Irradiation Hardening Research Articles

Establishment of multi-beam irradiation facility at Wuhan University

A multiple ion beam irradiation facility has been established at Wuhan University to simulate severe irradiation conditions to materials in advanced nuclear energy systems. The platform consists of a 2 × 1.7 MV tandem accelerator, a 200 kV ion implanter, and a 50 kV ion implanter. A single high-energy irradiation chamber and a triple-beam irradiation chamber were set up. China Low Activation Martensitic Steel (CLAM) samples were irradiated with single beam of iron ions and simultaneous dual-beam of iron ions and helium ions at 450 °C. A large number of dislocation loops were observed in the samples after both irradiation conditions. The number density and average size of dislocation loops in the dual-beam irradiated samples were significantly larger than those in the single-beam irradiated samples. Irradiation hardening in dual-beam irradiated samples is more severe than that in single-beam irradiated samples.

Interface-controlled mechanical properties and irradiation hardening in nanostructured Cr75Al25/Zr multilayers

As an efficient strategy to design accident-tolerant fuel (ATF) materials, nanostructured metallic multilayers (NMMs) have attracted significant research interest. In this work, Cr75Al25/Zr (CA/Zr) NMMs were prepared with equal individual layer thickness h spanning from 5 to 75 nm, and subjected to He+ implantation with a total fluence of 1 × 1017 ions cm−2 at room temperature. With decreasing h, the interface changed from crystalline/crystalline to crystalline/amorphous along with size-driven strengthening to softening behavior of CA/Zr NMMs at the critical thickness h = 25 nm. The size-dependent hardness consists well with the evolution of the back stress with respect to h, which is elucidated by the statistical dislocation absorption propensity by the interface. Moreover, notable irradiation hardening emerges in the CA/Zr NMMs, which is caused by He bubble-dislocation interactions at large h and by the cooperation of He bubble hardening and increased back-stress hardening at small h. The hardness of both as-deposited and irradiated CA/Zr NMMs was quantitatively explained by dislocation models operative at different length scales.

Unusual He-ion irradiation strengthening and inverse layer thickness-dependent strain rate sensitivity in transformable high-entropy alloy/metal nanolaminates: A comparison of Fe50Mn30Co10Cr10/Cu vs Fe50Mn30Co10Ni10/Cu

In this work, we prepare transformable HEA/Cu nanolaminates (NLs) with equal individual layer thickness (h) by the magnetron sputtering technique, i.e., Fe50Mn30Co10Cr10/Cu and Fe50Mn30Co10Ni10/Cu, and comparatively study He-ion irradiation effects on their microstructure and mechanical properties. It appears that the as-deposited HEA/Cu NLs manifest two size h-dependent hardness regimes (i.e., increased hardness at small h and hardness plateau at large h), while the He-implanted ones exhibit monotonically increased hardness. Contrary to the fashion that smaller h renders less irradiation hardening in bimetal NLs, the Fe50Mn30Co10Cr10/Cu NLs manifest the trend that smaller h leads to greater irradiation hardening. By contrast, the Fe50Mn30Co10Ni10/Cu NLs exhibit the maximum irradiation hardening at a critical h = 50 nm. Below this critical size, smaller h results in lower radiation hardening (similar to bimetal NLs), while above this size, smaller h results in greater radiation hardening (similar to Fe50Mn30Co10Cr10/Cu NLs). Moreover, these transformable HEA/Cu NLs display inverse h-dependent strain rate sensitivity (SRS m) before and after He-ion irradiation. Nevertheless, compared with as-deposited samples, the irradiated Fe50Mn30Co10Cr10/Cu NLs display reduced SRS, while the irradiated Fe50Mn30Co10Ni10/Cu NLs display enhanced SRS. Such unusual size-dependent irradiation strengthening and inverse h effect on SRS in irradiated samples were rationalized by considering the blocking effects of He bubbles on dislocation nucleation and motion, i.e., dislocations shearing or bypassing He bubbles.

Recent progress in experimental investigation of neutron irradiation response of tungsten

In order to facilitate its application as plasma facing material, neutron irradiation response of tungsten (W) has been extensively studied by employing fission reactors. In this article, the major findings of the microstructural evolution and thermo-mechanical property degradation of W following neutron irradiation in various reactors are summarized by reviewing the papers published in 2019–2021 from the US-Japan joint projects on fusion materials research and the EUROfusion program. Microstructural development of W following irradiation has been well documented as functions of irradiation temperature, dose, and neutron spectrum. Detailed characterization of the primary defect features, i.e., dislocations, voids, and precipitates, enables a better understanding of the interactions between displacement damage and solid transmutant elements. Irradiation hardening and embrittlement are the primary degradation phenomena of W exposed to neutrons, manifested by hardness and yield strength increase, ductility loss at high doses, and shift of the ductile to brittle transition temperature (DBTT). It is believed that the segregation and precipitation of the solid transmutant elements dominantly contributes to the property deterioration. Predicting W performance in real fusion neutron irradiation environments based on the current understanding of W response to fission neutrons remains an open question.

Revealing the synergistic effect of invisible helium clusters in helium irradiation hardening in tungsten

Irradiation hardening caused by helium (He) ion bombardment is directly related to the pinning effect of He-related defects on dislocation glide in metals. Taking tungsten as an example, we have systematically investigated the pinning effect of different size He-related defects on the 1/2<111> dislocation glide by molecular dynamics simulations. Interestingly, it is found that not only the nanoscale He bubbles, but also the invisible He clusters play a critical role in He-irradiation hardening. Under the multiple invisible He clusters synergy, the critical resolved shear stress (CRSS, t) will linearly increase with the increasing of He cluster number (m), following the equation τCRSS=τdis+m·ΔτHen−Vi. Accordingly, the CRSS of the multiple invisible He clusters is easy to catch up or even surpass that for nanoscale He bubbles. Consequently, the synergistic effect of invisible He clusters must be considered for He-irradiation hardening in metals.

Study on irradiation effects of refractory bcc high-entropy alloy

• Refractory HEAs exhibit high thermal stabilities after heat-treatment. • The irradiation hardening of MoNbTaTiW is smaller than that of MoNbTaTi. • Refractory HEAs have higher irradiation resistance than pure tungsten. We developed body-centered cubic refractory high-entropy alloys, MoNbTaTi, MoNbTaW, and MoNbTaTiW, and explored their microstructure, and mechanical properties in order to develop a new high-irradiation resistance material suitable for fusion reactor components. Three HEAs were annealed at 1200 °C for 48 h to form single face BCC alloys, while MoNbTaW was also annealed at 1500 °C. SEM inspection and EDS mapping have both validated this result. The Vickers hardness of three HEAs was investigated, and the stable results indicate that they are all stable at elevated temperatures. Three HEAs are ion irradiated with 6.4 MeV Fe 3+ at a temperature of 500 °C, and the irradiation hardening is examined using nanoindentation. In comparison to pure tungsten, the three HEAs exhibit increased resistance to irradiation.

Characterization of microstructure and hardening of SLM nickel-based alloy irradiated by He ions

For the application of additively manufactured nickel-based alloys in molten salt reactors (MSRs), it is crucial to evaluate their performance in nuclear environments, especially under He-induced damage. Therefore, in this study, the irradiation damage behaviour of selective laser melted (SLM) nickel-based GH3536 alloy was studied by irradiating it with He ions at various doses. Traditionally manufactured (TM) materials were used as reference materials to evaluate the resistance of SLM GH3536 to irradiation. Transmission electron microscopy (TEM) was performed to reveal the depth distribution of irradiation-induced He bubbles, as well as the characteristics of dislocation loops, using both on-zone scanning transmission electron microscopy (STEM) and rel-rod modes. Nanoindentation tests were performed on both alloys to investigate their irradiation hardening. The TEM graphs and quantitative analysis showed that, owing to the cellular sub-grain structures, the He bubbles in the SLM alloy exhibited a lower number density but larger size than those in the TM alloy. Additionally, the SLM alloy showed inferior He tolerance but better resistance to irradiation hardening, as revealed by the nanoindentation test results. Consistency between the experimental hardness increment and calculated yield stress increment showed that the irradiation hardening of the sample was mainly caused by the presence of irradiation-induced defects. A comparison between the on-zone STEM and rel-rod modes suggested that the rel-rod mode is more effective if only the effect of irradiation-induced defects on mechanical properties is studied. Furthermore, this study provides insight into the design of SLM nickel-based alloys for nuclear industry applications.

Effect of Post-irradiation Annealing on Microstructure Evolution and Hardening in GH3535 Alloy Irradiated by Au Ions

The Au ion-irradiation experiments of GH3535 alloy, a candidate alloy structural material for molten salt reactor, was carried out in this study. Herein, isochronous annealing experiments were conducted from 200 to 850 °C to clarify the evolution behavior of damage defects with increasing temperature. The coarsening of dislocation loops and formation and dissolution of precipitates with increasing annealing temperature were characterized by transmission electron microscopy. Nanoindentation was performed to measure the variation of hardness caused by irradiation. Additionally, the relationship between irradiation hardening and microstructure evolution was established. This study lays a foundation for the evaluation of irradiation damage properties of GH3535 alloy at different annealing temperatures.

Characterizing heavy ions-irradiated Zr/Nb: Structure and mechanical properties

In this work, the radiation responses of Zr/Nb nanostructured metallic multilayers (NMMs) are studied. The nanostructures with different layer thicknesses were deposited on Si (111) substrate by using magnetron sputtering and were subjected to heavy-ion irradiation at room temperature with different fluences. Nanoindentation, XRD, DFT, SIMS, and Variable Energy Positron Annihilation Spectroscopy (VEPAS) techniques were used to study the type and distribution of defects, and strain within the material as well as the changes in the hardness of the structures as a function of damage. Our results suggest that the strain and the irradiation hardening are layer thickness- and damage-dependent while they are independent of the type of irradiated ions. The magnitude of hardening decreases with decreasing individual layer thickness indicating that the number of interfaces has a direct effect on the radiation tolerance enhancement. For thin layers with a periodicity of 27 nm (Zr/Nb27), a transition from hardening to softening occurs at high fluence, and a saturation point is reached in thick layers with a periodicity of 96 nm (Zr/Nb96). The as-deposited thin multilayers presented a significantly higher atomic-scale disorder which increases with ion irradiation compared to the thick multilayers. VEPAS reveals the vacancy defects before and after irradiation that contribute to the presented strain. Based on the findings, thin nanostructured Zr/Nb multilayered structures possess excellent radiation resistance due to the high density of interfaces that act as sinks for radiation-induced point defects.

Insights into hardening, plastically deformed zone and geometrically necessary dislocations of two ion-irradiated FeCrAl(Zr)-ODS ferritic steels: A combined experimental and simulation study

As one of promising candidate materials for fuel claddings and structural components in the Gen-IV fission reactors, FeCrAl(Zr)-ODS ferritic steels were studied to well understand the radiation hardening behavior. Nanoindentation (NI) hardness and plastically deformed zone (PDZ), geometrically necessary dislocations (GNDs) as well as microstructures were investigated for two FeCrAl(Zr)-ODS ferritic steels irradiated with 6.4 MeV Fe3+ at room temperature (RT) up to the nominal damages of 2, 10 and 50 dpa. Irradiation-induced hardening, which was estimated by using the Nix-Gao model regardless of the damage gradient effect (DGE), increased continuously with increasing the nominal displacement damage. When taking the DGE into account, the dependence of the irradiation hardening (in MPa) on a local damage level (dpa) obtained by finite element method (FEM) simulations was 153.65 × (dpa)0.26 and 158.35 × (dpa)0.25 for the non-Zr steel and the Zr-added one, respectively. The hardening caused by ion-irradiation was discussed in terms of the loss of oxide particles, the formation of dislocation loops and the solid solution hardening mainly by dissolved oxygen. The addition of Zr reduced the ion-irradiation hardening of steel to some extent mainly by suppressing the formation of dislocation loops. Meantime, as the FEM simulations revealed, the PDZ size underneath a conical indenter at the equivalent plastic strain εeq > 1.9% can reasonably estimate the density of GNDs. According to the strain gradient plasticity (SGP) theory, the densities of GNDs at an indentation depth of 200 – 300 nm were slightly higher in the case of the 50 dpa/nominal than the unirradiated, in spite of similarity for the two FeCrAl(Zr)-ODS steels.

Localized Backside Etching Structure of SOI Substrates on Total Ionizing Dose Effect Hardening for RF Applications

In this work, the localized backside etching (LBE) structure is introduced as a strategy for total ionizing dose (TID) irradiation hardening. Immunity to TID-irradiation-induced radio frequency (RF) property degradation is observed for the first time in SOI substrates with the LBE structure. Identical co-planar waveguide transmission lines (CPW TLines), crosstalk characterization structures, and planar spiral inductors are designed and fabricated on different 8-in SOI wafers to characterize the RF properties of substrates. Electrical characterization in the frequency ranges from 100 MHz to 40 GHz reveals that the LBE structure is beneficial to suppressing the parasitic surface conduction (PSC) effect, which will lead to better RF properties of SOI substrate. It is shown that under <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">60</sup> Co <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula> -ray irradiation, RF losses are significantly increased in the case of conventional substrates. In contrast, LBE substrates are almost insensitive to TID irradiation due to suppression of PSC. A developed physical model is proposed to characterize the behavior of inductors under PSC and TID effects, and the improvement by LBE structure on RF TID tolerance is separated through model parameter extraction.

Tungsten–tantalum alloys for fusion reactor applications

Tungsten–tantalum (W-Ta) alloys containing 1–5%Ta were developed for fusion reactor applications. For the mechanical properties and resistance against recrystallization in the non-irradiated state, Ta-alloying improved the performance of W. Grain refining and solute Ta improved the recrystallization temperature and strength. In addition, W-1%Ta exhibited a superior ductility, including ductile-to-brittle transition temperature (DBTT), which might be attributed to the strength of sub-grain boundary. Proton irradiation study at 1500 °C to 0.5 dpa revealed a decrease in irradiation hardening (increase in hardness) in recrystallized specimens and a decrease in hardness reduction in as-received specimens with increasing Ta concentration. With respect to the high temperature irradiation of the as-received specimens, microstructural recovery followed by recrystallization could be one of the dominant factors affecting the hardness change. As a material performance required for the actual components of fusion reactors, resistance against high heat flux exposure have been studied through thermal shock tests. Thorough this experiment, it was clarified that the degradation by high heat flux exposure could be suppressed due to Ta-alloying.

Machine learning predictions of irradiation embrittlement in reactor pressure vessel steels

Irradiation increases the yield stress and embrittles light water reactor (LWR) pressure vessel steels. In this study, we demonstrate some of the potential benefits and risks of using machine learning models to predict irradiation hardening extrapolated to low flux, high fluence, extended life conditions. The machine learning training data included the Irradiation Variable for lower flux irradiations up to an intermediate fluence, plus the Belgian Reactor 2 and Advanced Test Reactor 1 for very high flux irradiations, up to very high fluence. Notably, the machine learning model predictions for the high fluence, intermediate flux Advanced Test Reactor 2 irradiations are superior to extrapolations of existing hardening models. The successful extrapolations showed that machine learning models are capable of capturing key intermediate flux effects at high fluence. Similar approaches, applied to expanded databases, could be used to predict hardening in LWRs under life-extension conditions.

Evaluating the irradiation hardening of reactor pressure vessel steels by nanoindentation hardness test and micropillar compression test

ABSTRACT Ultra-small testing technologies (USTTs), such as nanoindentation hardness tests and micropillar compression tests, can be potentially used to measure the mechanical properties of irradiated materials. However, the hardness and yield strengths of USTTs are different from those of standardized methods on bulk samples. In addition, irradiation-induced defects can affect the size dependence on the mechanical properties. This study examined A533B reactor pressure vessel steels before and after 5 MeV electron irradiation using USTTs. The depth dependence of the nanoindentation hardness was analyzed using the Nix-Gao model to evaluate the bulk-equivalent nanoindentation hardness and confirm its good linear relationship with the Vickers hardness. The compression yield strength was evaluated with a truncated square pyramid shape micro-pillars where the edge length (d) of the top surface and the height (L) of the micropillars were made similar: (d, L) = (1,2), (2,4), (3,6), (4,8), and (8,16) in μm. The size dependence of the yield stress in the micropillars has a minimum at d = 2 µm, which remained unaffected by electron beam irradiation up to 22 mdpa. The average yield stress of the d = 4 and 8 µm specimens was similar to that of the bulk tensile specimens.

The effect of interstitial carbon atoms on defect evolution in high entropy alloys under helium irradiation

In this work, a design strategy based on a typical multi-principal element alloy (Fe50Mn30Co10Cr10, at. %) was proposed to tune the irradiation behavior by adding interstitial carbon atoms (Fe49.5Mn30Co10Cr10C0.5, at. %). Advanced scanning transmission electron microscopy was employed to study the evolution of dislocation loops and helium bubbles in alloys after 400 keV He+ ion irradiation at three different temperatures ranging from 350 °C to 450 °C. The results showed that the interstitial carbon atoms inhibited the growth of He bubbles and irradiation hardening. The high-resolution energy dispersive spectroscopy mapping revealed that the segregation at the dislocation loops in Fe49.5Mn30Co10Cr10C0.5 was efficiently lessened compared to that in Fe50Mn30Co10Cr10, which shows an even superior performance than the equiatomic FeMnCoCrNi Cantor alloy. Radiation-induced “W-shaped” elemental oscillation patterns with a width of about 1 nm around dislocation loops, were observed for the first time in high-entropy alloys. Such nanoscale local elemental fluctuations are suggested to modify the lattice friction and thus have an impact on the mechanical properties of the material. First-principles calculations suggest that the interstitial carbon atoms can effectively enhance the site-to-site local lattice distortion and modify vacancy migration energy near carbon atoms. Our systematic experiments demonstrate that adding interstitial carbon atoms into multi-principal element or high-entropy alloy is beneficial for the reduction in irradiation hardening and volumetric swelling, as well as radiation-induced segregation.

Orientation-selected micro-pillar compression of additively manufactured 316L stainless steels: Comparison of as-manufactured, annealed, and proton-irradiated variants

Irradiation response and deformation mechanisms of additively manufactured (AM) 316 L stainless steel were studied by atomic scale characterization and micro-pillar compression. The AM 316 L stainless steels were fabricated by direct energy deposition, a laser-based additive manufacturing process. Irradiation with 2 MeV protons at 360 °C was performed to create ∼1.8 displacements-per-atom (dpa) damage in AM 316 L. Deformation behaviors of the as-manufactured, annealed, and proton-irradiated variants were studied, focusing on the effects of manufacturing-induced pores, residual stress, and irradiation-introduced defects (dislocation loops and voids). Micro-pillars were prepared from grains of pre-selected orientation, avoiding contributions of grain boundaries and allowing determination of resolved shear stress on {111} glide planes. Transmission electron microscopy was used to characterize the pre- and post-deformation microstructure. It was found that in the as-manufactured alloy variant, moving dislocations were the major deformation carrier, with noticeable blocking by fabrication-induced pores, In the annealed variant, hardness was reduced, and deformation was also accomplished by dislocation gliding. In the proton-irradiated variant, significant twinning was observed. Comparing measured resolved shear stress and predicted critical stress for dislocation dissociation, we conclude that irradiation hardening became high enough to activate twinning. Therefore, the deformation mechanism changes from dislocation gliding to twinning. The study is important for both processing optimization and performance evaluation of AM alloys for reactor applications.

In-situ TEM investigation of dislocation loop reaction and irradiation hardening in H2+-He+ dual-beam irradiated Mo

• Irradiation can directly cause the formation of 1/2<111> loops and <100> loops. • Irradiation can directly induce 1/2<111>loop to transform into <100> loop. • Two 1/2<111> loops react to form <100> loop without strict size similarity condition. • Yield strength increment caused by irradiation-induced loops is analyzed. • Loop evolution and its relationship with irradiation dose are established. Through in-situ TEM observation during 30 keV H 2 + -He + dual-beam irradiation at 723 K, the reaction and transformation of dislocation loops in pure Mo were investigated, especially for <100> loops. Irradiation could directly cause the formation of 1/2<111> loops and <100> loops, but 1/2<111> loops were dominant. In-situ observation confirmed the formation mechanism of <100> loops, including direct irradiation induced mechanism, 1/2<111> loop direct conversion mechanism, and reaction mechanism of two 1/2<111> loops. Meanwhile, the reaction of two 1/2<111> loops to produce <100> loop should not require the strict size similarity condition. The reaction between 1/2<111> loops could also produce 1/2<111> loop, which was essentially a process in which one loop absorbed another one. The yield strength increment caused by irradiation-induced loops was analyzed, and its saturation value reached 0.48 GPa at 0.06 dpa. Compared with single He + irradiation, the number density and average diameter of loops increased significantly and more serious damage was caused under the synergistic effect of hydrogen and helium. The mechanism based on in-situ experimental observation was discussed in depth. Graphical Abstract .

A comparison study of change in hardness and microstructures of a Zr-added FeCrAl ODS steel irradiated with heavy ions

FeCrAl oxide dispersion strengthened (ODS) steel is regarded as one of the most promising candidate materials for nuclear fission reactors such as supercritical pressurized water reactor (SCPWR), sodium-cooled fast reactor (SFR), and lead bismuth-cooled fast reactor (LFR) and so on, since it has excellent irradiation resistance and good corrosion resistance of super-critical water and Pb–Bi coolants. In the present work, irradiation response of a Zr-added FeCrAl ODS steel (15Cr–4Al-0.1Ti-0.6Zr) together with an Al-free ODS (16Cr-0.1Ti) steel and a non-ODS steel T91 was studied. Electron backscattered diffraction (EBSD) and transmission electron microscopy (TEM) were used to characterize the microstructures including grains, dislocations, and oxides in the pristine specimens and irradiation defects in the irradiated specimens. Nano-indentation tests were used to evaluate the change in hardness after irradiation. The oxides in the 15Cr–4Al-0.1Ti-0.6Zr specimen are slightly larger than those in the 16Cr-0.1Ti specimen, but with a similar number density. After irradiated with 9.45 MeV 83Bi35+ to 26 dpa at ambient temperature, hardening was observed in all the specimens. Irradiation hardening of the 15Cr–4Al-0.1Ti-0.6Zr specimen is slightly higher than the 16Cr-0.1Ti specimen but significantly lower than the non-ODS steel T91. The irradiation defects in the ODS steel specimens are apparently smaller than those in T91 with a similar number density. By calculating sink strength of the initial microstructures such as grains, dislocations and oxides, the evolution of irradiation defects was discussed. A linear relationship between the irradiation hardening and the sink strength was established. A similar linear relationship between the calculated irradiation hardening using the TEM data (based on dispersion barrier hardening model (DBH) and Friedel-Kroupa-Hirsch (FKH) model, respectively) and sink strength was also found. The calculated irradiation hardening is relatively lower than the experimental value, indicating the possible contribution of Cr segregation to hardening which needs further study.

Irradiation hardening of advanced tungsten grades: An experimental and inverse finite element method study

The effect of neutron irradiation on mechanical properties of various advanced W grades (pure, doped by K, alloyed by Re) was assessed by means of mechanical testing using miniaturized tensile and sub-miniaturized three-point bending (3PB) specimens. Different irradiation conditions in terms of temperature (400, 600, 800 and 1100 °C) and irradiation dose (0.2–0.25 dpa, 1.0–1.1 dpa) were considered. In order to extract the yield stress from the flexural load-displacement data, the empirical correlation between the flexural stress at a specific reference point on the flexural stress-strain curve and uniaxial tensile yield stress was established. The empirical ratio identified can be applied for the 3PB testing in a wide range of temperatures and plastic properties of un-irradiated and irradiated W alloys. The accuracy of the yield stress determination deduced using the empirical correlation from the 3PB tests was compared with the inverse FEM modelling procedure applied to the 3PB tests carried along with reference data extracted from the dedicated tensile tests. The information on the tensile yield stress and work hardening rate of all studied materials at different irradiation conditions was summarized and discussed in the view of the irradiation hardening.

Experiments and/or crystal plasticity finite element modeling of the mechanical properties of pristine and irradiated tungsten single crystal

Tungsten is used for vital applications in nuclear industry, particularly in experimental fusion reactors like ITER. Experimental evaluation of the irradiation-induced hardening is usually carried out by indentation tests. However, accurate prediction of irradiation hardening by indentation simulations faces two challenges: (1) compression data of single-crystal tungsten for indentation model calibration are lacking, since most available data are from tension tests, but tungsten exhibits a tension-compression asymmetry; (2) the irradiation hardening model for tungsten needs to be improved by incorporating more accurate irradiation defect strengthening mechanisms and defect density evolution laws that are quantitatively extracted from the lower scale modeling such as dislocation dynamics (DD) and molecular dynamics (MD). In this work, we present a comprehensive solution for the neutron irradiation hardening prediction of tungsten based on the experiments on pristine single-crystal tungsten and lower scale simulations. First, we conduct compression and Vickers indentation experiments to calibrate and validate the model for indentation simulations, and extend the model with the irradiation defect hardening law and the defect density evolution law informed by DD and MD simulations. Then, we use the irradiation hardening results from the literature to benchmark the model for its effectiveness, and find that the predicted hardness increases match well with the experimental results for different doses. Moreover, we discuss the drawback to the widely-used conventional hardening models, and find that these models tend to overestimate the defect strengthening coefficients.