-chromium carbide–carbon composite

Cladding materials play the essential role in the safe, efficient, and economic operation of a nuclear reactor. In this work, a silicon diffused nanostructured ferritic alloy-chromium carbide–carbon (NFA|Si-Cr3C2-C) composite is studied as a nuclear fuel cladding material and ion irradiation with highly energetic Fe++ ions has been carried out to induce a peak damage level of ∼250 dpa. Elemental distribution, phase composition, and defect structure are investigated in order to understand the ion irradiation response of the NFA|Si-Cr3C2-C composite. SRIM simulation reveals displacement damage profiles for different phase compositions. The NFA|Si phase has good ion irradiation resistance with newly induced dislocation loops. The C phase in the NFA|Si matrix is prone to being partially amorphized whereas the Cr3C2 phase in the NFA|Si matrix is completely amorphized. This work offers important guidance for nuclear cladding material development.

An Improved Near Term 14 MeV Neutron Test Facility for Fusion Power Plant Materials

A near term, low cost 14 MeV neutron materials test facility has been designed that allows significant radiation damage (dpa, appm He, etc.) levels to be achieved typical of those that will be experienced in DT Demonstration or commercial DT power plants. The design described in this paper produces peak damage levels of ≈4–6 dpa/fpy in 15 cm3 and has ≈600 cm3 test volume covering the damage range from 1 to 6 dpa/fpy. The total active tritium inventory in the test facility is less than 1 g and the overall construction costs are also roughly unchanged from an earlier (2015) design. The time to initial operation remains at ≈4 years from the start of construction because it builds on an on-going project for radioisotope production already under construction. This latest facility design has the possibility to provide a 2 MW-y/m2, 14 MeV neutron exposure to first wall materials in less than 4 fpy’s of operation.

Radiation response of alloy T91 at damage levels up to 1000 peak dpa

Ferritic/martensitic alloys are required for advanced reactor components to survive 500–600 neutron-induced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 °C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500–600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

The influence of the annealing temperature on deuterium retention in self-damaged tungsten

The influence of the annealing temperature on deuterium retention was studied for self-ion damaged tungsten in the range from 600–1200 K. Samples were damaged by 20 MeV W ions at room temperature to the peak damage level of 0.5 dpa. Samples were then annealed at the desired temperature for 1 h and exposed to deuterium atom beam with the flux of for 144 h to populate the remaining defects. An unannealed sample was also used as a reference. Nuclear reaction analysis technique was used for deuterium depth profile analysis and thermal desorption was performed on the same samples to measure the amount of total retained D. Scanning transmission electron microscopy was used for the calculation of dislocation densities in the samples. After annealing at 1200 K approximately 66% of those initial defects which retain deuterium were annealed.

Transformation of YSZ under high fluence argon ion implantation

In this work, we present the effect of extremely high fluence ion implantation on microstructure of single crystalline YSZ samples with three major low index orientations: (100), (110) and (111). The samples were implanted at room temperature with 150keV Ar+ ions to a fluence of 1×1017Ar/cm−2 corresponding to the peak damage level of ∼120dpa and peak Ar atom concentration of ∼12at.%. Rutherford backscattering/channeling spectrometry (RBS/C), scanning electron microscopy (SEM), transmission electron microscopy (TEM) and associated analytical tools were used to determine the orientation dependent damage, surface morphology, and microstructure modifications of the implanted layers. Ar+ ion implantation resulted in formation of severely damaged layers, which however remained crystalline. The damage peak maximum, determined by RBS/C, indicated that the fourth damage accumulation stage, previously predicted for Ar-implanted YSZ, was achieved. The (110) oriented YSZ demonstrated slightly better radiation tolerance, as observed by RBS/C, compared to the other low index orientations. Microstructural studies revealed large cavities aligned parallel to the specimen surface, which emerged in a form of circular blisters on the surface. The origin of the cavities was related to the segregation of Ar atoms into pressurized gas filled bubbles. The crystallographic anisotropy of microstructural parameters (thickness of the damages layer, surface blister density and diameter, cavity dimensions) remains uncertain.

Effect of high fluence Au ion irradiation on nanocrystalline tungsten film

Irradiation-induced microstructure evolution and phase transformation have been studied on the nanocrystalline tungsten film with simple cubic structure (β phase). The samples were irradiated by 4MeV Au ions at different fluences ranging from 6×1014 to 1×1016ions/cm2, corresponding to the peak damage levels from 6 to 100dpa (displacements per atom). Analysis by GIXRD and TEM reveals that the structure of tungsten film undergoes a distinct transition from simple cubic phase to bcc phase when the fluence is above 3×1015ions/cm2 (30dpa) and grain growth occurs with the increasing fluence. Radiation-induced small voids are clearly observed in the interior of coarsened grains at the highest fluence, 1×1016ions/cm2 (100dpa).

Characterization of irradiation damage distribution near TiO2/SrTiO3 interfaces using coherent acoustic phonon interferometry

We apply ultrafast coherent acoustic phonon interferometry to characterize the distribution of the radiation damage near the TiO2/SrTiO3 interfaces. We show that the optical and mechanical properties of anatase TiO2 remain unaffected by the radiation dosages in the 0.1÷5 dpa (displacements per atom) range, while the degraded optical response indicates a significant defect accumulation in the interfacial region of SrTiO3 at 0.1 dpa and subsequent amorphization at 3 dpa. Comparison between the theoretical simulations and the experimental results reveals an almost threefold reduction of the sound velocity in the irradiated SrTiO3 layer with peak damage levels of 3 and 5 dpa.

Hydrogenic retention in irradiated tungsten exposed to high-flux plasma

Two sets of identical tungsten (W) targets are irradiated at 300 K with 12.3 MeV W4+ ions to peak damage levels ranging from 0.5 to 10 displacements per atom (dpa). This results in a damage profile that is peaked at ∼0.8 µm and extends to a depth of ∼1.5 µm. Both sets of targets are exposed to high-density (ne,center = 3 × 1020 m−3), low-temperature (Te,center = 1.6 eV) deuterium (D) plasma in Pilot-PSI. One set of irradiated targets is exposed at high surface temperatures (TW = 950–680 K) and the other at low surface temperatures (TW = 480–340 K). The surface temperature is determined by the local plasma conditions. Nuclear reaction analysis (NRA) is used to determine the D depth profiles at specific radial locations, thus giving a surface temperature scan of the D retention in the damaged W. Global retention is determined by thermal desorption spectroscopy, which yields total D retained in the target and also gives information of the different types of lattice defects that are trapping the D in the W lattice. The main results are that there is no measurable difference between the different dpa levels, implying a saturation of the retention enhancement at a level ⩽0.5 dpa. For both irradiated and unirradiated tungsten, a peak in the retention is seen at TW = 480 K; however, the W4+ irradiation clearly enhances the retention. This enhancement is also temperature dependent and increases with increasing surface temperature up to an enhancement by a factor of 15–23 at TW = 950 K. At the lowest surface temperatures, a fluence dependence appears since the implanted deuterium is diffusion limited to only a small fraction of the irradiated zone. TDS spectra show an enhancement of both low-energy trap sites and high-energy trap sites. For these conditions, diffusion-limited, low fill fraction trapping determines the hydrogenic retention of the W.

Defect production in silicon carbide irradiated with Ne and Xe ions with energy of 2.3 MeV/amu

In the present work specimens of mono-crystalline silicon carbide (4H polytype) were irradiated to three successively increasing ion fluences ranging from 7.2×1014 to 6.0×1016ions/cm2 (corresponding to the peak displacement damage of 1, 4 and 13dpa) with Ne and Xe ions respectively with the energy of 2.3MeV/amu. The irradiated specimens were subsequently annealed at temperatures of 1173 and 1273K. Defect structure was investigated with transmission electron microscopy (TEM) using a cross-sectional specimen preparation technique.The typical microstructures of the annealed specimens irradiated with Ne or Xe ions to high fluences are characterized by small gas bubbles in high concentration in the peak damage region and black dots and dislocation loops (located in the basal plane) in a shallower and broader depth region. Larger dislocation loops were observed in the Xe-ion irradiated specimen than in the Ne-ion irradiated specimen at the same peak damage level. The enhanced formation of dislocation loops in the case of Xe-ion irradiation is understandable by assuming stronger inclination of heavier inert-gas atoms to occupy substitute site in the peak damage region.

L-glutamine and transforming growth factor-alpha enhance recovery of monoacylglycerol acyltransferase and diacylglycerol acyltransferase activity in porcine postischemic ileum.

Recovery of the ability to digest and absorb lipids is essential to the maintenance of normal nutrition in infants with bowel damage. Two intrinsic microsomal enzymes, monoacylglycerol acyltransferase (MGAT) and diacylglycerol acyltransferase (DGAT), catalyze the major pathway for intestinal triacylglycerol biosynthesis. This study describes the effects of intestinal ischemia on epithelial DGAT and MGAT activities and their recovery in response to two luminal treatments: L-glutamine (Gln), the primary intestinal fuel, and transforming growth factor-alpha (TGF-alpha), a mitogenic hormone similar to epidermal growth factor present in breast milk. Ischemic damage and recovery were analyzed in mucosa from Thiry-Vella loops in the mid-ileum of 7-wk-old pigs. Loops were subjected to 2-h occlusion of local mesenteric arteries, followed by 6 or 72 h of recovery in the presence of luminal glucose (control), Gln, or TGF-alpha. Ischemic tissue followed by 6-h recovery exhibited an approximate 50% decrease in both MGAT and DGAT activities compared with nonischemic loop tissue. At 72 h, MGAT and DGAT recovery in Gln plus TGF-alpha-treated loops was significantly greater than their corresponding 6-h peak damage levels (p < 0.05). From 6 to 72 h, MGAT increased 4-fold and DGAT increased 3.6-fold after Gln plus TGF-alpha treatment. With other treatments, MGAT and DGAT activities increased <2.5-fold from 6 to 72 h. This study shows that intestinal MGAT and DGAT activities decrease after ischemic damage, yet recover rapidly in bowel exposed to Gln and/or TGF-alpha. By stimulating the rate of recovery of the villi and lipid synthesizing enzymes, these treatments could improve the efficacy of enteral feeding in infants recovering from bowel damage.

The irradiation damage response of MgO · 3Al 2O 3 spinel single crystal under high-fluence ion-irradiation

Non-stoichiometric single crystal spinel samples with composition MgO · 3Al 2O 3 were irradiated near (001) orientation with 400 keV Xe ++ ions at 100 K to a dose of 1 × 10 16 Xe/cm 2, representing a peak damage level of 25 displacements per atom (dpa). Damage accumulation in the irradiated surface was assessed by cross-sectional transmission electron microscopy. A continuous crystalline layer was observed in the implanted region. The layer is oriented with the underlying substrate, but is highly defected and possesses a symmetry different than that of the bulk. This change in symmetry is revealed by the absence of diffracted intensity at first-order spinel reciprocal lattice positions. These results are compared with earlier observations on equimolar (001) MgO · Al 2O 3 single crystals irradiated with 400 keV Xe ++ ions to the same dose and at the same temperature [N. Yu, K.E. Sickafus and M. Nastasi, Phil. Mag. Lett. 70(4) (1994) 235 (Ref. [1])].