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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]

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]

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]

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 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]

• 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]