Irradiated Reactor Pressure Vessel Steels Research Articles

Cluster dynamics study on nano damage of RPV steels under proton irradiation at 290°C

Irradiation-induced defects such as dislocation loops, cavities or solute clusters and chemical composition segregation of reactor pressure vessel (RPV) steel are the root causes of irradiation embrittlement. Combining two nucleation mechanisms, namely, the uniform nucleation and non-uniform nucleation of solute clusters (such as Cu-rich phase), a cluster kinetic simulation was established based on the reaction rate theory, and the co-evolution of matrix damage and Cu-rich phase in low-copper RPV steel was simulated under irradiation. And the average size and number density of defective clusters and solute clusters were established with irradiation dose. Compared with the average size and number density of dislocation loops observed by transmission electron microscopy (TEM) of proton irradiated RPV steel at 290°C, the verification results show that the cluster dynamics model considering both the nucleation mechanism of interstitial dislocation loops and vacancy clusters can well simulate the irradiation damage behavior of materials.

Experimental and modeling study on irradiation effect of A508-Ⅲ steel

Irradiation can induce microscopic defects in reactor pressure vessel (RPV) steel, resulting in macroscopic hardening and embrittlement. This paper proposes a micro‑meso-macroscale calculation framework for the mechanical properties of irradiated RPV steel based on experimental and numerical investigations. The mean size and number density of Fe-ion-irradiation-induced dislocation loops were measured using transmission electron microscopy (TEM). These parameters were described by a power-function relationship with the irradiation dose at various temperatures and were used as input parameters for the crystal plasticity finite element model (CPFEM). Nano-indentation tests of A508-Ⅲ steel under different irradiation conditions were successfully simulated. In addition, microstructural characterization and molecular dynamics (MD) simulations revealed the phenomenon of dislocation pinning and the sweeping of irradiation defects during dislocation migration, implying that the dislocation-loop interactions could be responsible for the formation of dislocation channels and curved dislocation lines. Furthermore, increase in yield stress and the ductile-brittle transition temperature (DBTT) at various temperatures were investigated. Results indicated that yield stress, DBTT, and the degree of irradiation hardening and embrittlement increased with the irradiation dose and decreased with irradiation temperature. Finally, the adjusted reference temperature (ART), a significant design parameter for nuclear reactors, was calculated and evaluated using the multiscale framework. The framework can serve as a reference text for extending the life of nuclear reactors.

A multiscale study on the microstructure and hardening models of the irradiation defects on reactor pressure vessel steels: Modelling and experiment

Irradiation defects are known to impede the dislocation motion in irradiated reactor pressure vessel (RPV) steels. However, consistent experimental studies of the respective irradiation defects and relevant theoretical studies of the interaction mechanism and hardening model are still lacking. In this study, the microstructure and irradiation-induced hardening of RPV steels are investigated using transmission electron microscopy (TEM)/energy-dispersive X-ray spectrometry (EDS) and molecular dynamics (MD) simulations. In A508-Ⅲ steel, 9R, twinned fcc, and other typical Cu-rich cluster structures are observed using TEM after the irradiation experiment. Based on the experimental results, MD simulations are performed to study the interactions between the dislocations and Cu clusters or voids. A mechanism diagram of the interaction mechanism coupling the effects of temperature and defect size is shown for both the Cu clusters and voids. A relevant hardening model is proposed based on the MD data, which is further discussed in relation to the macroscopic mechanical properties. Furthermore, the irradiation-induced hardening of the RPV steel is successfully predicted, which further demonstrates the reliability of the proposed unified hardening model.

Development of pressure-temperature limit curves considering unified constraint for reactor pressure vessel

In this work, the pressure-temperature (P-T) limit curves considering unified constraint for a typical large-sized reactor pressure vessel (RPV) with axial surface cracks have been investigated and developed based on the KJ-Ad* two-parameter and the Master Curve (MC) method. The ASME fracture toughness KIc curve and the MC with and without considering unified constraint for irradiated RPV steel were used in the constructions of the P-T limit curves. The effects of crack sizes (constraint levels) and fracture toughness curves on the P-T limit curves have been analyzed for the RPV under heatup and cooldown conditions. The results further show that the P-T limit curves based on traditional ASME KIc curve has the highest degree of conservatism and the use of the MC method can significantly increase the safety margin of P-T limit curves. With increasing crack sizes (increasing constraint levels), the safety region in P-T limit curves reduces, which may lead to the decrease of operating flexibility for nuclear power plants. The accuracy of the P-T limit curves can be improved by considering unified constraint. For smaller size cracks with lower constraint in the RPV, the consideration of constraint can reduce conservatism of the P-T limit curves, and for larger size cracks with higher constraint, it can avoid the non-conservatism. It is recommended to obtain more accurate P-T limit curves based on the MC with considering unified constraint and finite element analysis for RPVs under various operating conditions.

The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels

Embrittlement of light water reactor pressure vessel (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. The steels were irradiated to a fluence of ∼ 1.38×1020 n/cm2 at ∼ 292 °C in the UCSB ATR-2 experiment. In nominally Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promote precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.

Atom probe characterisation of segregation driven Cu and Mn–Ni–Si co-precipitation in neutron irradiated T91 tempered-martensitic steel

The T91 grade and similar 9Cr tempered-martensitic steels (also known as ferritic-martensitic) are leading candidate structural alloys for fast fission nuclear and fusion power reactors. At low temperatures (300–400 °C) neutron irradiation hardens and embrittles these steels, therefore it is important to investigate the origin of this mode of life limiting property degradation. T91 steel specimens were separately neutron irradiated to 2.14 dpa at 327 °C and 8.82 dpa at 377 °C in the Idaho National Laboratory Advanced Test Reactor. Atom probe tomography was used to investigate the segregation driven formation of Mn–Ni–Si-rich (MNSPs) and Cu-rich (CRP) co-precipitates. The precipitates increase in size and, slightly, in volume fraction at the higher irradiation temperature and dose, while their corresponding compositions were very similar, falling near the Si(Mn,Ni) phase field in the Mn–Ni–Si projection of the Fe-based quaternary phase diagram. While the structure of the precipitates has not been characterised, this composition range is distinctly different than that of the typically cited G-phase. The precipitates are composed of CRP with MNSP appendages. Such features are often observed in neutron irradiated reactor pressure vessel (RPV) steels. However, the Si, Ni, Mn, P and Cu solutes concentrations are lower in the T91 than in typical RPV steels. Thus, in T91 precipitation primarily takes place in solute segregated regions of line and loop dislocations. These results are consistent with the model for radiation induced segregation driven precipitation of MNSPs proposed by Ke et al. Cr-rich alpha prime (α’) phase formation was not observed.

Radiation Damage of Reactor Pressure Vessel Steels Studied by Positron Annihilation Spectroscopy—A Review

Safe and long term operation of nuclear reactors is one of the most discussed challenges in nuclear power engineering. The radiation degradation of nuclear design materials limits the operational lifetime of all nuclear installations or at least decreases its safety margin. This paper is a review of experimental PALS/PLEPS studies of different nuclear reactor pressure vessel (RPV) steels investigated over last twenty years in our laboratories. Positron annihilation lifetime spectroscopy (PALS) via its characteristics (lifetimes of positrons and their intensities) provides useful information about type and density of radiation induced defects. The new results obtained on neutron-irradiated and hydrogen ions implanted German steels were compared to those from the previous studies with the aim to evaluate different processes (neutron flux/fluence, thermal treatment or content of selected alloying elements) to the microstructural changes of neutron irradiated RPV steel specimens. The possibility of substitution of neutron treatment (connected to new defects creation) via hydrogen ions implantation was analyzed as well. The same materials exposed to comparable displacement damage (dpa) introduced by neutrons and accelerated hydrogen ions shown that in the results interpretation the effect of hydrogen as a vacancy-stabilizing gas must be considered, too. This approach could contribute to future studies of nuclear fission/fusion design steels treated by high levels of neutron irradiation.

The mechanistic implications of the high temperature, long time thermal stability of nanoscale Mn-Ni-Si precipitates in irradiated reactor pressure vessel steels

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 (rc). 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 rc for PIA at 425 °C up to 52 weeks. Atom probe tomography and energy dispersive x-ray spectroscopy showed MNSPs with r < rc dissolved, while those with r > rc slightly coarsened, consistent with thermodynamic predictions.

Complex study of grain boundary segregation in long-term irradiated reactor pressure vessel steels

Grain boundary (GB) embrittlement of reactor pressure vessel steel occurs at the long-term operation through the radiation-enhanced segregation of impurities. All the known techniques of studying the GB segregation have their advantages and disadvantages, and complete information can be obtained only through a comprehensive analysis by different methods.In this paper the level of GB segregation in VVER-1000 RPV steels with high nickel content was assessed by combined methods of Auger electron spectroscopy (by the P monolayer coverage and atomic concentration of other elements); atom probe tomography (by the Gibbsian interfacial excess) and fractographic analysis (by the maximum fraction of brittle intergranular fracture).The results obtained by different methods are in good agreement and show that phosphorus GB segregation increases with the increase of fast neutron fluence, GB concentration of Ni and Mn, the bulk concentration of Ni and Mn, and correlates with the fraction of brittle intergranular fracture.

Understanding the importance of the energetics of Mn, Ni, Cu, Si and vacancy triplet clusters in bcc Fe

Numerous experimental studies have found the presence of (Cu)-Ni-Mn-Si clusters in neutron irradiated reactor pressure vessel steels, prompting concerns that these clusters could lead to larger than expected increases in hardening, especially at high fluences late in life. The mechanics governing clustering for the Fe-Mn-Ni-Si system are not well-known; state-of-the-art methods use kinetic Monte Carlo (KMC) parameterized by density functional theory (DFT) and thermodynamic data to model the time evolution of clusters. However, DFT-based KMC studies have so far been limited to only pairwise interactions due to lack of DFT data. Here, we explicitly calculate the binding energy of triplet clusters of Mn, Ni, Cu, Si, and vacancies in bcc Fe using DFT to show that the presence of vacancies, Si, or Cu stabilizes cluster formation, as clusters containing exclusively Mn and/or Ni are not energetically stable in the absence of interstitials. We further identify which clusters may be reasonably approximated as a sum of pairwise interactions and which instead require an explicit treatment of the three-body interaction, showing that the three-body term can account for as much as 0.3 eV, especially for clusters containing vacancies.

On the Elevated Temperature Thermal Stability of Nanoscale Mn-Ni-Si Precipitates Formed at Lower Temperature in Highly Irradiated Reactor Pressure Vessel Steels

Atom probe tomography (APT) and scanning transmission electron microscopy (STEM) techniques were used to probe the long-time thermal stability of nm-scale Mn-Ni-Si precipitates (MNSPs) formed in intermediate and high Ni reactor pressure vessel steels under high fluence neutron irradiation at ≈320 °C. Post irradiation annealing (PIA) at 425 °C for up to 57 weeks was used to determine if the MNSPs are: (a) non-equilibrium solute clusters formed and sustained by radiation induced segregation (RIS); or, (b) equilibrium G or Γ2 phases, that precipitate at accelerated rates due to radiation enhanced diffusion (RED). Note the latter is consistent with both thermodynamic models and x-ray diffraction (XRD) measurements. Both the experimental and an independently calibrated cluster dynamics (CD) model results show that the stability of the MNSPs is very sensitive to the alloy Ni and, to a lesser extent, Mn content. Thus, a small fraction of the largest MNSPs in the high Ni steel persist, and begin to coarsen at long times. These results suggest that the MNSPs remain a stable phase, even at 105 °C higher than they formed at, thus are most certainly equilibrium phases at much lower service relevant temperatures of ≈290 °C.

Evolution of microstructures and hardening property of initial irradiated, post-irradiation annealed and re-irradiated Chinese-type low-Cu reactor pressure vessel steel

Investigation of microstructures and hardening property of initial irradiated, post-irradiation annealed, and re-irradiated Chinese-type low-Cu (0.01 wt%) reactor pressure vessel (RPV) steel were conducted by slow positron beam (SPB), TEM, and nanoindentation. Results of the SPB measurements indicated that a large number density of open volume defects such as vacancies, vacancy clusters, vacancy-solute/H complexes, and dislocation loops were introduced in both initial-irradiated and re-irradiated specimens. The TEM results indicated that interstitial-type dislocation loops with a number density of ∼1022 m−3 were formed and the mean size (∼3 nm) of the dislocation loops was almost unchanged in both initial-irradiated and re-irradiated specimens. The open volume defects such as vacancy-type defects and dislocation loops were almost annealed out and some stable defects still existed in the post-irradiation annealed specimen. The nanoindentation results identified that the obvious hardening phenomena were found in the initial irradiated, post-irradiation annealed and re-irradiated specimens. The quantitative analysis suggested that irradiation hardening of Chinese-type low-Cu RPV steel were attributed to matrix defects including large-size vacancy clusters, vacancy-solute/H complexes, and dislocation loops produced during initial irradiation and re-irradiation. On the other hand, the existed stable defects such as solute-atom (Mn, Ni, and Si) complexes/precipitates, which cannot be distinguished by SPB or TEM, might be responsible for the rest fraction of irradiation hardening of highly initial irradiated and re-irradiated Chinese-type low-Cu RPV steel.

Multiscale modeling of crystal plasticity in Reactor Pressure Vessel steels: Prediction of irradiation hardening

The plastic behavior of irradiated Reactor Pressure Vessel (RPV) steels is described by constitutive equations capturing the temperature and strain rate sensitivities. The flow stress is decomposed into its fundamental components associated with the microstructure features peculiar to RPV steels, such as carbides, dislocation network and deformation confinement inside grains. Dislocations are assumed to move on the {110} and {112} crystallographic planes and a simplified interaction matrix is proposed. The predicted yield stress is obtained without adjustable parameters and found in close agreement with a large number of experimental results over a large temperature range. Finally, the contribution of radiation defects is accounted for using atomistic and dislocation dynamics results. The effect of solute cluster is analyzed in details in terms of the cluster size and density and strength. Results are discussed and compared with an experimental database on neutron-irradiated RPV steels.

Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels

The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature (To) was obtained with a relatively small number of specimens. Comparing the fracture toughness and To values determined using other larger specimens with those determined using the Mini-C(T) specimens, To values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41 J transition temperature (T41J) and the To values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.

Infrastructure development for radioactive materials at the NSLS-II

The X-ray Powder Diffraction (XPD) Beamline at the National Synchrotron Light Source-II is a multipurpose instrument designed for high-resolution, high-energy X-ray scattering techniques. In this article, the capabilities, opportunities and recent developments in the characterization of radioactive materials at XPD are described. The overarching goal of this work is to provide researchers access to advanced synchrotron techniques suited to the structural characterization of materials for advanced nuclear energy systems. XPD is a new beamline providing high photon flux for X-ray Diffraction, Pair Distribution Function analysis and Small Angle X-ray Scattering. The infrastructure and software described here extend the existing capabilities at XPD to accommodate radioactive materials. Such techniques will contribute crucial information to the characterization and quantification of advanced materials for nuclear energy applications. We describe the automated radioactive sample collection capabilities and recent X-ray Diffraction and Small Angle X-ray Scattering results from neutron irradiated reactor pressure vessel steels and oxide dispersion strengthened steels.

The fatigue behavior of irradiated Reactor Pressure Vessel steel

Abstract Fatigue specimens of A508-3 steel were irradiated in the swimming-pool test reactor in China Institute of Atomic Energy, the fluence was 3 × 10 19 n/cm 2 at 300 °C, then low-cycle fatigue tests were carried out at ambient temperature, with the fatigue strain range is 0.32–1.8%. The results indicate that, irradiated A508-3 specimens exhibit cyclic softening and instability behavior during the test, and the cyclic softening rate increased with strain range increased; fatigue life decreased from 1.7 × 10 5 to about 5 × 10 2 , as the strain range increased from 0.32% to 1.8%, the fatigue life of A508-3 steel increased after the neutron irradiation; fatigue fracture initiated at the surface of specimen, and more individual cracks formed on the specimens of higher strain range compared with the specimens of lower strain range.

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

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 60years 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.

Simulation of magnetic hysteresis loops and magnetic Barkhausen noise of α-iron containing nonmagnetic particles

The magnetic hysteresis loops and Barkhausen noise of a single α-iron with nonmagnetic particles are simulated to investigate into the magnetic hardening due to Cu-rich precipitates in irradiated reactor pressure vessel (RPV) steels. Phase field method basing Landau-Lifshitz-Gilbert (LLG) equation is used for this simulation. The results show that the presence of the nonmagnetic particle could result in magnetic hardening by making the nucleation of reversed domains difficult. The coercive field is found to increase, while the intensity of Barkhausen noise voltage is decreased when the nonmagnetic particle is introduced. Simulations demonstrate the impact of nucleation field of reversed domains on the magnetization reversal behavior and the magnetic properties.

Warm PreStress effect on highly irradiated reactor pressure vessel steel

This study investigates the Warm Prestress (WPS) effect on 16MND5 (A508 Cl3) RPV steel, irradiated up to a fluence of 13·1023n.m-2 (E>1MeV) at a temperature of 288°C, corresponding to more than 60years of operations in a French Pressurized Water Reactor (PWR). Mechanical properties, including tensile tests with different strain rates and tension–compression tests on notched specimens, have been characterized at unirradiated and irradiated states and used to calibrate constitutive equations to describe the mechanical behavior as a function of temperature and fluence. Irradiation embrittlement has been determined based on Charpy V-notch impact tests and isothermal quasi-static toughness tests. Assessment of WPS effect has been done through various types of thermomechanical loadings performed on CT(0.5T) specimens. All tests have confirmed the non-failure during the thermo-mechanical transients. Experimental data obtained in this study have been compared to both engineering-based models and to a local approach (Beremin) model for cleavage fracture. It is shown that both types of modeling give good predictions for the effective toughness after warm prestressing.