G. Robert Odette

Significant embrittlement in reactor pressure vessel (RPV) steels can be caused by the formation of nanometer-scale MneNieSi precipitates (MNSPs) and annealing is a promising technique for reducing embrittlement. To achieve better understanding of the evolution of these precipitates at the atomic scale, a kinetic lattice Monte Carlo (KLMC) model, parameterized using CALPHAD and recent atom probe tomography (APT) data, is used to simulate post-irradiation annealing of MNSPs. The model predicts MNSP volume fractions, number densities and sizes that agree well with the experimental observations. The model also predicts that the initial structure of the precipitates may be B2 bcc phases with one sublattice occupied by Ni and the other sublattice occupied by Mn and Si, as well as shows a modest temperature dependence of the MNSP composition. The results show that the simple approach can be used to model MNSP evolution and supports that these precipitates are stable thermodynamic phases.

This report is the product of a nearly decade long effort, beginning with a NSUF proposal for an ATR pressure vessel steel irradiation The overarching goal of the UCSB ATR-2 irradiation experiment was to provide a foundational database for developing new physical models to predict embrittlement of RPV steels at high fluence pertinent to extended life operation. A total of 1625 specimens of various types, composed of 172 alloys, were neutron irradiated over a range of flux (, fluence (t and temperature (Ti). Here we focus on the highest t capsules irradiated at 290°C. Most notably, embrittlement as manifested by irradiation hardening (Δσy) and corresponding 41 J temperature shifts (ΔTc), are systematically and significantly under-predicted by current regulatory models, including both the Eason-Odette- Nanstad-Yamamoto (EONY) model and the American Society for Testing and Materials (ASTM) E900 Standard Practice. Note, the E900 model greatly over-predicts embrittlement in very high 1.6% Ni steels, while under-predicting hardening at < 1% Ni pertinent to the US RPV fleet. Microstructural characterization by atom probe tomography, small angle neutron scattering and x-ray scattering and diffraction shows, that the high fluence hardening includes a significant contribution from so-called late blooming phases (LBP), in the form of nano-scale Mn-Ni-Si G and 2-phase precipitates (MNSP). The MNSPs emerge at high fluence in Cu bearing and Cu free steels, as well as at intermediate and high Ni contents. The ATR-2 results were combined with various UCSB and other databases to develop both detailed and reduced order thermokinetic physically-microstructurally based models of Δσy and ΔTc that are in good semi to fully quantitative agreement on effects of 12 the combination of all significant embrittlement variables (Cu, Nu, MN, Si, P, , t, and Ti). However, here we use the insight gained from both the models and various databases to develop a simple empirical method to predict Tc at high extended 80 y life t for low  vessel service conditions. The method relies on three critical elements:  Derivation of a new, accurate composition dependent chemistry factor (CF) for y at the high ATR-2 t.  Demonstration that the t dependence of y (and Tc) between ≈ 3 and 14x1023 n/m2, is approximately linear, allowing interpolation between the ATR-2 data and information at lower t based on surveillance data or model predictions of EONY or E900.  Discovery and confirmation that the effects of, that are strong at low t, decrease at high t. The new method is compared to the existing regulatory models and used to predict the Tc and the irradiated Tc at a t = 1024 n/m2 for the US RPV fleet, using composition and unirradiated Tc data in the US PREDB. The margin that should be applied to the new ATR-2 based predictions is estimated ±20°C. In contrast to the EONY and ATM E900 treatments, the initial microstructural and mechanical property PIE are in remarkable quantitative agreement with previously developed physically based models, showing the predicted formation of late blooming Mn-Ni-Si phases and supporting the treatment of flux effects. The combined database will quantify the “when (flux and fluence), where (alloy compositions) and how much (hardening-shift contributions) of late blooming phases that are not treated in current 13 regulatory models. In addition, UCSB ATR-2 will provide: (a) a test of the validity of the Master Curve fracture toughness assessment method in highly embrittled steels; (b) a large database on post irradiation annealing as a foundation for developing embrittlement remediation methods; and, (c) support for the development of a new class of high Ni advanced RPV alloys. The ATR-2 PIE program is also developing new mechanical property and microstructural characterization tools, such as an automated shear punch test tool and small angle x-ray scattering, respectively. Finally, the large ATR-2 experimental effort is being closely coupled to fundamental models of embrittlement in way that experiments truly inform advanced modeling and uses modeling to guide efficient and targeted experiments.

Changes in the chemistry of internal interfaces, particularly grain boundaries, are known to affect the macroscopic properties of a wide range of material systems. Solute segregation to grain boundaries is dependent on, amongst other factors, the physical structure of the grain boundary. We demonstrate how complementary use of transmission Kikuchi diffraction (TKD) and atom probe tomography (APT) can provide a more holistic characterisation of grain boundaries in a variety of materials. Structural information is reported from TKD data for a model steel, a titanium alloy, and a multicrystalline silicon sample. Complementary APT analyses are used to determine the segregation behaviour to these interfaces. A novel specimen preparation protocol allows for the grain boundary to be positioned more reliably within the apex of an APT specimen. Meanwhile, a method that allows a grain boundary’s five macroscopic degrees of freedom to be determined from TKD data alone is also proposed.

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.

Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

A series of Fe-Cr alloys, with 3 at.% - 18 at.% Cr, was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Labs (INL), USA, up to a dose of ~6.7 dpa at a temperature of ~456 ◦C. Transmission electron microscopy (TEM) samples were extracted using a focused ion beam (FIB) instrument, and the resulting microstructural defects, such as voids, dislocation loops, network dislocations, Cr rich precipitates, etc., were characterized using a TEM. It was found that the size and number density of these defects varied widely over the different alloys with varying Cr content. As expected, there were no Cr rich precipitates in samples with Cr up to 9 %, and they started appearing only in samples with 12 at.% Cr and above. The particle size decreased from about 15 nm at 12 % Cr to 8 nm at 18 % Cr, while the number density increased from ~7e20 /m3 to 6e22 /m3 for the same Cr contents. Grain boundary segregation of Cr, along with a precipitate free zone, was observed in the cases where a boundary was present in the sample. Large voids (>1–2 nm) were almost invisible in the Fe-3at.% Cr sample, while the average void size remained almost constant between 15 and 18 nm for samples with 6–15 at.% Cr and increased slightly at 18 at.% Cr to ~22 nm. Fe-3 %Cr showed a high density of small voids(<2 nm), estimated to be about 1e23/m3. The number density of large voids increased from ~0 at 3 % Cr to a peak of ~6.4e20 /m3 at 12 % Cr, then decreased to about 1.1e20 /m3 at 18 % Cr. The dislocation loops, which appeared in linear arrays in the Fe-9 %Cr sample, were analysed in detail using both invisibility criteria and image simulations using the Oxford University TEMACI software, and it was found that they are most likely to be ½ 〈111〉 type loops on {100} planes. These loops seem shaped like sections of helices in some places and are most likely formed by loops near screw dislocations climbing into a helix and then collapsing into a loop array. Similar dislocation analysis was performed in the other samples as well wherever feasible, and it was found that there is a mixture of ½ 〈111〉 and [100] dislocation loops.

High chromium ferritic-martensitic (F-M) steels are one of the promising structural material classes for future nuclear power plants. These steels are designed to combine corrosion resistance, conferred by chromium, with low swelling, high resistance to irradiation damage as well as to retain adequate toughness and elevated-temperature strength during service [1]. However, the long-term use of these steels in intense neutron irradiation environments requires reliable predictions of the evolution of their microstructures and mechanical properties. Binary Fe-Cr alloys constitute a model system for high Cr ferritic/martensitic steels and have therefore generated lot of interest by allowing the systematic study on irradiation induced microstructural changes. In the present study, microstructural changes in neutron irradiated Fe-Cr binary alloys are investigated using atom probe tomography (APT). A series of six Fe-Cr alloys of nominal compositions 3, 6, 9, 12, 15, and 18 at.%Cr were irradiated at a neutron fluence (E>1 MeV) of 1.1 x 1021 n/cm2 at 563 ± 15K and to a damage level of 1.82 displacements per atom (dpa). Solute distributions revealed a' precipitation for alloys containing more than 9at.%Cr (Figure 1). Both the Cr concentration dependence of a' precipitation and the measured matrix compositions are in agreement with the recently published Fe-Cr phase diagrams [2]. An irradiation-accelerated precipitation process is strongly suggested for a' precipitation. Along with homogenously distributed Cr-enriched clusters of the a' phase, few clusters involving Si, P, Ni, and Cr, are observed in the matrix [3]. For Fe-6, 9, 12 at.%Cr, Si and Cr are found segregated to dislocation loops and information pertaining to number density, size, and habit plane were analyzed for Fe-6at.%Cr alloy[4]. Grain boundary chemistry for Fe-Cr alloys are quantitatively compared between the as-received and the neutron irradiated alloys. Zones depleted of a' clusters and Si are found at the interfaces of carbide and nitride precipitates and along grain boundaries in the vicinity of these precipitates. To study stability of clusters and observed features in irradiated samples, annealing is carried out at high temperatures. The results are discussed in the context of equilibrium segregation, radiation-enhanced diffusion, and/or radiation induced segregation.

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 × 1020 and ~1.1 × 1021 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 × 1020 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 × 1021 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.

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 mechanical behaviour of Fe6%Cr in the un-irradiated, self-ion irradiated and neutron irradiated conditions was measured and compared. Irradiations were performed to the same dose and at the same temperature but to very different damage rates for both methods. The materials were tested using nanoindentation and micromechanical testing, and compared with microstructural observations from Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) reported elsewhere. Irradiated and un-irradiated micro-cantilevers with a wide range of dimensions were used to study the interrelationships between irradiation hardening and size effects in small-scale plasticity. TEM and APT results identified that the dislocation loop densities were ∼2.9 × 1022m−3 for the neutron irradiated material and only 1.4 × 1022m−3 for the ion irradiated material. Cr segregation to loops was only found for the neutron-irradiated material. The nanoindentation hardness increase due to neutron irradiation was 3 GPa and that due to ion irradiation 1 GPa. The differences between the effects of the two irradiation types are discussed, taking into account inconsistencies in damage calculations, and the differences in PKA spectra, dose rate and transmutation products for the two irradiation types.

Microstructural changes in a Fe–15 at.%Cr model alloy neutron irradiated to 1.82 dpa at 290 °C were characterized by atom probe tomography. Homogenously distributed a' precipitates as well as fewer clusters containing Si, P, Ni, and Cr, were observed in the matrix. Grain boundary analyses before and after irradiation revealed segregation of Cr, with W-shape concentration profiles developing in the vicinity of grain boundary carbide and nitride particles. After irradiation, impurities such as C, Si and P were segregated to the grain boundaries. Zones depleted of a' clusters, and Si were found at the interfaces of carbide and nitride precipitates and along grain boundaries in the vicinity of these precipitates.

The microstructural and chemical changes in a Fe–6 at.%Cr binary model alloy neutron irradiated to 1.82 dpa at 290 °C were investigated using atom probe tomography. After irradiation, Si and Cr are found segregated to dislocation loops, which were analyzed in terms of number density, size, and habit plane. Grain boundary chemistry was quantitatively compared between the as-received and the neutron irradiated alloys. The results are discussed in the context of equilibrium segregation, radiation-enhanced diffusion, and/or radiation induced segregation.

Ferritic-martensitic steels are leading candidates for many nuclear energy applications. However, formation of nanoscale a' precipitates during thermal aging at temperatures above 450?°C, or during neutron irradiation at lower temperatures, makes these Fe-Cr steels susceptible to embrittlement. To complement the existing literature, a series of Fe-9 to 18 Cr alloys were neutron-irradiated at temperatures between 320 and 455?°C up to doses of 20 dpa. In addition, post-irradiation annealing treatments at 500 and 600?°C were performed on a neutron-irradiated Fe-18 Cr alloy to validate the a-a' phase boundary. The microstructures were characterized using atom probe tomography and the results were analyzed in light of the existing literature. Under neutron irradiation and thermal annealing, the measured a' concentrations ranged from ~81 to 96?at.% Cr, as influenced by temperature, precipitate size, technique artifacts, and, possibly, cascade ballistic mixing.

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 Ni16Mn6Si7 or Ni3Mn2Si 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, Ni2-3Si 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

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 RPVtype steels, with an unusually wide range of dissolved Mn (0.06e1.34 at.%) and Ni (0.19e3.50 at.%) contents, were irradiated at z 290
C to z 1.4 1020 n/cm2 at an accelerated test reactor flux of z3.6 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 z 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

This article summarizes the recent progress on developing a class of potentially transformational structural materials called nanostructured ferritic alloys, which are leading candidates for advanced fission and fusion energy applications. Here, we focus on Fe-Cr-based ferritic stainless steels containing a very high concentration of Y-Ti-O nano-oxide features that enable a host of outstanding high-temperature properties, along with unique irradiation tolerance and thermal stability. Perhaps most notably, these alloys have an unprecedented capability to manage very high helium concentrations, pertinent to fusion service, in a way that transforms this element from a severe liability to a potential asset. In addition to providing some necessary background, we update progress on: (I) the character of the nanofeatures; (II) some unifying insights on key mechanical properties; (III) a quantitative model for nanofeature coarsening; (IV) recent irradiation experiments of the effects of helium on cavity evolution and void swelling; and (V) a powerful new mechanism controlling the transport, fate, and consequences of helium.

An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 58Ni(nth,γ) 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

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.

A set of low alloy model reactor pressure vessel steels, with systematic variations in their Mn, Ni, and Si contents, were neutron-irradiated to high fluence (1.4 × 1020 n/cm2) in the Advanced Test Reactor at Idaho at 290°C and a flux of 3.6 × 1012 n/cm2s. The alloys were analysed using atom probe tomography and solute clusters were observed in each alloy, including in one alloy that contained low nominal levels of Mn (0.04 at. %) and Si (0.06 at. %). Changes in the mechanical properties of the alloys were correlated with cluster volume fractions. Whilst the effect of nominal composition was observed to influence cluster composition, cluster nucleation site was not observed to affect composition. Several grain boundaries were also analysed and the segregation behaviour of certain elements is discussed.

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.

Formation of large volume fractions of Mn-Ni-Si precipitates (MNSPs) causes excess irradiation embrittlement of reactor pressure vessel (RPV) steels at high, extended-life fluences. Thus, a new and unique, semi-empirical cluster dynamics model was developed to study the evolution of MNSPs in low-Cu RPV steels. The model is based on CALPHAD thermodynamics and radiation enhanced diffusion kinetics. The thermodynamics dictates the compositional and temperature dependence of the free energy reductions that drive precipitation. The model treats both homogeneous and heterogeneous nucleation, where the latter occurs on cascade damage, like dislocation loops. The model has only four adjustable parameters that were fit to an atom probe tomography (APT) database. The model predictions are in semi-quantitative agreement with systematic Mn, Ni and Si composition variations in alloys characterized by APT, including a sensitivity to local tip-to-tip variations even in the same steel. The model predicts that heterogeneous nucleation plays a critical role in MNSP formation in lower alloy Ni contents. Single variable assessments of compositional effects show that Ni plays a dominant role, while even small variations in irradiation temperature can have a large effect on the MNSP evolution. Within typical RPV steel ranges, Mn and Si have smaller effects. The delayed but then rapid growth of MNSPs to large volume fractions at high fluence is well predicted by the model. For purposes of illustration, the effect of MNSPs on transition temperature shifts are presented based on well-established microstructure-property and property-property models.

Large volume fractions of Mn–Ni–Si (MNS) precipitates formed in irradiated light water reactor pressure vessel (RPV) steels cause severe hardening and embrittlement at high neutron fluence. A new equilibrium thermodynamic model was developed based on the CALculation of PHAse Diagrams (CALPHAD) method using both commercial (TCAL2) and specially assembled databases to predict precipitation of these phases. Good agreement between the model predictions and experimental data suggest that equilibrium thermodynamic models provide a basis to predict terminal MNS precipitation over wider range of alloy compositions and temperatures, and can also serve as a foundation for kinetic modeling of precipitate evolution.

A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed a' precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of a' precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested.