Production behavior of irradiation defects in vitreous silica under ion beam irradiation

The production behavior of irradiation defects in vitreous silica was studied by an in situ luminescence measurement technique under ion beam irradiation of H⁺ and He⁺. No apparent difference was observed in the luminescence spectra of specimens of different OH contents. The temperature dependence of the luminescence intensity at 280 and 460 nm was measured, and analyzed by considering the production mechanisms and kinetics of the irradiation defects of oxygen deficiency centers.     

Water formation in graphite and boron-doped graphite under simultaneous O and H irradiation

When graphite is exposed to simultaneous irradiation by H⁺ and O⁺, in addition to the H⁺-induced hydrocarbon and O⁺-induced CO and CO₂ formation, water is also formed. The present investigation explores the effect of the presence of boron in graphite on water formation. The results show that B-doped graphite specimens (∼15 at.% B) exposed to simultaneous O⁺ and H⁺ irradiation produce less water when compared to pure graphite under similar irradiation. The concept of the formation of ‘water-precursors’ at the end of the O⁺ range is proposed to explain the observed effect.     

Ab initio formation energies of Fe-Cr alloys

We have calculated ab initio lattice parameters, formation energies, bulk moduli and magnetic moments of Fe-Cr alloys. The results agree well with available experimental data. In addition to body centered cubic (bcc) alloys, which are representative of ferritic steels used in fast neutron reactors, face centered cubic (fcc) and hexagonal close packed (hcp) phases were considered in order to complete a theoretical database of thermodynamic properties. Calculations were done for the ferromagnetic phase, as well as for a phase with local moment disorder, simulating the magnetic structure at high temperatures. For the latter case, the formation energy of the alloy is strictly positive smooth function of chromium concentration, in agreement with experiments performed at high temperature. In the ferromagnetic case, a negative mixing enthalpy is found for chromium concentrations below 6%. Our observation is consistent with the experimentally observed inversion of the ordering trend, as well as with formation of the chromium rich α^′ phase at Cr-concentrations above 9%, occurring at T<900 K.     

The response of a chromium doped alumina screen to keV and MeV ions

We have characterised the response of a chromium-doped alumina screen ‘Chromox’ to light ions (H⁺ and He⁺) accelerated to keV and MeV energies. In particular, we have determined the absolute luminosity in terms of the number of photons emitted per incident ion from the front and back faces of such a screen. This work has been motivated by the application of this material to a diagnostic for measuring fast ion losses, close to the plasma edge, from the hot plasmas in fusion devices, where its radiation hardness, compared to that of standard phosphors, makes it very attractive. We also discuss the persistent afterglow observed after removal of the ion beam in terms of its cause and possible repercussions for this diagnostic.     

A cubic-to-monoclinic structural transformation in the sesquioxide Dy2O3 induced by ion irradiation

Polycrystalline pellets of the sesquioxide Dy₂O₃ were irradiated at cryogenic temperature with Kr⁺⁺ ions to a fluence of 1 × 10²⁰ Kr/m². The crystal structure of the irradiated Dy₂O₃ was observed to change from a cubic, so-called C-type rare-earth sesquioxide structure to a monoclinic, B-type rare-earth sesquioxide structure upon ion irradiation. This transformation is accompanied by a decrease in molecular volume (or density increase) of approximately 9%.     

Heats of formation of (U,Mo)Al3 and U(Al,Si)3

The heats of formation of (U,Mo)Al₃ intermetallic compounds were obtained by measuring the reaction heats of U-Mo/Al dispersion samples by differential scanning calorimetry. Based on literature data for the reaction heats of U₃Si/Al and U₃Si₂/Al dispersion samples, the heats of formation of U(Al,Si)₃ as a function of the Si content were calculated. The heat of formation of (U,Mo)Al₃ becomes less negative as the Mo content increases. Conversely, the heat of formation of U(Al,Si)₃ becomes more negative with increasing Si content.     

Effects of natural aging on the tensile properties of water-quenched U-6%Nb alloy

Uranium-6 wt% niobium (U-6%Nb) alloy has been in use for many years in the water-quenched (WQ) condition. The purpose of this work was to determine the effect of natural aging on tensile properties of the WQ U-6%Nb alloy. The materials studied were hemispherical shells after 15 and 20 years in storage. The alloy was successfully tested in the original curved configuration, using the specially designed tensile test apparatus. Finite element analysis confirmed the validity of the test method. The results of the tensile tests clearly indicated that in the WQ condition, the material is changing and after 15 and 20 years, the yield strength exceeds the original maximum allowable specification. The fracture mode transitions from highly ductile, microvoid coalescence in new material to a mixed mode of shallow dimples and inclusion-induced voids in the naturally aged material.     

Detection of point defects upon ion irradiation by means of precipitate coherency

Transmission electron microscopy has been widely used to investigate the radiation-damage microstructures, but has limitations when observing point defects due to its resolution limit. In the present study, a dilute Cu-Co alloy, which is a typical precipitation-hardened alloy, has been selected to investigate the formation and the mobility of point defects upon ion irradiation, during which the coherent precipitates lose their coherency and exhibit incoherent strain contrast, and the coherency loss can be used to detect the point defects.     

Laser annealing in combination with mass spectroscopy, a technique to study deuterium on tokamak carbon samples, a tool for detritiation

In this study, a method is presented based on mass spectroscopy to measure the areal density of deuterium on a graphite surface exposed to tokamak discharges. The studied sample was cut from a bumper limiter exposed in the TEXTOR tokamak and annealed by a 1 J Excimer laser (KrF). The energy used was 400 mJ cm⁻², which is below the threshold for ablation, 1 J cm⁻². The release of HD and D₂ was measured by a mass spectroscopy set-up and no other species released from the sample were detected in this experiment. The amount of D released from the sample after 20 laser pulses was measured to 7 × 10¹⁶ D atoms per cm⁻² (for this particular sample) and most of the hydrogen at the surface was released in the first pulse, as checked by nuclear reaction analysis (NRA) techniques, which gave changes of the amount of deuterium before and after laser annealing. The sensitivity in this experiment was 5 × 10¹⁴ atoms per cm⁻² for HD and 5 × 10¹³ atoms per cm⁻² for D₂.     

Stability of nanometer-sized oxide clusters in mechanically-alloyed steel under ion-induced displacement cascade damage conditions

Oxide dispersion strengthened ferritic steels are being considered for a number of advanced nuclear reactor applications because of their high strength and potential for high temperature application. Since these properties are attributed to the presence of a high density of very small (nanometer-sized) oxide clusters, there is interest in examining the radiation stability of such clusters. A novel experiment has been carried out to examine oxide nanocluster stability in a mechanically alloyed, oxide dispersion strengthened ferritic steel designated 12YWT. Pre-polished specimens were ion irradiated and the resulting microstructure was examined by atom probe tomography. After ion irradiation to ∼0.7 dpa with 150 keV Fe ions at 300 °C, a high number density of ∼4 nm-diameter nanoclusters was observed in the ferritic matrix. The nanoclusters are enriched in yttrium, titanium and oxygen, depleted in tungsten and chromium, and have a stoichiometry close to (Ti + Y):O. A similar cluster population was observed in the unirradiated materials, indicating that the ultrafine oxide nanoclusters are resistant to coarsening and dissolution under displacement cascade damage for the ion irradiation conditions used.     

Measurement of implanted helium particle transport by a flowing liquid lithium film

Due to its low atomic number, low sputtering yield, high sputtered ion fraction and excellent thermal properties, liquid lithium has been proposed as a potential candidate for advanced plasma-facing components (PFC). Using a liquid material opens the possibility of a continuously flowing, self-regenerating plasma-facing surface with a small residence time. This would allow such component to handle very high heat loads that are expected. There are, however, multiple unanswered questions regarding how such a liquid PFC would interact with the plasma in the reactor. The issue of particle control is critical, and it can be a factor to determine the feasibility of these advanced concepts. Hydrogen and helium are important in this regard: hydrogen transport by a flowing PFC impacts the reactor fuel recycling regime and tritium inventory; helium transport can help quantify ash removal by the flowing PFC. The flowing liquid-metal retention experiment (FLIRE) was built at the University of Illinois to answer some of the questions regarding particle transport by flowing liquid films exposed to plasmas. Experimental results regarding helium transport by a flowing lithium film after irradiation with an energetic He ion beam are presented in this work. Retained fraction values up to 2% were measured for the experimental conditions, and the retention was found to increase linearly with implanted ion energy. A pure diffusion model was used to describe the helium transport by the Li film, and it was found that such model predicts a diffusion coefficient of (2.8 ± 0.6) × 10⁻¹¹ m²/s, based on the experimental retention measurements. Preliminary evidence of long-term trapping of helium will also be presented.     

Ab initio calculations of self-interstitial interaction and migration with solute atoms in bcc Fe

The embrittlement of pressure vessel steels under radiation has been long ago correlated with the presence of solute Cu. Indeed the atom probe and the small angle neutron scattering, principally, have revealed the formation of Cu clusters under neutron flux in reactor pressure vessel (RPV) steels and dilute FeCu alloys. Other solutes such as Ni, Mn and Si which are also found within the clusters, are now suspected to contribute to the embrittlement. The interactions of these solutes with radiation induced point defects need thus to be characterized properly in order to understand the elementary mechanisms behind the formation of these clusters. We have investigated by ab initio calculations based on the density functional theory the interactions of self-interstitials with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si). Different possible configurations of solute-dumbbell complexes have been studied. Their binding energies are discussed, as well as their relative stability. The migration of dumbbells with a solute atom in their vicinity was also investigated. All these results are compared to some experimental ones obtained on dilute FeX model alloys. Our results indicate that for Mn solute atoms, diffusion via an interstitial mechanism is very likely.     

Object kinetic Monte Carlo study of sink strengths

The sink strength for three-dimensionally (3D) versus one-dimensionally (1D), or mixed 1D/3D, migrating defects in irradiated materials has attracted much attention in the recent past, because many experimental observations cannot be interpreted unless 1D or mixed 1D/3D migration patterns are assumed for self-interstitial atom clusters produced in cascades during irradiation. Analytical expressions for the sink strengths for defects migrating in 3D and also in 1D have been therefore developed and a ‘master curve’ approach has been proposed to describe the transition from purely 1D to purely 3D defect migration. Object kinetic Monte Carlo (OKMC) methods have subsequently been used to corroborate the theoretical expressions but, although good agreement was generally found, the ability of this technique to reach the 1D migration limit has been questioned, the limited size of the simulation box used in OKMC studies having been mainly blamed for the inadequacies of the model. In the present work, we explore the capability of OKMC to reproduce the sink strengths of spherical absorbers in a wide range of volume fractions, together with the sink strength of grain boundaries, for defects characterised by different migration dimensionality, from fully 3D to pure 1D. We show that this technique is not only capable of reproducing the theoretical expressions for the sink strengths in the whole range of conditions explored, but is also sensitive enough to reveal the necessity of correcting the theoretical expressions for large sink volume fractions. We thereby demonstrate that, in spite of the limited size of the OKMC simulation box, the method is suitable to describe the microstructure evolution of irradiated materials for any defect migration pattern, including fully 1D migrating defects, as well as to allow for the effect of extended microstructural features, much larger than the simulation box, such as grain boundaries.     

Damage inhomogeneity in the core region of displacement cascades in simplified nuclear glasses

Displacement cascades at energies ranging from 16 keV to 70 keV were simulated by classical molecular dynamics. Damage inhomogeneity was observed in each case: the atomic density was diminished by the incident projectile to a variable extent depending on the regions concerned. The regions near the initial projectile position are largely annealed, and regions near the end of the cascade are relatively unaffected because of the low residual projectile energy. However, maximum damage occurs in intermediate regions from collisions with incident projectiles at energies ranging from about 10 keV to 25 keV. This phenomenon illustrates the competition between structure annealing and projectile-induced damage: both increase with the local energy, but with different dynamics. At the highest energies, annealing wins out over damage, restoring the glass structure to its pristine state; hence the good structural behaviour in the zones closest to the initial projectile position, which are subjected to the greatest local temperature rise.     

Dissolution of magnetite using an environmentally friendly chelant: an electrochemical impedance spectroscopy study

Ethylenediaminedisuccinic acid (EDDS), a biodegradable substitute for EDTA, was used as a chelant for dissolving magnetite and magnetite formed on iron metal surface. Dissolution was found to increase in presence of ferrous ions and depend on pH of solution, concentration of ferrous ion, EDDS concentration, applied cathodic potential and temperature. The impedance spectrum for the dissolution of magnetite film formed on iron exhibited two time constants. The first at high frequency range is ascribed to the reductive dissolution of magnetite. It is very crucial to carry out the dissolution process at the appropriate temperature to insure complete removal of oxide layer.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

Thermal variation of the optical absorption of UO2: determination of the small polaron self-energy

The temperature variation of UV-VIS-NIR optical spectra of UO₂ have been investigated from room temperature up to 1173 K with careful in situ oxygen partial pressure control. The deduced optical absorption edge exhibits a strong temperature dependence. Its value decreases from ∼2 eV at room temperature to ∼0.8 eV at 1173 K. Such thermal behaviour is interpreted as the consequence of the existence of a strong electron-phonon coupling (small polaron). In the temperature range 300-1173 K, the model yields a hopping radius of ∼2 Å and a polaron self-energy of E_p=−0.38 eV.     

Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10¹⁸ and 1.02 × 10¹⁹ n/cm² (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.     

Detection of intergranular cracking susceptibility due to hydrogen in irradiated austenitic stainless steel with a superconducting quantum interference device (SQUID) sensor

To investigate the detection method of intergranular (IG) cracking susceptibility by hydrogen in irradiated austenitic stainless steel (SS), magnetic and mechanical properties were examined after two repeats of hydrogen charging and discharging (hydrogen treatment) in Type 304 SS which had been irradiated during use in different reactor cores. The residual magnetic flux density (Br) was measured with a superconducting quantum interference device sensor and Br increased with increased neutron fluence and repeated hydrogen treatments. Elongation decreased with an increase of Br and IG cracking appeared above Br of 2×10⁻⁵ T for this measuring method after repeated hydrogen treatments. These phenomena would be caused by hydrogen-induced martensite phase being formed on grain boundaries. It was thought the appearance of IG cracking susceptibility due to hydrogen in irradiated SS could be predicted by measuring the Br of the steel.     

Experimental investigation of the Zr corner of the ternary Zr-Nb-Fe phase diagram

Intermediate phases in the Zr-rich region of the Zr-Nb-Fe system have been investigated by X-ray diffraction, optical and electron microscopy and electron microprobe analysis. The chemical composition ranges covered by the alloys studied here are: (41-97) at.% Zr, (32-0.9) at.% Nb and (0.6-38) at.% Fe. The phases found in this region were: the solid solutions α-Zr and β-Zr, the intermetallic Zr₃Fe with less than 0.2 at.% Nb in solution, two new ternary intermetallic compounds (Zr+Nb)₂Fe `λ<sub>1</sub>' with a cubic Ti<sub>2</sub>Ni-type structure in the composition range (2.4-13) at.% Nb and (31-33) at.% Fe, and (Fe+Nb)<sub>2</sub>Zr `λ₂' indexed as hexagonal Laves phase MgZn₂ type (C14) with a wide range of compositions close to (35-37) at.% Zr, (12-31) at.% Nb and (32-53) at.% Fe.