Irradiation-induced Dislocation Loops Research Articles

In-situ TEM characterization of dislocation loops in tungsten under simultaneous helium and hydrogen irradiation

The interaction among He, H, and irradiation-induced dislocation loops in pure tungsten was examined through 30 keV simultaneous He+ and H2+ irradiation at 1073 K. Three different He/H concentration ratios (0.1, 1, and 5) were utilized. This study involved the characterization of the number density, average size, and size distribution of loops by varying the dose percentage of He and H. Increasing the dose percentage of He facilitated the loop nucleation. Increasing the total irradiation dose not only increased loop nucleation sites but also changed the loop size distribution from unimodal to bimodal. The specific Burgers vectors and habit planes of loops were carefully determined using the inside-outside method in combination with the g·b invisibility criterion and crystallographic map. The proportion of interstitial 1/2<111> loops exceeded 95 %, in which four Burgers vector variants of 1/2[111], 1/2[1¯11], 1/2[1¯1¯1], and 1/2[11¯1] were identified. Increasing the dose percentage of He led to an even distribution of Burgers vector variants. The {110} habit planes were found to deviate at an angle of 35° from the {111} ones. A majority of {111} habit planes were found compared to {110} for 1/2<111> loops. Altering the dose percentage of He or H showed no apparent effect on the habit planes of {111} or {110} for 1/2<111>loops, nor did it affect their average size when situated on {111} or {110}.

The effect of microstructure on recovery and recrystallization after annealing of two neutron irradiated ITER specification tungsten

The study of the recrystallization resistance of engineering grade tungsten (W) prior to and after neutron irradiation as well as subsequent thermal annealing provides valuable insight into neutron induced defect interactions and their kinetics as well as their correlation with recrystallization mechanisms. The outcome of such investigations would be of assistance for appropriate grade selection for a fusion reactor and the design of healing processes enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification W, forged bar and cold-rolled plate, were neutron irradiated to a dose of 0.18 dpa at 600 °C in the Material Test Reactor BR2, Mol, Belgium, and subsequently isochronally annealed from 700 to 1550 °C for 24 h with a 50 °C step with two non–irradiated counterparts. X-ray diffraction, optical microscopy and positron annihilation lifetime spectroscopy, combined with Vickers Hardness are used to correlate the recrystallization process and recovery mechanisms. It is found that the non-irradiated plate has a higher recrystallization resistance compared to that of the bar, with the former requiring 100 °C higher temperature for 24 h annealing time to recrystallize. Despite the irradiation induced dislocation loops and voids being of similar sizes and densities for both grades, the different initial microstructure affects the recrystallization process upon annealing, delaying it by 50 °C on the plate. Further in the plate a complete void dissolution is observed before recrystallization, which is not the case for the bar, for which the two effects occur at the same temperature.

Unveiling deformation behavior and damage mechanism of irradiated high entropy alloys

The oxide dispersion strengthening (ODS) high entropy alloy (HEA) exhibits the high elevated temperature performance and radiation resistance due to severe atomic lattice distortion and oxide particles dispersed in matrix, which is expected to become the most promising structural material in the next generation of nuclear energy systems. However, microstructure and damage evolution of irradiated ODS HEA under loading remain elusive at submicron scale using the existing simulations owing to a lack of atomic-lattice-distortion information from a micromechanics description. Here, the random field theory informed discrete dislocation dynamics simulations based on the results of high-resolution transmission electron microscopy are developed to study the dislocation behavior and damage evolution in ODS HEA considering the influence of severe lattice distortion and nanoscale oxide particle. Noteworthy, the damage behavior shows an unusual trend of the decreasing-to-increasing transition with the continuous loading process. There are two main types of damage micromechanics generated in irradiated ODS HEA: the dislocation loop damage in which the damage is controlled by irradiation-induced dislocation loops and their evolution, the strain localization damage in which the damage comes from the dislocation multiplication in the local plastic region. The oxide particle hinders the dislocation movement in the main slip plane, and the lattice distortion induces the dislocation sliding to the secondary slip plane, which promotes the dislocation cross-slip and dislocation loop annihilation, and thus reduces the material damage in the elastic damage stage. These findings can deeply understand atomic-scale damage mechanism and guide the design of ODS HEA with high radiation resistance.

Evolution of extended defects in UO2 during high temperature annealing

The evolution of thermophysical characteristics of uranium oxide (UO2) under irradiation is a critical aspect of nuclear fuel performance of light water reactors. This study examines the effect of high-temperature annealing at 1000 °C - 1600 °C on defect evolution in krypton irradiated UO2 at room temperature. Transmission electron microscopy analysis reveals that the size of irradiation-induced bubbles and dislocation loops increases but their density decreases with annealing temperatures. Perfect dislocation loops with b⇀ = <110> dominate, regardless of the annealing temperature. A rate theory model for extended defect evolution is applied to conclude that current understanding of dynamical loop and bubble evolution at high-temperature annealing is governed by defect coalescence.

Microstructure-Informed Prediction of Hardening in Ion-Irradiated Reactor Pressure Vessel Steels

Ion irradiation combined with nanoindentation is a promising tool for studying irradiation-induced hardening of nuclear materials, including reactor pressure vessel (RPV) steels. For RPV steels, the major sources of hardening are nm-sized irradiation-induced dislocation loops and solute atom clusters, both representing barriers for dislocation glide. The dispersed barrier hardening (DBH) model provides a link between the irradiation-induced nanofeatures and hardening. However, a number of details of the DBH model still require consideration. These include the role of the unirradiated microstructure, the proper treatment of the indentation size effect (ISE), and the appropriate superposition rule of individual hardening contributions. In the present study, two well-characterized RPV steels, each ion-irradiated up to two different levels of displacement damage, were investigated. Dislocation loops and solute atom clusters were characterized by transmission electron microscopy and atom probe tomography, respectively. Nanoindentation with a Berkovich indenter was used to measure indentation hardness as a function of the contact depth. In the present paper, the measured hardening profiles are compared with predictions based on different DBH models. Conclusions about the appropriate superposition rule and the consideration of the ISE (in terms of geometrically necessary dislocations) are drawn.

Nested loops explain low irradiation-induced swelling rate

In the quest of radiation and oxidation-resistant materials for the fuel claddings of nuclear power stations, chromium (Cr) emerges as a good candidate. It shows a significantly lower irradiation-induced swelling rate of nearly an order magnitude lower than other body-centered cubic materials, such as W and Fe. This phenomenon indicates that there is an unsuspected aspect, beyond the crystal structure that affects the swelling of materials. To understand it, the structure and distribution of irradiation-induced dislocation loops in Cr were investigated after ion irradiation at 550 ℃ from 0.1 to 15 dpa. It appears that after irradiation Cr exhibits nested loops, either vacancy-type loops called ‘trough’ or interstitial-type loops called ‘island’, which are composed of a large dislocation loop containing multiple small and coplanar loops with inverse nature. Our results indicate that the nested trough and island dislocation loops stem from the 1D migration of dislocation loops. This loop formation mechanism is an alternative mechanism of recombination and annihilation of point defects, which in turn can explain the low swelling rate in Cr. This mechanism provides a new insight into the irradiation-induced swelling rate of metals.

Nano-indentation deformation behavior of heavy ion irradiated reduced activation ferritic/martensitic (RAFM) steels

The Nano-indentation deformation behavior of the heavy ion irradiated RAFMs is investigated in the present work. Specimens of the CLF-1 steel were irradiated with 123.4 MeV 20Ne ion at both -50 °C and 400 °C, respectively, to achieve a mean displacement damage level of 0.15 dpa. A quasi-uniform distribution of the displacement damage was produced in the specimens by utilizing an energy degrader at the irradiation chamber. The results of nano-indentation experiments indicate that during the loading stage, the irradiated materials exhibit a lower strain rate, but during the holding stage after reaching the maximum load, the irradiated materials show a higher strain rate than the un-irradiated material. Transmission electron microscopy observation showed that the low-temperature irradiation produced a high density of irradiation-induced dislocation loops, which impeded the glide of dislocation lines during the loading stage. However, as the load increases gradually to the holding stage, the dislocation lines can overcome the pinning by the dislocation loops. The strain that did not occur due to the pinned effect of the dislocation loops in the loading stage was released during the holding stage, resulting in a higher strain rate in the irradiated specimen during the holding stage. Based on the Peierls mechanism, a quantitative relationship between the number density of the irradiation-induced dislocation loops and the strain rate during the holding stage was established. The calculated results agree well with the experimentally obtained strain rate during the holding stage.

Characterization of irradiation-induced dislocation loops and helium bubbles via electron channeling contrast imaging

Characterizing irradiation-induced defects is essential to understand damage mechanisms and develop irradiation-resistant materials. Prevailing strategies rely on transmission electron microscopy (TEM) observations which require rigorous prerequisite to prepare high-quality foils, making it difficult for rapid screening of irradiation-resistant materials. Here, we propose that the electron channeling contrast imaging (ECCI) is expected to provide a novel strategy for characterizing surface irradiation defects of ion-irradiated materials. The comparison between ECCI and conventional cross-sectional TEM for the observation of dislocation loops and helium (He) bubbles is analyzed in a common FeCoNiCr high-entropy alloy. The results indicate that the ECCI images of irradiation-induced defects are optimal at the extra-high tension of 20 kV, with a limit resolution of ∼1.5 nm for visible defects. The irradiation-induced dislocation loops show sharper contrast than bubbles in ECCI images, and the acceptable misorientation was within 4° of sample tilt. These findings offer a promising solution to the challenges for rapid assessment of irradiation damage in ion-irradiated materials.

TEM investigation of helium bubble evolution in tungsten and ZrC-strengthened tungsten at 800 and 1000°C under 40keV He+ irradiation

Helium-induced defect nucleation and accumulation in polycrystalline W and W0.5 wt%ZrC (W0.5ZrC) were studied in-situ using the transmission electron microscopy (TEM) combined with 40 keV He+ irradiation at 800 and 1000°С at the maximum damage level of 1 dpa. Radiation-induced dislocation loops were not observed in the current study. W0.5ZrC was found to be less susceptible to irradiation damage in terms of helium bubble formation and growth, especially at lower temperature (800 °C) when vacancies were less mobile. The ZrC particles present in the W matrix pin the forming helium bubbles via interaction between C atom and neighbouring W atom at vacancies. This reduces the capability of helium to trap a vacancy which is required to form the bubble core and, as a consequence, delays, the bubble nucleation.At 1000 °C, significant bubble growth occurred in both materials and all the present bubbles transitioned from spherical to faceted shape, whereas at 800 °C, the faceted helium bubble population was dominated in W.

Fractional densities and character of dislocations in different slip modes from powder diffraction patterns

Irradiation induced dislocation loops and lattice dislocations produced by plastic deformation have very different dipole characters, dislocation contrasts and range of strain fields. To determine partial dislocation densities of the different dislocation types from X-ray or neutron diffraction patterns, all differences in dislocation properties should be considered. We have extended the convolutional multiple whole profile line profile analysis method to convolve different strain profiles with different effective outer cut-off radii, corresponding to different dislocation type, into a single size profile. The extended procedure is applied to determine the 〈a〉 loop and lattice dislocation densities in a neutron irradiated and tensile deformed Zr-2.5 %Nb alloy and in a proton or neutron irradiated Zircaloy-2 alloy. We show that when the dislocation densities are very different the effective outer cut-off radius of dislocations is a better descriptor of dislocation character than the dislocation arrangement parameter, which was used previously in line profile analysis. Our results show that the combination of line profile analysis and electron microscopy methods provides a detailed and comprehensive description of defect structures in deformed and irradiated materials.

Irradiation hardening of FeCrAl ODS alloys with Zr/Hf addition after high-energy self-ions irradiation

FeCrAl ODS steel is the promising candidate cladding material for the accident tolerant fuel (ATF) of the light water reactors (LWR), due to the formation of Al2O3 in alloys which could suppress the oxidation reaction with hot steam. In the present work, change in microstructures and the corresponding irradiation hardening of two kinds of FeCrAl ODS alloys (15.5Cr-4Al), with 0.6 wt% Zr (denoted as P1) or 0.8 wt% Hf addition (denoted as P2), were investigated. Electron backscattered diffraction (EBSD) and transmitted electron microscopy (TEM) were used to characterize grains and dispersion morphology of oxides, respectively. Compared to P1, P2 has finer and denser oxides. 56Fe17+ ions with kinetic energy of 352.8 MeV were used to produce a quasi-uniformly distributed damage region from the near-surface to a depth of 24 μm in the specimens by using an energy degrader. A damage level of 0.8 dpa (displacement per atom) was attained in the specimens of both alloys simultaneously at 200℃. The Vickers hardness test shows that both specimens exhibited an apparent irradiation hardening. The extent of hardening of P2 (6 %) is significantly lower than P1 (12 %). TEM results show that irradiation-induced dislocation loops occur in both specimens but with a relatively smaller average size in P2. The results demonstrate that compared to the 0.6 wt% Zr addition, the 0.8 wt% Hf addition improves more significantly the hardening resistance, which was discussed in terms of the calculated sink strength originating from the oxides/matrix interface.

Irradiation damage and corrosion performance of proton irradiated 304 L stainless steel fabricated by laser-powder bed fusion

The irradiation-induced microstructure and corrosion performance of proton irradiated 304 L stainless steel (SS) fabricated by laser powder bed fusion (LPBF) is investigated. Irradiation-induced dislocation loops are observed in both LPBF and traditionally manufactured (TM) 304 L SS following 2 MeV proton irradiation at 360 °C. A comparison analysis of the oxide scales formed on proton irradiated LPBF and TM 304 L SS is conducted along with the unirradiated counterparts. The irradiation-induced defects accelerate the corrosion by providing a rapid diffusion path to form a thicker and more porous film with a double-layer structure and herein the oxide film thickness of TM 304 L SS is approximately twice that of the LPBF 304 L SS. The unique microstructure of high dislocation densities, cellular sub-grains boundaries and dispersed nanoscale inclusions provides large amounts of interfaces in LPBF samples, which is benefit to absorb the irradiation-induced point defects. The irradiation induced fast diffusion paths and the nucleation points of oxides are thus reduced, resulting in a thinner oxide film thickness and a better corrosion resistance revealed by the more Cr enrichment of inner scale in LPBF 304 L SS.

Irradiation accelerated corrosion of alumina-forming austenitic steels in supercritical CO2: The oxide scale formed within an individual grain or affected by grain boundary

The irradiation accelerated corrosion within monolayer grains of a new AFA steel exposed to sCO2 are studied. The presence of irradiation induced dislocation loops accelerate element diffusion and reduces corrosion resistance by ∼44%. Many oxide scale features are firstly found, including a two-phase inner layer consisting of Ni-rich austenite and Cr-rich spinel, and Al2O3/SiO2 oxide precipitates. Irradiation leads to an additional Cr-rich oxide layer at the bottom of inner layer and a narrower Al2O3 band in the internal oxidation layer. Irradiation also induces Cr depletion at grain boundaries, thereby impeding the formation of a protective Cr-rich oxide scale.

Radiation-induced segregation at grain boundaries of alloy 800H: Experimentally-informed atomistic simulations

The irradiation of alloys can alter the spatial distribution of solute and impurity elements. The redistribution of alloying elements can cause local chemical changes at surfaces and interfaces, which may degrade the mechanical and chemical properties of alloys. In this paper, a combination of microscopic experiments and atomistic simulations are employed to study the effect of irradiation damage on radiation-induced segregation (RIS) at grain boundaries (GBs) of the alloy 800H. The grain boundary character distribution in alloy 800H was determined by electron backscatter diffraction, and proton-irradiation-induced defects were identified by transmission electron microscopy. The experimentally-observed GBs were recreated in atomistic simulations to further study their atomistic structure, grain boundary energies, and local element concentration. The effects of radiation-induced defects on the redistribution of alloying elements at various GBs were investigated. A mechanism is proposed to explain the local chemical changes at GBs after the interaction with radiation-induced dislocation loops, which is of importance to nuclear reactor performance.

Investigation of the Y Effect on the Microstructure Response and Radiation Hardening of PM V-4Cr-4Ti Alloys after Irradiation with D Ions

In the current work, an analysis of the effects of Y on the radiation hardening and microstructure response of a V-4Cr-4Ti alloy has been conducted after 30 keV D ion irradiation at room temperature using transmission electron microscopy (TEM) and nanoindentation. The results show that the formation of large Y2O3 and small Y2V2O7 nanoparticles was confirmed, indicating that the addition of Y reduces the amount of dissolved oxygen. The addition of Y has been shown to affect the radiation-induced dislocation loops, radiation hardening, and Ti-rich segregation of the V-4Cr-4Ti alloy. With the addition of Y, the mean size of the radiation-induced dislocation loop decreased, which may result from the strong sink strength of the nanoparticle/matrix interface, interactions between Y atoms and SIA clusters, and the strong binding energy of vacancy–oxygen pairs. Some particles with core–shell structures were observed after ion irradiation, where Ti-rich segregations at the nanoparticle/matrix interface were confirmed. These results indicate that Y might promote abnormal segregation. Possible causes for this include the lower interface energy at the particle/matrix interface and the interaction between oxygen and solute atoms.

Surface blistering and void swelling of α-Al2O3 irradiated with H2+ ions followed by 1000, 1200 °C annealing

Surface morphology and defect structures in α-Αl2O3 irradiated with 106 keV H2+ and post-irradiation-annealed at 1000, 1200 °C were investigated. The ion irradiations and annealing cause modifications to the Raman scattering peaks such as the peak disappearance and peak shifts. Surface blisters with several micrometers in diameter are visible. The mean blister sizes increase with increasing annealing temperatures and decrease with increasing irradiation fluences. Transmission electron microscopy characterizations reveal large numbers of spherical or oval shaped voids around the peak-damage zone. Faceted voids are also observed. The irradiation-induced dislocation loops are mostly annealed out with only some dislocation lines left inside. Nevertheless, small dislocation loops with sizes of ∼20–30 nm are characterized under different diffraction conditions, which are found to be associated with the void structures in the 1.5 × 1017 H2+/cm2 irradiated and annealed sample. These loop structures are considered to be induced through the ‘loop punching’ mechanism due to a high gas pressure inside during the bubble growth period.

Analysis of dislocation characteristics and inside-outside contrasts in irradiated and annealed tungsten as a fusion reactor material

Tungsten is an important candidate of plasma-facing material for fusion reactors. Its irradiation response, especially the post-irradiation annealing (PIA) behavior needs further investigating. In addition, the practice of the “inside-outside” contrast method of determining the characteristics of irradiation induced dislocation loops has not been utilized frequently, and the present research serves as an example to present some practical considerations. In the present work, a tungsten thin-foil specimen is irradiated at 400 ℃ with 58 keV D&lt;sup&gt;+&lt;/sup&gt; to a final fluence of 1× 10&lt;sup&gt;17&lt;/sup&gt; cm&lt;sup&gt;–2&lt;/sup&gt;, corresponding to a dose of about 0.1 dpa. The specimen is prepared through the electro-polishing method by using a NaOH based electrolyte. The ion irradiation is carried out directly on the electro-polished specimen. The irradiated specimen is followed by isothermal annealing at 900 ℃ for 1 h. The as-irradiated and post-irradiation annealing modified defects are investigated with a transmission electron microscope (TEM) operated at 200 kV. The irradiation defects are characterized by using TEM bright-field imaging for the same imaging field with different &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vectors around the three major zone axes: the [001], [111], and [011] zone axis of the body-center cubic lattice of W. For each &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vector, the ±&lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; are characterized, and the corresponding contrast-extinctions and “inside-outside” contrasts of selected dislocation loops are identified. The indices of the &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vectors around different zone axes are assigned consistently with the Kikuchi map. As a result, the D&lt;sup&gt;+&lt;/sup&gt; irradiation increases a fine distribution of dislocation loops with an average size of (11.10±5.41) nm and a bulk density of 2.40×10&lt;sup&gt;22&lt;/sup&gt; /m&lt;sup&gt;3&lt;/sup&gt;. Voids are not observed obviously. The post-irradiation annealing causes the loop size to increase and the loop density to decrease, with numbers of (18.25±16.92) nm and 1.19×10&lt;sup&gt;22&lt;/sup&gt; /m&lt;sup&gt;3&lt;/sup&gt;, respectively. Through the contrast-extinction analysis, the dislocation loops in the irradiated and annealed specimen are identified to be &lt;i&gt; &lt;b&gt;a&lt;/b&gt; &lt;/i&gt;/2&lt;inline-formula&gt;&lt;tex-math id="Z-20230218170937"&gt;\begin{document}$\langle {111} \rangle$\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="5-20222124_Z-20230218170937.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="5-20222124_Z-20230218170937.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;-type dislocation loops. The post-irradiation annealing also causes the coalescence of large loops and forms large irregular-shaped dislocation loops. Voids with typical sizes of 1–2 nm are also observed in the annealed specimen. The PIA modified microstructure is consistent with the stage IV or stage V characterization of classical PIA induced microstructures. Through the “one-step” inside-outside contrast method, the dislocation loops are identified as an interstitial type. The influcing factors for the “one-step” method are discussed and the importance of consistent indexing is also noted.

Effect of different pre-existing dislocation densities on the microstructure evolution of W-0.5ZrC alloy during in-situ He+ irradiation

The evolution of dislocation loops and helium bubbles in W-0.5ZrC alloy with different pre-existing dislocation densities was investigated by in-situ transmission electron microscopy during 30 keV He+ irradiation at 800 °C. The nucleation, migration, aggregation, and growth of dislocation loops and helium bubbles occurred with the whole irradiation process. The variation trends of average size and volume number density of dislocation loops and helium bubbles with the increase of irradiation dose in the regions with different pre-existing dislocation densities were obtained. It was found that the growth of dislocation loops was inhibited in the region with a low pre-existing dislocation density (about 5.2 × 1013 /m2), and the nucleation of dislocation loops was inhibited in the region with the dispersed medium density (about 1.6 × 1014 /m2, dislocations were crossed but dispersed), while both the nucleation and growth of dislocation loops were inhibited in the region with the compact medium density (about 1.5 × 1014 /m2, dislocations were crossed and compact). During the irradiation process, the pre-existing dislocations would interact with the irradiation-induced dislocation loops and helium bubbles, but their influence weakened with the increase of irradiation dose, and finally resulted in the disappearance or deformation of pre-existing dislocations. The irradiation hardening caused by dislocation loops and helium bubbles was increased with the increase of irradiation dose, and the corresponding mechanism was analyzed.

Gamma-ray irradiation induced dislocation loops in hexagonal zirconium

ABSTRACT With doses simulating those received by spent nuclear fuel (SNF) facilities, gamma-ray (γ-ray) irradiation experiments were conducted on hexagonal zirconium. γ-ray irradiation–induced crystalline defects were observed in the hexagonal zirconium and characterized. Monte Carlo calculations combined with Transmission Electron Microscopy characterizations revealed that the energy of primary knock-on atoms could reach the threshold energy () of zirconium atoms and result in dislocation loops within the α-zirconium grains. The influence of γ dose on dislocation loop size and number density was also evaluated.

Measurement of hydrogen trapping in cold-work dislocations using synchrotron X-ray diffraction

In water-cooled nuclear power plants containing zirconium alloys as the clad material, embrittlement of the cladding can occur from hydrides produced by the water-zirconium alloy interaction, thus limiting the life of the fuel assembly (Motta et al., 2019). The hydrogen produced at the water-zirconium interface can also be trapped in neutron irradiation-induced microstructural defects, reducing the amount of hydride present at any given time/temperature. Knowledge of the extent of hydrogen trapping by various microstructural features in zirconium and its alloys is therefore key for accurate predictions of hydride-induced embrittlement. A novel method for quantifying both trap capacity and trap binding energy using synchrotron X-ray diffraction is described and demonstrated for a cold-worked, Zircaloy-4 sample in which line dislocations provide an analogue for irradiation-induced dislocation loops as hydrogen trap sites. This study has experimentally determined the following parameters for cold-work dislocations: Thermal stability; Binding energy; Maximum capacity for hydrogen; Number of equivalent hydrogen trap sites. These parameters may be used in modelling hydrogen embrittlement of zirconium alloys, as experimental estimates for their analogue of irradiation-induced dislocation loops. More importantly, this pilot study has successfully demonstrated the methodology and results of a novel technique that may be applied generally to irradiated material to quantify hydrogen trapping at other irradiation-induced microstructural defects.