Determination of the Fe content of embedded Cu-rich particles in ferritic alloys using energy-filtered TEM

A novel technique for the quantification of the iron content of copper precipitates in ferritic steels is presented. Energy-filtered (EF) imaging has been used to extract elemental maps with high spatial resolution. These maps contain enough information to attempt the quantification of the signal produced by the precipitates when either a line profile is measured across them or the whole image signal is integrated. Assumptions such as sphericity of the precipitates and composition variations are discussed. Special attention to the assessment of drift on the information extracted from EF images has been taken. Minimum detectability and optimum acquisition conditions are discussed.     

Oxidation behavior of Incoloy 800 under simulated supercritical water conditions

For a correct design of supercritical water-cooled reactor (SCWR) components, data regarding the behavior of candidate materials in supercritical water are necessary. Corrosion has been identified as a critical problem because the high temperature and the oxidative nature of supercritical water may accelerate the corrosion kinetics. The goal of this paper is to investigate the oxidation behavior of Incoloy 800 exposed in autoclaves under supercritical water conditions for up to 1440 h. The exposure conditions (thermal deaerated water, temperatures of 723, 773, 823 and 873 K and a pressure of 25 MPa) have been selected as relevant for a supercritical power plant concept. To investigate the structural changes of the oxide films, X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive X-ray spectrometry (EDX) and electrochemical impedance spectroscopy (EIS) analyses were used. Results show changes in the oxides chemical composition, microstructure and thickness versus testing conditions (pressure, temperature and time). The oxide films are composed of two layers: an outer layer enriched in Fe oxide and an inner layer enriched in Cr and Ni oxides corresponding to small cavities supposedly due to internal oxidation.     

Structural materials challenges for advanced reactor systems

Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials mechanical properties and corrosion resistance, as well as component mock-up tests on technology loops to validate potential applications while accounting for mechanical design rules and manufacturing processes. The selection, assessment and validation of materials necessitate a large number of experiments, involving rare and expensive facilities such as research reactors, hot laboratories or corrosion loops. The modelling and the codification of the behaviour of materials will always involve the use of such technological experiments, but it is of utmost importance to develop also a predictive material science. Finally, the paper stresses the benefit of prospects of multilateral collaboration to join skills and share efforts of R&D to achieve in the nuclear field breakthroughs on materials that have already been achieved over the past decades in other industry sectors (aeronautics, metallurgy, chemistry, etc.).     

Pressurized heavy water reactor fuel behaviour in power ramp conditions

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

Measurement and calculation of high-energy neutron spectra behind shielding at the CERF 120 GeV/c hadron beam facility

Neutron energy spectra were measured behind the lateral shield of the CERF (CERN-EU High Energy Reference Field) facility at CERN with a 120 GeV/c positive hadron beam (a mixture of mainly protons and pions) on a cylindrical copper target (7-cm diameter by 50-cm long). An NE213 organic liquid scintillator (12.7-cm diameter by 12.7-cm long) was located at various longitudinal positions behind shields of 80- and 160-cm thick concrete and 40-cm thick iron. The measurement locations cover an angular range with respect to the beam axis between 13 and 133°. Neutron energy spectra in the energy range between 32 MeV and 380 MeV were obtained by unfolding the measured pulse height spectra with the detector response functions which have been verified in the neutron energy range up to 380 MeV in separate experiments. Since the source term and experimental geometry in this experiment are well characterized and simple and results are given in the form of energy spectra, these experimental results are very useful as benchmark data to check the accuracies of simulation codes and nuclear data.Monte Carlo simulations of the experimental set up were performed with the FLUKA, MARS and PHITS codes. Simulated spectra for the 80-cm thick concrete often agree within the experimental uncertainties. On the other hand, for the 160-cm thick concrete and iron shield differences are generally larger than the experimental uncertainties, yet within a factor of 2. Based on source term simulations, observed discrepancies among simulations of spectra outside the shield can be partially explained by differences in the high-energy hadron production in the copper target.     

Study by EELS of helium bubbles in a martensitic steel

This work presents measurements of the helium density and pressure in small bubbles in a martensitic steel, which is a vital first step towards identifying their role in the microstructural mechanisms determining the macroscopic properties of the material. Electron Energy-Loss Spectroscopy in the Scanning Transmission Electron Microscope has been used to analyse individual bubbles. The energy shift of the 1s → 2p transition and the helium density have been measured for each bubble analysed. The pressure inside the bubbles has been calculated from the helium density using an equation of state. In these bubbles, the values for the helium pressure seem to be smaller than the equilibrium pressure, and agree in trend with the findings of previous studies, although our results extend to smaller radii and higher pressures.     

Visual Observation of a Sound-Induced Bubble in Liquid 3He–4He Mixtures

We report visual observation of a sound-induced bubble in superfluid ³He–⁴He liquid mixtures using a high-speed camera at a rate of 1 msec/frame. The experiments were performed in the ³He dilute phase of phase-separated mixtures at 300 mK. The resonant frequency of the piezoelectric transducer was 9.36 MHz and the diameter of the active electrode was about 4 mm. When an acoustic wave pulse of sufficient magnitude was applied to the dilute phase from the transducer under saturated vapor pressure, a single bubble was nucleated on the active area. The bubble expanded almost spherically on the transducer, as it reached maximum size, it started shrinking, detached from the transducer, and collapsed. We also investigated the motion of the bubble in mixtures with a ³He concentration of 25% at 750 mK. In this case, the bubble grew elliptically on the transducer and detached from it without much change in shape.     

Microprobe analysis of neutron irradiated and autoclaved zirconium niobium claddings using synchrotron-based hard X-ray imaging and spectroscopy

Two zirconium alloys (Zr-2.5%Nb) - one oxidized in a pressurized water reactor, the other oxidized in autoclave and used as reference - are analyzed by combining synchrotron-based scanning transmission and fluorescence X-ray microscopy and micro-X-ray absorption spectroscopy (micro-XAS). Two-dimensional zirconium distribution maps recorded on the neutron irradiated and the non-irradiated autoclaved Zr-2.5%Nb alloys clearly allow the localization of the oxide and the metal parts of the interface with a micrometer spatial resolution. Micro-XAS investigations make possible the determination of the speciation of zirconium and niobium both in the oxide and the metal parts of the interface for the irradiated and non-irradiated samples. The coordination environment and/or the valency of zirconium and niobium in the metal and the oxide parts of the interface have been determined for both materials, and interpreted on the basis of comparison with metal and oxide reference compounds.     

Irradiation effects in helium implanted silicon carbide measured by X-ray absorption spectrometry

Silicon carbide (SiC) is investigated as a possible structural material for future nuclear power plants. It is utilized as fibre and/or as matrix in ceramic composite materials. The fibre reinforcement is necessary to provide the required ductility. In this work, the behaviour of pure SiC under irradiation by He implantation is studied. Samples are investigated by means of the extended X-ray absorption fine structure (EXAFS) spectroscopy, performed at the Si K-edge. The Fourier transforms of the EXAFS data indicate a decrease of the Si-Si bond related shells around the absorbing Si. The possible damage features are discussed and the three most probable ones for the irradiation conditions are selected for future modelling work.