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Solution annealing and age hardening are important processes for achieving good engineering and chemical properties of alloy 718. The composition of alloy 718 also plays an important role as it can affect both mechanical (γ”- and γ’-phase formation) and irradiation behaviors. Therefore, five different sets of alloy 718 samples with two different chemical compositions were fabricated using different processing conditions and irradiated up to 4 dpa. Based on the effects of irradiation on the microstructural data of γ”- and γ’-phases, irradiation-induced dislocation statistics, and crystallographic data, it can be inferred that the samples processed at high solution annealing temperature (1093°C) behave better under irradiation compared to the samples processed at low solution annealing temperatures (945 and 1065°C).

Chemical phase and microstructural investigations of alloy 718 solution-annealed and age-hardened were performed in this study. Focus was made in the effects of solution annealing temperature, aging temperature and holding time, and the amount of intermediate cold work on the alloy. The formation of secondary phases such as γ’-phase, γ”-phase, and δ-phase, grain sizes, and any deformations of the microstructure with respect to the processed conditions have been studied. Statistics such as size and number densities of these precipitates with respect to the processing conditions were evaluated and a discussion on optimum conditions in obtaining finer and higher density of γ’- and γ”-phase precipitates is also presented.

Nuclear power currently provides a fraction of non-carbon-emitting power generation in the United States and around the world. The existing fleet of water-cooled reactors continues to improve on operations and reliability, as ensuring public safety and environmental protection is a prerequisite, whether water reactor, advanced reactor, or fusion. Materials are important during all phases of a reactor's lifetime and must perform reliably for their entire expected lifetime. Unfortunately, nuclear reactors of all designs present a challenge for component service and material performance, as exemplified in water reactors. Components within a reactor must tolerate the harsh environment of high-temperature water, stress, vibration, and, for those components in the reactor core, an intense neutron field. Degradation of materials in this environment can lead to reduced performance and, in some cases, sudden failure. This chapter will provide an overview of the key water reactor designs, their features, and materials of construction. In addition, this chapter will introduce some of the key degradation modes facing materials for nuclear power reactors, including irradiation effects and corrosion. Finally, this chapter will provide an introduction to the key materials of construction, their use, and key characteristics.

This work presents a detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe–Cr–Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. The calculated fluxes up to 10 dpa suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.

Cast stainless steels (CASSs) have been extensively used for the large components of light water reactor (LWR) power plants such as primary coolant piping and pump casing. The thermal embrittlement of CASS components is one of the most serious concerns related to the extended-term operation of nuclear power plants. Many past researches have concluded that the formation of Cr-rich α′-phase by Spinodal decomposition of δ-ferrite phase is the primary mechanism for the thermal embrittlement. Cracking mechanism in the thermally-embrittled duplex stainless steels consists of the formation of cleavage at ferrite and its propagation via separation of ferrite–austenite interphase. This article intends to provide an introductory overview on the thermal aging phenomena in LWR-relevant conditions. Firstly, the thermal aging effect on toughness is discussed in terms of the cause of embrittlement and influential parameters. An approximate analysis of thermal reaction using Arrhenius equation was carried out to scope the aging temperatures for the accelerated aging experiments to simulate the 60 and 80 years of services. Further, an equilibrium precipitation calculation was performed for model CASS alloys using the CALPHAD program, and the results are used to describe the precipitation behaviors in duplex stainless steels. These results are also to be used to guide an on-going research aiming to provide knowledge-based conclusive prediction for the integrity of the CASS components of LWR power plants during the service life extended up to and beyond 60 years.

Alloy 718 is generally considered a highly corrosion-resistant material but can still be susceptible to stress corrosion cracking (SCC). The combination of factors leading to SCC susceptibility in the alloy is not always clear enough. In the present work, alloy 718 leaf spring (LS) materials that suffered stress corrosion damage during two 24-month cycles in pressurized water reactor service, operated to >45 MWd/mtU burn-up, was investigated. Compared to archival samples fabricated through the same processing conditions, little microstructural and property changes occurred in the material with in-service irradiation, contrary to high dose rate laboratory-based experiments reported in literature. Though the lack of delta phase formation along grain boundaries would suggest a more SCC resistant microstructure, grain boundary cracking in the material was extensive. Crack propagation routes were explored through focused ion beam milling of specimens near the crack tip for transmission electron microscopy as well as in polished plan view and cross-sectional samples for electron backscatter diffraction analysis. It has been shown in this study that cracks propagated mainly along random high-angle grain boundaries, with the material around cracks displaying a high local density of dislocations. The slip lines were produced through the local deformation of the leaf spring material above their yield strength. The cause for local SCC appears to be related to oxidation of both slip lines and grain boundaries, which under the high in-service stresses resulted in crack development in the material.

Analytical equations were developed to calculate the strength factors of precipitates, Frank loops, and cavities in austenitic alloys, which strongly depend on barrier type, size, geometry and density, as well as temperature. Calculated strength factors were successfully used to estimate radiation hardening using the broadly employed dispersed barrier-hardening model, leading to good agreement with experimentally measured hardening in neutron-irradiated type 304 and 316 stainless steel variants. The formulated strength factor provides a route for more reliable hardening predictions and can be easily incorporated into component simulations and design.

Understanding the thermal stability of intermetallic phases in Fe-rich Fe-Cr-Ni-Mo alloys is critical to alloy design and application of Mo-containing austenitic steels. Coupled with thermodynamic modeling, the thermal stability of intermetallic Chi and Laves phases in two Fe-Cr-Ni-Mo alloys was investigated at 1273 K, 1123 K, and 973 K (1000 °C, 850 °C, and 700 °C) for different annealing times. The morphologies, compositions, and crystal structures of the precipitates of the intermetallic phases were carefully examined by scanning electron microscopy, electron probe microanalysis, X-ray diffraction, and transmission electron microscopy. Two key findings resulted from this study. First, the Chi phase is stable at high temperature, and with the decreasing temperature it transforms into the Laves phase that is stable at low temperature. Secondly, Cr, Mo, and Ni are soluble in both the Chi and Laves phases, with the solubility of Mo playing a major role in the relative stability of the intermetallic phases. The thermodynamic models that were developed were then applied to evaluating the effect of Mo on the thermal stability of intermetallic phases in type 316 and NF709 stainless steels.

Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling with mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were used to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibiting unique defect sink characteristics depending on their local structure. Such interactions were found to be consistent across all doses investigated and to have larger global implications, including precipitation of Ni–Si clusters near different grain boundary types.

The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young’s modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in “soft” grains with a high Schmid factor located near “stiff” grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to 〈0 0 1〉||TA (tensile axis) and 〈1 0 1〉||TA were twin free and grain with 〈1 1 1〉||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

Le prolongement de la duree de fonctionnement des reacteurs au-dela de 60 ans et l’amelioration de leur performance sont essentiels pour satisfaire les futurs besoins energetiques des Etats-Unis, tout en reduisant les emissions de gaz a effet de serre. Pour cela, l’evaluation critique des connaissances des materiaux des structures et des composants d’une centrale nucleaire doit etre menee. Bien que l’accent ait ete mis sur les performances et les mecanismes de degradation possible des metaux en raison de l’accroissement de leur temps d’exposition en temperature, les sollicitations sur le beton creees par le liquide de refroidissement et les champs de radiation doivent aussi etre examinees. C’est l’objet d’un plan d’etude mene par l’EPRI et le laboratoire national d’Oakridge, soutenu par le Ministere americain de l’energie (Department of Energy - DOE).

A specific phenomenon – highly localized regions of deformation – was found and investigated at the free surface and near-surface layer of a neutron irradiated AISI 304 stainless steel bend specimen deformed to a maximum surface strain of 0.8%. It was shown that local plastic deformation near the surface might reach significant levels being localized at specific spots even when the maximum free surface strain remains below 1%. The effect was not observed in non-irradiated steel of the same composition at similar strain levels. Cross-sectional EBSD analysis demonstrated that the local misorientation level was highest near the free surface and diminished with increasing depth in these regions. (S)TEM indicated that the local density of dislocation channels might vary up to an order of magnitude. These channels may contain twins or may be twin free depending on grain orientation and local strain levels. BCC-phase (α-martensite) formation associated with channel-grain boundary intersection points was observed using EBSD and STEM in the near-surface layer.

The effect of thermal aging on coarsening kinetics of alloy 617, a candidate material for heat exchanger of the very high temperature reactor, was experimentally studied at 750 and 950 °C for up to 5300 h. Formation of various precipitates such as μ-phase, M23C6 and γ′ phases and significant coarsening of the γ′ phase have been observed in the microstructure of the aged samples. Experimental observation was compared to alloy thermodynamic calculation and γ′-phase precipitation kinetics simulation. Thermal aging effect on the microstructural evolution and mechanical behavior of alloy 617 was then discussed based on experimental and microstructural modeling results.

This work aims at utilizing modern computational microstructural modeling tools to accelerate the understanding of phase stability in austenitic steels under extended thermal aging. Using the CALPHAD approach, a thermodynamic database OCTANT (ORNL Computational Thermodynamics for Applied Nuclear Technology), including elements of Fe, C, Cr, Ni, Mn, Mo, Si, and Ti, has been developed with a focus on reliable thermodynamic modeling of precipitate phases in AISI 316 austenitic stainless steels. The thermodynamic database was validated by comparing the calculated results with experimental data from commercial 316 austenitic steels. The developed computational thermodynamics was then coupled with precipitation kinetics simulation to understand the temporal evolution of precipitates in austenitic steels under long-term thermal aging (up to 600,000 h) at a temperature regime from 300 to 900 °C. This study discusses the effect of dislocation density and difusion coefficients on the precipitation kinetics at low temperatures, which shed a light on investigating the phase stability and transformation in austenitic steels used in light water reactors.

Iron-based austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with the radiation-induced acceleration of ferrite phase formation. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593 K. An increase in magnetic phase amount for all alloys was observed over the 0–12 dpa dose range. However, magnetic phase (ferrite according to TEM results) did not appear to continuously increase at higher doses (above 12 dpa) but did tend to saturate. The formation of a magnetic phase in austenitic stainless steels during irradiation at 593 K appeared to be sensitive to alloy composition. It was found that silicon and manganese accelerated ferrite accumulation in the dose range of 0–12 dpa, whereas carbon and probably molybdenum resisted it. Also, an increase in grain size resisted ferrite formation, but cold work was found to stimulate it.

Surface relief due to localized deformation in a 4.4-dpa neutron-irradiated AISI 304 stainless steel was investigated using scanning electron microscopy coupled with electron backscattering diffraction and scanning transmission electron microscopy. It was found a body-centered-cubic (BCC) phase (deformation-induced martensite) had formed at the surface of the deformed specimen along the steps generated from dislocation channels. Martensitic hill-like formations with widths of ∼1 μm and depths of several microns were observed at channels with heights greater than ∼150 nm above the original surface. Martensite at dislocation channels was observed in grains along the [0 0 1]–[1 1 1] orientation but not in those along the [1 0 1] orientation.

Twinning and martensitic transformation have been investigated in nickel-enriched AISI 304 stainless steel subjected to tensile and indentation deformation. Using electron backscatter diffraction (EBSD), the morphology of α- and e-martensite and the effect of grain orientation to load axis on phase and structure transformations were analyzed in detail. It was found that the twinning occurred less frequently under indentation than under tension; also, twinning was not observed in [001] and [101] grains. In tensile tests, the martensite particles preferably formed at the deformation twins, intersections between twins, or at the twin-grain boundary intersections. Conversely, martensite formation in the indentation tests was not closely associated with twinning; instead, the majority of martensite was concentrated in the dense colonies near grain boundaries. Martensitic transformation seemed to be obstructed in the [001] grains in both tensile and indentation test cases. Under a tensile stress of 800 MPa, both α- and e-martensites were found in the microstructure, but at 1100 MPa only α-martensite presented in the specimen. Under indentation, α- and e-martensite were observed in the material regardless of the stress level.

RAFM steels are one of the candidate structural materials for fusion reactors, in which tantalum (Ta) and tungsten (W) are alloyed to replace niobium (Nb) and molybdenum (Mo) in conventional FM steels, respectively. This paper, using three RAFM heats, presents the effects of Ta and the primary austenite stabilizer carbon (C) on microstructure and strength. Thermomechanical treatment (TMT) was also applied to the heats, leading to significant increases in strength, attributable to the TMT-refined sub-grains and precipitates. The Ta-alloying favored the formation of (V, Ta)(N, C) and (Ta, V)C and exhibited greater strength. Fractographs also revealed the beneficial effects of TMT and Ta-alloying. However, extra C content, favoring a larger amount of M23C6 precipitates, did not show strengthening effect.

Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ∼320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

The objective of this work is to explore possible thermodynamic correlations between the degradation modes of austenitic and ferritic–martensitic alloys observed in high temperature He–CO–CO 2 environments with oxygen and carbon chemical potentials equivalent to that in a liquid sodium environment containing 2–5 molppm oxygen and 0.02–0.2 molppm carbon at temperatures 500–700 °C. Two He–CO–CO 2 environments (Pco/Pco 2 = 1320, Pco = 1980 molppm, and Pco/Pco 2 = 9, Pco = 13.5 molppm) were selected to test alloys NF616 and 316L at 700 and 850 °C. Upon exposure to He environments at 850 °C, 316L samples exhibited thick surface Cr 2 O 3 scales and substantial internal oxidation; however at 700 °C no significant internal oxidation was observed. NF616 samples exhibited relatively thinner surface Cr 2 O 3 scales compared to 316L samples at both temperatures. NF616 samples exposed to liquid sodium at 700 °C and He–Pco/Pco 2 = 9 at 850 °C showed decarburization. No surface oxide formation was observed on the sample exposed to the Na environment. Results obtained from He exposure experiments provide insight into what may occur during long exposure times in a sodium environment.

High-Cr (9 wt.%) ferritic–martensitic steels are important materials for use in nuclear reactors. This study shows a development activity for this category of steels via thermomechanical treatment (TMT) optimization and alloying element adjustment based on Grade 92 steels. Vickers microhardness and tensile tests were employed to assess the mechanical properties of the materials in the normalized-tempered (N&T) and optimized TMT conditions. The treatment of one of the modified heats produced ∼29% and ∼47% increases in hardness and yield strength, respectively, compared to the Grade 92 in the N&T condition. The TMT-treated alloys showed comparable or superior strength relative to the oxide-dispersion-strengthened steel PM2000. Microstructure analyses by optical and transmission electron microscopy together with thermodynamic calculations identified the strengthening mechanisms of the TMT and precipitates.

This work developed thermodynamic models for describing phase stability and thermodynamic property of the Fe–Cr–Zr system using the Calphad approach coupled with experimental study. Thermodynamic descriptions of the Fe–Cr and Cr–Zr systems were either directly adopted or slightly modified from literature. The Fe–Zr system has been remodeled to accommodate recent ab-initio calculation of formation enthalpies of various Fe–Zr compounds. Reliable ternary experimental data and thermodynamic models were mainly available in the Zr-rich region. Therefore, selected ternary alloys located in the vicinity of the eutectic valley of β(Fe,Cr,Zr) and (Fe,Cr)2Zr laves phase in the Fe-rich region have been experimentally investigated in this study. Microstructure has been examined by using scanning electron microscope, energy-dispersive X-ray spectroscopy and X-ray diffraction. These experimental results, along with the literature data were then used to develop thermodynamic models for phases in the Fe–Cr–Zr system. Calculated phase equilibria and thermodynamic properties of the ternary system yield satisfactory agreements with available experimental data.

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe–21Cr–32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and γ′-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly γ′ precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

Ferritic/Martensitic (F/M) steels containing 9 wt.% Cr are candidates for structural and cladding components in the next generation of advanced nuclear fission and fusion reactors. Although it is known these alloys exhibit radiation-induced segregation (RIS) at grain boundaries (GBs) while in-service, little is known about the mechanism behind RIS in F/M steels. The classical understanding of RIS in F/M steels presents a mechanism where point defects migrate to GBs acting as perfect sinks. However, variation in grain boundary structure may influence the sink efficiency and these migration processes. A proton irradiated 9 wt.% Cr model alloy steel was investigated using STEM/EDS spectrum imaging and GB misorientation analysis to determine the role of GB structure on RIS at different GBs. An ab initio based rate theory model was developed and compared to the experimental findings. This investigation found Cr preferentially segregates to specific GB structures. The preferential segregation to specific GB structures suggests GB structure plays a key role in the mechanism behind radiation-induced segregation, showing that not all grain boundaries in F/M steels act as perfect sinks. The study also found how irradiation dose and temperature impact the radiation-induced segregation response in F/M steels.

Creep–fatigue testing of Ni-based superalloy Inconel 617 and Haynes 230 were conducted in the air at 850 °C. Tests were performed with fully reversed axial strain control at a total strain range of 0.5%, 1.0% or 1.5% and hold time at maximum tensile strain for 3, 10 or 30 min. In addition, two creep–fatigue life prediction methods, i.e. linear damage summation and frequency-modified tensile hysteresis energy modeling, were evaluated and compared with experimental results. Under all creep–fatigue tests, Haynes 230 performed better than Inconel 617. Compared to the low cycle fatigue life, the cycles to failure for both materials decreased under creep–fatigue test conditions. Longer hold time at maximum tensile strain would cause a further reduction in both material creep–fatigue life. The linear damage summation could predict the creep–fatigue life of Inconel 617 for limited test conditions, but considerably underestimated the creep–fatigue life of Haynes 230. In contrast, frequency-modified tensile hysteresis energy modeling showed promising creep–fatigue life prediction results for both materials.

This paper summarizes an investigation of the deformation hardening behavior of neutron-irradiated stainless steels and copper in terms of true stress(σ)–true strain(e) curves. It is commonly accepted that the σ–e curves are more informative for describing plastic flow, but there are few papers devoted to using the true curves for describing constitutive behavior of irradiated materials. This study uses previously published true and engineering curves for stainless steel and copper irradiated to different damage level. The most appropriate constitutive equation has been identified, and it is shown that for the strain range 0–0.6 the true curves can be well described by the Swift equation: σ = k(e − e0)0.5. The influence of irradiation on the parameters of the Swift equation is investigated in detail. It is found that in most cases the k-parameter of this equation is not changed significantly by irradiation. Since large data scatter was observed for the e0-parameter, a modified Swift equation σ = k(e − σ02/k2)0.5 was proposed and evaluated. This equation is based on the concept of an initial stress σ0, which is, in general, close to the yield stress. The relationships among k, e0, and damage dose, influence of test temperature and grain size are discussed.

In order to provide a scientific basis for proposed life extension of current light water reactors, the radiation-induced degradation of stainless steel reactor internals will be discussed. A brief review of the basic radiation damage effects in stainless steels at LWR relevant conditions will be presented. It will be discussed how these basic effects result in the key degradation modes that have been identified by light water reactor experience to date, as well as the possibility for more severe degradation or new forms of degradation under extended service conditions. The forms of degradation that will be discussed include radiation hardening, embrittlement, dimensional stability (e.g., creep and swelling), radiation-induced segregation and precipitation, transmutation effects and irradiation-induced stress corrosion cracking.

Ferritic–martensitic (F–M) steels with 9 wt.%Cr are important structural materials for use in advanced nuclear reactors. Alloying composition adjustment, guided by computational thermodynamics, and thermomechanical treatment (TMT) were employed to develop high strength 9Cr F–M steels. Samples of four heats with controlled compositions were subjected to normalization and tempering (N&T) and TMT, respectively. Their mechanical properties were assessed by Vickers hardness and tensile testing. Ta-alloying showed significant strengthening effect. The TMT samples showed strength superior to the N&T samples with similar ductility. All the samples showed greater strength than NF616, which was either comparable to or greater than the literature data of the PM2000 oxide-dispersion-strengthened (ODS) steel at temperatures up to 650 °C without noticeable reduction in ductility. A variety of microstructural analyses together with computational thermodynamics provided rational interpretations on the strength enhancement. Creep tests are being initiated because the increased yield strength of the TMT samples is not able to deduce their long-term creep behavior.

Radiation-induced segregation in ferritic–martensitic alloy T 91 was studied to understand the behavior of solutes as a function of dose and temperature. Irradiations were conducted using 2 MeV protons to doses of 1, 3, 7 and 10 dpa at 400 °C. Radiation-induced segregation at prior austenite grain boundaries was measured, and various features of the irradiated microstructure were characterized, including grain boundary carbide coverage, the dislocation microstructure, radiation-induced precipitation and irradiation hardening. Results showed that Cr, Ni and Si segregate to prior austenite grain boundaries at low dose, but segregation ceases and redistribution occurs above 3 dpa. Grain boundary carbide coverage mirrors radiation-induced segregation. Irradiation induces formation of Ni–Si–Mn and Cu-rich precipitates that account for the majority of irradiation hardening. Radiation-induced segregation behavior is likely linked to the evolution of the precipitate and dislocation microstructures.

Casting of austenitic stainless steels offers the possibility of directly producing large and/or relatively complex structures, such as the first wall shield modules or the divertor cassette for the ITER, which may lead to simpler component fabrication and major cost savings. Past efforts to use cast steel for these large components were unsuccessful due to lower than acceptable strength in the test components. To improve and validate cast stainless steel as a substitute for wrought stainless steel for shield module applications, a series of test cast steels based on the commercially available CF3M specification have been designed and fabricated. These modifications utilize combinations of Mn and N, which result in significant increases in strength, fracture toughness, and impact properties. These mechanical performance improvements have been achieved without any loss of irradiation performance, corrosion performance, or weldability.

The microstructural evolution, specifically of grain boundaries, precipitates, and dislocations in thermomechanically processed (TMP) Alloy 800H samples was characterized by scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), electron backscattered diffraction (EBSD), transmission electron microscopy (TEM), and atomic force microscopy (AFM). The TMP not only significantly increased the fraction of low-Σ coincidence site lattice boundaries, but also introduced nanoscale precipitates in the matrix and altered the distribution of dislocations. Statistical analysis indicates that the morphology and distribution of grain boundary precipitates were dependent on grain boundary types. The microstructure optimization played a synergistic effect on the significantly increased strength with comparable ductility and enhanced intergranular corrosion resistance and creep-fatigue life compared to the as-received samples.

Ferritic–martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350–700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y–Ti–O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster–matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.

Thermophysical and mechanical properties of high purity chemically vapor-deposited (CVD) SiC and chemically vapor-infiltrated SiC matrix, pyrocarbon/SiC multilayered interphase composites with Hi-Nicalon™ Type-S and Tyranno™-SA3 SiC fibers were evaluated following neutron irradiation. Specimens including statistically significant population of tensile bars were irradiated up to 5.3 displacement-per-atom at ∼220 to ∼1080 °C in the Advanced Test Reactor at Idaho National Laboratory and High Flux Isotope Reactor at Oak Ridge National Laboratory. Thermal diffusivity/conductivity of all materials decreased during irradiation. The reciprocal thermal diffusivity linearly increased with temperature from ambient to the irradiation temperature. The magnitude of defect thermal resistance was distinctively different among materials and its ranking was Hi-Nicalon™ Type-S > Tyranno™-SA3 > CVD SiC regardless of irradiation condition. Dynamic Young’s modulus decrease for the irradiated CVD SiC exhibited explicit correlation with swelling. No significant effects of neutron irradiation on tensile properties of the composites were revealed, except for an anomaly case for the Hi-Nicalon™ Type-S composite irradiated in a specific condition. According to the single filament tensile evaluation, fibers of both types retained the original strength during irradiation at intermediate temperatures but significantly deteriorated during bare fiber irradiation at ∼910 °C. However, fiber strength deterioration was not observed when irradiated in composite form. Irradiation effects on the fiber–matrix interface properties were discussed based on results from the composite and single filament tensile tests, the hysteresis analysis, and the fracture surface examination.

Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural materials will be critical for the future success of proposed fusion energy reactors, which will subject the structures to unprecedented fluxes of high-energy neutrons along with intense thermomechanical stresses. Advanced materials can enable improved reactor performance via increased safety margins and design flexibility, in particular by providing increased strength, thermal creep resistance and superior corrosion and neutron radiation damage resistance. In many cases, a key strategy for designing high-performance radiation-resistant materials is based on the introduction of a high, uniform density of nanoscale particles that simultaneously provide good high temperature strength and neutron radiation damage resistance.

A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.

Greenhouse gas emission is a mounting problem that threatens the future production of electricity from both coal and natural gas. In 2006, 70% of domestic electricity generation relied upon fossil fuels. Projections by the Energy Information Agency1 show U.S. demand for electricity increasing 30% to 40% by 2030. Today nuclear power is the largest source of non-greenhouse-gas emitting energy generation and will be an important source of energy production in the future. This paper considers potential construction of new plants as well as the long-term prospects of existing nuclear power plants in the United States.

The objective of this paper is to examine the possible forms of irradiation damage that may impact materials performance over an extended service period in light water nuclear reactors. This paper will explore the different forms of irradiation damage that may be of concern under extended operation. Radiation-induced segregation, precipitation, hardening, embrittlement, and dimensional changes all will be discussed.

Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3 dpa at 400 °C and 1 dpa at 500 °C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3 dpa at 400 °C and above 1 dpa at 500 °C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS.

The proposed uses of fission reactors for manned or deep space missions have typically relied on the potential use of refractory metal alloys as structural materials. Throughout the history of these programs, a leading candidate has been Nb-1Zr, due to its good fabrication and welding characteristics. However, the less-than-optimal creep resistance of this alloy has encouraged interest in the more complex FS-85 (Nb-28Ta-10W-1Zr) alloy. Despite this interest, only a relatively small database exists for the properties of FS-85. Database gaps include the potential microstructural instabilities that can lead to mechanical property degradation. In this work, changes in the microstructure and mechanical properties of FS-85 were investigated following 1100 hours of thermal aging at 1098, 1248, and 1398 K. The changes in electrical resistivity, hardness, and tensile properties between the as-annealed and aged materials are compared. Evaluation of the microstructural changes was performed through optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). The development of intragranular and grain-boundary precipitation of Zr-rich compounds as a function of aging temperature was followed. Brittle tensile behavior was measured in the material aged at 1248 K, while ductile behavior occurred in samples aged above and below this temperature. The effect of temperature on the under- and overaging of the grain-boundary particles is believed to have contributed to the mechanical property behavior of the aged materials.

Ferritic–martensitic (FM) alloys are expected to play an important role as cladding or structural components in Generation IV systems operating in the temperature range 350–700 °C and to doses up to 200 dpa. Oxide dispersion strengthened (ODS) ferritic–martensitic steels have been developed to operate at higher temperatures than traditional FM steels. These steels contain nanometer-sized Y–Ti–O nanoclusters as a strengthening mechanism. Heavy ion irradiation has been used to determine the nanocluster stability over a temperature range of 500–700 °C to doses of 150 dpa. At all temperatures, the average nanocluster size decreases but the nanocluster density increases. The increased density of smaller nanoclusters under radiation should lead to strengthening of the matrix. While a reduction in size under irradiation has been reported in some other studies, many report oxide stability. The data from this study are contrasted to the available literature to highlight the differences in the reported radiation response.

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.