Irradiated Steel Research Articles

Ion Irradiation-Induced Microstructural Evolution of Ni⁻Mo⁻Cr Low Alloy Steels.

As leading candidates of sheet steels for advanced nuclear reactors, three types of Ni–Mo–Cr high-strength low alloy (HSLA) steels named as CNST1, CNST2 and CNSS3 were irradiated by 400 keV Fe+ with peak fluence to 1.4 × 1014, 3.5 × 1014 and 7.0 × 1014 ions/cm2, respectively. The distribution and morphology of the defects induced by the sample preparation method and Fe+ irradiation dose were investigated by transmission electron microscopy (TEM) and positron-annihilation spectroscopy (PAS). TEM samples were prepared with two methods, i.e., a focused ion beam (FIB) technique and the electroplating and twin-jet electropolishing (ETE) method. Point defects and dislocation loops were observed in CNST1, CNST2 and CNSS3 samples prepared via FIB. On the other hand, samples prepared via the ETE method revealed that a smaller number of defects was observed in CNST1, CNST2 and almost no defects were observed in CNST3. It is indicated that artifact defects could be introduced by FIB preparation. The PAS S-W plots showed that the existence of two types of defects after ion implantation included small-scale defects such as vacancies, vacancy clusters, dislocation loops and large-sized defects. The S parameter of irradiated steels showed a clear saturation in PAS response with increasing Fe+ dose. At the same irradiation dose, higher values of the S-parameter were achieved in CNST1 and CNST2 samples when compared to that in CNSS3 samples. The mechanism and evolution behavior of irradiation-induced defects were analyzed and discussed.

Slow positron beam study of highly irradiated RPV steel under proton and ion impact

Ion irradiation combined with a slow positron beam could be used to evaluate irradiation damage correlated with embrittlement of nuclear materials. In this study, the evolutions of irradiation-induced defects in reactor pressure vessel steel exposed to 240-keV proton and 3-MeV Fe13+ ion irradiations with the same dose of 0.05, 0.14, 0.35, 1.37, and 2.26 displacements per atom (dpa) at a low temperature are investigated. Vacancy-type defects were produced by both proton and Fe13+ ion irradiation. The S parameters of proton-irradiated pure Fe and RPV steel increased with the dose and did not tend to a saturated value up to a high dose (2.26 dpa). The S parameters showed a saturated value at low doses after Fe13+ ion irradiation. This is because heavy ions can more easily produce larger open-volume vacancy-type defects than light ions at low dose.

Mechanical properties of ion irradiated steel EK-181 investigated by dynamic nanoindentation

Heavy ion irradiation is a unique tool that can be used to simulate and study radiation processes in materials. It can reproduce in a small near the surface layer various microstructural features (dislocation loops, precipitates, voids, etc.) observed in neutron-irradiated materials. Because the irradiation damage is restricted by a few micrometres near surface layer, mechanical properties investigation of the samples was carried out by dynamic nanoindentation technique. This method allows obtaining continuously the hardness and elastic modulus depth profile by one indent. Nanoindentation results acquired in this work showed increase in irradiated layer hardness caused by the increase in irradiation temperature and damage dose.

Construction of Theoretical Curves of the Long-Term Strength for Neutron-Irradiated Austenitic Steels Kh18N9 and 08Kh16N11M3

Abstract—Having analyzed experimental data on the long-term strength of unirradiated austenitic materials and predictive dependences obtained in the context of a physical and mechanical model of intergranular fracture, the authors have determined the assurance factor for plotting the long-term strength curves for unirradiated and irradiated materials. The above model, as well as experimental results, has made it possible to calculate the normative curves for the initial and irradiated steels Kh18N9 and 08Kh16N11M3. Their validity has been verified on the basis of external and intrareactor tests.

Surface structure and properties of high-chromium steel irradiated with a submillisecond pulsed electron beam

In this paper we analyze the surface structure and properties of high-chromium steel (steel 420) irradiated with a submillisecond (0.45 ms) intense pulsed electron beam on the COMPLEX vacuum setup at an Ar pressure of 3⋅10−2 Pa. The analysis suggests that electron beam treatment with surface melting dissolves the initial carbide phase in the steel, saturates its surface layer with Cr and C atoms, and form dendrites and martensite structure. The microhardness and the wear resistance of the irradiated steel are respectively 1.5 and 50 times higher compared to the initial material, and their maximum values are attained at a beam energy density of 53 J/cm2.

Study on the irradiation effect of mechanical properties of RPV steels using crystal plasticity model

Abstract In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.

Predicting recrystallized grain size in friction stir processed 304L stainless steel

A major dilemma faced in the nuclear industry is repair of stainless steel reactor components that have been exposed to neutron irradiation. When conventional fusion welding is used for repair, intergranular cracks develop in the heat-affected zone (HAZ). Friction stir processing (FSP), which operates at much lower peak temperatures than fusion welding, was studied as a crack repair method for irradiated 304L stainless steel. A numerical simulation of the FSP process in 304L was developed to predict temperatures and recrystallized grain size in the stir zone. The model employed an Eulerian finite element approach, where flow stresses for a large range of strain rates and temperatures inherent in FSP were used as input. Temperature predictions in three locations near the stir zone were accurate to within 4%, while prediction of welding power was accurate to within 5% of experimental measurements. The predicted recrystallized grain sizes ranged from 7.6 to 10.6 μm, while the experimentally measured grains sizes in the same locations ranged from 6.0 to 7.6 μm. The maximum error in predicted recrystallized grain size was about 39%, but the associated stir zone hardness from the predicted grain sizes was only different from the experiment by about 10%.

Carbon-14 release and speciation from carbon steel in highly alkaline conditions

ABSTRACTThe release and the speciation of carbon species from irradiated JRQ carbon steel samples, representative of the reactor pressure vessel of Belgian nuclear power plants, were studied in a saturated portlandite aqueous solution, relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. In addition, the corrosion rate (which determines the 14C release) was estimated by measuring the release of 60Co. Gas chromatography showed that during the static corrosion test, the carbonaceous species methane, carbon dioxide, ethene, and ethane were produced. Under the hypothesis that all the carbon released from the JRQ steel was transformed into carbon-base gaseous compounds, this corresponds to a corrosion rate of approximately 100 nm/yr, which is in good agreement with literature data.

Quantification of dissolved organic14C-containing compounds by accelerator mass spectrometry in a corrosion experiment with irradiated steel

ABSTRACTLow- and intermediate-level (L/ILW) radioactive waste produced in Switzerland consists of large amounts of14C-containing irradiated steel.14C will be released during the anoxic corrosion of the steel in the cementitious near field of an L/ILW repository. In this study, a corrosion experiment with irradiated steel was carried out to determine the speciation of14C released during the corrosion process in conditions similar to those anticipated in the near field of a cement-based repository. The development of the experimental setup, including installation of the reactor and development of suitable analytical methods based on compound-specific14C analysis with accelerator mass spectrometry (CSRA AMS) is reported. Time-dependent increase in the total content of14C-bearing organic compounds in solution (TO14C) was determined by AMS and the main organic corrosion products that are14C-bearing formate, acetate and lactate were identified by CSRA AMS after a pre-concentration step. The concentration of the14C-bearing organic compounds was found to be very low (fmol to pmol14C/L). Stable carbon compounds were identified and quantified while the source of stable carbon in the system has not yet been identified and the temporal evolution of the concentration of these carbon species is presently not understood.

Study of vacancy defects and their thermal stability in MeV Fe ion irradiated RAFM steel using positron beam Doppler broadening spectroscopy

ABSTRACTIndian reduced activation ferritic-martensitic steel was irradiated with 1.1 MeV Fe ions to various doses from 1 to100 dpa at room temperature. The depth profiling of irradiation-induced vacancy-type defects and the defect-recovery under post-irradiation annealing was studied using variable low-energy positron beam Doppler broadening spectroscopy. The influence of irradiation-induced defects on the microstructural properties was studied by glancing incidence x-ray diffraction (GIXRD) and nanoindentation technique. Positron annihilation study showed the signatures of reduced vacancy concentration at the peak damage region due to injected interstitial effect from 30 to 100 dpa and the widening of vacancy-interstitial recombination-rich region towards the end of ion range with the increase in dose. The GIXRD results were analysed by modified Williamson–Hall plot method, and the variation of coherent domain size and micro-strain with irradiation dose was studied. Irradiation-induced hardening was observed in the nanoindentation study. The features observed in the GIXRD and nanoindentation study are correlated with the depth-resolved defect distribution observed in the positron annihilation study.

Instability of MX and M23C6 type precipitates in F82H steels under 2.8 MeV Fe2+ irradiation at 673 K

The purpose of this study is to investigate the response of the precipitates against irradiation in F82H steels which are considered as main candidate structure material for the fusion DEMO reactor blanket. M23C6-type precipitates contained F82H-IEA steel, and MX- and M23C6-type precipitates contained F82H-BA12 steel were selected as experimental specimens. Iron irradiation at 673 K using ion accelerator was performed on both steels up to 50 dpa. After irradiation, the instability behavior of precipitates was characterized using transmission electron microscopy. It is found that in F82H-IEA specimens, the number density of radiation defects at 20 dpa was 3.26 × 1021 m−3, and the amorphization of M23C6 was undetected up to 50 dpa. Although Ostwald ripening occurred in M23C6 precipitates, the average size of M23C6 mildly decreased with irradiation dose within this study. For F82H-BA12 specimens, the shrinkage rate of MX- and M23C6-type precipitates was estimated respectively as 0.06 and 0.03 nm/dpa. Hence, the MX and M23C6-type precipitates in F82H steel exhibit an excellent irradiation resistance under the present irradiation condition. Moreover, the behaviors like amorphization and weakening in contrast were not observed in M23C6 precipitates in both specimens, which is different from electron irradiation, presumably due to the factors such as different damage rate.

Simulation of impact toughness with the effect of temperature and irradiation in steels

Abstract One of the important requirements for the application of reduced activation ferritic/martensitic steel is to retain proper mechanical properties in irradiation and high temperature conditions. In order to simulate the impact toughness with the effect of temperature and irradiation, a simulation model based on energy balance method consisted of crack initiation, plastic propagation and cleavage propagation stages was established. The effect of temperature on impact toughness was analyzed by the model and the trend of the simulation results was basicly consistent with the previous experimental results of CLAM steels. The load-displacement curve was simulated to express the low temperature ductile-brittle transition. The effect of grain size and inclusion was analyzed by the model, which was consistent with classical experiment results. The transgranular-intergranular transformation in brittle materials was also simulated.

Micro-structural effects of irradiation temperature and helium content in neutron irradiated B-alloyed Eurofer97-1 steel

The micro-structural effects of different neutron irradiation temperatures and helium contents, for 16 dpa dose, have been investigated by means of small-angle neutron scattering (SANS) in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2 V, 1.08 W wt%, B concentrations up to 1000 ppm); due to B transmutations, fusion relevant He/dpa values are expected to be produced under neutron irradiation. SANS measurements have been carried out on a sample irradiated at 350 °C, with estimated helium content of 5600 appm, and compared to previous SANS results, obtained on two other irradiated samples of this same B-alloyed steel. These new measurements confirm that for such high helium contents the SANS cross-section increases in order of magnitude and the magnetic SANS component is strongly reduced, compared to lower helium content (400 appm). Such effects are attributed to increase in helium bubbles density and to the presence of micro-cavities, produced after dissolution of large B-carbides. The SANS data analysis procedure has been improved, also thanks to the additional information provided by the new measurements, and more accurate helium bubble size distributions have been obtained for all the investigated samples. For 5600 appm helium content, bubble volume fractions are found of 0.025 for the sample irradiated at 350 °C and of 0.041 for the previously investigated sample irradiated at 400 °C, significantly increasing with the irradiation temperature. These values are approximately one order of magnitude larger than the value of 0.003 previously found for the sample with 400 appm helium. The size distributions are compared with electron microscopy observations of these same samples. It appears that the occurrence of complex micro-structural changes in irradiated Eurofer97-1 steel should be taken in due account when considering its application under high He/dpa ratio values.

Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels

The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature (To) was obtained with a relatively small number of specimens. Comparing the fracture toughness and To values determined using other larger specimens with those determined using the Mini-C(T) specimens, To values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41 J transition temperature (T41J) and the To values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.

A hierarchical theoretical model for mechanical properties of lath martensitic steels

Lath martensitic steels possess excellent mechanical properties because of the special microstructure, i.e. laths, blocks and packets with particular crystallography in a prior austenite grain. To theoretically analyze the mechanical behavior of lath martensitic steels, a hierarchical model is proposed combining the crystal plasticity theory and micromechanical method. Within the crystal plasticity theory, an interfacial dislocation model is proposed at the block level to physically describe the effect of lath boundary on the deformation behavior. Furthermore, The scale transition among block, prior austenite grain and lath martensitic steel is accomplished by the elastic-viscoplastic self consistent theory. Based on the proposed model, the deformation behavior of lath martensitic steel with hierarchical structures has been theoretically analyzed. It is revealed that the deformation anisotropy of prior austenite grain is weak due to the particular crystallography of microstructures, and the influence of temperature on yield stress is dominated by the thermally related intrinsic lattice friction and elastic modulus. Moreover, the hierarchical model is generalized to study the mechanical behavior of irradiated lath martensitic steel, and it is found that the change of temperature has a limited effect on the irradiation hardening behavior, and the increase of yield stress induced by irradiation-induced defects can accelerate the evolution rate of dislocation density. Numerical results with/without irradiation effect can both match well with corresponding experimental data, indicating the good accuracy and rationality of the proposed model.

The elasticity of the ½ a0 and a0 dislocation loop in α-Fe thin foil

Dislocation loops in irradiated ferritic steels have a Burgers vector of the type ½ a0 <111> or a0 <100>. When they are located in a thin foil such as the one used in transmission electron microscopy, the presence of the two free surfaces modify the elastic field by the action of the so-called ‘image forces’, which can in turn influence the loops. In this work, a general analytical method was deployed to calculate the image forces in an anisotropic bcc α-Fe thin foil containing a nanometric dislocation loop. We observe that image forces induce an out of plane displacement that doubles the bulging of the thin foil surfaces induced by the loop. The elastic field and energy induced by the image forces become remarkable when the depth of the dislocation loop is comparable to its size. Moreover, there is large difference in the response to image forces between the ½ a0 <111> and the a0 <100> loop, with a stronger one for the ½ a0 <111> loop, which relates to the anisotropy of the α-Fe.

Effect of heat load on microstructural development in an irradiated low alloy steel

To investigate the influence of heat load on microstructure change, we conducted in-situ experiments of annealing and irradiation for a model alloy (Fe-0.75Mn-0.47Mo-0.45Ni) using a multi-beam high voltage electron microscope. Electron-irradiation resulted in the formation of black dots in the model alloy. It is noted that some of black dots grew and changed to dislocation loops, but some did not. Furthermore, the 500 °C annealing to the electron-irradiated sample resulted in the shrinkage and/or the disappearance of black dots and small loops (<10 nm). It is suggested that the annealing would cause the release of interstitial atoms from small loops, which could flow into adjacent larger loops or diffuse to the surface.

In-situ high-energy X-ray characterization of neutron irradiated HT-UPS stainless steel under tensile deformation

The tensile deformation behavior of a high-temperature, ultrafine-precipitate strengthened (HT-UPS) stainless steel was characterized in-situ with high-energy X-ray diffraction at 20 and 400 °C. The HT-UPS samples were neutron irradiated to 3 dpa at 400 °C. Significant irradiation hardening and ductility loss were observed at both temperatures. Lattice strain evolutions of the irradiated samples showed a strong linear response up to near the onset of the macroscopic yield, in contrast to the unirradiated HT-UPS which showed a pronounced non-linear behavior well below the macroscopic yield. While the room-temperature diffraction elastic moduli in the longitudinal direction increased after irradiation, the 400 °C moduli were similar before and after irradiation. The evolution of the {200} lattice strain parallel to the loading axis (ε{200}L) showed unique characteristics: in the plastic regime, the evolution of ε{200}L after yield is temperature-dependent in the unirradiated specimens but temperature-independent in the irradiated specimens; and the value of ε{200}L at the yield is an irradiation-sensitive, temperature-independent parameter. The evolution of ε{200}L corresponds well with the dislocation density evolution, and is an effective probe of the deformation-induced long-range internal stresses in the HT-UPS steel.

Bubble-loop complexes induced by helium irradiation and effect of pre-irradiation and post-irradiation of hydrogen in reduced-activation ferritic/martensitic steel

ABSTRACTIn the present work, the synergistic effect of high concentration hydrogen and helium on the dislocation loops and bubbles as well as their correlations in reduced-activation ferritic/martensitic (RAFM) steels is investigated. Such an effect was transmuted from 14 MeV neutron irradiation and has been one of the most challenging issues for RAFM steels for future fusion reactors. After low dose (0.18 dpa) high concentration (5000 appm) single-ion helium irradiation at 723 K, very large dislocation loops were observed, and the majority of bubbles were inside dislocation loops, forming bubble-loop complexes. These bubble-loop complexes defects were also present in hydrogen/helium and helium/hydrogen sequential-ion irradiated steels. Pre-irradiated hydrogen ion effectively inhibited the later growth of loops induced by helium post-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of inhibition. At high fluence of hydrogen pre-irradiation, the structure of bubble-loop complexes disappeared. On the other hand, hydrogen post-irradiation promoted the growth of loops induced by helium pre-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of promotion. The mechanisms for hydrogen/helium synergistic effects are discussed.

Strain evaluation using a non-contact deformation measurement system in tensile tests of irradiated F82H and 9cr ODS steels

We developed a non-contact deformation measurement system to accurately evaluate strain for post irradiation tensile testing, since conventional strain gages cannot be used for small size specimens. The strain calculated from cross-head displacement generally includes deformation from specimen shoulders, fixtures, and the test frame in addition to the deformation from the specimen gauge section. In our system, the distance between painted marks within the specimen gauge section was measured using a high resolution video camera to evaluate the specimen deformation during room temperature tensile testing. The test materials were F82H and 9Cr ODS steels irradiated up to ≈71 displacements per atom (dpa) at about 573 K in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Our system yielded accurate stress strain curves without deformations other than the specimen gage section, and the elongation was less than that calculated from cross-head displacement. This system can contribute to expanding the technically reliable database for the design activity of fusion reactor blanket, including the effects of irradiation on tensile properties.