Irradiation Hardening Research Articles

Effect of niobium content on irradiation microstructure and hardening in FeCrAl-based alloys

• The evolution of dislocation loop with niobium content in FeCrAl alloys was discussed. • The formation of laves phase resulted in the decrease of Mo content and the increase of Cr content in the matrix of the FeCrAl alloys. • Nb addition increased nanoindentation hardness and reduced irradiation hardening in the FeCrAl alloys . Iron-chromium-aluminum (FeCrAl) alloys with different content of niobium (Nb)—0, 0.4 wt%, 0.8 wt%, and 1.2 wt%—were designed and prepared. All samples were then irradiated with 2.4 MeV Fe 2+ ion to the dose of 1 and 15 displacements per atom (dpa) at 400 °C. The formations of dislocation loops induced by self-ion irradiation in these alloys were investigated by transmission electron microscopy (TEM). Nano-indentation tests were used to assess the hardness and irradiation hardening of samples. For the samples before irradiation, the (Fe, Cr) 2 (Nb, Mo) Laves phases density and the nano-indentation hardness increased with increasing Nb content of the samples. After irradiation to 1 and 15 dpa, both of a /2<111> and a <100> dislocation loops were produced but no voids or αˊ phase were found in all samples. With increasing Nb content of the samples, the size of dislocation loops increased first and then decreased, while the total volume number density decreased and then increased. The fraction of a <100> dislocation loops increased first and then decreased with increasing Nb content, and increased with increasing irradiation dose. Dislocation networks and the amorphization of the Laves phases were observed in the samples with irradiation dose of 15 dpa. Irradiation hardening of Nb free samples was two to four times that of Nb containing samples, and the irradiation hardening increased with increasing Nb content of Nb containing samples. The experimental results indicate that the increase of Nb content in FeCrAl alloys can increase the density of Laves phases, leading to the decrease of Mo content and increase of Cr content in the matrix. The competition between the two types of solutes affects the nucleation and growth of the dislocation loops.

A statistical model of irradiation hardening induced by non-periodic irradiation defects

Numerous efforts have been made to develop the irradiation hardening models, due to their important role in nuclear reactor design and service lifetime prediction. Most of the models are based on the dislocation line tension approximation, by assuming a periodic array of irradiation defects. The random distribution effect is also generally taken into account from a fully deterministic perspective, such as the well-known Bacon-Kocks-Scattergood model. Different from these previous studies, this work is aimed at developing a statistical irradiation hardening model by systematically coarsening the interaction mechanism between dislocation and non-periodic irradiation defects. The cumulative distribution function of the critical shear stress to overcome the irradiation defect barrier is obtained through the analytical derivation, and is lumped into a 2D Monte Carlo model to predict the mechanical response of irradiated material. The prediction results agree very well with three-dimensional discrete dislocation dynamics results and experimental data.

Influence of Ni-Mn contents on the embrittlement of PWR RPV model steels irradiated to high fluences relevant for LTO beyond 60 years

Understanding the irradiation induced degradation of the reactor pressure vessel (RPV) steels at high fluences becomes more important in view of safe long term operation (LTO) of existing nuclear power plants. Specifically, the role of Ni, Mn and Si and associated synergetic effects on embrittlement behaviour need to be further investigated to understand accelerated embrittlement at high neutron fluences (ca. 1 × 10 20 n.cm −2 , E>1 MeV) observed in certain types of RPV steels. In addition, it is required to verify the validity of existing embrittlement trend curves (ETCs) at high fluence values. To address these issues, NRG and JRC performed a unique irradiation experiment called LYRA-10 in the High-Flux Reactor (HFR) Petten. In LYRA-10 a variety of specimens (e.g. tensile, Charpy, mini-Charpy and microscopy) of RPV steels (base metal and welds) with systematic variations in Ni, Mn and Si contents, while keeping the base composition representative of VVER-1000 and PWR RPV steels, have been irradiated for 16 HFR cycles to achieve the nominal fast neutron fluence of 1.11 × 10 20 n.cm −2 at a nominal temperature of 286 °C. A large post-irradiation examination campaign on LYRA-10 specimens has been initiated within the STRUMAT-LTO, which is an international collaboration project co-funded by EURATOM. The research reported in this paper includes results from tensile and mini-Charpy impact testing of four PWR RPV model steel specimens irradiated in LYRA-10, which is performed as part of STRUMAT-LTO in-kind pilot project before the start of the EURATOM co-funded project. The combined influence of Ni and Mn on embrittlement behaviour, characterized by irradiation hardening and shifts in transition temperature (T k ) properties, has been evaluated. The measured shifts in T k values for all four model steels have been compared with predicted values from two selected ETCs in order to check their validity for these steels at the above high neutron fluence values.

Effect of Au-ion irradiation on the surface morphology, microstructure and mechanical properties of amorphous AlCrFeMoTi HEA coating

The amorphous AlCrFeMoTi high-entropy alloy (HEA) coating with near-equal molar ratio was successfully deposited on F/M substrate by magnetron co-sputtering technique and then was irradiated by 6 MeV Au ions with different irradiation damage doses from 10 dpa to 90 dpa. XRD, SEM, AFM, TEM and nanoindentation tests were carried out to investigate the surface morphology, microstructure and hardening effect of the coating. The results showed that the coating surface became more featureless after irradiation, where the surface roughness decreased with increasing damage doses. The crystallization phenomena occurred in amorphous AlCrFeMoTi coatings at high irradiation damage doses (45 dpa and 90 dpa), where the crystallization zone were much closer to the coating surface than the calculated peak damage region. However, the irradiation-induced crystallization did not obviously change the uniformity of elements distribution. Meanwhile, the irradiation hardening effect can be observed, and the nanohardness of the coating increased with increasing damage doses. The AlCrFeMoTi HEA coating exhibits excellent ion-irradiation resistance and has potential applications in lead cooled fast reactor. Meanwhile, this work can help us to further investigate the performance of the AlCrFeMoTi HEA coating in the coupled environment of irradiation and corrosion, and evaluate its serviceability.

Impact of neutron irradiation on hardening of baseline and advanced tungsten grades and its link to initial microstructure

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.

Constructing high-performance radiation-resistant ternary YSZ-MgO-CNT nanocomposites via tailored nanostructures

Developing long-lifetime bulk-form ceramic-based materials with high irradiation resistance is crucial for advanced nuclear systems. Here, we incorporated carbon nanotubes (CNTs) into yttria-stabilized zirconia (YSZ) and magnesia (MgO) nanocrystals to fabricate bulk YSZ-MgO-CNT nanocomposites with abundant ternary nanostructures by spark plasma sintering. To understand the role of tailored ternary nanostructure on irradiation, we investigated the microstructure and mechanical properties evolutions of the YSZ-MgO-CNT nanocomposites irradiated by multi-energy He+ ions at high temperature to different fluences. Compared with the single-phase YSZ and ultrafine-grained YSZ-MgO composites, the YSZ-MgO-CNT nanocomposites possessed higher ability to manage irradiation-induced He bubbles/defects via the defect-interface interactions of proposed “loading-unloading” and “loading-transporting-unloading” mechanisms for controlling the dynamical behaviors of He atoms/defects in the CNT-doped ternary nanostructures, thereby presenting more stable microstructure and better performance in resisting irradiation hardening. This work provides insight into the design of advanced inert matrix nuclear fuel and new nuclear waste management materials.

Effect of irradiation damage and indenter radius on pop-in and indentation stress-strain relations: Crystal plasticity finite element simulation

In this work, a crystal plasticity finite element model is developed to simulate the spherical nano-indentation of ion-irradiated metallic materials. Two critical features are addressed by the proposed theoretical framework, which include the pop-in event observed in the loading force-indentation depth (P−h) relations and irradiation hardening characterized by the transformed indentation stress-strain (ISS) curves. For the former, the pop-in event is dominated by the dislocation nucleation mechanisms, which are affected by both the density of irradiation-induced dislocation nucleation sites and indenter radii. For the latter, irradiation hardening is closely related to the heterogeneously distributed irradiation-induced defects within the irradiated layer with a limited depth. By applying the developed model to the spherical nano-indentation of helium-irradiated single crystal tungsten, it is informed that the simulated results can match well with corresponding experimental data, which include the unirradiated and irradiated P−h and ISS relations at different indenter radii. Moreover, the evolution of different hardening mechanisms during the indentation process is systematically analyzed, which can help comprehend the fundamental deformation mechanisms of ion-irradiated materials under spherical nano-indentation.

An atomistic study of defect energetics and diffusion with respect to composition and temperature in γU and γU-Mo alloys

• Vacancy and self-interstitial formation energies were calculated in γ U and γ U-Mo. • Self-diffusion and interdiffusion coefficients in γ U- x Mo ( x = 7 , 10, 12 wt.%) were calculated. • Results can be used as input parameters in mesoscale and engineering scale fuel modeling. Uranium-molybdenum (U-Mo) alloys are promising candidates for high-performance research and test reactors, as well as fast reactors. The metastable γ phase, which shows acceptable irradiation performance, is retained by alloying U with Mo with specific quenching conditions. Point defects contribute to the atomic diffusion process, defect clustering, creep, irradiation hardening, and swelling of nuclear fuels, all of which play a role in fuel performance. In this work, properties of point defects in γ U and γ U- x Mo ( x = 7, 10, 12 wt. % ) were investigated. Vacancy and self-interstitial formation energies in γ U and γ U- x Mo were calculated with molecular dynamics (MD) simulations using an embedded atom method interatomic potential for the U-Mo system. Formation energies of point defects were calculated in the temperature range between 400 K and 1200 K. The vacancy formation energy was higher than the self-interstitial formation energy in both γ U and γ U- x Mo in the evaluated temperature range, which supports the previous results obtained via first-principles calculations and MD simulations. In γ U- x Mo, the vacancy formation energy decreased with increasing Mo content, whereas the self-interstitial formation energy increased with increasing Mo content in the temperature range of 400 K to 1200 K. The self-diffusion and interdiffusion coefficients were also determined in γ U- x Mo as a function of temperature. Diffusion of U and Mo atoms in γ U- x Mo were negligible below 800 K. The self-diffusion and interdiffusion coefficients decreased with increasing Mo concentration, which qualitatively agreed with the previous experimental observations. Point defect formation energies, self-diffusion coefficients, and interdiffusion coefficients in γ U- x Mo calculated in the present work can be used as input parameters in mesoscale and engineering scale fuel performance modeling.

Effect of helium bubbles on irradiation hardening of additive manufacturing 316L stainless steel under high temperature He ions irradiation

To evaluate the effects of irradiation on additive manufacturing materials which are expected to be applied in advanced nuclear energy systems, selective laser melting 316L stainless steel (SLM 316L SS) and traditionally manufactured 316L stainless steel (TM 316L SS), were irradiated with 500 keV He ions at 700°C, respectively. SEM, TEM and nanoindentation measurements were performed to survey the effects of helium (He) bubbles on the irradiation hardening. It was found that the degree of irradiation hardening of the two materials increased with an increase of ions fluences. The hardening range of SLM 316L SS was 6-47% while that of TM 316L SS was 32-75%. Moreover, the hardening saturation phenomenon was more pronounced in TM 316L SS than in SLM 316L SS. These results suggested that SLM 316L SS owned higher resistance to irradiation hardening than TM 316L SS. TEM observations showed that the density of He bubbles in TM 316L SS was 7.06 × 1023/m3 while that in SLM 316L SS was 3.33 × 1023/m3. Further, the average sizes of bubbles were 2.2 and 2.4 nm, respectively. The differences of the number density might be why SLM 316L SS had a slighter irradiation hardening than TM 316L SS, which was consistent with Orowan model. It was suggested that the numerous intrinsic microstructures such as entangled dislocation network walls and subgrain boundaries in SLM 316L SS could greatly reduce the density of He bubbles.

Microstructure and hardening induced by neutron irradiation in single crystal, ITER specification and cold rolled tungsten

In this work, we have performed neutron irradiation and sub-sequent microstructural and micro-hardness measurements on a series of tungsten grades. The neutron irradiation was performed in BR2 reactor up to 0.2 dpa at 600, 800 and 1200 °C. The selected irradiation temperatures correspond to a large range of the operation of tungsten in ITER as armour for the divertor. The applied fast neutron flux is representative of ITER irradiation conditions in terms of damage rate. The primary purpose of the study was to establish the relation between the initial microstructure, the resulting irradiation-induced microstructure and corresponding irradiation hardening. The microstructure was studied using transmission electron microscopy (TEM) and hardening was assessed by micro-hardness measurements. The obtained TEM results were used to predict the hardening employing the dispersed barrier model and strength coefficients established for the single crystal from the open literature. Based on the performed analysis, it appeared that the main contribution to the hardening at high irradiation temperature originates from the voids, while the dislocation loops provide a comparable contribution (30–50%) only under irradiation at 600 °C. Excellent agreement between the model prediction and experimentally measured hardness increase was obtained for the single crystal and ITER specification grade without any additional adjustment, while the hardening induced in the cold rolled plate was essentially overestimated by the model. This result suggests that the cold rolled tungsten plate exhibits an alternative mechanisms of the plastic deformation besides the conventional dislocation glide and multiplication, and this alternative mechanism is impacted by the irradiation less severely compared to the prediction coming from the dispersed barrier model.

Effects of Helium Production and Displacement Damage on Microstructural Evolution and Mechanical Properties in Helium-Implanted Austenitic Stainless Steel and Ferritic/Martensitic Steel

The effects of helium concentration and displacement damage on microstructural evolution at low dpa and low helium concentration were mainly investigated in specimens of austenitic stainless steel 316FR or SUS304 and a high chromium martensitic steel (HCM12A). The 316FR and HCM12A specimens were implanted uniformly with helium at 823 K up to 30 appm-He or 50 appm-He by 50 MeV cyclotron accelerator using energy degraders. After the helium implantation, the microstructures were examined by a transmission electron microscopy and positron annihilation lifetime measurements. Irradiation hardening behaviors were analyzed using SUS304 and HCM12A steels at 823 K implanted with He ion up to 100 appm with different He/dpa ratios in the HIT ion irradiation experiments and the hardening behaviors were examined by nano indentation method. In the irradiation and annealing specimens, these mechanical properties and microstructures were examined to understand the effects of helium production, displacement damage and annealing on microstructural development, and kinetic Monte Carlo (kMC) simulations were also performed to understand the microstructural development, and the results were compared with the results of TEM observation and positron annihilation lifetime measurements. Important some differences in the microstructural developments such as cavity formation and growth between austenitic stainless steel and martensitic steel were observed in low dpa and low helium concentration conditions.

Heavy ion irradiation induced defects in Zircaloy-4

Radiation damage induced in Zircaloy-4 by 20 MeV Au ions was investigated using Grazing x-ray diffraction (GIXRD), positron annihilation techniques, micro-and nano-hardness techniques. The irradiations were performed at room temperature in the fluence range 1 × 1013–5 × 1014 ions cm−2 corresponding to 0.07–3.63 displacements per atom (dpa). The grazing incidence angle x-ray diffraction (GIXRD) revealed the presence of the hydride precipitates in the analyzed zircaloy-4. According to our experimental data it is found that the amount of these precipitates decreases after irradiation as emphasized by GIXRD and Positron Annihilation Lifetime Spectroscopy (PALS) results. The same data revealed also a zircaloy-4 swelling above a certain dose. From the combination of domain size/microstrain with the micro- and nano-hardness results, we conclude that the Zircaloy hardening is mainly due to the dislocation formation indicating the effect of the microstructure on the mechanical properties. Moreover, from the dose dependence of irradiation hardening we conclude that the transition point from low-dose to high-dose regime is observed around 0.07 dpa.

Fatigue Reliability Analysis Method of Reactor Structure Considering Cumulative Effect of Irradiation

The influence of irradiation should be considered in fatigue reliability analyses of reactor structures under irradiation conditions. In this study, the effects of irradiation hardening and irradiation embrittlement on fatigue performance parameters were quantified and a fatigue life prediction model was developed. Based on this model, which takes into account the cumulative effect of a neutron dose, the total fatigue damage was calculated according to Miner’s linear cumulative damage law, and the reliability analysis was carried out using the Monte Carlo simulation method. The case results show that the fatigue life acquired by taking into account the cumulative effect of irradiation was reduced by 24.3% compared with that acquired without considering the irradiation effect. Irradiation led to the increase of the fatigue life at low strains and its decrease at high strains, which is in accordance with the findings of an irradiation fatigue test. The rate of increase in the fatigue life decreased gradually with the increase of the neutron dose. The irradiation performance parameters had a small influence on fatigue reliability, while the fatigue strength coefficient and the elastic modulus had a great influence on the fatigue reliability. Compared with the current method, which uses a high safety factor to determine design parameters, a fatigue reliability analysis method taking into account the cumulative effect of irradiation could be more accurate in the reliability analysis and life prediction of reactor structures.

Irradiation behaviors of two novel single-phase bcc-structure high-entropy alloys for accident-tolerant fuel cladding

High-entropy alloys (HEAs) are potential alternative materials for accident-tolerant fuel cladding due to their excellent irradiation resistance and high-temperature corrosion resistance. In this work, two novel body-centered cubic (bcc) structured Mo0.5NbTiVCr0.25 and Mo0.5NbTiV0.5Zr0.25 HEAs were fabricated. Helium-ion irradiation was performed on the two HEAs to simulate neutron irradiation, and the crystal structure, hardness, and microstructure evolution were investigated. The crystal structure of the Mo0.5NbTiVCr0.25 HEA remained stable at low fluences, while amorphization may occur at high fluences in the two HEAs. The irradiation hardening value of the Mo0.5NbTiVCr0.25 was 0.77 GPa at fluences of 1 × 1017 ions/cm2 and 1.49 GPa at fluences of 5 × 1017 ions/cm2, while the hardening value of the Mo0.5NbTiV0.5Zr0.25 was 1.36 GPa at ion fluences of 5 × 1017 ions/cm2. In comparison with most of the conventional alloys, the two HEAs showed slight irradiation hardening. The helium bubbles and dislocation loops with small size were observed in the two HEAs after irradiation. This is the first time to report the formation of a dislocation loop in bcc-structure HEAs after irradiation. However, voids and precipitates were not observed in the two HEAs which could be ascribed to the high lattice distortion and compositional complexity of HEAs. This research revealed that the two HEAs show outstanding irradiation resistance, which may be promising accident-tolerant fuel cladding materials.

IASCC of neutron irradiated 316 stainless steel to 125 dpa

The irradiated microstructure and irradiation-assisted stress corrosion cracking (IASCC) susceptibility were determined for neutron-irradiated cold-worked (CW) 316 stainless steel tested in 320 °C simulated PWR primary water with various damage levels of 46.9, 67.4, and 125.4 dpa. The size and density of dislocation loops and precipitates were found to vary little over the dose range 46.9–125.4 dpa and dislocation loops were found to control the irradiation hardening. The %IG in constant extension rate tensile (CERT) tests and the stress to initiate a crack in four-point bend tests were consistent in establishing that the IASCC susceptibility was essentially unchanged across the dpa range. GB oxidization was observed prior to crack initiation, and localized deformation and triple junctions where local stresses were high were preferred crack initiation sites. IASCC cracks initiated preferentially on grain boundaries oriented nearly normal to the applied stress. An apparent threshold stress threshold for IASCC was determined to be ~45% of the irradiated yield strength for neutron-irradiated CW 316 SS between 46.9 and 125.4 dpa.

Evaluation of irradiation hardening in ODS-Cu and non ODS-Cu by nanoindentation hardness test and micro-pillar compression test after self-ion irradiation

Abstract Oxide dispersion strengthened (ODS) Cu alloys are expected to have high strength and superior irradiation resistance that are necessary characteristics for the divertor components of fusion reactors. The present study investigated the irradiation hardening after a self-ion irradiation experiment that was carried out at 100 °C with 5.1 MeV Cu2+ ions up to three displacements per atom in ODS-Cu and compared it to that of non-ODS Cu materials such as mechanically alloyed (MAed) Cu and single crystal (SC) Cu. Nanoindentation hardness tests and micro-pillar compression tests on 1 μm cubic pillars were carried out to evaluate the mechanical strength before and after the ion-irradiation. Both MAed-Cu and ODS-Cu showed good resistance against irradiation hardening, while SC-Cu showed a larger increase in the nanoindentation hardness and in the micro-pillar compression yield stress. Relationships between the nanoindentation hardness, Vickers hardness and micro-pillar yield stress were discussed considering the effects of both specimen/indentation size and irradiation hardening.

The Studies of Irradiation Hardening and RIS on IASCC of Reactor Internals Bolts

IASCC of stainless steel has been the most important issue for internals BFBs. The inspection data analysis indicates that there is a closed relation between irradiation fluence and cracked BFBs distribution. Then the nanoindentation and 3DAP tests were carried out to study the hardening and radiation induced segregation (RIS) behaviors of the reactor internals stainless steel specimens irradiated with 6 MeV Xe ions at room temperature. It is indicated that higher irradiation damage will cause more significant hardening and RIS and consequently increase the IASCC susceptibility.

Irradiation Hardening Behavior of He-Irradiated V–Cr–Ti Alloys with Low Ti Addition

A set of V–(4–8)Cr–(0–4)Ti alloys was fabricated to survey an optimum composition to reduce the radioactivity of V–Cr–Ti alloys. These alloys were subjected to nano-indenter tests before and after 2-MeV He-ion irradiation at 500 °C and 700 °C with 0.5 dpa at peak damage to investigate the effect of Cr and Ti addition and gas impurities for irradiation hardening behavior in V–Cr–Ti alloys. Cr and Ti addition to V–Cr–Ti alloys for solid–solution hardening remains small in the unirradiated V–(4–8)Cr–(0–4)Ti alloys. Irradiation hardening occurred for all V–Cr–Ti alloys. The V–4Cr–1Ti alloy shows the highest irradiation hardening among all V–Cr–Ti alloys and the gas impurity was enhanced to increase the irradiation hardening. These results may arise from the formation of Ti(CON) precipitate that was produced by He-ion irradiation. Irradiation hardening of V–Cr–1Ti did not depend significantly on Cr addition. Consequently, for irradiation hardening and void-swelling suppression, the optimum composition of V–Cr–Ti alloys for structural materials of fusion reactor engineering is proposed to be a highly purified V–(6–8)Cr–2Ti alloy.

Helium ions irradiation analysis of W0.5(TaTiVCr)0.5 for application as a future fusion plasma-facing material

The irradiation resistance of a reduced activation tungsten-containing high-entropy alloy-based material W0.5(TaTiVCr)0.5 was analyzed using He ions irradiation. Mirror-polished samples were exposed to 200 keV He+ ions at a maximum fluence of 1 × 1021 ions/m2. The nano-indentation hardness of W0.5(TaTiVCr)0.5 increased from 11.6 GPa to 14.9 GPa (at 200 nm depth) following He ion irradiation. Examination of the sample's internal microstructure, using transmission electron microscopy, revealed some randomly dispersed black spots and precursors of He bubbles and dislocations causing irradiation hardening. Relatively better irradiation resistance of W0.5(TaTiVCr)0.5 compared with W, under the given experimental conditions suggests its potential fusion plasma facing applications.

Helium implantation damage resistance in nanocrystalline W-Ta-V-Cr high entropy alloys

Nanocrystalline W-Ta-Cr-V high entropy alloys have shown promising properties as nuclear fusion materials with enhanced radiation resistance to heavy ion irradiation and negligible radiation hardening. In this work, we investigate the performance of the alloy under low energy helium (He) implantation up to a fluence of 1.25 × 1017 cm−2 at 1223 K. We observe a uniform high density of very small (~2–3 nm) bubbles grown at a slow rate along with enhanced He bubble damage resistance, further marked by no preferential bubble formation on the grain boundaries, even at much higher fluences compared to previously implanted tungsten grades. First principle calculations of He formation and migration energies in this alloy indicate deep energetic wells on the potential landscape and low diffusivity of He compared to pure W. The results imply higher overall (considering both grain matrices and grain boundaries) implantation resistance due to slow He diffusion and accumulation, and confirm the enhanced vacancy-self interstitial recombination argument in these alloys.