U–Zr–Fe–O Melt density

A method of evaluating the density of U–Zr–Fe–O melts, which is of interest for predicting the spatial configuration of stratified melt during a serious accident in water moderated and cooled reactors, has been proposed previously. In the present work, the density of Fe–U–Zr–O metallic melt at 2023 K, (U, Zr)O₂ melt with atomic ratio U/Zr = 0.9 at 2973 K, and zirconium oxide melt at 3073 K are determined experimentally in order to verify the proposed model. The agreement between the calculations and the experimental results is satisfactory as a whole; this makes it possible to recommend the proposed method for evaluating the spatial configuration of the melt in a VVER vessel during a serious accident.     

The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2 pseudo-ternary system

Specimens of (U, Pu, Zr)O₂ were prepared as simulated corium debris that were assumed like debris generated in the severe accident of the Fukushima Daiichi Nuclear Power Plant and their melting temperatures were measured by the thermal arrest technique in order to evaluate the influence of plutonium and zirconium content on the melting temperature of the corium debris. From the evaluation, it was found that the influence of zirconium on the melting temperatures of both (U, Pu, Zr)O₂ and (U, Zr)O₂ was similar and that the melting temperature of (U, Pu, Zr)O₂ had a local maximum value in the Pu-content between 0 and 20 mol%. The UO₂–PuO₂–ZrO₂ pseudo-ternary phase diagram at 2900 and 3000 K was evaluated from the present experimental results and previously reported results.     

Development of thermodynamic databases in the system U–Zr–Ce–Cs–Fe–B–C–I–O–H for application to simulating phase equilibria in severe nuclear accidents

In case of severe nuclear accidents involving melt down of nuclear fuels at high temperatures, it is of considerable importance to accurately evaluate the highly-volatizing behavior of fission products (FPs) over multicomponent debris. Particularly, cesium (Cs)- and iodine (I)- bearing chemical species are regarded as notable FPs. In the present work, the authors have generated original thermodynamic databases for the system U–Zr–Ce–Cs–Fe–B–C–I–O–H featuring Cs- as well as I-bearing subsystems, which are contained in oxide, iodide, and metal (including borides and carbides) sub-databases. It has been confirmed that the phase diagrams calculated by the present set of the databases reproduce the corresponding literature data well in various kinds of subsystems of the above multicomponent system. The present set of databases has subsequently been applied to simulate phase equilibria and volatizing behavior of Cs- and I-including species, respectively, in multicomponent debris under specific temperature and atmospheric conditions corresponding to severe nuclear accidents.     

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.     

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.     

Development of Thermodynamic Database for U—Zr—Fe—O—B—C—FPs System

A thermodynamic corium database using ionic two-sublattice model for liquid was developed and stratification of molten corium, supposed to occur in in-vessel retention accident management, was analyzed. The database consists of U—Zr—Fe—O—C—B–(FP oxides) system. Fundamentally, data were obtained from existing assessed databases, such as SGTE's. The liquid phase data were reconstructed based on the ionic model and lacking data including excess energies were assessed to be consistent with existing phase diagrams. Liquidus temperatures measured under OECD RASPLAV project were analyzed with the database. In addition, an analysis of corium under a severe accident condition was carried out and demonstrates that the database gives an improved method based on thermodynamics to analyze the corium stratification.     

Simulation of multi-atomic interactions in H–O–W system with the MD code CADAC

For future tokamak reactors, chemical erosion of tungsten armour surfaces under impact of hot deuterium–tritium plasma that contains impurities, for instance oxygen, is an important issue. Oxygen can form volatile molecular complexes O_xW_y at the surface, and the retained H-atoms form the volatile complexes H_xO_y, which mitigates the erosion (H states for hydrogen isotopes). The plasma impact can substantially destroy the complexes.To describe this H–O–W system, the molecular dynamics (MD) code CADAC was earlier developed using only pair–atomic interactions. Now CADAC is extended for multi-body forces to simulate molecular organization of atoms near the tungsten surface. The approach uses the Abell's model of empirical bond-order potentials in addition combined, for the first time, with a valence concept. CADAC simulates chemical features using atomic valences and the Morse potentials. The new model is introduced and model parameters are estimated.     

Creep studies on U–7%Zr,U–7%Nb and U rich U–Nb–Zr alloys

Creep behavior of U–7%Zr, U–5%Zr–2%Nb, U–3.5%Zr–3.5%Nb, U–2%Zr–5%Nb, U–7%Nb alloys (composition in wt.%) was investigated by the impression creep technique at 630 and 700 °C at a stress of 22.2 MPa. Creep rate was found to be two orders lower in the binary U–7%Nb than the U–7%Zr alloy. In ternary U–Nb–Zr alloys, the creep rate was found decreasing drastically with the increase in Nb content.     

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.     

Structural studies of the phase separation in the UO2–PuO2–Pu2O3 ternary system

In the oxygen hypo-stoichiometric range of (U_1?yPu_y)O_2?x mixed oxide MOX fuels, the U–Pu–O phase diagram is known to exhibit a large biphasic domain depending on the Pu content. However, the phase equilibria are still to be fully described as various representations are proposed in the literature.In the present work, we notify new insights into the phase separation occurring in the UO₂–PuO₂–Pu₂O₃ domain at room temperature. Our microstructural and X-ray diffraction results are compared to the different representations reported in the literature. We provide, for the first time in the hypo-stoichiometric domain, an indisputable experimental observation of a triphasic region at high Pu content, composed of two fluorite-type structures and of one α-Pu₂O₃ sesquioxyde type structure. These results are in contradiction with previous experimental representations of the U–Pu–O ternary system.     

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.     

Understanding the phase equilibrium and irradiation effects in Fe–Zr diffusion couples

We have studied the radiation effects in Fe–Zr diffusion couples, formed by thermal annealing of a mechanically bonded binary system at 850 °C for 15 days. After irradiation with 3.5 MeV Fe ions at 600 °C, a cross sectional specimen was prepared by using a focused-ion-beam-based lift out technique and was characterized using scanning/transmission electron microscopy, selected-area diffraction and X-ray energy dispersive spectroscopy analyses. Comparison studies were performed in localized regions within and beyond the ion projected range and the following observations were obtained: (1) the interaction layer consists of FeZr₃, FeZr₂, Fe₂Zr, and Fe₂₃Zr₆; (2) large Fe₂₃Zr₆ particles with smaller core particles of Zr-rich Fe₂Zr are found within the α-Fe matrix; (3) Zr diffusion is significantly enhanced in the ion bombarded region, leading to the formation of an Fe–Zr compound; (4) grains located within the interaction layer are much smaller in the ion bombarded region and are associated with new crystal growth and nanocrystal formation; and (5) large α-Fe particles form on the surface of the Fe side, but the particles are limited to the region close to the interaction layer. These studies reveal the complexity of the interaction phase formation in an Fe–Zr binary system and the accelerated microstructural changes under irradiation.     

Characteristics of hydriding and hydrogen embrittlement of the Ti–Al–Zr alloy

The characteristics of hydriding and hydrogen embrittlement of the Ti–Al–Zr alloy were evaluated. The amount of hydrogen absorbed into the alloy at 500 °C was continuously monitored using a hydrogen pressure measurement. The rate of decrease in hydrogen pressure indicated a high absorption rate of hydrogen into the alloy, following a linear rate law. X-ray diffraction studies showed the formation of δ-phase titanium hydride (TiH_1.97) after hydriding. At room temperature, the alloy showed much sensitivity to embrittlement in ductility by hydrogen. The δ-hydrides in the grain boundaries promoted the crack propagation in the presence of stress, leading to the cleavage failure mode. However, the tensile strengths were almost independent of the hydrogen content up to 1174 ppm. It is thus concluded that the δ-hydride acts to decrease the ductility without affecting tensile strengths.     

Studies on physics parameters of metal (U–Pu–Zr) fuelled FBR cores

For faster growth of nuclear power in India, it is essential to shift to the use of metal-fuels in fast breeder reactors (FBR), which gives a higher breeding ratio (BR) and lower doubling time (DT). Also, future commercialization of the FBR fuel cycle necessitates the use of metallic fuel along with the pyro-process recycling, which can be less costly than oxide fuel reprocessing. Two-dimensional diffusion calculations have been performed to investigate the various physics parameters of metal (U–Pu–Zr) fuelled FBR cores as a function of reactor parameters like reactor power, smear density, zirconium content in the fuel and the number of rows in radial blankets. A 1000 MWe fast reactor with U–Pu fuel (i.e. metal-fuel with no zirconium – which is a theoretical possibility now, due to the lack of irradiation experience) can attain a breeding ratio of 1.61 and a reactor fuel doubling time of 6.6 yrs. Two methods to reduce the sodium void reactivity, which is high and positive in metal-fuelled FBR cores, are suggested.     

Evolution of microstructure and second-phase particles in Zr–Sn–Nb–Fe alloy tube during Pilger processing

Evolution of microstructure and second-phase particles (SPPs) in Zr–Sn–Nb–Fe alloy tube were investigated during Pilger process using electron backscatter diffraction, secondary electron and transmission electron microscopy imaging techniques. Results show that the Pilger rolled tubes present heterogeneous structures with the C axes of less deformed grains mostly concentrated in the axial direction. During the Pilger rolling, the increase of deformation caused weakening of linear distribution of second-phase particles. The mean diameters of the precipitates are in the range of 70–100 nm in all specimens, and the growth mechanism of SPPs follows second-order kinetics. The grain growth is controlled by Zener pinning in the Pilger rolling–annealing specimens. Clusters containing the Zr(Nb,Fe)₂ and β_Nb precipitates formed in the Zr–1.0Sn–1.0Nb–0.12Fe alloy. Most of the particles located in grain boundaries are the Zr(Nb,Fe)₂ Laves phase with hexagonal structure, and stacking faults have been found in the Zr(Nb,Fe)₂ precipitates. The types, morphology and distribution of precipitates depend on the constituent and structural fluctuations of the nucleation area.     

A calorimetric study of high temperature phase stability in Fe–U alloys

A differential scanning calorimetry study of high temperature phase equilibria and phase transformation kinetics in Fe_100?xU_x binary alloys, with x varying from 0 to 95 mass% U is undertaken. Accurate measurement of transformation temperatures pertaining to: (i) α-Fe → γ-Fe → δ-Fe polymorphic phase change, (ii) UFe₂ + γ-Fe → L and U₆Fe + UFe₂ → L transformations and (iii) melting has been made as a function of uranium concentration. The measured transformation temperatures are used to construct the binary Fe?U phase diagram, which showed general agreement with the latest assessment. The L → UFe₂ + γ-Fe eutectic reaction is found to occur at 1357 ± 5 K, with the eutectic composition of 47 mass%. The heat of transformation associated with this invariant reaction is estimated to be 19,969 ± 1736 J mol of atoms^?1. The L → U₆Fe + UFe₂ reaction occurs at 89 mass%, and at 1001 ± 5 K, with a heat of transformation 20,250 ± 2113 J mol of atoms^?1. The heat of melting of stoichiometric UFe₂ is estimated to be 20,983 ± 2098 J mol of atoms^?1, which is higher than the currently assessed value by 30%. A non-faceted microstructural morphology is found to accompany the eutectic solidification process of all the alloy compositions.     

A study of the structure of Zr(CrFe)2 laves phase precipitates in a ZrCrFe alloy

C15 type Zr(CrFe)₂ Laves phase precipitates have been found in Zr-1.15 wt% Cr-0.1 wt% Fe alloy. Twinned, multiple twinned and dislocation structures have been found in the precipitates. Comparison of calculated and measured precipitate size show the growth of the Zr(CrFe)₂ Laves phase is controlled by diffusion.The orientation relationships (1̄11̄)_L//(112̄0)_α, [110]_L//[0001]_α between the Zr(CrFe)₂ Laves phase precipitates and ga-Zr matrix in the ZrCrFe alloy give the same type of model for the transformation as previously suggested for Zircaloy-4.     

Thermophysical properties of molten core materials: Zr–Fe alloys measured by electrostatic levitation

In order to investigate the progression of a core meltdown accident, it is necessary to understand the behavior of molten core materials. Zr–Fe alloys are one of the low-melting-temperature liquid phases that are thought to form in the early stages of bundle degradation. The objective of this study is to measure the thermophysical properties of Zr–Fe liquid alloys. Alloy samples with a composition of Zr_1?xFe_x (x = 0.12, 0.24, and 0.50) were synthesized by arc melting, and their density, viscosity, and surface tension were measured using an electrostatic levitation technique. The results indicate that the density of Zr–Fe liquid alloys can be estimated by a linear combination of the measured or extrapolated densities of pure Zr and Fe. The viscosities of the Zr–Fe liquid alloys can be roughly estimated by extrapolating those of Zr to lower temperatures, although this method tends to underestimate the viscosity of alloys, especially for eutectic compositions. The values of the Zr–Fe liquid alloys’ surface tensions are close to those of pure Zr.     

Creep behaviour of δ-phase of U-Zr system by impression creep technique

In this study, the impression creep behaviour of δ-phase of U-50 wt.% Zr (U-72.29 at.% Zr) system was studied in the temperature range 525-575 °C at different stresses. The velocity of the punch at different stresses and temperatures were evaluated for the above alloy. The stress exponents and thermal activation parameters of the above alloy were determined. A power law behaviour is displayed with the stress exponents range from 6.5 to 7. The activation enthalpy for the δ-UZr₂ was found to be independent of stress with an average value of 106 kJ/mol.