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The brittle–ductile transition temperature (BDTT) and fracture toughness are useful indicators of the brittleness of reactor pressure vessel materials. However, current methods to analyze these parameters are expensive and time-consuming; moreover, the relevant mechanism is not very clear. Thus, in this study, a body-centered cubic crystal plasticity finite element model (CPFEM) based on the densities of dislocation and irradiation defects, which are full- and partial-absorption dislocation loops, was numerically implemented on A508-3 steel. Irradiation embrittlement was investigated in the CPFEM framework. The cleavage failure model for a representative volume element was proposed to calculate the BDTT, the results of which agreed well with experimental results in the literature for both unirradiated and irradiated samples. The risk for brittle failure is higher for irradiated samples at constant temperature. Furthermore, the proposed model combined with the Beremin cleavage fracture local approach, could estimate the fracture toughness of A508-3 steel at different temperatures before and after irradiation. The results obtained were consistent with reported experimental data, indicating that irradiation exacerbates the embrittlement. This work is beneficial to the calculation and evaluation of irradiation effect of RPV steel, providing reference for the research of reactor life extension. [ABSTRACT FROM AUTHOR]

Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals, and steam generator, which are the lifetime-limiting components besides the containment structure. Some of the corrective actions, improvements of in-service inspections, maintenance, and ageing management are also considered in the paper. It is shown that the selection of materials and unavoidable material degradation processes do not limit the operation of the WWER-440 reactors for up to 60–70 years under the established operational environment and ageing management practice. [ABSTRACT FROM AUTHOR]

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21. [ABSTRACT FROM AUTHOR]

The fitness for service code of the Japan Society of Mechanical Engineers (JSME) provides a fracture evaluation methodology based on linear elastic fracture mechanics (LEFM) for boiling water reactor (BWR) internal components irradiated to a neutron fluence of 3.0 × 1024 n/m2 or more. On the other hand, it is known that austenitic stainless steel maintains its ductility and toughness even after neutron irradiation up to the fluence expected for BWR internal components. Accordingly, it is preferable to apply the ductile fracture evaluation methodology for BWR internal components. The present study examined the applicability of damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel. The changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. As a result, in terms of material strength properties the applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2. • Damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel was investigated. • Changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. • Fracture tests and analyses were performed with plate specimens. • Applicability of GTN model was examined by comparing the maximum loads and fracture behaviors between the tests and analyses. • Applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2. [ABSTRACT FROM AUTHOR]

The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works. [ABSTRACT FROM AUTHOR]

• The changes in fracture behavior of stainless steel with in neutron dose can be predicted from the presented method. • The multiaxial fracture strain model is determined by simulating tensile and fracture toughness test. • The neutron irradiation embrittlement constant is determined by simulating tensile test. • The neutron irradiation embrittlement constant is multiplied with the multiaxial fracture strain model to simulate fracture toughness tests, enabling the prediction of degradation in fracture toughness due to neutron irradiation. This paper proposes a numerical method for predicting the reduction in the fracture toughness with neutron dose using the concept of an embrittlement constant. Tensile and fracture toughness test data for type 304 stainless steel in the literature [ 41 , 42 ] are used to validate the proposed method. Fracture toughness tests are simulated using a stress-modified fracture strain damage model implemented in an ABAQUS user-subroutine. The true stress-strain curve and one point of the fracture strain locus are obtained from the tensile test simulation. The remaining parameters of the fracture strain locus are determined by simulating a fracture toughness test. The fracture strain locus is modified to reflect the irradiation effect on the fracture toughness using the proposed irradiation embrittlement constant C I. The function of the constant with the neutron dose is calculated from the tensile tests and simulations. Subsequently, the modified damage model is used to predict the reduction in the fracture toughness. The predicted simulation results are compared with experimental data, showing good agreement. [Display omitted] [ABSTRACT FROM AUTHOR]

Life extension of light water reactors requires robust models to predict the embrittlement of reactor pressure vessel (RPV) steels under long-term neutron irradiation. Embrittlement is due to nanoscale precipitation hardening by phases containing Cu, Mn, Ni, and Si, resulting in increases of the ductile to brittle transition temperature. Embrittlement data for low flux, high fluence extended life conditions, up to 80 years or more, are largely not available. Thus, higher flux, accelerated test reactor irradiations have been used to help develop models to predict low flux embrittlement at high fluence. Here flux, or dose rate, is expressed in units of displacements per atom per second (dpa/s). The dose rate can significantly influence precipitate evolution in a way that depends on fluence (dpa), temperature, and alloy composition, as well as flux itself. Here, we explore precipitation in surveillance steels with varying compositions at high fluxes (10−5–10−6 dpa/s) produced using 70 MeV Fe2+ ion irradiations. Atom probe tomography was used to characterize the dose dependence of precipitation at very high ion irradiation dose rates, and to evaluate corresponding similarities or differences compared to neutron irradiated steels at a much lower dose rate of ≈ 5 × 10−9 dpa/s. The data show that, on average, the ion irradiation precipitate volume fractions increase by factors of ∼ 1.5 to 2 between 0.3 dpa and 1.2 dpa. Limited data show that the major effect of neutron versus ion irradiation is the former has lower precipitate number densities and larger sizes. Otherwise, the precipitates are remarkably similar in composition and volume fraction. [ABSTRACT FROM AUTHOR]

• 9 chemical contents and 3 irradiation conditions were used as input factors. • The probability density distribution of Δ RT NDT was used to evaluated the model performance. • The GBDT model has the best performance, its prediction is close to that of ASTM E900-15. • FNF, T and the content of Cu, Ni, C are the most important 5 factors for irradiation embrittlement. A reliable irradiation embrittlement model of reactor pressure vessel (RPV) is critical to the safe operation and life extension of nuclear power plants (NPP). In this study, 474 datasets from NUREG/CR-6551 and 6 machine learning (ML) methods were adopted to accurately predict the irradiation embrittlement behavior (Δ RT NDT) of RPV with comprehensive input factors involving irradiation conditions and metal composition. The GBDT model has the best prediction performance. The calculated result by GBDT model is close to those of JEAC-4201 and ASTM E900-15. Cu content is the most sensitive factor for Δ RT NDT , followed by neutron fluence. This work demonstrates a successful application of ML in promoting NPP safety and RPV materials design. [ABSTRACT FROM AUTHOR]

The comparison of experimental values of the critical brittle temperature ΔTF and reference temperature ΔT0 of VVER-1000 reactor vessel weld metal with an elevated content of manganese and nickel is performed. ΔTF and ΔT0 values are defined proceeding from the standard impact bend Charpy and Charpy cracked fracture toughness specimen tests, respectively. Specimens were irradiated in industrial reactors in the frame of surveillance specimen program up to the fast (E ≥ 0.5 MeV) neutron fluences corresponding to the NPP long term operation period. The research results showed the shifts ΔTF and ΔT0 to agree with each other. Besides, it was discovered that in the range of over-design fluences the design embrittlement model has a tendency to underestimate the critical brittle temperature shift. [ABSTRACT FROM AUTHOR]

The behavior of solute atoms in an undersaturated Fe-3.3 at.%Ni model alloy upon heavy and light particle irradiations was investigated by Atom Probe Tomography. Whereas first irradiations, performed using 27 MeV Fe ions at 400 °C, highlight significant Ni enrichment on point defect clusters, solute-rich features detected after 2.8 MeV H + ion (400 °C) and 1 MeV electron (320 °C) irradiations are only associated with pre-existent microstructural defects. Nanoindentation tests show no hardening in light particle irradiated samples, strongly suggesting that no dislocation loop formed. These observations put forward the importance of point defect clusters resulting from displacement cascades in Ni clusters' formation. Image, graphical abstract [ABSTRACT FROM AUTHOR]

In situ irradiation in transmission electron microscope following pre-irradiation was performed on Fe-0.6Ni alloy to investigate the growth of dislocation loops under irradiation. The observations showed two types of loop growth behaviour. One type was almost linear growth with irradiation dose, possibly caused by loop absorbing freely migrating defects. The other was abrupt growth with rapid increase in diameter. It was found that nanometre-sized loop might migrate one-dimensionally, following which the large loop could absorb the small loop. This process can result in the abrupt growth of loop and broadening of loop size distribution. Image, graphical abstract [ABSTRACT FROM AUTHOR]

The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications. Image, graphical abstract [ABSTRACT FROM AUTHOR]

Post irradiation annealing (PIA) clarified the induced versus enhanced controversy regarding nanoscale Mn-Ni-Si precipitate (MNSP) formation in pressure vessel steels. Radiation induced MNSPs would dissolve under high temperature PIA, while radiation enhanced precipitates would be stable above a critical radius (r c). A Cu-free, high Ni steel was irradiated with 2.8MeV Fe2+ ions at two temperatures to generate MNSPs with average radii (r ¯) above and below an estimated r c for PIA at 425°C up to 52 weeks. Atom probe tomography and energy dispersive x-ray spectroscopy showed MNSPs with r < r c dissolved, while those with r > r c slightly coarsened, consistent with thermodynamic predictions. Image, graphical abstract [ABSTRACT FROM AUTHOR]

The study of irradiation hardening and embrittlement is critically important for the development of next-generation structural materials tolerant to neutron irradiation, and could dramatically affect the approach to the design of components for advanced nuclear reactors. In addition, a growing interest is observed in the field of research and development of irradiation-resistant materials. This review aims to provide an overview of the theoretical development related to irradiation hardening and embrittlement at moderate irradiation conditions achieved in recent years, which can help extend our fundamental knowledge on nuclear structural materials. After a general introduction to the irradiation effects on metallic materials, recent research progress covering theoretical modelling is summarized for different types of structural materials. The fundamental mechanisms are elucidated within a wide range of temporal and spatial scales. This review closes with the current understanding of irradiation hardening and embrittlement, and puts some perspectives deserving further study. [ABSTRACT FROM AUTHOR]

An improved Cluster Dynamics (CD) model of Cu rich and Mn–Ni–Si phase co-precipitation was developed to provide insights on the combined effects of the flux, fluence, temperature and alloy composition on irradiation enhanced precipitation leading to the embrittlement of reactor pressure vessel steels. The CD model was calibrated using a large microstructural database, and key parameters (e.g., interfacial energies) were fitted to minimize the predicted versus measured errors. The CD model was further validated against data not used in the fitting. The CD model predicts that: a) even 0.05% Cu reduces the Mn, Ni, Si precipitation threshold fluence; b) precipitate number densities increase, while their sizes and volume fractions, decrease with increasing flux; c) precipitate number densities and volume fractions increase with decreasing temperature; and, d) most of the matrix Cu precipitates in the early years of vessel service, while MnNiSi precipitates continue to grow up to very high extended life fluence. Image 1 [ABSTRACT FROM AUTHOR]

Mn-Ni-Si intermetallic precipitates (MNSPs) that are observed in some Fe-based alloys following thermal aging and irradiation are of considerable scientific and technical interest. For example, large volume fractions (f) of MNSPs form in reactor pressure vessel low alloy steels irradiated to high fluence, resulting in severe hardening induced embrittlement. Nine compositionally-tailored small heats of low Cu RPV-type steels, with an unusually wide range of dissolved Mn (0.06–1.34 at.%) and Ni (0.19–3.50 at.%) contents, were irradiated at ≈ 290 °C to ≈ 1.4 × 1020 n/cm2 at an accelerated test reactor flux of ≈3.6 × 1012 n/cm2-s (E > 1 MeV). Atom probe tomography shows Mn-Ni interactions play the dominant role in determining the MNSP f, which correlates well with irradiation hardening. The wide range of alloy compositions results in corresponding variations in precipitates chemistries that are reasonably similar to various phases in the Mn-Ni-Si projection of the Fe based quaternary. Notably, f scales with ≈ Ni1.6Mn0.8. Thus f is modest even in advanced high 3.5 at.% Ni steels at very low Mn (Mn starvation); in this case Ni-silicide phase type compositions are observed. Image 1 [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) of pressurized water reactor undergoes severe neutron irradiation during normal operation and life-extension phase, thus leading to the formation of numerous nanometric irradiation-accelerated and irradiation-induced solute-enriched clusters among the ferritic matrix, including so-called late blooming phases (LBP), which was believed as Mn-Ni-Si enriched stable precipitates to emerge lately and grow rapidly and show great threat to the mechanical performance of RPV steels. However, the mechanism of nucleation and growing for LBP remains not in light, especially the effect of inherent defects such as dislocation and grain boundary on that. We investigated the atomic configurations of Fe-Cu, Fe-Mn, Fe-Ni, Fe-Mn-Ni, Fe-Mn-Cu, Fe-Ni-Cu and Fe-Mn-Ni-Cu as RPV model alloys at 600 K, implemented firstly by constructing an enough big simulation box containing edge dislocation and random-distributed alloy atoms and secondly evolving to equilibrium status after essential simulation steps with Metropolis Monte Carlo (MMC) method. The number of clusters, average solute numbers contained in each cluster and the volume atomic fraction of all clusters were characterized, as well as the structure of typical big clusters among simulation analyzed. [ABSTRACT FROM AUTHOR]

Solute cluster induced hardening plays a dominant role in the irradiation embrittlement of reactor pressure vessel. For Cu-rich clusters, the Russell and Brown model (1972) is often applied for their correlation with hardening, but many parameters of the model are taken as adjustable to fit specific experimental data. In this work, a correlation is attempted and discussed between Cu-rich clusters and hardening in a series of reactor pressure vessel model alloys. Compared with the Cu clusters of the same size, Cu-rich clusters containing Ni and Mn may have lower obstacle strength and cause lower hardening. [ABSTRACT FROM AUTHOR]

Neutron irradiation following thermal ageing experiments were performed using Fe-Cu, Fe-Cu-Ni, Fe-Cu-Ni-Mn, and Fe-Cu-Si alloys to investigate the effects of Ni, Mn, and Si on the formation of Cu-rich clusters and their chemical composition. Atom probe tomography showed that Ni and Mn increased the number density of Cu-rich clusters. This can be caused by that Ni and Mn reduce the formation energy of Cu nuclei, thus enhancing the Cu nucleation rate. Mn may also enhance the heterogeneous nucleation of Cu because of its strong interaction with self-interstitial atoms. In addition, Mn can accelerate solute diffusion during thermal ageing. A high Ni alloying content enhanced the Mn concentration in the clusters as well as the Ni concentration, which is believed to be caused by the thermodynamic relationship between Ni and Mn. In Fe-Cu-Si alloys, the Si concentration in the clusters was very low and was comparable to the Si alloying content. This indicates that Si atoms hardly cluster with Cu. [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) is the critical un-changeable component of the reactor during its service lifetime, which prevents the radioactive leak of the nuclear power plant core. The irradiation test (about 10 × 10 19 cm − 2 , E > 1 MeV) in research reactor of the pressure vessel material was carried out, and the charpy impact test has been carried out before and after the neutron irradiation. Analysis of the impact energy and the fracture morphology has been done to estimate the embrittlement due to neutron irradiation. It shows that the main effects of neutron irradiation to fracture are the crack initiation and stable expansion process. And there also are a small amount of intergranular fracture in the unstable crack expansion after neutron irradiated which aware us pay more attention to the increasing intergranular fracture behavior of higher neutron fluence level after 60a nuclear power plant operation. [ABSTRACT FROM AUTHOR]

A titanium aluminide (TiAl) alloy 45XD has been irradiated by a He ion beam with an energy of 5 MeV on a tandem accelerator at the Australian Nuclear Science and Technology Organization (ANSTO). The total fluence of He ions was 5 × 10 17 ion cm −2 . A 17 μm uniform damage region from the material surface with a helium concentration of about 5000 appm was achieved by using an energy degrading wheel in front of the TiAl target. The micro-size test specimens from the damage layer were fabricated using a focused ion beam & scanning electron microscope (FIB-SEM) system. The in situ SEM micromechanical compressive testing was carried out inside an SEM and the results indicated irradiation embrittlement in the helium affected region. Electron back scatter diffraction (EBSD) analysis has been applied to reveal the orientation of the lamellae in the TiAl specimens, and used to understand the deformation processes in the sample. The irradiation damage of gallium ion beam from FIB on the surface of TiAl sample was also investigated. [ABSTRACT FROM AUTHOR]

A theoretical study is conducted to evaluate the nucleation free energy of copper-vacancy clusters in Fe as a function of the numbers of copper atoms and of vacancies in a cluster. Using this free energy value, cluster nucleation processes during irradiation are investigated. The results clearly show that there are two different types of cluster nucleation paths on the free energy surface; one is the formation of empty voids by jumping over the ridge of the free energy surface, and the other corresponds to a path for the formation of copper clusters by going around the ridge. The dependence of easy nucleation paths on the damage rate is discussed. [ABSTRACT FROM AUTHOR]

A multiscale modeling approach was applied to investigate the nucleation process of CRPs (copper rich precipitates, i.e., copper-vacancy clusters) in α-Fe containing 1 at.% Cu during irradiation. Monte-Carlo simulations were performed to investigate the nucleation process, with the rate theory equation analysis to evaluate the concentration of displacement defects, along with the molecular dynamics technique to know CRP thermal stabilities in advance. Our MC simulations showed that there is long incubation period at first, followed by a rapid growth of CRPs. The incubation period depends on irradiation conditions such as the damage rate and temperature. CRP’s composition during nucleation varies with time. The copper content of CRPs shows relatively rich at first, and then becomes poorer as the precipitate size increases. A widely-accepted model of CRP nucleation process is finally proposed. [ABSTRACT FROM AUTHOR]

Samples harvested from an HT-9 fuel test assembly (ACO-3) irradiated for six years in the Fast Flux Test Facility (FFTF) reaching 2–147 dpa at 382–504 °C were deformed in-situ while collecting high-energy X-ray diffraction data to monitor microstructure evolution. With the initiation of plastic deformation, all samples exhibited a clear load transfer from the ferrite matrix to carbide particulate. This behavior was confirmed by modeling of the control material. The evolution of dislocation density in the material as a result of deformation was characterized through full pattern line profile analysis. The dislocation densities increased substantially after deformation, the level of dislocation evolution observed was highly dependent upon the irradiation temperature of the sample. Differences in both the yield and hardening behavior between samples irradiated at higher and lower temperatures suggest the existence of a transition in tensile behavior at an irradiation temperature near 420 °C dividing regions of distinct damage effects. [ABSTRACT FROM AUTHOR]

Massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that grows with dose (fluence), as manifested by an increasing ductile-to-brittle fracture transition temperature. Extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with beyond 60 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize highly embrittling nm-scale Mn–Ni–Si precipitates that develop in the irradiated steels at high fluence. These precipitates lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementary techniques has, for the very first time, successfully identified the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions. [ABSTRACT FROM AUTHOR]

Zr/Nb multilayers with periodicities of 10, 30 and 60 nm were prepared by magnetron sputtering and irradiated for prolonged time (1311 h) by γ-rays with energy of 1.25 MeV and a dose of 510 kGy. A qualitative comparison between XRD patterns acquired before and after irradiation revealed a progressive increase of compressive stress, especially in Nb layers, for smaller periodicities with a consequent increase in hardness measured by nanoindentation. The combination of smaller grain size and radiation-induced defect density distribution, primarily in Nb layers, was found to be responsible for the observed radiation hardening effect. [ABSTRACT FROM AUTHOR]

To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8. [ABSTRACT FROM AUTHOR]

The irradiation of high-energy particles always leads to the embrittlement of metallic materials. Here, we report an abnormal irradiation toughening phenomenon in a hierarchical structured ZrTi, which results from the irradiation-induced lattice relaxation that yields an enhanced ductility with a slight decrease of strength. Positron annihilation measurements reveal that the lattice relaxation dominantly originates from a recombination of irradiation-induced interstitials (i.e., Ti atoms) with preexisting dislocation-associated vacancies. This study provides an approach to suppress the irradiation-induced brittleness in metallic materials. [ABSTRACT FROM AUTHOR]

Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ∼295 °C to high and very high neutron fluences of ∼1.3 × 10 20 and ∼1.1 × 10 21 n cm −2 . Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ∼1.3 × 10 20 n cm −2 , while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 10 21 n cm −2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa. [ABSTRACT FROM AUTHOR]

In this study, APT and TEM analyses were used to characterise two Zr-based alloys neutron-irradiated to 13.7 dpa. The high spatial and chemical resolution of these techniques has enabled the irradiation-induced nanoscale distribution of solute elements to be characterised. The results on both Zircaloy-2 and low-Sn ZIRLO support previous observations of Fe segregation to planar dislocation arrays with enhanced Sn between. In Zircaloy-2, the APT data in this study have also revealed short-range (∼2 nm) clustering of Fe and Sn atoms both within the Fe-enriched planes and in the regions between dislocation arrays. The resultant microstructure is reminiscent of spinodal decomposition. In low-Sn ZIRLO Nb-rich precipitates formed in the Sn-rich planes. The data in this article enhance our understanding of microstructural evolution in Zr alloys under neutron irradiation and are relevant for predicting changes in the properties of Zr alloys in-service in fission reactors. [Display omitted] [ABSTRACT FROM AUTHOR]

• The role of Si content on irradiation embrittlement was clarified. • Solute atom clusters of highly neutron-irradiated Japanese pressurized water reactor surveillance test specimens were analyzed by atom probe tomography. • The cluster radius and number density decreased and increased, respectively, with increasing Si content, resulting in a constant volume fraction of SCs. • Increase in Si reduces the degree of irradiation embrittlement. This is consistent with the results of our previous study based on Bayesian statistical analysis. Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 × 1020 n/cm2 were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Mn, Ni, and Si atoms like a core-shell structure. In low-Cu bearing materials, Mn, Ni, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius (r) and number density (N d) decreased and increased with increasing nominal Si content, respectively. The shift in the reference temperature for nil-ductility transition (ΔRT NDT) showed a good correlation with the square root of volume fraction (V f) multiplied by ( V f × r). The negative relation between the nominal Si content and ΔRT NDT indicated that increasing of nominal Si content reduces the degree of embrittlement. [ABSTRACT FROM AUTHOR]

[Display omitted] • Modest variations in SA508 steel chemistry have a marked impact on EB weld residual stresses. • Significant stresses remain after an ASME-compliant post-weld heat treatment. • Single-pass EB welds are likely to have thickness-dependent residual stress distributions. • Phase transformation kinetics must be incorporated into models for stress development. Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels. [ABSTRACT FROM AUTHOR]

Most of reactor vessel internals (RVIs) have been constructed by austenitic stainless steels (ASSs) to support and protect core from high temperature coolant and radiation particles. However, if they are operated in the harsh environment for a long time, ASS materials may undergo significant changes in micro-structural features, macroscopic deformation and strengths due to age-related degradation mechanisms (ARDMs). In this study, user subroutines were developed based on material constitutive models of recently revised technical reports to reflect irradiation embrittlement, irradiation enhanced creep and void swelling behaviors. Benchmark analyses for a simple rod and detailed numerical analyses for a RVI sub-components assembly with welds were conducted via coupling the subroutines. With regard to the latter analyses, not only typical normal operating sequences but also three fuel cycles associated with specific core loading patterns were taken into account with a constant initial weld residual stress distribution. Adequacy of the subroutines was verified through analytical solutions generated from temperature and radiation damage dependent equations. Complex neutron dose, temperature and pressure profiles of the assembly were successfully determined from thermal and mechanical finite element analyses. Structural integrity assessment in terms of the ductility, embrittlement and irradiation assisted stress corrosion cracking susceptibility ratio after 40-years of operation led to acceptable in spite of the ARDMs. Contributions of factors and interaction effects were further quantified and discussed via design of experiment approach. • Five subroutines were developed to reflect irradiation induced degradations. • Uniform operating, fuel cycle loads and weld residual stress were incorporated. • The contribution of irradiation creep was the most influential to loss of ductility. • Stress relief reduced irradiation assisted stress corrosion cracking susceptibility. [ABSTRACT FROM AUTHOR]

• The grain-boundary P segregation for A533B RPV steels irradiated to high fluences in PWRs or a MTRs was analyzed using AES. • An increase in irradiation-induced grain-boundary P segregation was confirmed by a simulation based on a rate theory model as well as AES analysis. • No flux effect on grain-boundary P segregation was confirmed for A533B RPV steels with bulk P contents equivalent to that in RPV steels employed in Japanese nuclear power plants. • Intergranular embrittlement is not likely to occur for RPV steels with a bulk P content, which is higher than that of most U.S. A533B steels. Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.2 × 1020 n/cm2 (E > 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation, which was analyzed using Auger electron spectroscopy (AES) on intergranular facets, increased with increasing neutron fluence. A rate theory model based on four diffusion-reaction equations for substitutional P atoms, octahedral interstitial P atoms, vacancies, and self-interstitial atoms was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%, using parameters optimized by the present AES data. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.%, which is the maximum P concentration in RPV steels used in Japanese nuclear power plants intended for restart, was estimated to be less than 0.1 in monolayer coverage at 1 × 1020 n/cm2 (E > 1 MeV). A comparison of the PWR data with the MTR data, including that from the literature, showed that neutron flux had no effect upon grain-boundary P segregation for A533B steels. The relationships of the ductile-brittle transition temperature (DBTT) shifts to grain-boundary P segregation and to yield strength were also discussed. A linear relationship between the yield strength and the DBTT shift with a slope of 0.63 was obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs. [ABSTRACT FROM AUTHOR]

• Chemical compositions of solute atom clusters in surveillance materials from four nuclear power plants were compared. • The balance of Mn-Ni-Si in solute clusters in a BWR material is similar to that of the Γ 2 phase. • Relative Si concentrations in solute clusters in PWR materials are higher than the Si concentration in the G phase. This suggests the contribution of radiation-induced segregation of Si. • The "catalyst effect" of Cu precipitates on the gathering of Mn and Ni may assist in Mn-Ni-Si cluster formation. In a highly irradiated reactor pressure vessel (RPV), solute Mn, Ni, and Si (MNS) atoms gather to form nanometer-sized microstructures, generally called MNS clusters. MNS often gather with Cu-rich precipitates, which can form in RPVs following lower dose irradiation. In this study, surveillance specimens provided from four nuclear power plants in Japan were analyzed using three-dimensional atom probe tomography (APT), and the nature of the solute enrichment was carefully compared. When analyzing the chemical composition of each cluster, a clear negative correlation was found between Si and Cu in all materials, but conversely, Mn was likely present in clusters with a high Cu concentration. Moreover, in a boiling water reactor material with high Cu, the ratio of MNS was shown to be similar to that of the Γ 2 phase [Mn (Ni, Si) 2 ]. In pressurized water reactor materials with medium and low Cu, however, Ni and Si enrichment was demonstrated to be higher than the ratio of the expected intermetallic compounds; such as a Γ 2 phase and a G phase [Mn 6 Ni 16 Si 7 ]. Ni and Mn atoms, once enriched in a copper-rich region, may elute out, and form an intermetallic compound with Si atoms within the vicinity. Particularly in highly irradiated RPV materials, such a structure may tend to be decorated by irradiation-induced lattice defects with Si segregation. Image, graphical abstract [ABSTRACT FROM AUTHOR]

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. In this paper, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58Ni, and F82H doped with 1.4% 60Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness. [ABSTRACT FROM AUTHOR]

Nuclear reactor lifetimes may be limited by nano-scale Cu-Mn-Ni-Si precipitates (CRPs and MNSPs) that form under neutron irradiation (NI) of pressure vessel (RPV) steels, resulting in hardening and ductile to brittle transition temperature increases (embrittlement). Physical models of embrittlement must be based on characterization of precipitation as a function of the combination of metallurgical and irradiation variables. Here we focus on rapid and convenient charged particle irradiations (CPI) to both: a) compare to precipitates formed in NI; and, b) use CPI to efficiently explore precipitation in steels with a very wide range of compositions. Atom probe tomography (APT) comparisons show NI and CPI for similar bulk steel solute contents yield nearly the same precipitate compositions, albeit with some differences in their number density, size and volume fraction (f) dose (dpa) dependence. However, the overall precipitate evolutions are very similar. Advanced high Ni (>3 wt%) RPV steels, with superior unirradiated properties, were also investigated at high CPI dpa. For typical Mn contents, MNSPs have Ni 16 Mn 6 Si 7 or Ni 3 Mn 2 Si phase type compositions, with f values that are close to the equilibrium phase separated values. However, in steels with very low Mn and high Ni, Ni 2-3 Si silicide phase type precipitate compositions are observed; and when Ni is low, the precipitate compositions are close to the MnSi phase field. Low Mn significantly reduces, but does not eliminate, precipitation in high Ni steels. A comparison of dispersed barrier model predictions with measured hardening data suggests that the Ni-Si dominated precipitates are weaker dislocation obstacles than the G phase type MNSPs. • Nuclear reactor life times are limited by nano precipitates formed under irradiation. • Ion irradiations provide a rapid & affordable way to gain insight into precipitation. • Atom Probe data estimates hardening and embrittlement at lower service dose. • Advanced high Ni steels, with superior unirradiated properties, were investigated. • Simple thermodynamic models are able to predict volume fraction for alloys. [ABSTRACT FROM AUTHOR]

It is known that irradiation damage can cause a reduction in fracture toughness and an increase in the strength of austenitic stainless steels used for the internal structures of nuclear power plant reactors. A change in the mechanical properties of material, such as fracture toughness and strength, affects the fracture behavior of the components. Actual reactor internal structures, such as the top guides and core shrouds in BWR plants, are irradiated from the core in a single direction. This causes the attenuation of neutron flux, which results in the distribution of neutron fluence in the components, meaning that the distribution of mechanical properties is in the through-wall direction of the components. It is therefore possible for the fracture mode of the components to change in the through-wall direction. It would be valuable for more accurate assessment of the structural integrity of the nuclear components to make the effect of the mechanical properties distribution on the fracture mode clear. However, it is difficult to prepare large-scale test specimens with the mechanical properties distribution induced by uni-directional neutron irradiation. In this study, the mechanical properties distribution was simulated by applying the gradient of the cold working ratio using Type 316L stainless steel plate with an uneven thickness. The targeted range of the cold working ratio was designed by assuming BWR reactor internal structures, such as the top guides. The mechanical properties distribution was evaluated based on the results of hardness, tensile and fracture toughness tests. It was found that the macroscopic mechanical properties of irradiated stainless steels can be simulated by using the cold working and the macroscopic distribution of the mechanical properties of irradiated stainless steels can also be simulated by using the cold working ratio with a gradient. Furthermore, fracture tests were performed using test plates with the mechanical properties distribution assuming irradiated components with a crack. In some cases, the maximum loads obtained by the tests were almost the same value in spite of the different cold working ratio in the ligament. Based on the results of the fracture tests, it was considered that the dominant mechanical properties for the fracture load were those at the tip of the slit, not those in the ligament. [ABSTRACT FROM AUTHOR]