Irradiation-induced Microstructural Changes Research Articles

Use of LANSCE Spallation Neutrons to Drive a Fusion Prototypic Neutron Source

A high-power neutron-producing spallation source is one option for meeting the mission of a fusion prototypic neutron source (FPNS). The Los Alamos Neutron Science Center at Los Alamos National Laboratory is uniquely suited to host such a source. A target concept has been developed that satisfies the initial goals established for an FPNS, as has a beam transport design that satisfies beam-on-target requirements for the concept. We discuss the potential impact of pulsed beam operation on radiation-induced changes in microstructure, as well as the ingrowth of calcium in steel alloys at a rate well beyond that expected in a fusion reactor first wall. A preliminary thermomechanical assessment shows the need to modify the target design to reduce temperatures and stresses in the tungsten target.

X-ray diffraction line profile analysis of defects in neutron-irradiated austenitic stainless steels at low displacement damage levels

Neutron irradiation-induced microstructural changes in variants of austenitic stainless steels were characterized using the Angle Dispersive X-Ray Diffraction (ADXRD) technique with a synchrotron X-ray source of energy 14.968 keV (wavelength ∼0.82833 Å). SS 304 L(N), SS 316 L(N), and SS 316 materials were neutron-irradiated to low displacement damage levels (1–6.75 dpa) in the Fast Breeder Test Reactor at Kalpakkam at temperatures in the range ∼380 °C-400 °C. The effect of the irradiation-induced defects on X-ray diffraction peak profile parameters such as peak broadening, peak shift, and asymmetry were quantified and related to dislocation density. Peak broadening is found to increase as a function of dpa in most of the cases. Considering the highest dpa sample, lattice contraction has been shown by SS 304 L(N) and SS 316 whereas lattice expansion is exhibited by SS 316 L(N). Asymmetry in peak broadening is found to decrease in case of the highest dpa sample. Ferrite volume percent is seen to have increased as a function of dpa in SS 304 L(N) and SS 316 L(N) but not observed in SS 316. SS 304 L(N) has shown a greater propensity towards the austenite to ferrite transformation and 5 dpa sample has about 18 vol% of α-ferrite. However, in case of 5 dpa SS 316 L(N), ferrite content has increased to ∼4 vol% . The increase in yield strength was also related to the dislocation density amongst all three grades of stainless steel. SS 304 L(N) showed saturation in dislocation density and yield strength around 2 dpa whereas SS 316 L(N) showed linear behaviour up to the highest dpa.

Impact of pre-existing crystal lattice defects on the accumulation of irradiation-induced damage in a C/C composite

A carbon-fibre reinforced carbon-matrix (C/C) composite was irradiated with 30 MeV C6+ ions to a peak damage of ∼25 dpa. Ion irradiation-induced microstructural changes were mainly studied using Raman spectroscopy. The irradiation-induced crystal lattice defect accumulation in the C/C composite was compared with a reference of PCIB graphite (nuclear-grade). It shows that a high concentration of pre-existing crystal lattice defects in the studied C/C composite have a significant impact on the unexpectedly high disordering of the crystal lattice observed along the entire ion range. In comparison, PCIB graphite with much less pre-existing crystal lattice defects behaves in a more predictable manner with the irradiation damage accumulated in a narrow high dpa region. We rationalised that a large number of pre-existing crystal lattice defects in the C/C composite lead to a stronger electron-phonon coupling and play an important role on the formation of stable crystal lattice defects due to electronic energy loss during ion irradiation. The present results have implications for the development of C/C composites for radiation-tolerant applications, in terms of the crystal lattice defect elimination in the as-manufactured microstructure. Additionally, this investigation identifies a fundamental knowledge gap in the electronic energy loss effect on the irradiation damage produced in carbon-based materials at intermediate ion energies.

Characterization of hardening behavior of carbon steels at low temperature irradiations

• Embrittlement factors of carbon steels irradiated at low temperature were fine dislocation loops and solute clusters of Mn-Si-Cu atoms. • Carbon steels and their weld metals containing high concentrations of Cu, Mn and Si had relatively high radiation hardening due to promotion of solute atom clustering. • Low temperature irradiation resulted in higher radiation hardening due to the higher number density of dislocation loops, despite the lower number density of solute clusters. • Low dose rate irradiation (electron irradiation) resulted in greater radiation hardening due to the higher number density of solute clusters than high dose rate irradiation (ion irradiation) did. Overpacks used for geological disposal of high-level radioactive waste will be irradiated by neutrons and γ rays with very low dose rate for a long time. A possible way to predict effects of such radiations is that microstructure changes under the expected irradiation conditions are estimated by both experiments and analytical modeling and then the changes in hardness and fracture toughness from the microstructure changes are estimated. In the present study, carbon steels, which are one of the candidate materials for overpacks, model alloys and a weld of carbon steel were irradiated with either Fe ions or electrons at 90 or 290 °C, and then the microstructure and hardness were investigated to gain an understanding about embrittlement factors of carbon steels and effects of both material composition and irradiation conditions (temperature, dose rate) on radiation-induced changes in microstructure and hardness. Formations of fine dislocation loops and solute clusters containing Mn, Si and Cu atoms were confirmed by microstructure analysis using transmission electron microscopy and atom probe tomography. Hardness increase was well correlated with the evolution of dislocation loops and solute clusters, indicating that these defects are embrittlement factors of carbon steels. Higher radiation hardening was confirmed for the materials containing a high density of Mn, Si and Cu, and the irradiation conditions of low temperature and low dose rate.

Effect of crystallographic orientation on structural response of silicon to femtosecond laser irradiation

Femtosecond laser has been widely utilized for modification of crystal structure to achieve desired functions. So far, however, the effect of crystallographic orientation on the induced structure by femtosecond laser processing has yet been comprehensively studied. The present work is undertaken in an attempt to fill this gap in our knowledge. To this end, commercial-purity Si is used as a target material and high-resolution transmission electron microscopy as well as electron backscatter diffraction are applied to examine the irradiation-induced microstructural changes. The structural response of the pulsed material is found to be principally influenced by the crystallographic orientation of the target surface. Specifically, at the surface orientation close to {111}, a pronounced amorphization effect is observed whereas no disordered material is detected at the orientations close to {100}. This phenomenon could be explained by the lowest crystallization speed required by the (111) surface due to its smallest surface energy. Compared with nanosecond laser, non-thermal melting induced by femtosecond laser induces mild thermal gradient and favors recrystallization.

Combined ab-initio and empirical model for irradiated metal alloys with a focus on uranium alloy fuel thermal conductivity

In this work, we developed a model framework of thermal conductivity of irradiated metal and metal alloys based on density functional theory calculations and experimental data. The model incorporates the contributions of point defects (vacancies and transmutation products), grain boundaries, and noble gas bubbles (inter- and intra-granular) resulting from irradiation into the overall thermal conductivity. The model is demonstrated on U-Mo alloys and displays semi-quantitative agreement with experimental data, with a root-mean-square error of about 1 W/m-K (≈9% error versus experiment) over the typical operating temperature and burn-up range for U-Mo alloys. This work provides insight into the relationships of the thermal conductivity changes resulting from the different irradiation-induced changes in microstructure over a wide range of temperatures, and shows that scattering of electrons from point defects produces the majority of the thermal conductivity reduction of U-Mo alloys under irradiation. Specifically, the effect of point defects decreases the thermal conductivity by about 20% when burn-up >20% (e.g., 50% burn-up), while grain boundary and gas bubble formation contribute only about another 10% reduction at 50% burn-up. The ability of our model to individually quantify the effect of different irradiation-related defects on the thermal conductivity as a function of temperature, composition, and irradiation conditions offers both a powerful tool for quantitative semi-empirical modeling of thermal conductivity in irradiated metals and materials design insight for the control of thermal conductivity in metal alloys in applications experiencing irradiation, such as next-generation nuclear reactors.

Oxygen-18 Tracer Measurements of Anion Diffusion in Uranium Dioxide Thin Films

Oxygen 18 was used as a tracer to quantify anion diffusion in thin-film UO2 using secondary ion mass spectroscopy to measure one-dimensional depth profiles. Both thermal and heavy ion bombardment (1.8 MeV Kr+) treatments were employed over a temperature range from 295 K to 623 K. Textured and single-crystal thin-film samples were grown using reactive-gas magnetron sputtering at ambient temperature. Both microstructures resulted in similar thermal activation energies, Ea = 0.2 eV. This activation energy is significantly lower than the known value for intrinsic anion vacancy-self diffusion in stoichiometric UO2.00 (Ea = 2.5 eV). We attribute this to an interstitialcy mechanism in our hyper-stoichiometric films. The activation energy for irradiated textured films was approximately half that of thermal diffusion, consistent with the chemical rate theory. The opposite was true for the single-crystal microstructure (irradiated Ea = 0.36 eV). This may be due to radiation-induced changes in the microstructure. The mixing parameter was quantified on the anion sublattice as well, ξ = 2.1 ± 0.2 A5eV−1.

TEM studies of 1 MeV Fe+ ion-irradiated W alloys by wet chemical method: high-temperature annealing and deuterium retention

In this study, the microstructural changes in irradiated W–1 wt%Pr2O3, W–1 wt%La2O3, W–1 wt%TiC and W–1 wt%ZrC specimens by 1 MeV Fe+ ion irradiation to doses of 1 dpa at room temperature have been characterized by transmission electron microscopy (TEM) to understand the radiation damage. The specimens of 1 dpa were consecutively annealed at temperatures of 600 °C, 800 °C, 900 °C and 1100 °C for 1 h, and the post-irradiation annealing effect on the microstructure was studied. Microstructural features such as dislocations, dislocation loops and dense dislocation networks were observed by the means of TEM. There was a significant change in the microstructure of the tungsten (W) after irradiation and post-irradiation annealing. W–1 wt%TiC likely consisted of fewer radiation-induced defects and hence exhibits higher resistance to radiation-induced microstructural changes than the others. The general trend is that the density of defect clusters decreases while the average size of the clusters increases with the annealing temperature. A complete removal of dislocation structures at 1100 °C occurred in all materials. The irradiation damage effect of Fe+ ions on the deuterium retention of as-synthesized W-based composites was investigated. Thermal desorption spectroscopy of W implanted with ions was performed using the low-energy ion irradiation system of Kyoto University. Compared with the samples without Fe+ pre-irradiation, the deuterium retention of the four composites increased after Fe+ pre-irradiation. The retained amount of deuterium was also increased with the increase in the Fe+ irradiation dose. This is of vital importance to the scientific community working on these materials for extreme irradiating environments.

A Review on the Radiation Response of Nanoporous Metallic Materials

High-dose, high-energy particle radiation induces significant microstructural damage in metallic materials. Free surfaces are ideal sinks for radiation-induced defects. Hence, nanoporous materials with abundant free surfaces may have outstanding radiation tolerance. This review summarizes recent studies on the interactions of defect clusters with free surfaces, and the influence of free surfaces and temperature on radiation-induced microstructural changes, such as shrinkage of nanopores, in nanoporous metallic materials. The mechanical properties of irradiated nanoporous metals are also briefly summarized.

Initial examination of fuel compacts and TRISO particles from the US AGR-2 irradiation test

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second irradiation test. These compacts were selected for examination because there were indications that they may have contained particles that released cesium through a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzer (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using X-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and X-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers.

Tailoring the structural and optical properties of TiN thin films by Ag ion implantation

Titanium nitride (TiN) thin films thickness of ∼260nm prepared by dc reactive sputtering were irradiated with 200keV silver (Ag) ions to the fluences ranging from 5×1015ions/cm2 to 20×1015ions/cm2. After implantation TiN layers were annealed 2h at 700°C in a vacuum. Ion irradiation-induced microstructural changes were examined by using Rutherford backscattering spectrometry, X-ray diffraction and transmission electron microscopy, while the surface topography was observed using atomic force microscopy. Spectroscopic ellipsometry was employed to get insights on the optical and electronic properties of TiN films with respect to their microstructure. The results showed that the irradiations lead to deformation of the lattice, increasing disorder and formation of new Ag phase. The optical results demonstrate the contribution of surface plasmon resonace (SPR) of Ag particles. SPR position shifted in the range of 354.3–476.9nm when Ag ion fluence varied from 5×1015ions/cm2 to 20×1015ions/cm2. Shift in peak wavelength shows dependence on Ag particles concentration, suggesting that interaction between Ag particles dominate the surface plasmon resonance effect. Presence of Ag as second metal in the layer leads to overall decrease of optical resistivity of TiN.

Simulation of C-Component Dislocation Loops by Ions and Protons Irradiation. A Tool to Understand the Breakaway Growth of Recrystallized Zr-Based Alloys

During normal in-service operation, the PWR fuel assemblies exhibit a macroscopic elongation. This dimensional change is the result of three different phenomena: thermal creep, irradiation-creep and irradiation growth which occurs in the absence of any applied stress. This last phenomenon, referred as stress-free growth, is related to irradiation-induced microstructural changes taking place in the hexagonal close packed grains and corresponding to an elongation along the basal plane and a shortening along the c-axis, the volume remaining constant (Fig 1). Under neutron irradiation, recrystallized Zr-based alloys undergo stress-free growth which accelerates for high irradiation doses (Fig 2). This acceleration is correlated to specific c-component vacancy dislocation loops.

Fuel Assembly Design for Supercritical Water-Cooled Reactor

The supercritical water-cooled reactor (SWCR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to the state-of-the-art light water reactors (LWRs). However, there are a large number of potential problems that must be addressed, particularly the fuel assembly design of the SCWR. SCWRs are a kind of high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (374°C, 22.1 MPa). Corrosion and degradation of materials used in supercritical water environments are determined by several environment- and material-dependent factors. In particular, irradiation-induced changes in microstructure and microchemistry are major concerns in a nuclear reactor. Many structural materials including alloys and ceramics have been proposed for use as SCWR components or materials for applying protective coatings in SCWRs. In this paper, the present status of supercritical fuel assembly design at home and abroad is reported. According to the special requirements of supercritical core design, a kind of configuration design of fuel assembly with two-flow core and using SiC as cladding material are proposed. The analysis results have shown that the design basically meets the requirements of fuel assembly design, which has good feasibility and performance.

Effect of simultaneous helium implantation on the microstructure evolution of Inconel X-750 superalloy during dual-beam irradiation

This study focuses on investigation into the effect of helium implantation on microstructure evolution in Inconel X-750 superalloy during dual-beam (Ni+/He+) irradiation. The 1 MeV Ni+ ions with the damage rate of 10−3 dpa/s as well as 15 keV He+ ions using rate of 200 appm/dpa were simultaneously employed to irradiate specimens at 400 °C to different doses. Microstructure characterization has been conducted using high-resolution analytical transmission electron microscopy (TEM). The TEM results show that simultaneous helium injection has significant influence on irradiation-induced microstructural changes. The disordering of γ′ (Ni3 (Al, Ti)) precipitates shows noticeable delay in dose level compared to mono heavy ion irradiation, which is attributed to the effect of helium on promoting the dynamic reordering process. In contrast to previous studies on single-beam ion irradiation, in which no cavities were reported even at high doses, very small (2–5 nm) cavities were detected after irradiation to 5 dpa, which proved that helium plays crucial role in cavity formation. TEM characterization also indicates that the helium implantation affects the development of dislocation loops during irradiation. Large 1/3 〈1 1 1〉 Frank loops in the size of 10–20 nm developed during irradiation at 400 °C, whereas similar big loops detected at higher irradiation temperature (500 °C) during sole ion irradiation. This implies that the effect of helium on trapping the vacancies can help to develop the interstitial Frank loops at lower irradiation temperatures.

From solid solution to cluster formation of Fe and Cr in α-Zr

To understand the mechanisms by which the re-solution of Fe and Cr additions increase the corrosion rate of irradiated Zr alloys, the solubility and clustering of Fe and Cr in model binary Zr alloys was investigated using a combination of experimental and modelling techniques — atom probe tomography (APT), x-ray diffraction (XRD), thermoelectric power (TEP) and density functional theory (DFT). Cr occupies both interstitial and substitutional sites in the α-Zr lattice; Fe favours interstitial sites, and a low-symmetry site that was not previously modelled is found to be the most favourable for Fe. Lattice expansion as a function of Fe and Cr content in the α-Zr matrix deviates from Vegard's law and is strongly anisotropic for Fe additions, expanding the c-axis while contracting the a-axis. Matrix content of solutes cannot be reliably estimated from lattice parameter measurements, instead a combination of TEP and APT was employed. Defect clusters form at higher solution concentrations, which induce a smaller lattice strain compared to the dilute defects. In the presence of a Zr vacancy, all two-atom clusters are more soluble than individual point defects and as many as four Fe or three Cr atoms could be accommodated in a single Zr vacancy. The Zr vacancy is critical for the increased apparent solubility of defect clusters; the implications for irradiation induced microstructure changes in Zr alloys are discussed.

Effects of helium and hydrogen on radiation-induced microstructural changes in austenitic stainless steel

Microstructural changes in austenitic stainless steel by helium, hydrogen, and iron ion irradiation were investigated with transmission electron microscopy. Typical radiation-induced changes, such as the formation of Frank loops in the matrix and radiation-induced segregation (RIS) or depletion at grain boundaries, were observed after ion irradiation. The helium ion irradiation led to the formation of cavities both at grain boundaries and in the matrix, as well as the development of smaller Frank loops. The hydrogen ion irradiation generated stronger RIS behavior at the grain boundaries compared to irradiation with helium and iron ions. The effects of helium and hydrogen on radiation-induced microstructural changes were discussed.

Applications of Transient Grating Spectroscopy to Radiation Materials Science

The ability to study radiation damage in situ will directly enable the rapid innovation and qualification of materials for nuclear applications by allowing direct observation of the effects of radiation damage accumulation. This is a challenging task, as the measurement technique must be noncontact, nondestructive, rapid, and still allow for online irradiation without interference. Applicable methods of mechanical spectroscopy are surveyed, noting their potential usefulness for characterizing radiation-induced microstructural changes in situ. The transient grating (TG) spectroscopy technique appears most suited for these studies, due to its noncontact, nondestructive nature, its ability to rapidly probe materials to the depth of ion irradiation, and the large number of deconvolvable components extractable from its signal. Work is proposed to separate the individual mechanisms of irradiation damage using in situ and ex situ TG spectroscopy, through a suite of single-effect and integrated experiments.

Effect of helium implantation on SiC and graphite**Project supported by the ITER-National Magnetic Confinement Fusion Program, China (Grant Nos. 2010GB109000, 2011GB108009, and 2014GB123000) and the National Natural Science Foundation of China (Grant No. 11075119).

Effects of helium implantation on silicon carbide (SiC) and graphite were studied to reveal the possibility of SiC replacing graphite as plasma facing materials. Pressureless sintered SiC and graphite SMF-800 were implanted with He+ ions of 20 keV and 100 keV at different temperatures and different fluences. The He+ irradiation induced microstructure changes were studied by field-emission scanning electron microscopy (FESEM), atomic force microscopy (AFM), and transmission electron microscopy (TEM).

Electron Beam Effects during In-Situ Annealing of Self-Ion Irradiated Nanocrystalline Nickel

ABSTRACTTransmission electron microscopy (TEM) is a valuable methodology for investigating radiation-induced microstructural changes and elucidating the underlying mechanisms involved in the aging and degradation of nuclear reactor materials. However, the use of electrons for imaging may result in several inadvertent effects that can potentially change the microstructure and mechanisms active in the material being investigated. In this study, in situ TEM characterization is performed on nanocrystalline nickel samples under self-ion irradiation and post irradiation annealing. During annealing, voids are formed around 200 °C only in the area illuminated by the electron beam. Based on diffraction patterns analyses, it is hypothesized that the electron beam enhanced the growth of a NiO layer resulting in a decrease of vacancy mobility during annealing. The electron beam used to investigate self-ion irradiation ultimately significantly affected the type of defects formed and the final defect microstructure.

Evolution of microstructure and mechanical properties of VVER-1000 RPV steels under re-irradiation

This is a comprehensive study of microstructure and mechanical properties evolution at re-irradiation after recovery annealing of VVER-1000 RPV weld and base metals as well as the effect of annealing on the microstructure and properties of base metal in the zone of the temperature gradient that is implemented during annealing using special heating device. It is shown that the level of radiation-induced microstructural changes under accelerated re-irradiation of weld and base metal is not higher than for the primary irradiation. Thus, we can predict that re-embrittlement of VVER-1000 RPV materials considering the flux effect will not exceed the typical embrittlement rate for the primary irradiation.