Investigations of the melting behaviour of the U–Zr–Fe–O system

During a severe nuclear accident, the UO₂ fuel rods, Zircaloy cladding, guide tubes, absorber and steel structural components inside the reactor pressure vessel overheat and a series of interactions between these elements and the steam atmosphere occur. These produce more heat in addition to the decay heat and result in a liquid corium of oxidic and metallic phases depending on the exact conditions and processes. A major systems resulting from this is the U–Zr–Fe–O system. High-temperature data for this system is important in order to be able to model these interactions. The Joint Research Centre, Institute for Transuranium Elements (JRC-ITU) has been examining the melting ranges for this system over the whole FeO range by means of a specialized laser flash technique that achieves very high temperatures and avoids crucible contamination. The melted zones were examined for their structure, composition and for estimation of the liquidus and solidus temperatures. The results showed that with FeO contents of over 20mol% there was a very large melting range that would permit long liquid cooling times and extend the relocation of fuel material within the reactor pressure vessel. Based on these results, the main phase regimes expected under severe accident conditions could be identified.     

Dynamics of deuterium retention and sputtering of Li–C–O surfaces

Chemistry as well as sputtering and reflection dynamics of lithiated carbon material, bombarded by slow hydrogen atoms are studied. We present a realistic method for computational simulation of the dynamics of the polar Li–C–O–H material dynamics. It is based on an approximate, semi-empirical quantum mechanics of electrons and classical mechanics of nuclei. Results are validated qualitatively by comparison with experiments and with a first principle DFT computations. In particular, we explain observed details of the hydrogen bonding chemistry in lithiated carbon, showing that incoming hydrogen interacts preferably with Li-C rather than C structures.     

Metallurgical characterizations of Fe–Cr–Ni–Zr base alloys developed for geological disposal of radioactive hulls

Alloy melting route is currently being considered for radioactive hulls immobilization. Towards this, wide range of alloys, belonging to Zirconium–Iron binary and Zirconium–Stainless steel pseudo-binary systems have been prepared through vacuum arc melting route. Detail microstructural characterization and quantitative phase analyses of these alloys along with interaction study between Zirconium and Stainless steel coupons at elevated temperatures identify Zr(Fe,Cr)₂, Zr(Fe,Cr), Zr₂(Fe,Cr), Zr₃(Fe,Ni), Zr₃(Fe,Cr), Zr₃(Fe,Cr,Ni), β-Zr and α-Zr as the most commonly occurring phases within the system for Zirconium rich bulk compositions. Nano-indentation studies found Zr(Fe,Cr)₂ and Zr(Fe,Cr) as extremely hard, Zr₃(Fe,Ni) as moderately ductile and β-Zr, Zr₂(Fe,Cr) as most ductile ones among the phases present. Steam oxidation studies of the alloys, based on weight gain/loss procedure and microstructural characterization of the mixed oxide layers, suggest that each of the alloys responded to the corrosive environment differently. Fe₂O₃, NiFe₂O₄, NiO, monoclinic ZrO₂ and tetragonal ZrO₂ are found to be most common constituents of the oxide layers developed on the alloys. Integrating the microstructural, mechanical and corrosion properties, ZrFeCrNi3 (Zr: 84.00, Fe: 11.20, Cr: 3.20, Ni: 1.60, in wt.%) is identified as the acceptable base alloy for disposal of radioactive hulls.     

Effect of thermal exposure time on tellurium-induced embrittlement of Ni–16Mo–7Cr–4Fe alloy

The embrittlement of nickel-based structural alloys by fission-produced tellurium(Te) is a major challenge for molten salt reactors(MSR). In this study, the effects of thermal exposure time on tellurium diffusion in a candidate MSR structural alloy(Ni–16 Mo–7 Cr–4 Fe) and the consequent mechanical property degradation of the alloy were investigated through surrogate diffusion experiments at 700 °C. The results show that some tellurium reacted with the alloy to form tellurides on the surface,while some tellurium diffused into the alloy along grain boundaries. Ni_3Te_2 and CrTe were the most stable reaction products at the tested temperature, and the formation of CrTe on the surface induced the Cr depletion at grain boundaries of the alloy. The diffusion depth of Te increased gradually with thermal exposure time, and thediffusion rate kept stable within the test duration of up to3000 h. The Te diffusion in the alloy caused the embrittlement of grain boundaries, inducing crack formation and strength degradation in tensile test at room temperature.     

Enhanced radiation tolerance of ultrafine grained Fe–Cr–Ni alloy

The evolutions of microstructure and mechanical properties of Fe–14Cr–16Ni (wt.%) alloy subjected to Helium ion irradiations were investigated. Equal channel angular pressing (ECAP) process was used to significantly reduce the average grain size from 700 μm to 400 nm. At a peak fluence level of 5.5 displacement per atom (dpa), helium bubbles, 0.5–2 nm in diameter, were observed in both coarse-grained (CG) and ultrafine grained (UFG) alloy. The density of He bubbles, dislocation loops, as well as radiation hardening were reduced in the UFG Fe–Cr–Ni alloy comparing to those in its CG counterpart. The results imply that radiation tolerance in bulk metals can be effectively enhanced by refinement of microstructures.     

Analytical developments in the Wong–Fung–Tam–Gao radiation model of thermal diffusivity

When the thermal diffusivity, χ, of a thin film on a substrate is measured by means of the mirage method, the photothermal deflection of the probe beam is determined by the heat radiation field contributed by the film and the substrate, heated by the pump beam. A two-dimensional algorithm is here presented in order to deduce the measure of the diffusivities of the film and the substrate in one set of mirage detection from the experimental data.     

Influence of precipitate density on the nodular corrosion resistance of Zr–Sn–Fe–Cr alloys at 500°C

In a Zr–1.3% Sn base alloy, both the addition of increasing amounts of iron and chromium, conserving a constant Fe/Cr ratio, and the reduction of the cumulative annealing parameter ΣA have beneficial effects on the corrosion resistance in 500°C steam. It is shown that these two observations can be rationalized by considering that the important metallurgical factor is the number of precipitates per unit volume rather than their size.     

Thermodynamic assessments of the Al–Th and Th–Zn systems

The phase diagrams of the Al–Th and Th–Zn systems have been evaluated by using the Calculation of Phase Diagrams (CALPHAD) method with the experimental data including the phase equilibria and thermodynamic properties. The Gibbs free energies of the liquid, bcc and hcp phases were described by the subregular solution model with the Redlich–Kister equation, and those of the stoichiometric compounds of the Th₂Al, Th₃Al₂, ThAl, Th₂Al₃, ThAl₂, ThAl₃, Th₂Al₇, Th₂Zn, ThZn₂, ThZn₄ and Th₂Zn₁₇ were described by the two-sublattice model. The calculated phase equilibria and thermodynamic properties are in good agreement with the experimental data.     

Thermal and mechanical properties of hydrides of Zr–Hf alloys

Polycrystalline bulk samples of δ-phase Hf hydrides with various Zr contents were prepared and their high-temperature stability and thermal and mechanical properties were investigated. The phase structure was examined between room temperature and 973 K using high-temperature X-ray diffraction and thermogravimetric–differential thermal analysis. From room temperature to 673 K, the coefficient of linear thermal expansion, specific heat capacity, and thermal conductivity were evaluated. The Vickers hardness and sound velocity were measured at room temperature, and the elastic modulus was evaluated. The effect of the Zr content on the high-temperature stability and the thermal and mechanical properties of Hf hydrides was studied.     

Measurement of Irradiation Creep Coefficient of Fine–Grained Isotropic Graphite

In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and high-pressurization-rate burst tests in the Japan Atomic Energy Research Institute (JAERI).     

Electrorefining of nickel from nickel–chromium alloy in molten LiCl–KCl

Electrorefining of nickel in LiCl–KCl melt was investigated using electrochemical techniques. Nickel products after electrorefining were characterized by X-ray diffraction, X-ray fluorescence, and scanning electron microscopy. Both cyclic voltammetry and square wave voltammetry results suggested that Ni~(2+) was directly reduced to Ni metal in Li Cl–KCl. Based on a preliminary study on the electrochemical behavior of nickel and chromium, electrorefining was carried out under constant potential, whereupon deposits were formed on the cathode.The purity of nickel increased from 72.62% in the original alloy to 99.83% in cathodic deposits, as determined by inductively coupled plasma atomic emission spectroscopy analysis. Almost all the nickel in the alloy could be recovered during the electrochemical process with [ 90%current efficiency. A lower concentration of Ni Cl_2 in Li Cl–KCl was found to be favorable for nickel electrorefining, as increased Ni Cl_2 concentration caused severe corrosion of the nickel anode at the gas–liquid interface due to the accumulation of Cl_2 gas.     

Thermodynamic modelling of the C–U and B–U binary systems

The thermodynamic modelling of the carbon–uranium (C–U) and boron–uranium (B–U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium–concrete interaction (MCCI). The key O–U–Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe–U, Cr–U and Ni–U were modelled as a preliminary work for modelling the O–U–Zr–Fe–Cr–Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver–indium–cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B₄C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B₂O₃ and C with the major components of TDBCR, O–U–Zr–Fe–Cr–Ni–Ag–In–Ba–La–Ru–Sr–Al–Ca–Mg–Si + Ar–H. The critical assessment of the very numerous experimental information available for the C–U and B–U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.     

Numerical simulation of gas–liquid–solid three-phase flow using particle methods

We want to simulate, based on particle methods, the dynamic behavior of multi-phase flows in a gas–solid–liquid mixture system. With the governing equations discretized within the finite volume particle method, the effects of contact and collision between solid particles were modeled by the distinct element method. Applicability of the viscosity model and an empirical drag force model were confirmed for the hydrodynamic interactions between solid particles and fluid. Simulations were performed of a single bubble rising in a tank of stagnant solid particle–liquid. The results for the dynamic behavior indicate that the present computational framework of particle-based simulation method may be useful for numerical simulations of multi-phase flow behavior in a solid particle–fluid mixture system.     

Injection casting of U–Zr and U–Zr–RE fuel slugs and their characterization

U–Zr fuel slugs containing rare-earth elements can be difficult to cast because of the high reactivity of rare-earth elements. In this study, U–Zr and U–Zr–RE (RE: a rare-earth alloy comprising 53% Nd, 25% Ce, 16% Pr, and 6% La by weight) fuel slugs were prepared by injection casting, and their characteristics were evaluated. The as-cast fuel slugs were fabricated to the full length of the mold, and they showed no thin sections or cracks. Compared to the theoretical density, the measured density of the U–Zr and U–Zr–RE fuel slug was lower and higher, respectively. Chemical analysis revealed that the Zr and RE compositions in the U–10Zr and U–10Zr–3RE fuel slugs matched the target composition within 1.0 wt%. However, the RE composition in the U–10Zr–7RE fuel slug differed from the target composition by over 4 wt%. The melting crucible was further deteriorated and the casting yield was lower for the casting of a high rare-earth bearing fuel slug.     

Aging behavior of large area MCP–PMT

The Institute of High Energy Physics, Chinese Academy of Sciences have designed a new type of photomultiplier tube(PMT) based on microchannel plates(MCPs) with large area photocathode, known as large area microchannel plate photomultiplier tube(MCP–PMT). The aging characteristics of the large area MCP–PMTs are different from dynode PMTs and small proximity-focus MCP–PMT. In this work, a prototype large area MCP–PMT was aged by operating with nearly 1000 photoelectrons per pulse for 3 months, and aging process of the MCP–PMT was discussed based on the aging curve.     

Fundamental experiments on phase stabilities of Fe–B–C ternary systems

To understanding the control blade degradation mechanism of a boiling-water reactor (BWR), a thermodynamic database for the fuel assembly materials is a useful tool. Although the iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data are not sufficient for the region in which boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples such as Fe_0.68B_0.06C_0.26 (at%), Fe_0.68B_0.16C_0.16 (at%), and Fe_0.76B_0.06C_0.18 (at%) were analyzed by X-ray diffraction, scanning electron microscopy, and energy–dispersive X-ray spectrometry. The results indicate that the Fe₃(B,C) phase exists only in the intermediate region at 1273 K and that the solidus temperature widely maintains at approximately 1400 K for all three samples; these results differ from the calculated data using the previous thermodynamic database. The difference might originate from the overestimation of the interaction parameter between boron and carbon in Fe₃(B,C). Proper titling was performed using the present data.     

Tensile properties of K-doped W–3%Re

In this work, the tensile properties of K-doped W–3%Re were investigated. This material was fabricated by powder metallurgy and hot rolling on an industrial scale. It is expected that there would be improvement of the high-temperature strength, an increase of the recrystallization temperature, and a decrease in the ductile–brittle transition temperature (DBTT) of pure tungsten due to the dispersion of K bubbles and the addition of 3% Re. In addition, suppression of the formation of irradiation-induced defect clusters is also expected. Tensile tests in the temperature range from room temperature to 1800 °C were conducted. After the tensile tests, fracture surface observations were carried out using a scanning electron microscope (SEM). The tensile strength decreased with increasing test temperature. Elongation of K-doped W–3%Re was observed above 500 °C. The results of fracture surface observation showed that delamination of the layered structure occurred at 500, 700, and 900 °C and cracking along the grain boundaries occurred at 1500 and 1800 °C.     

Numerical simulations of gas–liquid–particle three-phase flows using a hybrid method

For the analysis of debris behavior in core disruptive accidents of liquid metal fast reactors, a hybrid computational tool was developed using the discrete element method (DEM) for calculation of solid particle dynamics and a multi-fluid model of a reactor safety analysis code, SIMMER-III, to reasonably simulate transient behavior of three-phase flows of gas–liquid–particle mixtures. A coupling numerical algorithm was developed to combine the DEM and fluid-dynamic calculations, which are based on an explicit and a semi-implicit method, respectively. The developed method was validated based on experiments of water–particle dam break and fluidized bed in systems of gas–liquid–particle flows. Reasonable agreements between the simulation results and experimental data demonstrate the validity of the present method for complicated three-phase flows with large amounts of solid particles.     

Effect of He-pre-injection on dislocation loop formation and irradiation-induced segregation of Fe–12%Cr–15%Mn austenitic steel

The effect of He-injection on irradiation-induced segregation of aging treated Fe–12%Cr–15%Mn austenitic steels, which are candidate materials as the reduced radio-activation of structure material for nuclear and/or fusion reactors was investigated by using the 1250 kV high voltage electron microscope (HVEM) connected with an ion accelerator. The Fe–Mn–Cr steel has been irradiated at 573 K by three irradiation modes of single electron-beam irradiation, electron-beam irradiation after He-injection and electron/He⁺-ion dual-beam irradiation in a HVEM. Irradiation-induced segregation analyses were carried out by an energy dispersive X-ray analyzer (EDX) in a 200 kV FE-TEM with beam diameter of about 0.5 nm. Dislocation loops with strain contrast were formed during irradiation and the loop numbers density increased rapidly with irradiation dose for He-pre-injected specimens. Voids were not observed after irradiations with three irradiation modes up to 5.4 dpa at 573 K. Irradiation-induced segregations of Cr and Mn near grain boundary were observed in each irradiation condition, but the amounts of Mn segregation decreased in the cases of electron/He⁺-ion dual-beam irradiation compared with single electron-beam and electron-beam irradiation after He-injection conditions.     

Gamma and X-ray radiation compatibility of Ti–Ta–Hf–Zr alloys used for coronary stent applications

We have studied the gamma and X-ray radiation compatibility of Ti-based alloys such as Ti–37 Ta–26 Hf–13 Zr-24(wt%) [Alloy 1], Ti–40 Ta–22 Hf–11.7 Zr-26.3(wt%) [Alloy 2], Ti–45 Ta–18.4 Hf–10 Zr-26.6(wt%) [Alloy3], Ti–50 Ta–15 Hf–8 Zr-27(wt%) [Alloy 4], Ti–55 Ta–12 Hf–7 Zr-26(wt%) [Alloy 5], and Ti–60 Ta–10 Hf–5 Zr-25(wt%) [Alloy 6]. Gamma and X-ray radiation compatibility is studied by evaluating the mass attenuation coefficient,mean free path, HVL, TVL effective atomic number,effective electron density, exposure buildup factor, and relative dose. We have compared these parameters for studied alloys with that of arteries. The alloys Ti–55 Ta–12 Hf–7 Zr-26 and Ti–60 Ta–10 Hf–5 Zr-25 have added properties such as gamma/X-ray radiation compatibility,high elastic admissible strain, high mechanical strength,and excellent biocompatibility. Hence, we may suggest that, among Ti–Ta–Hf–Zr alloys, these alloys are best materials for coronary stent applications.