Engineering topics

Bhattacharya, Arunodaya

Publications and source records attributed to Bhattacharya, Arunodaya.

17 records

DOE OSTI2024

Development of PWHT-Free, Reduced Activation Creep-Strength Enhanced Bainitic Ferritic Steel for Large-Scale Fusion Reactor Components

A compositional modification has been proposed to validate an alloy design which potentially eliminates the requirement of post-weld heat treatment (PWHT) while preserving the advantage of mechanical properties in a reduced activation bainitic ferritic steel based on Fe-3Cr-3W-0.2V- 0.1Ta-Mn-Si-C, in weight percent, developed at Oak Ridge National Laboratory in 2007. The alloy design includes reducing the hardness in the as-welded condition for improving toughness, while increasing the hardenability for preserving the high-temperature mechanical performance such as creep-rupture resistance in the original steel. To achieve such a design, a composition range with a reduced C content combining with an increased Mn content has been proposed and investigated. Newly proposed “modified” steel successfully achieved an improved impact toughness in the as- welded condition, while the creep-rupture performance across the weldments without PWHT demonstrated ~50% improvement of the creep strength compared to that of the original steel weldment after PWHT. The obtained results strongly support the validity of the proposed alloy design.

Yamamoto, Yukinori↗

DOE OSTI2023

Degradation of electrical resistivity of tungsten following shielded neutron irradiation

A major challenge for heat transfer in nuclear materials is to ensure thermal mobility after high amounts of neutron irradiation. Tungsten is widely selected as a heat transfer material in fusion reactors. In metals, thermal conductivity is dominated by electrons’ ability to transfer energy. Neutron irradiation generates point defects, clusters, and solid transmutation (e.g.rhenium and osmium in tungsten), which inhibit electron motion. The purpose of this work is to quantify the irradiation-induced change in electron mobility and deconvolute transmutation and microstructural effects on observed changes to electron mobility. Single and polycrystalline tungsten were fast neutron irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to doses between 0.2 and 0.7 displacements per atom (dpa) and temperatures from 500 °C to 1000 °C. Grain growth was observed in all samples. Microstructure and transmutation were quantified. The geometric orientation of samples with elongated grains has been shown to affect electrical resistivity. A mathematical model was developed and used to deconvolute solid-solution transmutation, grain, and temperature-dependent lattice effects on resistivity. At ~0.4 dpa at ~590 °C, the combined resistivity degradation due to voids, vacancies, interstitials, and dislocations is estimated to be greater than the contribution from solid solution Re transmutation, which is greater than the contribution from grain boundaries. At doses of ~0.7 dpa at ~750 °C, solid solution Re contributions are greater than all other effects combined. As a result, this work establishes a basis to predict the effects of irradiation temperature and transmutation on thermal properties of tungsten and highlights the importance of irradiation temperature.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Analysis of position-dependent cavity parameters in irradiated metals to obtain insight on fundamental defect migration phenomena

Motion of point defects is a fundamental process that governs microstructure and properties of materials. Here, we examine the near-surface and grain boundary cavity swelling depth profiles in neutron- and ion-irradiated simple metals (Cu, Ni, and Fe-Cr) and investigate diffusional broadening of implanted Ni ions in Fe-Cr alloys. Vacancy migration energies and radiation-enhanced diffusion were experimentally estimated. Cavity denuded zone widths near planar sinks are shown to be dependent on temperature, damage rate, balance of point defects and sinks, and vacancy migration energies. An enhanced cavity swelling zone adjacent to the void-denuded zone was observed in specimens with low to moderate sink strength and interpreted as evidence of 1D gliding interstitial clusters. Radiation-enhanced diffusional broadening of implanted Ni ions (at 400–550 °C) in Fe and Fe-Cr was up to 250 nm toward the surface, which leads to a much broader near-peak-damage region where cavity swelling is suppressed. Diffusional broadening of implanted ions is calculated to be similarly pronounced for self-ion irradiations, particularly near the peak and higher swelling temperature regimes. Adequately high ion energies (~8–15 MeV) are recommended for ion-irradiation studies to provide a sufficiently broad midrange safe analysis region with minimized surface and injected ion effects.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Preparation & shipping of ten neutron irradiated Eurofer97 steel variants to HFIR for SANS experiments

At ORNL, ten variants of Eurofer97 steel were irradiated in the high flux isotope reactor (HFIR) to ~2.94 – 3.24 dpa at 300 ± 30 °C, as part of the EUROfusion Lot-IV collaboration. The irradiations were performed in non-instrumented rabbit capsules in the flux trap region, which included SS-J3 type tensile samples and M4-CVN multi-notch bend bars. Their full set of microstructure and mechanical properties were previously reported in multiple previous publications. In this project, half-broken irradiated and nonirradiated SS-J3 tensile samples from ten alloys, code-named H, I, J, K, L, M, N, O, P and E, were prepared and shipped for facilitating small angle neutron scattering (SANS) experiments at the HFIR general purpose (GP)-SANS beam line. This report summarizes the completed tasks which included canister moves at the Irradiated Materials Examination and Testing (IMET) hot-cell facility to retrieve the samples from long-term storage, loading of the ten irradiated samples at IMET inside lead (Pb) piglets that were specifically provided by ORNL for SANS experiments and radiological shipment from IMET facility to the HFIR hot cells for performing SANS experiments. In addition to the irradiated samples, ten nonirradiated half-broken pieces from the Eurofer97 variants were also provided to HFIR for SANS.

36 MATERIALS SCIENCE↗

DOE OSTI2022

The role of Cr concentration and temperature on cavity swelling with co-injected helium in dual-ion irradiated Fe and Fe-Cr alloys

The level of chromium plays an essential role in irradiation tolerance of Fe-Cr ferritic alloys. However, conflicting results have been reported regarding the dependence of cavity swelling under irradiation on Cr level and temperature. Here, we have performed a comprehensive set of simultaneous dual-ion (Ni + He) irradiations to high dose (~30 displacements per atom, dpa) at 400–550 °C on a series of ultra-high purity Fe and Fe-Cr binary alloys (3–14 wt.%Cr). Helium co-implantation rates of 0.1 and 10 appm He/dpa were selected to examine He synergistic effects relevant for fission and fusion reactor conditions, respectively. Cavities were observed in all irradiated samples by transmission electron microscopy. The results show that higher He implantation rate causes a shift in the swelling peak to higher temperatures in both Fe and Fe-Cr alloys. When assuming smaller cavities as biased sinks, the non-monotonic nature of the cavity swelling behavior is related to the ratio of biased to unbiased point defect sink strengths. Cr-enriched precipitates were observed in Fe-14Cr irradiated at 400 °C by atom probe tomography. Our analysis suggests the formation of Cr-enriched precipitates could suppress cavity swelling for Fe-Cr alloys with Cr content above 10 wt%.

36 MATERIALS SCIENCE↗

DOE OSTI2022

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

We report Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ~10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Data on Cu- and Ni-Si-Mn-rich solute clustering in a neutron irradiated austenitic stainless steel

The data presented in this article is supplementary to the research article “Phase instabilities in austenitic steels during particle bombardment at high and low dose rates” (Levine et al.). Needle-shaped samples were prepared with focused ion beam milling from a 304L stainless steel that was irradiated with fast neutrons (E 0.1 MeV) in the BOR-60 reactor at 318 °C to 47.5 dpa. Atom probe tomography (APT) experiments in voltage mode were then conducted on a Cameca LEAP 5000X HR. Atom position, range, and mass spectrum files after reconstruction with Cameca’s IVAS software are included. Cu- and Ni-Si-Mn-rich solute nanoclusters were identified and analyzed using the Open Source Characterization of APT Reconstructions (OSCAR) program. Python code for OSCAR, information on the program’s underlying algorithm, and sample output files are provided. A proximity histogram of a Ni-Si-Mn-rich cluster and a 1D density/solute concentration profile of a Cu-rich cluster are given to demonstrate OSCAR’s analytical functionalities. The provided APT dataset is valuable for benchmarking phase instabilities in neutron-irradiated austenitic stainless steels that occur at high doses. The OSCAR program can be reused to process other APT data sets where solute nanoclustering is of interest.

42 ENGINEERING↗

DOE OSTI2022

Effect of heavy ion irradiation dose rate and temperature on α' precipitation in high purity Fe-18%Cr alloy

Cr-rich alpha prime precipitates (CrRP) induce hardening and embrittlement of FeCr alloys, but the kinetics of CrRP formation due to particle irradiation are not well understood. In this study, Fe18wt.%Cr alloy in solid solution state and pre-aged to produce relatively coarse CrRP was irradiated with 8 MeV Fe ions. The irradiation conditions involved two midrange doses of 0.37 and 3.7 displacements per atom (dpa), a wide range of dose rates (10 -5 –10 -3 dpa/s) and temperatures (300–450 °C). The distributions of CrRP after irradiation were studied with atom probe tomography (APT). The critical irradiation conditions to suppress CrRP formation were identified as 300 °C and 10-3 dpa/s; CrRP formation occurred readily at lower dose rates or higher temperatures. From 0.37 to 3.7 dpa, CrRP were observed to slightly grow at 350 °C and strongly coarsen at 450 °C. Specimens with pre-existing CrRP evolved into a similar precipitate distribution as detected after ion irradiation on solution annealed specimens at 300–350 °C to 0.37 dpa, indicating that the precipitate microstructure approaches a quasi-equilibrium for doses <1 dpa. Limited shrinking of pre-existing CrRP was observed after irradiation at 450 °C to 0.37 dpa, indicating a higher recovery rate at this temperature. Additionally, the evolution of CrRP is quantitatively explained by employing corrections to the historic Nelson-Hudson-Mazey precipitate stability model, and a radiation modified precipitation mechanism is proposed to account for the competition between radiation enhanced diffusion and ballistic dissolution which results in the modifications on both size and solute concentration of CrRP.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Radiation induced amorphization of carbides in additively manufactured and conventional ferritic-martensitic steels: In-situ experiments on extraction replicas

In-situ irradiations using 1 MeV Kr 2+ ions in a transmission electron microscope were performed on extraction replica samples containing M 23 C 6 carbides and MX carbonitrides from two ferritic-martensitic (FM) steels: (i) 9%Cr-1%W-TaV based Eurofer97 and (ii) 9%Cr-1%Mo-VNb (all in wt.%) based additively manufactured (AM) Grade91 steel. The irradiations were performed between 100 and 773 K up to a maximum dose of ~2.4 displacements per atom (dpa). The M 23 C 6 carbides are highly susceptible to radiation induced amorphization (RIA), while the MX type nanoprecipitates are highly amorphization resistant across the entire irradiation temperature range. Between 100 and 423 K, RIA of M 23 C 6 carbides occurs very rapidly with critical amorphization doses ranging between ~0.35 and 0.9 dpa, increasing to higher doses at 573 K. Complete amorphization of the M 23 C 6 carbides up to doses of ~2.4 dpa is not possible at 773 K under the present irradiation conditions. The critical amorphization dose of M 23 C 6 carbides increases nearly exponentially with irradiation temperature. The critical temperature for the crystalline-to-amorphous phase transformation (T c→a ) of M 23 C 6 carbides irradiated as replica samples, i.e. without the surrounding metal matrix, was estimated to be ~812 K. Comparing the present results with neutron irradiation data on bulk samples reveals a decrease in the critical amorphization dose for M 23 C 6 and an increase of T c→a , highlighting the effect of dose rate on the amorphization behavior that is qualitatively consistent with literature on other non-metals. Changes in the minor chemistry of M 23 C 6 , such as presence or absence of W, V, Nb, Mo, seems to have little effect on the amorphization behavior. Here, in-situ irradiations on extraction replica samples provide a novel pathway to explore radiation tolerance of nanoprecipitates in nuclear structural materials

36 MATERIALS SCIENCE↗

DOE OSTI2022

STEM imaging of irradiation induced defects in Eurofer97 steel variants irradiated in the EUROfusion collaboration

Ten exploratory variants of Eurofer97 reduced activation ferritic martensitic (RAFM) steel were irradiated in HFIR to ~2.94 – 32.4 dpa, 300±30 °C as a part of the EUROfusion collaboration. The irradiations were performed in rabbit capsules. ES21-22 capsule were used for irradiating SS-J3 tensile samples while M4CVN bend bar samples were irradiated in ES31-35 capsules. The mechanical properties and multi-length scale characterization of the microstructures of these ten steels, designated as H, I, P, J, K, L, M, N, O, P and reference E, in the nonirradiated form was reported in "Mechanical properties and microstructure characterization of unirradiated Eurofer-97 steel variants for the EUROfusion project, ORNL/SPR-2018/882," while their irradiated properties and some microstructure analysis was reported in "Post-irradiation examination of Eurofer97 steel variants irradiated to 2.5 dpa, ~300 °C in HFIR for the EUROfusion program, ORNL/SPR-2020/1440." Here, additional microstructures using TEM/STEM and STEM-EDX of the irradiated steels is presented. Characterization was performed at the Low Activation Materials Development and Analysis (LAMDA) laboratory.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Irradiation hardening and ductility loss of Eurofer97 steel variants after neutron irradiation to ITER-TBM relevant conditions

Ten Eurofer97 steel variants, produced by non-standard fabrication-processing routes and modified alloying chemistries, were studied by neutron irradiations in the high flux isotope reactor. The irradiations were performed to ITER-TBM relevant conditions of ~255–350 °C, 2.94–3.24 dpa. We quantified the irradiation-induced degradation of the steels using mechanical property tests. All the steels suffered from irradiation hardening, where a significant increase in Vickers microhardness and yield stress (σ YS ) occurred, accompanied with severe loss of tensile elongation. The extent of hardening was material dependent. For T irr = 300±30 °C, most steels showed σ YS increase in the range of ~30% to as high as ~66%, except for a low temperature tempered steel with σYS increase below 15%. Despite large losses in elongation, most failures were ductile. Significant post-necking ductility was retained with reduction in area (RA) between 65–75%, but <50% for low temperature tempered steels. The ultimate tensile stress to yield stress (σ UTS /σ YS ) ratios decreased significantly after irradiation, highlighting irradiation-induced strain hardening capacity reduction. No major effect of irradiation on the plastic instability stress (σ PIS ) and true fracture stress of the steels was observed. By comparing the tensile stresses in true stress units and with literature, the results suggest that RAFM steel designing should target materials with a large separation between σ PIS and σ YS , to ensure the materials can maintain large work hardening and uniform deformation capability after irradiation. The tensile data of the steels additionally revealed a compelling evidence of an inverse trend between the change in RA and increase in σ YS of the neutron irradiated Eurofer97 type steels.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Helium causing disappearance of a/2<111> dislocation loops in binary Fe-Cr ferritic alloys

In this work, single and dual-beam self-ion irradiations were performed at 500°C on ultra-high purity Fe14%Cr alloy to ~0.33 displacements-per-atom (dpa), and 0 or 3030 atomic-parts-per-million (appm) helium/dpa, respectively. Using transmission electron microscopy, we reveal that helium can drastically modify the dislocation loop Burgers vector in Fe-Cr alloys. Helium co-implantation caused complete disappearance of a/2<111> type dislocation loops, and the microstructure consisted of only a<100> loops. Conversely, a/2<111> type loops were predominant without He co-implantation. The total loop density remained largely unaffected. The results strikingly contrast literature asserting that helium stabilizes a/2<111> type loops in bcc Fe alloys, based on low temperature irradiations. Collectively analyzing the results with literature suggest that the small positive interaction between helium and self-interstitial atoms (SIA) in Fe predicted by atomistic simulations maybe insufficient to holistically explain the dislocation loop microstructure development in presence of helium. Helium-SIA positive binding inadvertently implies elevated a/2<111> loop fraction and higher loop densities that the present results contradict. Helium induced high cavity density causing a preferential loss of highly glissile <111> clusters, leaving the matrix saturated with <100> type clusters is proposed as a potential mechanism. Further, the in-situ irradiations combined with Burgers vector analysis strengthened the evidence of Cr-induced dislocation loop mobility reduction that appears to stabilize the a/2<111> type loops and causes higher loop densities in Fe-Cr alloys.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Technological aspects in blanket design: Effects of micro-alloying and thermo-mechanical treatments of EUROFER97 type steels after neutron irradiation

Presently available data on neutron irradiation damage raise doubts on the feasibility of using EUROFER97 steel for a water-cooled starter blanket in a DEMO reactor, since the ductile-to-brittle transition temperature (DBTT) increases significantly for irradiation temperatures below 350°C. The additional DBTT shift caused by H and He transmutation can only be estimated based on very few results with isotopically tailored EUROFER97 steel. Conservative calculations show that the DBTT of EUROFER97 steel could exceed the operating temperature in water-cooled starter blankets within a relatively short time period. This paper presents results from a EUROfusion funded irradiation campaign that was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The paper compares ten newly developed reduced activation ferritic-martensitic (RAFM) steels irradiated to a nominal dose of 2.5 dpa at 300°C. The post-irradiation experiments using Small Specimen Test Technology included hardness, tensile, and fracture mechanics tests combined with fractography and microstructure analysis are presented. Results show that micro-alloying EUROFER97-type steels influenced the mechanical properties but a dominating impact on irradiation damage resistance could not be identified. In contrast, specific thermo-mechanical treatments lead to better DBTT behavior. Discussion about irradiation response to heat treatment conditions is also given. Here, despite requiring data also at high dpa values, the results indicate that with these modified materials an increased lifetime and potentially also an increased operating temperature window can be achieved compared to EUROFER97.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Temperature-dependent cavity swelling in dual-ion irradiated Fe and Fe-Cr ferritic alloys

Fe-Cr ferritic-martensitic (FM) steels are promising structural material candidates for fusion and advanced fission reactors due to their attractive mechanical properties and volumetric swelling resistance. However, significant discrepancies exist regarding the effect of solutes and irradiation temperature on cavity swelling under ion versus neutron irradiation conditions. In this study, simultaneous dual ion irradiations (8 MeV Ni 3+ ions and energy-degraded 3.5 MeV He 2+ ions) were used to quantify the cavity swelling behavior in ultra-high purity Fe and Fe-Cr alloys (3-14 wt.% Cr), Fe-10 wt.% Cr-780 wt.ppm C, and Eurofer97 FM steel. The irradiations were conducted over a wide temperature range (400-550°C) with a mid-range dose of ~30 displacements per atom (dpa) and 0.1 appm/dpa He implantation rate. Here, using state-of-the-art transmission electron microscopy (TEM), we reveal that pure Fe has a ~50°C lower peak swelling temperature difference than Fe-Cr alloys, which is attributed to higher vacancy mobility in pure Fe. Chromium solute appears to strongly suppress cavity swelling in Fe-Cr alloys for temperatures below ~470°C, but seems to have little effect or slightly enhances swelling above ~470°C. Cavities were observed in all the irradiated samples between 400-550°C. This indicates that the narrow temperature range of observable cavities reported in prior ion irradiated Fe-Cr ferritic alloy studies is likely an artifact associated with the use of low ion energies (<5 MeV), which leads to pronounced near-surface and implanted ion effects that suppress cavity swelling even at midrange depths (particularly at high temperatures).

36 MATERIALS SCIENCE↗

DOE OSTI2020

Cavity Swelling in Irradiated Materials

Cavity swelling, due to three-dimensional clustering of vacancies, is an undesirable isotropic volume expansion of materials under irradiation. This phenomenon occurs above stage-III recovery temperature where vacancies are mobile, between ~0.3–0.6 Tm where Tm is the absolute melting temperature. The primary mechanism for cavity swelling is “dislocation bias”, i.e., preferential absorption of interstitials by network dislocations, causing a vacancy supersaturation. Cavity swelling is highly temperature dependent and typically follows a bell-shaped curve with peak swelling occurring at an intermediate temperature. With increasing temperature, cavity size increases, and number density decreases logarithmically. With increasing dose, cavity swelling increases, but follows a hockey-stick type pattern, characterized by a low-swelling transient incubation period and then a steady-state regime. The incubation period, dominant in cavity nucleation, is dependent upon material’s chemistry and irradiation conditions such as dose rate/temperature. Cavity swelling increases linearly with dose in the steady-state regime, that is independent of chemistry and dose rate. This regime is dominated by cavity growth. Transmutation gases like He enhance cavity nucleation such that an early onset of cavity formation occurs. However, swelling magnitudes can be either increased or suppressed when He is present. Results in austenitic steels and Cu indicate that cavity swelling peaks at intermediate He/dpa ratios of ~10 appm He/dpa. The most effective method to mitigate cavity swelling is by increasing the fixed point-defect sink density, by promoting precipitation or dispersion strengthening.

Bhattacharya, Arunodaya↗

DOE OSTI2020

Post-Irradiation Evaluation of Eurofer97 Fracture Toughness Using Miniature Multinotch Bend Bar Specimens

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

Chen, Xiang↗