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).     

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.     

Contribution of thermal neutrons to radiation hardening of pure copper

The paper presents the results of an experiment the aim of which was to estimate directly the effect of the thermal neutron fluence on pure copper hardening. Identical specimens were irradiated in two reactors (SM-2 and RBT-6) in the dose range 10⁻³-10⁻¹ dpa at T_irr=80 °C under substantially different, by a factor of 5, thermal neutron fluences, with other irradiation parameters being close. The results show that the elevated thermal fluence in the SM-2 reactor increases the radiation hardening of pure copper by 50% at a dose of about 10⁻³ dpa as compared with specimens irradiated in the RBT-6 reactor. The contribution of thermal neutrons proved to be much more considerable than the theoretical estimates.     

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.     

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.     

Oxidation kinetics of Zircaloy-4 and Zr-1Nb-1Sn-0.1Fe at temperatures of 700-1200 °C

The oxidation characteristics for the Zircaloy-4 and Zr-1.0Nb-1.0Sn-0.1Fe alloys were investigated in the temperature ranges of 700-1200 °C for 3600 s under steam supply conditions, using a modified thermo-gravimetric analyzer. The oxidation at these temperatures generally complied with the parabolic rate law for the examined duration up to 3600 s. However, the parabolic rate was not obeyed in the temperature ranges of 800-1050 °C. The oxidation kinetics were changed depending on the oxidation temperatures due to the phase transformations of the base metal and its oxide. The oxidation rate exponents of the Zr-1.0Nb-1.0Sn-0.1Fe alloy at all the temperatures were higher than those of Zircaloy-4. Considering the data controlled by the parabolic rates at 700, 1100, 1150, and 1200 °C, the oxidation rate constants were the same slopes as the Baker-Just relationship. The rate transition at 800 °C could have resulted from the phase transformation of the base metal and those at 1000 and 1050 °C could have resulted from the lateral cracks in the oxide due to the ZrO₂ phase transformation from a monoclinic structure to a tetragonal structure.     

Impact properties of 304L stainless steel GTAW joints evaluated by high strain rate of compression tests

This paper presents an investigation into the high velocity impact of 304L stainless steel gas tungsten arc welded (GTAW) joints at strain rates between 10⁻³ and 7.5 × 10³ s⁻¹ using a compressive split-Hopkinson bar. The results show that the impact properties and fracture characteristics of the tested weldments depend strongly on applied strain rate. This rate-dependent behavior is in good agreement with model predictions using the hybrid Zerilli-Armstrong constitutive law. It is determined that the tested weldments fail as a result of adiabatic shearing. The fracture surfaces of the fusion zone and base metal regions are characterized by the presence of elongated dimples. The variation in the observed dimple features with strain rate is consistent with the results of the impact stress-strain curves.     

The neutron irradiation effect on mechanical properties of HIP joint material

Dispersion strengthened copper (DSCu) and stainless steel are the candidate material for the heat sink and the structural material of the ITER shielding blanket and these materials are joined by hot isostatic pressing (HIP). In this study, the neutron irradiation effect on mechanical properties of HIP joint material was examined by tensile and impact tests using specimens with irradiation damage of about 1.5 dpa. The results of tensile tests show that tensile strength of HIP joint material was about the same as that of DSCu base material, and this trend did not change after neutron irradiation. On the other hand, the impact value of HIP joint material was smaller than that of DSCu base material because of the diffusion of main elements at joint boundary. It was shown that embrittlement by the neutron irradiation effect is smaller than that of the effect by HIP joint.     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

Neural-network analysis of irradiation hardening in low-activation steels

An artificial neural network has been used to model the irradiation hardening of low-activation ferritic/martensitic steels. The data used to create the model span a range of displacement damage of 0-90 dpa, within a temperature range of 273-973 K and contain 1800 points. The trained model has been able to capture the non-linear dependence of yield strength on the chemical composition and irradiation parameters. The ability of the model to generalise on unseen data has been tested and regions within the input domain that are sparsely populated have been identified. These are the regions where future experiments could be focused. It is shown that this method of analysis, because of its ability to capture complex relationships between the many variables, could help in the design of maximally informative experiments on materials in future irradiation test facilities. This will accelerate the acquisition of the key missing knowledge to assist the materials choices in a future fusion power plant.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 °C

Six austenitic stainless steel heats (three heats each of 304SS and 316SS) neutron-irradiated at 275 °C from 0.6 to 13.3 dpa have been carefully characterized by TEM and their hardness measured as a function of dose. The characterization revealed that the microstructure is dominated by a very high density of small Frank loops present in sizes as small as 1 nm and perhaps lower, which could be of both vacancy and interstitial-type. Frank loop density saturated at the lowest doses characterized, whereas the Frank loop size distributions changed with increasing dose from an initially narrow, symmetric shape to a broader, asymmetric shape. Although substantial hardening is caused by the small defects, a simple correlation between hardness changes and density and size of defects does not exist. These results indicate that radiation-induced segregation to the Frank loops could play a role in both defect evolution and hardening response.     

Solid state synthesis of Mg-Ni ferrite and characterization by XRD and XPS

Single-phase magnesium-nickel ferrites with varying amounts of nickel and magnesium were characterized by X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) techniques. A plot of lattice parameter versus composition of the ferrites (Mg_xNi_(1−x)Fe₂O₄, x?1) showed an abrupt deviation of lattice parameter linearity near MgFeO₄. The deviation was explained in terms of the distribution of Mg²⁺ in the octahedral and tetrahedral sites of the oxygen lattice. In XPS spectra, a broadening of the Mg 1s peak in Ni rich Mg-Ni ferrites from that observed in pure MgFe₂O₄, was explained by changes in the distribution of Mg²⁺ ion in tetrahedral and octahedral sites. A depth distribution of Mg in Ni_0.5Mg_0.5Fe₂O₄ showed an enrichment of Mg on surface.     

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.     

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.     

Determining the shear properties of the PyC/SiC interface for a model TRISO fuel

The fracture behavior of TRISO-coated fuel particles is dependent on the shear strength of the interface between the inner pyrolytic carbon (PyC) and silicon carbide coatings. This study evaluates the interfacial shear properties and the crack extension mechanism for TRISO-coated model tubes using a push-out technique. The interfacial debond shear strength was found to increase with increasing sample thickness and finally approached a constant value. The intrinsic interfacial debond shear strength of ∼280 MPa was estimated. After the layer is debonded, the applied load is primarily transferred by interfacial friction. A non-linear shear-lag model predicts that the residual clamping stress at the interface is ∼350 MPa, and the coefficient of friction is ∼0.23, yielding a frictional stress of ∼80 MPa. These relatively high values are attributed to the interfacial roughness. Of importance in these findings is that this unusually high interfacial strength could allow significant loads to be transferred between the inner PyC and SiC in application, potentially leading to failure of the SiC layer.     

Steam oxidation of Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr and Zircaloy-4 at 900-1200 °C

The steam oxidation characteristics for the Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr (HANA-4) and Zircaloy-4 claddings were elucidated at LOCA temperatures of 900-1200 °C by using a modified thermo-gravimetric analyzer. After the oxidation tests, the oxidation behaviors, oxidation rates, surface appearances, and microstructures of the as-received, as-oxidized, and burn-up simulated claddings were evaluated in this study. The high-temperature oxidation resistance of the as-received HANA-4 cladding was superior to that of the Zircaloy-4. The superior oxidation resistance of the HANA-4 cladding could be attributed to the higher Nb and the lower Sn within its cladding. The pre-oxidized layer formed at the low temperatures below 500 °C could retard the oxidation rate at the high temperatures above 900 °C. And the soundness of the pre-oxidized layer formed at a lower temperature could influence the oxidation kinetics and the rate constants during a steam oxidation at LOCA temperatures from 900 to 1200 °C.     

Immobilisation of I by encapsulation in tin by hot-pressing at 200 °C

The volatility of iodine-129 and its soluble nature in anionic form makes it very difficult to incorporate in ceramic or glassy solids for the purpose of long-term immobilisation. Thus encapsulation in a low-melting metal such as tin is an attractive option, and we describe experiments in which we have hot-pressed AgI-bearing alumina beads surrounded by tin powder at 200 °C.     

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.     

Contribution to the understanding of the ZrNb(l%)O(0.13%) oxidation mechanism at 500 °C in dry air

The oxidation of ZrNb(l%)O(0.13%) at 500 °C in dry air was investigated in situ by thermogravimetric analyses and electrochemical impedance spectroscopy. Sheets of the alloy were coated with different noble metals (Pt, Au, Ag) as electrode material. After an initial sub-parabolic rate law, the kinetics of ZrNb(l%)O(0.13%) oxidation are characterized by a transition to another decreasing rate law for different times and thicknesses. Noble metals were observed to clearly modify the oxidation rate, even when a pre-oxidized zirconia film was formed before the deposit and the increase in the oxidation rate was always monitored for thick oxides (30 μm). The kinetic transition is hypothesized to be associated with the microstructural degradation of the oxide film. Localized oxidation rate increases were revealed by scanning electron microscopy at the tip of radial cracks distributed on more than 2% of the total area of the sample. Catalytic effects observed on the oxidation rate after the noble metal deposition suggest that the mechanism controlling the oxidation rate is not a solely one of oxygen diffusion through the oxide layer. The reaction of oxygen reduction at the oxide/metal/gas interface partially controls the oxidation kinetics of ZrNb(l%)O(0.13%). Complex electrical signatures monitored during the oxide growth corroborate this assumption and hence indicate that oxygen reduction is still partially controlling the oxidation rate when noble metal are present on ZrNb(l%)O(0.13%) surface. Finally, a mixed process of interfacial-diffusion mechanism is proposed to be the rate determining step for ZrNb(l%)O(0.13%) oxidation in this environment.