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.     

In-pile dimensional changes in neutron irradiated zirconium base alloys

A model is developed to predict in-pile growth in zirconium base alloys as a function of neutron flux, neutron fluence, temperature, dislocation density, and texture. The model is based on vacancy and interstitial behavior with respect to straight dislocations, dislocation loops, depleted zones and grain or sub-grain boundaries. Results indicate very little growth dependence on temperature or neutron flux at temperatures below ~320°C, at fluxes above ~10¹³ n/cm² sec, and at fluences below 10²¹ n/cm². As the flux is lowered and the temperature and fluence are raised, the temperature and flux dependencies increase. Comparison between theory and data is given for both growth and dislocation loop size.     

A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

Zirconium alloys used as fuel cladding tubes in the nuclear industry undergo important changes after neutron irradiation in the microstructure as well as in the mechanical properties. However, the effects of the specific post-irradiation deformation mechanisms on the mechanical behavior are not clearly understood and modeled. Based on experimental results it is discussed that the kinematic strain hardening is increased by the plastic strain localization inside the dislocation channels as well as the only basal slip activation observed for specific mechanical tests. From this analysis, the first polycrystalline model is developed for irradiated zirconium alloys, taking into account the irradiation induced hardening, the intra-granular softening as well as the intra-granular kinematic strain hardening due to the plastic strain localization inside the channels. This physically based model reproduces the mechanical behavior in agreement with the slip systems observed. In addition, this model reproduces the Bauschinger effect observed during low cycle fatigue as well as the cyclic strain softening.     

Effect of Fe content on positron annihilation in neutron irradiated V---Fe alloys

Neutron irradiated V---xFe alloys (with x from 0 to 5 at.%) have been studied by the conventional positron annihilation technique. A remarkable narrowing of angular correlation of annihilation radiation (ACAR) curves was observed for all alloys investigated. A specific feature of ACAR curves in pure vanadium is the presence of a narrow component attributed usually to the positronium (Ps) formation in voids, with inner surfaces covered by gaseous impurities such as oxygen. Significant changes in the ACAR curve component intensities with increase of iron content has been observed. At Fe concentration of about 1 at.% the narrow component disappears completely and the intensity of the middle one decreases significantly. It was concluded that the increase of Fe concentration in V---Fe alloys suppresses the void surface contamination by oxygen atoms and changes the positron work function from bulk materials into voids. Such behavior of the ACAR curve component intensities can be explained in terms of radiation-induced segregation of iron atoms at point defect sinks.     

Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10¹⁸ and 1.02 × 10¹⁹ n/cm² (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.     

Deformation and fracture properties in neutron irradiated pure Mo and Mo alloys

The effect of neutron irradiation on the mechanical properties of select molybdenum materials, unalloyed low carbon arc-cast (LCAC) Mo, Mo-0.5% Ti-0.1% Zr (TZM) alloy, and oxide dispersion-strengthened (ODS) Mo alloy, was characterized by analyzing the temperature dependence of mechanical properties. This study assembles the tensile test data obtained through multiple irradiation and post-irradiation experiments, in which tensile specimens were irradiated up to 13.1 dpa at 80-1000 °C and tested at −194 to 1000 °C. Irradiation at 80-609 °C increased yield stress significantly, up to 170%, while the increase of yield stress after irradiation at 784-936 °C was not significant. The plastic instability stress was strongly dependent on test temperature but was nearly independent of irradiation dose and temperature. The true fracture stress showed weak dependences on test temperature, irradiation dose and temperature when ductile failure occurred. Among the test materials the stress-relieved ODS material in the longitudinal direction (ODS-LSR) displayed the highest resistance to irradiation embrittlement due to its relatively high fracture stress. The critical temperature for shear failure (CTSF) was defined and evaluated for the test materials and the CTSF values were compared with the ductile-to-brittle transition temperatures (DBTT) based on ductility data.     

Tensile properties of EC316LN irradiated in SINQ to 20 dpa

The wide application of 316-type austenitic stainless steels in existing spallation targets requires a comprehensive understanding of their behavior in spallation irradiation environments. In the present study, EC316LN specimens were irradiated in SINQ targets to doses between 3 and 17.3 dpa at temperatures between about 80 °C and 390 °C. Tensile tests were conducted at room and irradiation temperatures. The results demonstrate that the irradiation induced significant hardening and embrittlement in the specimens. The irradiation hardening and embrittlement effects show a trend of saturation at doses above about 10 dpa. Although the ductility was greatly reduced, all specimens broke with strong necking, which indicates a ductile fracture mode.     

Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155 dpa at 443 °C

The majority of data on the irradiation response of ferritic/martensitic steels has been derived from simple free-standing specimens irradiated in experimental assemblies under well-defined and near-constant conditions, while components of long-lived fuel assemblies are more complex in shape and will experience progressive changes in environmental conditions. To explore whether the resistance of HT9 to void swelling is maintained under more realistic operating conditions, the radiation-induced microstructure of an HT9 ferritic/martensitic hexagonal duct was examined following a six-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). The calculated irradiation exposure and average operating temperature of the duct at the location examined were ∼155 dpa at ∼443 °C. It was found that dislocation networks were predominantly composed of (a/2)<1 1 1> Burgers vectors. Surprisingly, for such a large irradiation dose, type a<1 0 0> interstitial loops were observed. Additionally, a high density of precipitation occurred. These two microstructural characteristics may have contributed to the rather low swelling level of 0.3%.     

Effect of Fe content on positron annihilation in neutron irradiated VFe alloys

Neutron irradiated VxFe alloys (with x from 0 to 5 at.%) have been studied by the conventional positron annihilation technique. A remarkable narrowing of angular correlation of annihilation radiation (ACAR) curves was observed for all alloys investigated. A specific feature of ACAR curves in pure vanadium is the presence of a narrow component attributed usually to the positronium (Ps) formation in voids, with inner surfaces covered by gaseous impurities such as oxygen. Significant changes in the ACAR curve component intensities with increase of iron content has been observed. At Fe concentration of about 1 at.% the narrow component disappears completely and the intensity of the middle one decreases significantly. It was concluded that the increase of Fe concentration in VFe alloys suppresses the void surface contamination by oxygen atoms and changes the positron work function from bulk materials into voids. Such behavior of the ACAR curve component intensities can be explained in terms of radiation-induced segregation of iron atoms at point defect sinks.     

Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.     

Microstructural examination of V-(Fe or Cr)-Ti alloys after thermal-creep or irradiation-creep tests

Microstructural examinations have been performed on irradiation-creep and thermal-creep pressurized tube specimens of V-3Fe-4Ti-0.1Si in order to understand failure and creep mechanisms. There are no typical microstructural differences between unstressed and pressurized creep tube specimens irradiated in ATR-A1 in the irradiation temperature regime from 212 to 300 °C. Failed thermal creep specimens show dislocation structures dependent on the tube specimen geometry. This can be interpreted in terms of a large number of slip dislocations oriented for optimum slip.     

The effect of copper and manganese on magnetic minor hysteresis loops in neutron irradiated Fe model alloys

Changes of magnetic minor hysteresis loops in pure Fe, Fe-1 wt% Mn, Fe-0.9 wt% Cu, and Fe-0.9 wt% Cu-1 wt% Mn model alloys after neutron irradiation have been studied. Minor-loop coefficients which are obtained from scaling relations between minor-loop parameters and in proportion to internal stress, were found to decrease in all model alloys after the irradiation to a fluence of 3.32 × 10¹⁹ n cm⁻². The decrease of the coefficients is larger for alloys including Cu and is enhanced by 1 wt% Mn addition. Such decrease implying the reduction of internal stress during irradiation is in contrast with changes of yield strength after the irradiation that increase with Cu and Mn contents. A qualitative explanation was given on the basis of the preferential formation of Cu precipitates along pre-existing dislocations which reduces internal stress of the dislocations.     

Nano-cavities observed in a 316SS PWR flux thimble tube irradiated to 33 and 70 dpa

The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290 °C and 70 dpa at 315 °C were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.     

Isochronal recovery of fast neutron irradiated metals

The isochronal recovery of Al, Ni, Cu, Ag. Au, Pt, Fe, Mo and Co, irradiated at 4.5K with fast neutrons has been studied resistometrically. For the face-centered cubic elements, Al, Ni, Cu, Ag, and Au, the magnitude of stage I recovery is inversely proportional to the atomic mass of these elements, which is in excellent agreement with recent calculations on the spatial distribution of lattice defects in fast neutron irradiated metals. The total irradiation-induced resistivity and hardness of neutron spectra have more of an influence on the details of the recovery stages than does the initial residual resistivity and the initial resistivity ratio. Stage II recovery, at least in Cu, can be explained best by the rearrangement and/or annihilation of defects that are released from the stress fields of the vacancy clusters that exist in fast neutron irradiated metals.     

Microstructural studies of irradiated mixed-oxide fuels subjected to multiple,mild thermal transients

Microstructural changes and swelling were examined in irradiated mixed-oxide fuels subjected to multiple, mild (maximum fuel temperature ~2600°C) thermal transients in a direct-electrical-heating apparatus. Simulant quartz “cladding” was used during these tests. Significant intra- and intergranular fission-gas porosity was developed as a result of the transients; the intergranular porosity accounted for the majority of the swelling. The data analysis suggests that during thermal transients (particularly under cyclic conditions), the thermal-shock-induced cracking of grain boundaries and grain edges contributes significantly to the fuel swelling and fission-gas release behavior. It is recommended that such phenomena be included in the transient fuel-swelling and gas-release modeling efforts, which, to date, include only the mass-transport processes to account for the fuel swelling and the gas-release behavior. The data analysis also indicated that under mild transient conditions, the maximum volume swelling should not exceed 20%. However, in cases where the fuel temperature is >2600° C, the swelling may be greater.     

Tensile behaviour of 9Cr-1Mo tempered martensitic steels irradiated up to 20 dpa in a spallation environment

Tensile specimens of 9Cr-1Mo (EM10) and mod 9Cr-1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.