Effect of Nd and Pr addition on the thermal and mechanical properties of (U, Ce)O2

The thermal conductivity, Young’s modulus, and hardness of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.01, 0.08, 0.12) were evaluated and the effect of Pr and Nd addition on the properties of (U, Ce)O₂ were studied. The polycrystalline high-density pellets were prepared with solid state reactions of UO₂, CeO₂, Pr₂O₃, and Nd₂O₃. We confirmed that all Ce, Pr, and Nd dissolved in UO₂ and formed solid solutions of (U, Ce, Pr, Nd)O₂. We revealed that the thermal conductivity of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.12) was up to 25% lower than that of x = 0.01 at room temperature. The Young’s modulus of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ decreased with x, whereas the hardness values were constant in the investigated x range.     

Fabrication and characterization of uranium-thorium-zirconium hydrides

Two uranium-thorium-zirconium hydrides, (UTh₄Zr₁₀)H_1.9 and (U₄Th₂Zr₉)H_1.5, have been fabricated and characterized. Fabrication involved arc melting of the constituent pure metals to form homogenous alloys, followed by hydriding at elevated temperatures in a hydrogen gas environment. The compounds were characterized by X-ray powder diffractometry as well as scanning and transmission electron microscopy. These methods revealed a multi-phase mixture of δ-zirconium hydride (ZrH_1.6+x), thorium-zirconium hydride (ThZr₂H_7−x), and uranium metal. The elastic modulus was mapped across the microstructure using nanoscale dynamic stiffness mapping. The elastic modulus of ThZr₂H_7−x phase is found to be 172 GPa.     

Changes in hardness and elasticity of a Ti6Al4V alloy under helium irradiation

Ti₆Al₄V specimens have been irradiated at different temperatures with 200 keV He ions. Microhardness and elastic modulus of the unirradiated and irradiated specimens were measured by means of the nano-indentation technique and analyzed using the Oliver-Pharr method. The indentation depth of all samples is 700 nm, which is comparable in magnitude to the ion range. The subsurface structure of the Ti₆Al₄V specimens was investigated by the X-ray diffraction technique. The measurements indicate that the microhardness increased with the irradiation temperature from room temperature to 600 °C while the elastic modulus almost monotonically decreased. The Irradiation at 700 °C, however, caused softening and slight increase of the elastic modulus within the surface layer of the specimens. The hardening and reduction of the elastic modulus of the Ti₆Al₄V alloy under irradiation conditions used in this study is tentatively explained by a model based on the presence of point defects and dispersed obstacles of β-precipitates. The softening and slight increase of elastic modulus of helium-irradiated Ti₆Al₄V at 700 °C might be related to the coarsening of β-precipitates and formation of the hybrid γ-TiH phase in α-phase.     

Thermodynamic modeling of the Zr-H-Nb system on the Zr-rich side and application

A thermodynamic modeling of the Zr-H-Nb system on the Zr-rich side (wt.%Nb ? 2.5) was established using CALPHAD (CALculation of PHAse Diagrams) method. The calculated pressure-composition isotherms and the composition limits in the two-phase (bcc + fcc) region agree well with the experimental data. Meanwhile, some thermodynamic data of Zr-H-Nb system (wt.%Nb = 1, P_H2=101.325 kPa) was calculated, such as equilibrium hydrogen concentration, phase composition and phase transition temperature, which was applied to guide the preparation of the crack-free zirconium hydride. The hydriding of Zr-1 wt.%Nb alloys was carried out and the crack-free zirconium hydrides with hydrogen concentration of 1.6H/Zr (at.) and 1.8H/Zr were prepared respectively based on the calculated results.     

Synthesis and characterization of low-temperature precursors of thorium-uranium (IV) phosphate-diphosphate solid solutions

Several compositions of new precursor of thorium-uranium (IV) phosphate-diphosphate solid solutions (Th_4−xU_x(PO₄)₄P₂O₇, called β-TUPD) were synthesized in closed PTFE containers either in autoclave (160 °C) or on sand bath (90-160 °C). All the samples appeared to be single phase. From XRD data and TEM observations, the diffraction lines matched well with that of pure thorium phosphate-hydrogenphosphate hydrate (TPHPH), Th₂(PO₄)₂(HPO₄) · H₂O, which confirmed the preparation of a complete solid solution between pure thorium and uranium (IV) compounds. TGA/DTA experiments showed that samples of thorium-uranium (IV) phosphate-hydrogenphosphate hydrate (TUPHPH) prepared at 150-160 °C were monohydrated leading to the proposed formula Th_2−x/2U_x/2(PO₄)₂(HPO₄) · H₂O. The variation of the XRD diagrams versus the heating temperature showed that TUPHPH remained crystallized and single phase from room temperature to 200 °C. After heating between 200 °C and 800 °C, the presence of diphosphate groups in the solid was evidenced. In this range of temperature, the solid was transformed into the low-temperature monoclinic form of thorium-uranium (IV) phosphate-diphosphate (α-TUPD). This latter compound finally turned into well-crystallized, homogeneous and single-phase β-TUPD (orthorhombic form) above 930-950 °C for x values lower than 2.80. For higher x values, a mixture of β-TUPD, α-Th_1−zU_zP₂O₇ and U_2−wTh_wO(PO₄)₂ was obtained. By this new chemical route of preparation of β-TUPD solid solutions, the homogeneity of the samples is significantly improved, especially considering the distribution of thorium and uranium.     

Comparison of TSSD results obtained by differential scanning calorimetry and neutron diffraction

Due to their low absorption cross-section for neutrons, Zr alloys are used for reactor core components. The terminal solid solubility (TSS) for hydrogen in these alloys is very low - in Zr-2.5 wt% Nb, used to fabricate pressure tubes for CANDU (CANDU-CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Ltd.) power reactors, the TSS is ∼0.7 at.% H at 300 °C. The mechanical properties of the components may deteriorate when their hydrogen concentration exceeds TSS. Therefore, accurate values of the TSS are needed to assess the operating and end-of-life behaviours of these components. Differential scanning calorimetry (DSC) is used to measure the TSS of hydrogen in Zr alloys. Three distinct features are marked on a typical DSC heat flow curve when the material is being heated and the hydrides are dissolving; ‘peak temperature’, ‘maximum slope temperature’ and ‘completion temperature’. Usually, the maximum slope temperature, being about the average of the three temperatures, is interpreted as the TSS temperature for hydride dissolution (T_TSSD). A set of coordinated DSC and neutron diffraction measurements have been carried out to identify the features of the heat flow signal that closely correspond to the T_TSSD. Neutron diffraction was chosen because hydrides generate distinctive diffraction peaks whose intensities approach zero at the transition temperature - an unambiguous indication of dissolution. Neutron diffraction shows that the temperature of hydride dissolution correlates closely with the DSC peak temperature.     

Fabrication and mechanical characterization of zirconium and gadolinium hydrides

We prepared three kinds of metal hydrides: Zr hydrides, Gd hydrides, and the hydrides of Zr-Gd alloys (Zr:Gd = 10:1, 8:1, 6:1), and characterized their mechanical properties. It was confirmed that the hydrides of Zr-Gd alloys were composed of Zr hydrides and Gd hydrides mixtures. We evaluated the Vickers hardness and the Young’s modulus of the hydrides. We succeeded in proposing empirical equations describing the density, Vickers hardness, and Young’s modulus of the hydrides of Zr-Gd alloys, as functions of the hydrogen content and the Gd content.     

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.     

Stress-reorientation of hydrides and hydride embrittlement of Zr-2.5 wt% Nb pressure tube alloy

Hydrogen in excess of the terminal solid solubility precipitates out as a brittle hydride phase in zirconium alloys. The hydrides acquire platelet shaped morphology due to their accommodation in the matrix and can cause severe embrittlement, especially when these are oriented normal to the tensile stress axis. The precipitation of hydride platelets normal to the tensile stress when cooled under stress from a solution-annealing temperature is commonly referred to as ‘stress-reorientation’. The stress-reorientation is associated with a threshold stress below which no reorientation is observed. In this work, stress-reorientation of hydrides was investigated for unirradiated, cold worked and stress-relieved Zr-2.5 wt% Nb pressure tube material for a reorientation temperature of 423-723 K. The effect of the reoriented hydrides on the tensile properties of the Zr-2.5 wt% Nb pressure tube alloy was evaluated in the temperature range of 298-573 K.     

Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.     

Ion-induced amorphization in ceramic materials

Amorphization induced by swift heavy ions is discussed, the main features are reviewed and explained by the author’s model. In Al₂O₃ and MgAl₂O₄ the recrystallization reduces considerably the track diameters. The threshold electronic stopping power for amorphization S_et can be estimated reliably for these solids from the position of the amorphous-crystalline boundary formed at high ion fluences. The results are in good agreement with the predictions of the model. Experiments on CeO₂ and UO₂ are discussed. Estimates of S_et are made for β-Si₃N₄, CeO₂, pure ZrO₂, ZrSiO₄, and UO₂, and 10.1 > S_et > 6.2 keV/nm is obtained at room temperature for fission fragment energies. At about 1000 °C operation temperature, S_et is reduced by about 35-50%. No amorphization is expected by ion bombardment in AlN, SiC (semiconductors) and MgO (ionic crystal).     

Thermal studies on fluorite type ZryU1−yO2 solid solutions

Solid state reactions of UO₂ and ZrO₂ in mild oxidizing condition followed by reduction at 1673 K showed enhanced solubility up to 35 mol% of zirconium in UO₂ forming cubic fluorite type Zr_yU_1−yO₂ solid solution. The lattice parameters and O/M (M = U + Zr) ratios of the solid solutions, Zr_yU_1−yO_2+x, prepared in different gas streams were investigated. The lattice parameters of these solid solutions were expressed as a linear equation of x and y: a₀ (nm) = 0.54704 − 0.021x - 0.030y. The oxidation of these solid solutions for 0.1 ? y ? 0.2 resulted in cubic phase MO_2+x up to700 K and single orthorhombic zirconium substituted α-U₃O₈ phase at 1000 K. The kinetics of oxidation of Zr_yU_1−yO₂ in air for y = 0-0.35 were also studied using thermogravimetry. The specific heat capacities of Zr_yU_1−yO₂ (y = 0-0.35) were measured using heat flux differential scanning calorimetry in the temperature range of 334-860 K.     

In situ ERDA measurements of hydrogen isotope concentrations in palladium at atmospheric pressure

In situ elastic recoil detection analysis (ERDA) measurements in gases at atmospheric pressure have been carried out using 15 MeV ⁴He ion beams. The beams are extracted through a molybdenum foil having a thickness of 5 μm. The maximum depth of analysis is about 4 μm for the palladium hydride and palladium deuteride (PdH_x and PdD_x, x = 0.7-0.8) samples. The temperature of the samples rises stepwise from room temperature to 180 °C. ERDA spectra are obtained every 2 min. Hydrogen and deuterium in the samples are discharged in the temperature range of 120-140 °C in a vacuum. Decrease in the hydrogen concentration in the PdH_x sample heated in a vacuum follows a first order in the value of x and an apparent activation energy of discharge of hydrogen is 1.05 eV. On the other hand, the hydrogen and deuterium concentrations decrease at about 80 °C in air. No isotope effects are observed in both a vacuum and air. The temperature at which the hydrogen concentration decreases in helium gas is almost the same as that in a vacuum. It indicates that hydrogen and deuterium atoms are discharged by chemical reactions with air and that there are no effects of cooling of the thermocouple by convection of air.     

Corrosion behaviour of steels in lead-bismuth at 823 K

The corrosion behaviour of the martensitic T91 steel and the austenitic AISI 316L steel was analysed. The steels were immersed in stagnant molten Pb-55.2wt%Bi alloy at 823 K for different exposure times (t = 550-2000 h). The corrosion tests were carried out both under Ar and under Ar-5%H₂ mixture. Under the oxidising conditions (P_O2 = 6 × 10⁻³ Pa), the formation of oxide layers was observed which prevent the penetration of the liquid alloy into the matrix, while under the Ar-5%H₂ mixture (P_O2 = 3.2 × 10⁻²³ Pa), two phenomena occurred: a ‘reactive penetration’ at the liquid alloy/steel interface and the competition between oxidation and penetration.     

Dynamic Young’s moduli of tungsten and tantalum at high temperature and stress

Recently reported results of the long lifetime of the tungsten samples under high temperature and high stress conditions expected in the Neutrino Factory target have strengthened the case for a solid target option for the Neutrino Factory. In order to study in more detail the behaviour of the material properties of tungsten, a dynamic method has been used for measurement of Young’s modulus at high stress, high-strain-rates (>1000 s⁻¹) and very high temperatures (up to 2650 °C). The method is based on measurements of the surface vibration of thin wires, stressed by a pulsed current, using a Laser Doppler Vibrometer. The measured characteristic frequencies under the thermal excitation have been used to obtain Young’s modulus as a function of applied stress and temperature. The same procedure has been used to measure Young’s modulus of tantalum up to 2500 °C.     

Investigation of radiation swelling in the zirconium-hydrogen system

Conclusions The radiation swelling of the hydrides ZrH_1.78–ZrH_1.90 at a temperature of 50°C is due to the accumulation of zirconium vacancies and their fine pile-ups. At 430–560°C, swelling is determined by the accumulation of fine (10Å) vacancy complexes. In this case, micropores are also formed, with a diameter of 50 Å, but their total volume does not exceed one-tenth of the macroswelling.The considerable buildup of vacancies in the hydrides studied is related with the substructure of the hydride, formed by dislocation voids and the boundaries of twins, which promote an increase of concentration and dispersivity of the injection sinks or themselves serve as these sinks. In samples of the hydride obtained by dehydrogenation of the phase, the numerous dislocations remaining in the hydride after decay of the -phase substructure obviously fulfill this function. The weak tendency of the zirconium vacancies in the hydride to amalgamation in the pores during increased irradiation temperature (by comparison with the behavior of vacancies in metals) is explained by the reduction of their surface energy as a result of interaction with hydrogen atoms.The displacement of hydrogen atoms from the tetrahedral positions during irradiation and their return during annealing changes the periods of the crystal lattice of zirconium hydride in the same way as the change of hydrogen content, which enables the number of defects of the hydrogen sublattice to be estimated.Translated from Atomnaya Énergiya, Vol. 42, No. 1, pp. 16–19, January, 1977.     

Structural-phase transformations in near-surface layers of alumina-zirconium ceramics induced by low-energy high-current electron beams

The effect of low-energy high-current electron beams (LEHCEBs) at E = 15 keV on mechanical characteristics of near-surface layers of alumina-zirconium ceramics (AZC) of the 20Al₂O₃-80ZrO₂(Y) composition (in mass %) is investigated by the method of dynamic indentation. It is shown that electron-beam treatment reduces Young’s modulus and nanohardness of this ceramic material. The action of LEHCEBs on the structural-phase state of modified subsurface layers of AZC specimens is analyzed. Their elemental and phase compositions are examined using several techniques: electron-probe X-ray spectral microanalysis (EPMA), secondary-ion mass spectrometry (SIMS), and X-ray diffractometry. It is found out that irradiation of the ceramic material by LEHCEBs stimulates the transitions of an m-phase of zirconium dioxide into a tetragonal modification, and results in a considerable decrease in the alumina phase. Based on the analysis of the data obtained, a conclusion is drawn that it is the processes resulting from irradiation and giving rise to formation of a subsurface layer with a phase composition different from that of the bulk material, which are responsible for the effect of modification of the material mechanical property.     

Influence on the structural characteristic of defect solid solution GdxZr1−xO2−x/2 with 0.18 ? x ? 0.62 under longer term annealing studied by Gd Mössbauer spectroscopy

Three kinds of defect solid solution Gd_xZr_1−xO_2−x/2 with 0.18 ? x ? 0.62, including the three single crystal samples with x = 0.21, 0.26 and 0.30, were investigated by ¹⁵⁵Gd Mössbauer spectroscopy at 12 K. Difference in the structural characteristic under longer term annealing were confirmed by comparing the ¹⁵⁵Gd Mössbauer parameters of the polycrystalline samples sintered one time and twice at 1773 K for 16 h in air, respectively. The results indicated that the polycrystalline samples sintered twice have relatively equilibrated structure by comparing with the three single crystal samples. After being sintered twice, basically the local structure around the Gd³⁺ ions does not change, but the degree of the displacements of the six 48f oxygen ions from positions of cubic symmetry becomes slightly smaller, and distribution of the Gd³⁺ ions in the system becomes more homogeneous.     

Investigation of high temperature phase stability, thermal properties and evaluation of metallurgical compatibility with 304L stainless steel, of indigenously developed ferroboron alternate shielding material for fast reactor applications

Towards the cause of serving economic power production through fast reactors, it is necessary to bring in functionally more efficient and innovative design options, which also includes exploration of cheaper material alternatives, wherever possible. In this regard, the feasibility of using a commercial grade ferroboron alloy as potential alternate shielding material in the outer subassemblies of future Indian fast reactors has been recently investigated from shielding physics point of view. The present study explores in detail the high temperature thermal stability and the metallurgical compatibility of Fe-15.4B-0.3C-0.89Si-0.17Al-0.006S-0.004P-0.003O (wt.%) alloy with SS 304L material. In addition, the high temperature specific heat and lattice thermal expansion characteristics of this alloy have also been investigated as a part of the present comprehensive characterisation program. The Fe-15 wt.%B alloy is constituted of principally of two boride phases, namely tetragonal Fe₂B and orthorhombic FeB phases, which in addition to boron also contains some amount of C and Si dissolved in solid solution form. This Fe-B alloy undergoes a series of phase transformation as a function of increasing temperature; the major ones among them are the dissolution of Fe₂B-lower boride in the matrix through a eutectic type reaction, which results in the formation of the first traces of liquid at 1500 K/1227 °C. This is then followed by the dissolution of the major FeB boride phase in liquid and the melting process is completed at 1723 K/1450 °C. In a similar manner, the thermal stability studies performed on combined Fe-B + 304L steel reaction couples revealed that a pronounced pre-melting or liquid phase formation occurs at a temperature of 1471 K/1198 °C, which is lower than the melting onset of both Fe-B and SS 304L. It is found that within the limits of experimental uncertainty, this pre-melting phenomenon occurred at the same fixed temperature of 1471 K/1198 °C, irrespective of the mass ratios of Fe-B and 304L steel. Further, it is also found that SS 304L is completely soluble in Fe-B alloy and the fused product upon solidification formed a mixture of complex intermetallic borides, such as (Fe,Cr)(B,C), (Fe,Cr)₂(B,C) and (Fe,Ni)₃B. In the temperature range 823-1073 K (550-800 °C), the SS 304L clad is found to interact strongly with the Fe-B alloy. The diffusion layer thickness or the attack layer depth (x) is found to vary with time (t) up to about 5000 h, according to the empirical rate law, x² = k(T)t. The temperature sensitivity of the rate constant, k(T) is found to obey the Arrhenius law, k(T) = k_o exp(−Q/RT), with Q = 57 kJ mol⁻¹, being the effective activation energy for the overall diffusional interaction of Fe-B and SS 304L. The room temperature specific heat capacity of Fe-B alloy is found to be 538 kJ kg⁻¹ K⁻¹. The C_P values measured over 300-1350 K, is found vary smoothly with temperature according to the expression, C_P/kJ kg⁻¹ K⁻¹ = 0.62094 + 0.00012T + 10685.81T⁻². The lattice thermal expansion of both FeB and Fe₂B phases are found to be anisotropic in that the c-axis expansion is found to be more than that along a and b axes. The room temperature volume thermal expansivity of FeB and Fe₂B phases are found to be of the order of 48 × 10⁻⁶ K⁻¹ and 28 × 10⁻⁶ K⁻¹, respectively. The thermal expansion of FeB is found to be more temperature sensitive than that of Fe₂B.     

Identification and quantification of hydride phases in Zircaloy-4 cladding using synchrotron X-ray diffraction

Zirconium hydrides precipitate in fuel cladding alloys as a result of hydrogen uptake from the high-temperature corrosion environment of light water reactors. Synchrotron X-ray diffraction was performed at room temperature on stress-relieved Zircaloy-4 cladding with two distributions of hydrides - (1) uniformly distributed hydrides across the entire cladding wall and (2) hydride rim next to the outer surface. The δ-hydride phase was found to be the predominant hydride phase to precipitate for hydrogen contents up to 1250 weight parts per million (wt ppm). At a higher content, about 3000 wt ppm, although δ-hydride is still the majority phase, a significant amount of γ-hydride is also observed. At even higher hydrogen contents, in excess of approximately 6000 wt ppm, such as can occur in a highly dense hydride layer, peaks associated with the ε-hydride phase are also observed in the diffraction pattern. The volume fraction of hydrides was estimated as a function of hydrogen content using the integrated intensities of select diffraction peaks corresponding to the α-Zr matrix and the hydride phases. These estimated values agree well with calculated values from the independently measured concentrations. The results of this study indicate that hydride precipitation in Zircaloy-4 is a complex process of evolving hydride phases with increasing local hydrogen content.