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%.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.     

Tensile properties of ferritic/martensitic steels irradiated in STIP-I

Specimens of ferritic/martensitic (FM) steels T91, F82H, Optimax-A and the electron beam weld (EBW) of F82H were irradiated in the Swiss spallation neutron source (SINQ) Target-3 in a temperature range of 90-370 °C to displacement doses between 3 and 12 dpa. Tensile tests were performed at room temperature and the irradiation temperatures. The tensile test results demonstrated that the irradiation hardening increased with dose up to about 10 dpa. Meanwhile, the uniform elongation decreased to less than 1%, while the total elongation remained greater than 5%, except for an F82H specimen of 9.8 dpa tested at room temperature, which failed in elastic deformation regime. At higher doses of 11-12 dpa, the ductility of some specimens recovered, which could be due to the annealing effect of a short period of high temperature excursion. The results do not show significant differences in tensile properties for the different FM steels in the present irradiation conditions.     

Creep behaviour of modified 9Cr-1Mo ferritic steel

Creep deformation and fracture behaviour of indigenously developed modified 9Cr-1Mo steel for steam generator (SG) tube application has been examined at 823, 848 and 873 K. Creep tests were performed on flat creep specimens machined from normalised and tempered SG tubes at stresses ranging from 125 to 275 MPa. The stress dependence of minimum creep rate obeyed Norton’s power law. Similarly, the rupture life dependence on stress obeyed a power law. The fracture mode remained transgranular at all test conditions examined. The analysis of creep data indicated that the steel obey Monkman-Grant and modified Monkman-Grant relationships and display high creep damage tolerance factor. The tertiary creep was examined in terms of the variations of time to onset of tertiary creep with rupture life, and a recently proposed concept of time to reach Monkman-Grant ductility, and its relationship with rupture life that depends only on damage tolerance factor. SG tube steel exhibited creep-rupture strength comparable to those reported in literature and specified in the nuclear design code RCC-MR.     

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.     

Effect of sodium environment on the low cycle fatigue properties of modified 9Cr-1Mo ferritic martensitic steel

Modified 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of liquid metal cooled fast breeder reactors (LMFBRs). The steam generator has been designed to operate for 30-40 years. It is important to accurately determine the life of the components in the actual environment in order to consider the extension of life beyond the design life. With this objective in view, a programme has been initiated at our laboratory to evaluate the effects of flowing sodium on the LCF behaviour of modified 9Cr-1Mo steel. LCF tests conducted in flowing sodium environment at 823 K and 873 K exhibited cyclic softening behaviour both in air and sodium environments. The fatigue lives are significantly improved in sodium environment when compared to the data obtained in air environment under identical testing conditions. The lack of oxidation in sodium environment is considered to be responsible for the delayed crack initiation and consequent increase in fatigue life. Comparison of experimental lifetimes with RCC-MR design code predictions indicated that the design curve based on air tests is too conservative.     

Influence of dynamic sodium environment on the creep-fatigue behaviour of Modified 9Cr-1Mo ferritic-martensitic steel

The use of liquid sodium as a heat transfer medium for sodium-cooled fast reactors (SFRs) necessitates a clear understanding of the effects of dynamic sodium on low cycle fatigue (LCF), creep and creep-fatigue interaction (CFI) behaviour of reactor structural materials. Mod. 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of sodium cooled fast reactors. The steam generator has a design life of 30-40 years. The effects of dynamic sodium on the LCF and CFI behaviour of Mod. 9Cr-1Mo steel have been investigated at 823 and 873 K. The CFI life of the steel showed marginal increase under flowing sodium environment when compared to air environment. Hence, the design rules for creep-fatigue interaction based on air tests can be safely applied for components operating in sodium environment. This paper attempts to explain the observed LCF and CFI results based on the detailed metallography and fractography conducted on the failed samples.     

Dispersion hardening effects of Nb-V precipitates in Mod.9Cr-1Mo steels

The mechanism of high creep strength of high nitrogen Mod.9Cr-1Mo steels was metallurgically investigated by using an analytical high resolution TEM. The threshold stress in the constituent equation,

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, is strongly dependent on dispersion strengthening due to peculiar-shaped niobium-and-vanadium-precipitates, i.e. wing-like vanadium-nitrides “V-wings” adhering to spherical niobium-carbonitrides. Key factors of the strength are size and shape of the precipitates. Increase of nitrogen addition is effective to growth of V-wings leading to large threshold stress. Calculated stresses based on a dispersion strengthening showed a good fit with experimentally-measured stresses.     

The resistance of 9 Cr-1 MoVNb and 12 Cr-1 MoVW steels to helium embrittlement

Small amounts of helium (15–40 at. ppm) are known to cause embrittlement in a large number of alloys when tensile tested at elevated temperatures. To determine the effect of helium on ferritic steels of interest for fusion reactor applications, tensile specimens of normalized-and-tempered 9 Cr-1 MoVNb with 0.1 and 2% Ni and 12 Cr-1 MoVW with 0.4, 1, and 2% Ni were irradiated in the High Flux Isotope Reactor (HFIR) at ? 55°C. Helium up to ? 42 at. ppm was produced by a two-step reaction of ⁵⁸Ni with thermal neutrons. No embrittlement was detected in tensile tests at 700°C. These results are contrasted with the more severe embrittlement reported in the literature for several other types of alloys containing 15–30 at. ppm He.     

Calculation of cross-sections for p+ Mo reactions up to 160 MeV

Proton-induced reactions on ^{92,94,95,96,97,98,100}Mo have been studied in the energy range from threshold to 160 MeV. All cross-sections for neutrons, protons, deuterons, tritons, helium and alpha particles emission have been calculated by using nuclear theoretical models which integrate the optical model, the intra-nuclear cascade model, direct, pre-equilibrium and equilibrium reaction theories. Theoretical calculations have been compared with existing experimental data; in most cases, the calculated results are in good agreement with the experimental data.     

Calculations of neutron-induced production cross-sections of W up to 20 MeV

Neutron-induced cross-sections for the stable isotopes ^{180,182,183,184,186}W in the energy region up to 20 MeV have been calculated. Calculations were made with the codes CEM03.01 and ALICE/ASH, using the following models: the Dubna version of the intranuclear cascade model for the cascade stage of interaction; the hybrid, the geometry dependent hybrid and the exciton model for the pre-equilibrium component; the Hauser–Feshbach and the Weisskopf–Ewing statistical models for the equilibrium component. Effects of some important model parameters such as level density parameter and pairing correction were investigated. Calculated cross-sections were compared with available experimental data in the literature and with ENDF/B-VI T = 300 K and JENDL-3.3 T = 300 K evaluated data libraries.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Temperature dependence of lattice disorder in Ar-irradiated (1 0 0), (1 1 0) and (1 1 1) MgO single crystals

To better appreciate dynamic annealing processes in ion irradiated MgO single crystals of three low-index crystallographic orientations, lattice damage variation with irradiation temperature was investigated. Irradiations were performed with 100 keV Ar ions to a fluence of 1 × 10¹⁵ Ar/cm² in a temperature interval from −150 to 1100 °C. Rutherford backscattering spectroscopy combined with ion channeling was used to analyze lattice damage. Damage recovery with increasing irradiation temperature proceeded via two characteristic stages separated by 200 °C. Strong radiation damage anisotropy was observed at temperatures below 200 °C, with (1 1 0) MgO being the most radiation damage tolerant. Above 200 °C damage recovery was isotropic and almost complete recovery was reached at 1100 °C. We attributed this orientation dependence to a variation of dynamic annealing mechanisms with irradiation temperature.     

Tensile properties of the ferritic martensitic steel F82H after irradiation in a spallation target

Ferritic/Martensitic (FM) steel, F82H, was irradiated up to a displacement dose of 20 dpa (displacement per atom) at temperatures ranging from 510 to 1075 K in the third experiment of the SINQ Target Irradiation Program (STIP-III). Tensile testing was performed at 295 and 723 K. The tensile test results demonstrate that not only the specimen irradiated in the low temperature regime (<∼675 K) but also those irradiated at elevated temperatures ?710 K show significant hardening effect. After annealing at 873 K for 2 h the irradiated specimens still persist great hardening, which is usually not observed in FM steels after neutron irradiation at low temperatures and annealing at 873 K. The hardening observed in the specimens is believed to be due to the high-density He-bubbles formed in the specimens.     

Computer simulation of sputtering at the low index (1 0 0), (1 1 0) and (1 1 1) surfaces of Ni3Al in a STEM

The present study is relevant to the preferential Al sputtering and/or enhancement of the Ni/Al ratio in Ni₃Al observed by the scanning transmission electron microscopy fitted with a field emission gun (FEG STEM). Atomic recoil events at the low index (1 0 0), (1 1 0) and (1 1 1) surfaces of Ni₃Al through elastic collisions between electrons and atoms are simulated using molecular dynamics (MD) methods. The threshold energy for sputtering, E_sp, and adatom creation, E_ad, are determined as a function of recoil direction. Based on the MD determined E_sp, the sputtering cross-sections for Ni and Al atoms in these surfaces are calculated with the previous proposed model. It is found that the sputtering cross-section for Al atoms is about 7-8 times higher than that for Ni, indicating the preferential sputtering of Al in Ni₃Al, in good agreement with experiments. It is also found that the sputtering cross-sections for Ni atoms are almost the same in these three surfaces, suggesting that they are independent of surface orientation. Thus, the sputtering process is almost independent of the surface orientation in Ni₃Al, as it is controlled by the sputtering of Ni atoms with a lower sputtering rate.     

An assessment of carburization-decarburizatton behaviour of Fe-9Cr-Mo steels in a sodium environment

A critical assessment is made of the carburization-decarburization kinetics of Fe-9Cr-Mo steels exposed to a sodium environment, using the available information on carbide phase morphology, chromium activity in a ferrite matrix, chromium carbide activity in mixed carbides, carbon solubility in Cr-Mo ferritic steels, and activity-concentration relationships based on α-phase/M₆C or M₂₃C₆carbide equilibrium. Experimental data are presented on the decarburization of Fe-5Cr-Mo and Fe-9Cr-Mo steels at 973 K in a sodium environment to ascertain the long-term behaviour of these steels. The analysis shows that the decarburization of ferritic steels is largely dependent on the chemical reaction at the carbide/α interface and that at carbon activities <0.04, the rate is predominantly determined by the dissolution of (Fe, Mo)₆C carbides.     

Cavity formation during single- and dual-ion irradiation in a 9Cr-1Mo ferritic alloy

The dose and temperature dependence of cavity formation in a 9Cr-1Mo ferritic alloy irradiated simultaneously with Ni ⁺ and He⁺ has been studied with TEM. Comparisons are made with parallel experiments on Ni⁺-irradiated material that was preinjected with He. For dual-ion irradiation, both intergranular and intragranular cavities formed at all temperatures (450–600°C) and doses (5–25 dpa) investigated. The size of the intergranular cavities increased with increasing temperature, while the size of intragranular cavities decreased. In preinjected samples, cavities formed only at the lowest irradiation temperature (450°C). For 450°C single-ion irradiation and for 450 and 500°C dual-ion irradiation, there was a correlation between subgrain size and maximum cavity size, suggesting that the boundaries of the small (typically ~ 0.5 μm) subgrains act as the primary point-defect sink.