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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]

The development of prediction model on irradiation embitterment (PMIE) of reactor pressure vessel (RPV) is an important method for nuclear reactor long term operation. Based on the physical mechanism of RPV irradiation embrittlement, a preliminary model is determined and the critical threshold of Cu content of 0.072% is obtained according to this preliminary model. Then a prediction model named PMIE-2020 for low Cu RPV steels is developed. At last the residual, standard deviation and predicted values and test values distribution analysis are given. Simultaneously, a comparison between PMIE-2020 and other prediction model and irradiation data is provided. Results indicate that the predicted results of PMIE-2020 has no tendency with influence factors such as neutron fluence, flux, irradiation temperature, chemical elements Cu, P, Mn, Ni, Si. The residual standard deviation is 10.76 °C, which is lower than present prediction model. The distribution between predicted values of PMIE-2020 and test values are located the area near the 45° line. These results prove that the PMIE-2020 have high accuracy on irradiation embrittlement prediction., Competing Interests: The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (© 2023 The Authors.)

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

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]

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]

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]

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]

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 fractographic study of steels of nuclear reactor vessels after tests for crack resistance, two major sources of brittle destruction are detected: nonmetallic impurities and intergranular or subgranular boundaries. X-ray spectral microanalysis reveals that these are manganese sulfides and silicon oxides. The probable size of nonmetallic impurities (∼1 μm) which can initiate brittle destruction is detected. The interrelation between critical stress intensity coefficient K and distance from 'leader' to fatigue crack point CID (cleavage initiation distance) is considered. The influence of irradiation conditions on crack resistance of steels is demonstrated. [ABSTRACT FROM AUTHOR]

Highlights: [•] Ductile to brittle toughness predictions compared to experimental results. [•] Improved criterion for particle failure including measured particle distributions. [•] Predict experimentally measured locations of cleavage initiators. [•] Predicts toughness in transition region and shift with irradiation. [•] All results are obtained without changes in model parameters. [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]

Abstract: We present a microstructural study of a surveillance test specimen from a reactor pressure vessel steel of a Russian-type nuclear reactor after neutron irradiation, post-irradiation annealing and re-irradiation, using atom probe tomography (APT) and positron annihilation spectroscopy (PAS). The APT results showed the formation of Cu-rich solute nano-clusters (CRCs) during the initial irradiation and their subsequent coarsening during annealing. After re-irradiation, new CRCs have been observed. The irradiation-hardening almost recovered during annealing. However, by re-irradiation, hardening comparable to that by the initial irradiation was observed. The hardening due to the CRCs formed during the initial irradiation, estimated using the Russell–Brown model, was almost the same as that observed. However, the estimated hardening after the re-irradiation was about half of the measured one. The other hardening is attributed to the newly formed MDs by the re-irradiation, which was evidenced by PAS. [Copyright &y& Elsevier]

Reactor internals are subjected to three principal operation effects: neutron and gamma irradiation, static and dynamic mechanical stresses and coolant chemistry. In this study, we investigate the effect of these mechanisms on WWER 440 reactor vessel internals and present lifetime analysis, in order to extend the operational lifetime of reactor vessel internals. [ABSTRACT FROM AUTHOR]

We discuss irradiation tests performed at the Budapest research reactor operating with 10 MW using a new irradiation rig. Some examples of the use of the irradiation rigs are presented: results of reactor pressure vessel cladding, 9% Cr steels (Euroferr), application of Master Curve to irradiated reactor pressure vessel steels, as well as irradiation tests of Ti-alloys and tungsten. [ABSTRACT FROM AUTHOR]

Abstract: The purpose of this study is to investigate the internal friction (IF) behaviour of thermally aged Fe–Cu alloys that are model specimens simulating for the irradiation embrittlement of nuclear reactor pressure vessel (RPV) steels. Plate-shaped specimens of Fe–1wt.% Cu alloy were quenched from 1123K and thermally aged at 773K for 103 min. The IF of the specimens was measured without magnetic field using the free resonant bending vibration method. Magnetic hysteresis loops and conductivity were also measured. The IF value, hysteresis loss, and conductivity increased under thermal aging, which suggests that the IF characteristics are closely related to the change of electromagnetic eddy-current loss due to the depletion of Cu solute atoms. This study demonstrates the possibility of the application of IF measurement as one of the tools for nondestructive characterization of degradation processes of irradiated RPV steels. [Copyright &y& Elsevier]

• 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]

Abstract: The effects of intergranular P segregation and hardening on the ductile-to-brittle transition temperature (DBTT) in several neutron-irradiated reactor pressure vessel steels with different bulk contents of P and Cu have been investigated using a scanning Auger microbe, a local electrode atom probe and positron annihilation spectroscopy. Increasing the neutron fluence at 563K promotes intergranular P segregation, particularly in steels with high levels of P. The content of P (<570ppm) more significantly affects irradiation-hardening than that of Cu (<0.17wt.%) due to distinct formation of P-rich precipitates arising from the stabilization of vacancies. Analyzing the correlations between P segregation, hardening, fraction of intergranular fracture and DBTT, it is found neutron irradiation mitigates the embrittling effect of segregated P, and therefore the hardening more strongly affects the DBTT shift than the P segregation, with the exception of highly P-doped steel irradiated to high neutron fluence. [Copyright &y& Elsevier]

Abstract: Irradiation embrittlement is a limiting condition for the long-term safety of a nuclear reactor pressure vessel (RPV). The first PWR in Korea is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. The RPV was designed by an old construction code and its beltline circumferential welds have suffered from an irradiation shift problem like other Linde 80 welds. The master curve method is considered as the most promising tool to characterize irradiated structural steels by using a fracture mechanics basis. In order to implement the master curve method for the assessment of an irradiation embrittlement of old power reactors for a continued long-term operation, three practical issues were emphasized in this investigation, which are the specimen geometry effects on the master curve results, the specimen reconstitution techniques in an old existing surveillance program, and the index temperatures of an irradiation embrittlement when compared with the conventional Charpy data. [Copyright &y& Elsevier]

Abstract: The idea of an engineering version of the local approach to brittle fracture as well as the possibility for it to be employmed to estimate irradiation embrittlement of pressure vessel (PV) steels is considered. Unlike the conventional temperature-shift-based methodology, the approach presented utilises the concept of stability of the ductile state of the metal. A new characteristic “parameter of mechanical stability”P ms is proposed. This characteristic enables quantification of the level of stability of the ductile state of an irradiated PV steel in a specimen or in a reactor vessel with a crack at the specified level of loading. Within the framework of the proposed concept, a value for end-of-life fluence for a reactor PV is predicted by the condition of exhaustion of stability of a ductile state of a steel ahead of the crack (P ms=1). [Copyright &y& Elsevier]

Abstract: The correlation between recent PAS results and the outcomes from mechanical testing of RPV model alloys are presented, here significant changes due to different chemical composition and different irradiation levels are observed. The influence of alloying elements to the microstructure degradation process following irradiation was identified by analysis of the mean-lifetime parameter, since an interesting interdependency of this parameter with hardness was observed. [Copyright &y& Elsevier]

Abstract: Nuclear pressure vessel steels are subjected to irradiation embrittlement which is monitored using Charpy tests. Reference index temperatures, such as the temperature for which the mean Charpy rupture energy is equal to 56J (T 56J), are used as embrittlement indicators. The safety integrity evaluation is performed assuming that the shift of the nil-ductility reference temperature RT NDT due to irradiation is equal to the shift of T 56J. A material model integrating a description of viscoplasticity, ductile damage and cleavage brittle fracture is used to simulate both the Charpy test and the fracture toughness test (CT geometry). The model is calibrated on the Charpy data obtained on an unirradiated A508 Cl.3 steel. It is then applied to irradiated materials assuming that irradiation affects solely hardening. Comparison with Charpy energy data for different amounts of irradiation shows that irradiation possibly also affects brittle fracture. The model is then applied to predict the fracture toughness shifts for different levels of irradiation. [Copyright &y& Elsevier]

Mechanical properties of WWER-440 RPV weld joints have been studied with account of different states of the material: baseline (unirradiated), irradiated up to the average fast neutron fluence of $\backslash 6.5\; cdot\; 10^19$ n/$\backslash cm^2$, irradiated and eventually annealed, re-irradiated with the accumulated fast neutron fluence $\backslash 1.3\; cdot\; 10^20$ n/$\backslash cm^2$. Tensile, impact fracture, and fracture toughness tests were performed for each state of the material with the use of Charpy specimens (standard, reconstituted, and pre-cracked). [ABSTRACT FROM AUTHOR]

Radiation embrittlement and aging mechanisms for NPP reactor pressure vessels and vessel internals have been studied within NPP Plant Life Management activities for the evaluation, prediction, and monitoring of the critical components' service life. The main achievements of the SAFELIFE project, integrating various networks on PLIM issues, are given. Results of neutron embrittlement of model alloys are presented, and surveillance and research data on WWER reactor pressure vessel and other steels are analyzed. Projects for the development of destructive and non-destructive testing of irradiated materials are outlined. [ABSTRACT FROM AUTHOR]

EUROATOM research programs aimed at studying aging mechanisms and remedial procedures for structural materials of nuclear reactor components have been analyzed. Within this framework, projects are carried out focusing on the development of non-destructive techniques applied to thermal aging, neutron embrittlement monitoring, improved surveillance for WWER-440 reactors, dosimetry, and the use of various fracture mechanisms for NPP integrity assessment. Among major achievements are the ATHENA project activities on re-embrittlement model validation after annealing and the effect of chemical composition on the embrittlement rate in RPV steels. A brief description is given of the main results of current EUROATOM projects, as well as the goals of the next stage of research within the SAFELIFE network on the NPP Plant Life Management. [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 growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed. [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]

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]

• 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]