The crystal structure of oxides grown on Zr–20Nb alloy

Oxides that were grown on Zr–20Nb in water at 300°C for 3 d, or in air at 400°C for 2 h were characterized by analytical electron microscopy. In both oxides, a similar microstructure was observed and similar electron diffraction patterns and high resolution lattice images were obtained. Analyses of the results showed that the crystal structure of the oxides was identical to that of an incommensurate modulated Nb₂Zr_x−2O_2x+1 phase, with x ≈ 10.     

Recombination mechanism of point defect loss to coherent precipitates in alloys under irradiation

A new mechanism of defect loss by enhanced recombination inside coherent precipitates in alloys under irradiation is described. The mechanism is examined quantitatively to find the microstructural parameters responsible for resistance to dimensional instability. The proposed model explains why radiation properties of Zr–Nb alloys depend on density of fine-grained precipitates of β_Nb-phase.     

Enthalpy and heat capacity of LiAlO2 between 298 and 1700 K by drop calorimetry

The enthalpy of γ-LiAlO₂ was measured between 403 and 1673 K by isothermal drop calorimetry. The smoothed enthalpy curve between 298 and 1700 K results in H⁰(T) − H⁰(298 K)=−37 396 + 93.143 · T + 0.00557 · T² + 2 725 221 · T⁻¹ J/mol. The standard deviation is 2.2%. The heat capacity was derived by differentiation of the enthalpy curve. The value extrapolated to 298 K is C_p,298=(65.8 ± 2.0) J/K mol.     

Effects of hydrogen content and temperature on fracture toughness of Zircaloy-4

The influence of hydrogen content and temperature on the fracture toughness of a Zircaloy-4 commercial alloy was studied in this work. Toughness was measured on CT specimens obtained from a rolled material. The analysis was performed in terms of J-integral resistance curves. The specimens were fatigue pre-cracked and hydrogen charged before testing them at different temperatures in the range of 293–473 K. A negative influence of the H content on material toughness was important even at very small concentrations, being partially restored when the test temperature increased. Except for some specimens with high H concentration tested at room temperature, the macroscopic fracture behaviour was ductile. The role of Zr-hydrides and Zr(Fe,Cr)₂ precipitates in the crack growth and the dependence with hydrogen content were analysed by observation of the fracture surfaces and determination of the Zr(Fe,Cr)₂ precipitates density on them.     

Thermal stability and brazing characteristics of Zr0.7−xMxBe0.3 (M=Ti or Nb) ternary amorphous filler metals

The effects of Ti or Nb substitution on the thermal stability and brazing characteristics of Zr_0.7−xM_xBe_0.3 (M=Ti or Nb) ternary amorphous alloys were investigated in order to improve properties of Zr–Be binary amorphous alloy as a new filler metal for joining zirconium alloy. The Zr_0.7−xM_xBe_0.3 (M=Ti or Nb; 0x0.1) ternary amorphous alloys were produced by melt-spinning method. In the selected compositional range, the thermal stability of Zr_0.7−xTi_xBe_0.3 and Zr_0.7−xNb_xBe_0.3 amorphous alloys are improved by the substitution of titanium or niobium for zirconium. As the Ti and Nb content increases, the crystallization temperatures increase from 610°C to 717°C and 610°C to 678°C, respectively. These amorphous alloys were put into practical use in joining bearing pads on zircaloy cladding sheath. Using Zr–Ti–Be amorphous alloys as filler metals, smooth interface and spherical primary particles (proeutectic phase) appear in the brazed layer, which is the similar microstructure of using Zr_0.7Be_0.3 binary amorphous alloys. In the case of Zr–Nb–Be amorphous alloys, Ni-precipitated Zr phase that may cause some degradation in ductility and corrosion-resistance is formed at both sides of the brazed layer.     

Calculation of thermodynamic parameters of U–Pu–N system with carbon and oxygen impurities

Mixed nitride fuels are being considered for advanced FBR, but very little is known about the thermodynamic properties of these fuels. For an overall composition of the nitride fuel with small amounts of oxygen and carbon impurities, thermodynamic properties, e.g. carbon activity and partial pressures of nitrogen, carbon-monoxide, plutonium and uranium, were calculated in present work. These calculations were based on standard Gibbs free energies of the binary compounds, present in this multi-component system (U,Pu)–C–N–O. For an over all composition of the fuel, stable phase-field was determined by minimization of the Gibbs free energy of the system. The fabrication experiences of various workers, reported in literature, have shown that depending on the impurity content, nitride fuel can exist in two phase fields, mono-nitride phase in equilibrium with sesquinitride phase or mono-nitride phase in equilibrium with dioxide phase. Therefore, in present calculations special attention was given to the thermodynamic behavior of these two phase-fields. A comparison of calculated thermodynamic properties indicated that nitride fuel with dioxide as second phase will be superior to the one with sesquinitride.     

Thermal compatibility studies of U3Si2 dispersion fuels prepared with centrifugally atomized powder

The interaction between atomized U₃Si₂ and aluminum in dispersion fuel samples has been characterized and compared with that of comminuted U₃Si₂. Fuel samples with atomized powder showed a smaller volume increase compared to those with the comminuted powder, irrespective of heat treatment, and volume fraction of U₃Si₂ powder. The possible reasons for this seem to be as follows: (1) the smaller specific surface area of the atomized spherical powder compared to the irregular comminuted powder translating in a smaller U₃Si₂–Al interface area for the former affecting what appears to be a diffusion-controlled interaction process, (2) the atomized fuel samples also contain lower fraction of as-fabricated porosity than the comminuted fuel samples, which may enhance the restraint force in the swelling fuel meat, (3) the comminuted powder particles have distinctive aluminum penetration paths in the form of deformation zones that originated from the comminution process. There appear to be two pronounced penetration paths of aluminum into atomized U₃Si₂ powder; (1) through the phase interface, leaving a central unreacted island, (2) along grain boundaries, leaving several unreacted islands.     

Degradation of mechanical properties of structural reactor materials induced by formation of stress concentrators

The paper deals with the crack nucleation and stability in strain fields of stress concentrators (e.g. voids, gas bubbles, secondary phase precipitates). A general equation describes critical and subcritical crack length as a function of external (applied loading) and internal (stress concentrator type, normal traction, elastic properties of matrix, etc.) parameters. For the critical crack an analog of the Griffith criterion is found. The reduction of fracture stress due to different types of internal stress concentrators was evaluated.     

Temperature dependence of the 20% cold worked 316 stainless steel steady state irradiation creep rate

The irradiation creep data from four completed tests have been analysed to show that the steady state irradiation creep rate exhibits a moderate and complex temperature dependence. The irradiation creep tests were performed in the Experimental Breeder Reactor Number II (EBR-II) using beams and pressurized tubes, and in the Oak Ridge Reactor (ORR) and the High Flux Isotope Reactor (HFIR) using pressurized tubes. The data cover the temperature range from 200°C to 585°C, and show that from 200°C to 330°C, the steady state rate increases moderately with increasing temperature. At about 330°C, the steady state rate peaks and rapidly decreases with increasing temperature from 330°C to 370°C. From 370°C to 585°C the steady state rate moderately increases with increasing temperature.     

Thermal properties of redeposition layers in the JT-60U divertor region

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties.     

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

Fractographic and microstructural characterization of irradiated 304 stainless steel intergranularly fractured in inert gas

Fractographic and microstructural examinations were performed by scanning and transmission electron microscopy, respectively, and correlated, for the thermally sensitized 304 stainless steel (SS) irradiated to 1.2×10²¹ n/cm² (E>1 MeV) in BWR condition and fractured intergranularly in 290 °C inert gas. Intergranular (IG) cracks were present in the specimen surface region and the fracture surface periphery. The fractography showed IG facets decorated with various patterns of linear features/steps. The microstructures of the surface region revealed linear features/deformation twinning near grain boundaries and microtwins at grain boundaries. The linear features identified on the [1 1 1] habit plane varied depending on deformation levels. The high number density of microtwins evidences a high local stress and strain concentration, which may nucleate and initiate at the impingement of deformation twins and grain boundaries. Therefore we conclude that a mechanism causing the IG cracking mechanically in non-aqueous environment is present in the highly irradiated austenitic SS.