Manufacturing pressurized creep tubes from highly purified V-4Cr-4Ti alloys, NIFS-Heat2

A fabrication process for pressurized creep tubes (PCTs) for a highly purified V-4Cr-4Ti alloy, NIFS-heat2 was established. No increase in impurity contents in PCTs was detected during the tube manufacturing process. In a preliminary thermal creep test, homogeneous deformation was observed over entire tube length, which verifies reliability of creep measurement by using the present PCTs.     

Influence of Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb binary alloys

The influence of the Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb (x=0-0.6 wt%) binary alloys was evaluated using a static autoclave in the temperature range from 300 to 500 °C. Corrosion tests and precipitate analysis of Zr-xNb binary alloys showed that corrosion resistance increased with the increase of the Nb concentration in the α-matrix, and the best corrosion resistance was obtained when the Nb concentration was nearly at its equilibrium solubility limit at all test temperatures. The alloys containing a higher Nb concentration than their equilibrium solubility also showed good corrosion resistance, which could be attributed mainly to the formation of Nb-precipitates, resulting in an equilibrium Nb concentration in the α-matrix. These results imply that the corrosion resistance of Nb-containing Zr-alloys can be controlled by the Nb concentration in the α-matrix rather than the Nb-precipitates.     

Electrochemical characterisation of nickel-based alloys in sulphate solutions at 320 °C

Nickel alloy steam generator tubes of pressurized water reactors (PWR) are sensitive to stress corrosion cracking (SCC) and the possibility of predicting SCC from electrochemical measurements is of considerable interest for nuclear industry. The electrochemical properties of several nickel-based alloys were studied at 320 °C in sulphate solutions at neutral or slightly alkaline pH from corrosion potential measurements, polarisation curves and polarisation resistance (R_p) measurements by linear voltammetry and electrochemical impedance spectroscopy (EIS). The passive layers were much more stable in neutral conditions, due to the presence of chromium oxide, and alloys 600TT and 690 showed the best passivity. R_p measurements confirmed that alloys 600TT and 690 have the lowest corrosion rates. At alkaline pH, the passivation currents were higher than those obtained at neutral pH, and the alloys showed a close behaviour. Reduction of sulphates to sulphides seemed to be possible. Results are in agreement with thermodynamic and surface analysis data of literature. The electrochemical stability did not appear to be directly related to SCC susceptibility since it varied inversely with the pH dependance of SCC in sulphate medium.     

Recrystallization and precipitation behavior of low-activation V-Cr-Ti alloys after cold rolling

Recrystallization and precipitation behavior after cold working were investigated for three V-4Cr-4Ti alloys, NIFS-HEAT-1, NIFS-HEAT-2 and US 832665, which contain different levels of impurities. The decrease in hardness and the initiation of recrystallization against test temperature do not depend on the oxygen level. However, the rate of grain growth decreased with the increase in the impurity levels. Bimodal distribution of the precipitates was observed after annealing at 1273 K. The large precipitates were Ti-rich precipitates, and the small ones were composed mainly of Ti, C and O (Ti-C-O precipitates). The impurities were dissolved from Ti-C-O precipitates above 1273 K, and the level of impurities in the matrix increased and resulted in the increase in the hardness. Based on the results, the annealing temperature for V-4Cr-4Ti alloys is recommended not to exceed 1273 K for the purpose of maintaining good mechanical properties such as tensile and impact properties.     

Effects of small amount of additional elements on control of interstitial impurities and mechanical properties of V-4Cr-4Ti-Si-Al-Y alloys

In order to improve the mechanical properties of low activation vanadium alloys for fusion structural applications, effects of small addition of Si, Al and Y on the control of interstitial impurities (O, C and N) during the fabrication process were examined for several V-4Cr-4Ti-Si-Al-Y alloys produced by the levitation melting method. Charpy impact tests and tensile tests were carried out for five kinds of V-4Cr-4Ti-Si-Al-Y alloys using miniaturized specimens for the purpose of evaluating the effects of these elements on mechanical properties. Oxygen concentration decreased almost linearly with increasing loss of yttrium during melting. This oxygen reduction with yttrium loss during the melting process may have been achieved by two types of mechanisms, they are, (i) suppression of oxygen penetration into the molten materials from the environment and (ii) getting of oxygen from the matrix by forming Y₂O₃, which floats to the surface during the melting. There was no effect of Si and Al addition to control the concentration of interstitial impurities. V-4Cr-4Ti-0.1Si-0.1Al-0.1Y alloy showed the best impact properties out of the alloys investigated. Upper-shelf energy of the alloys decreased with increasing yttrium content. High number density of coarse inclusions containing yttrium could cause the degradation of impact properties, though they hardly affect tensile properties of the alloys. Even at higher yttrium contents, V-4Cr-4Ti-Y alloys without addition of Si and Al showed relatively high upper-shelf energy.     

Characterization and modeling of creep mechanisms in Zircaloy-4

The deformation microstructure and creep mechanisms of Zircaloy-4 have been investigated. Four Zircaloy-4 specimens were tested at different temperatures and stress levels and the deformation microstructures of these specimens were analyzed using transmission electron microscopy. On the basis of microstructural observation of a-type screw dislocations in prismatic slip systems, the modified jogged-screw model has been applied as a rate controlling mechanism for creep of Zircaloy-4. In addition, the stress dependency of dislocation density, jog spacing, and jog height has been evaluated via modeling and experimental observations. The purpose of this study is to provide a detailed understanding of the creep deformation of Zircaloy-4 and prediction of creep rates in this alloy based on the microstructural information obtained from TEM analysis.     

Effect of alloying elements (Cu, Fe, and Nb) on the creep properties of Zr alloys

The thermal creep behaviors of Zr-based alloys containing Cu, Fe and Nb were investigated under constant load stress at temperatures of 280 and 330 °C, and a stress range of 100-140 MPa. To evaluate an alloying effect on a creep, Zr-based alloys were selected as the binary and ternary systems of Zr-0.3Cu, Zr-0.3Fe, Zr-0.5Nb-0.3Cu and Zr-0.5Nb-0.3Fe. The final annealing of these alloys was performed at 510 °C for 8 h to obtain a recrystallization structure for all the tested alloys. A microstructure characterization test was carried out for the samples before and after the creep test by using TEM, and the results were used to understand the creep mechanism. Creep tests were performed for up to 70 h, which showed a steady-state secondary creep rate in all the alloys. The value of the stress exponent was about 5.5 in all the alloys. The dislocation density was increased by increasing the applied stress, regardless of the alloy system. From the results of this study, it was revealed that the Nb as an alloying element showed the strongest effect on the creep resistance among the added alloying elements, and Fe was more effective than Cu from the viewpoint of creep resistance.     

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.     

Microstructural development and radiation hardening of neutron irradiated Mo-Re alloys

Stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys with Re contents of 2, 4, 5, 10, 13 and 41 wt% were neutron irradiated up to 20 dpa at temperatures from 681 to 1072 K. On microstructural observation, sigma phase and chi phase precipitates were found in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimens, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874 K or below. From these results, the authors discuss about the relation between microstructure development and radiation hardening and embrittlement, and propose the optimum Re content and thermal treatment for Mo-Re alloys to be used under irradiation conditions.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

The effect of grain boundary misorientation on the intergranular M23C6 carbide precipitation in thermally treated Alloy 690

The precipitation characteristics of chromium carbides on various types of grain boundaries in Alloy 690 thermally treated at 720 °C for 10 h were studied through transmission electron microscopy. Precipitation of the intergranular chromium carbides, identified as Cr-rich M₂₃C₆, was retarded on the low angle grain boundaries, compared to that on the random high angle grain boundaries on which coarse and discrete ones were found. They were rarely found on the coherent twin boundaries, however, needle-like ones were evolved on the incoherent twin and twin related Σ9 boundaries. Precipitation of the chromium carbides was also suppressed on the nearly exact coincidence site lattice boundaries such as Σ11 and Σ15, for which the Brandon criterion was fulfilled. The results of the intergranular M₂₃C₆ carbide precipitation were explained in terms of the influence of the grain boundary energy.     

Transmission electron microscopy examination of oxide layers formed on Zr alloys

A transmission electron microscopy investigation was performed on oxides formed on three zirconium alloys (Zircaloy-4, ZIRLO and Zr-2.5Nb) in pure water and lithiated water environments. This research is part of a systematic study of oxide microstructures using various techniques to explain differences in corrosion rates of different zirconium alloys. In this work, cross-sectional transmission electron microscopy was used to determine the morphology of the oxide layers (grain size and shape, oxide phases, texture, cracks, and incorporation of precipitates). These characteristics were found to vary with the alloy chemistry, the corrosion environment, and the distance from the oxide/metal interface. These are discussed and used in conjunction with observations from other techniques to derive a mechanism of oxide growth in zirconium alloys.     

Post irradiation plastic properties of F82H derived from the instrumented tensile tests

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.     

Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.     

Chemistry control analysis of lead alloys systems to be used as nuclear coolant or spallation target

This study presents the lead alloy system chemistry analysis for use as nuclear coolant or spallation target in ADS related systems in order to set down the needs for purification processes and monitoring. The study is limited here to the two main impurities, oxygen and iron. The analysis of the various potential pollution sources that may occur during the various operating modes is given, as well as a first pollution rate assessment. In order to limit the consequences in term of contamination (clogging) and corrosion, it is necessary to define specifications for operation as regards oxygen and iron content in the fluid. As iron cannot be measured and controlled up to now, the best specification is to set the oxygen as high as possible, defined by the cold leg interface temperature to ensure tolerable contamination, in order to maximize the oxidation area to ensure corrosion protection by self-healing oxide layer for the entire system.     

Characterization of phases in ‘crud’ from boiling-water reactors by transmission electron microscopy

This paper reports phases identified in samples of crud (activated corrosion products) from two commercial boiling-water reactors using transmission and analytical electron microscopy and selected-area electron diffraction. Franklinite (ZnFe₂O₄) was observed in both samples. Hematite (α-Fe₂O₃), crystalline silica (SiO₂), a fine-grained mixture of iron oxides probably including magnetite (Fe₃O₄), hematite (α-Fe₂O₃), and goethite (α-FeOOH), and an unidentified high-Ba, high-S phase were observed in one of the samples. Willemite (Zn₂SiO₄), amorphous silica, and an unidentified iron-chromium phase were observed in the other. Chloride-bearing phases were found in both samples, and are assumed to represent sample contaminants. Because of the small sample volumes and numbers of particles studied and the possibility of contamination, it is not clear whether the differences between the phases observed in the two crud samples represent actual differences in the assemblages formed in the reactors.     

Characteristics of microscopic strain localization in irradiated 316 stainless steels and pure vanadium

Characteristics of localized dislocation glide were investigated for 316 and 316LN stainless steels and pure vanadium after ion or neutron irradiation near room temperature and deformation by a uniaxial tensile load or by a multiaxial bending load. In the irradiated 316 stainless steels, both the uniaxial tensile loading and the multiaxial bend loading produced straight localized bands in the form of channels and twins. In vanadium specimens, on the other hand, curved channels were observed after tensile deformation, and these became a common feature after multiaxial bend deformation. No twin was observed in vanadium. A river pattern of channels was observed in the bent samples after irradiation to a high dose of 0.69 dpa. A highly curved channel can be formed by successive cross slip of screw dislocations. Also, the channel width was not constant along the channels; channel widening occurred when weak defect clusters were removed by the gliding screw dislocations changing their paths by cross slip. It is believed that the dissociation of dislocations into partials and high angles between easy glide planes suppresses the formation of curved channels, while a multiaxial stress state, or a higher stress constraint, increases the tendency for channel bending and widening.     

In-reactor uniaxial tensile testing of pure copper at a constant strain rate at 90 °C

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10⁻⁸ dpa/s with a constant strain rate of 1.3 × 10⁻⁷ s⁻¹. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.     

Effect of the irradiation-annealing-irradiation cycle on the mechanical properties of pure copper and copper alloy

The report presents the results of two irradiation experiments. The first experiment was carried out in the SM-2 reactor with the aim to study the effect of single annealing after irradiation on mechanical properties of pure Cu and GlidCopAl25IG alloy. The aim of the second experiment performed in the RBT-6 reactor was to investigate the effect of the irradiation-annealing-irradiation (IAI) cycle. Pure Cu and GlidCopAl25IG alloy specimens were irradiated in the SM-2 and RBT-6 reactors to ?10⁻³, 10⁻² and 10⁻¹ dpa at T_irr=80 °C. The investigations performed revealed that IAI cycles do not cause an accumulation of embrittlement of pure copper and GlidCopAl25IG alloy in the cycles. The experiments lead to the conclusion that the regime of intermediate annealing produces the structure in the material (relatively low density of SFT), sufficiently insensitive to subsequent irradiation (at a low dose level ?10⁻² dpa).     

Detection of intergranular cracking susceptibility due to hydrogen in irradiated austenitic stainless steel with a superconducting quantum interference device (SQUID) sensor

To investigate the detection method of intergranular (IG) cracking susceptibility by hydrogen in irradiated austenitic stainless steel (SS), magnetic and mechanical properties were examined after two repeats of hydrogen charging and discharging (hydrogen treatment) in Type 304 SS which had been irradiated during use in different reactor cores. The residual magnetic flux density (Br) was measured with a superconducting quantum interference device sensor and Br increased with increased neutron fluence and repeated hydrogen treatments. Elongation decreased with an increase of Br and IG cracking appeared above Br of 2×10⁻⁵ T for this measuring method after repeated hydrogen treatments. These phenomena would be caused by hydrogen-induced martensite phase being formed on grain boundaries. It was thought the appearance of IG cracking susceptibility due to hydrogen in irradiated SS could be predicted by measuring the Br of the steel.