Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.     

Gaseous hydrogen embrittlement of PH 13-8 Mo steel

In this study, notched tensile and fatigue crack growth tests in gaseous hydrogen were performed on PH 13-8 Mo stainless steel specimens at room temperature. These specimens were susceptible to hydrogen embrittlement (HE), but at different degrees, depending on the aging conditions or the microstructures of the alloys. In hydrogen, the accelerated fatigue crack growth rate (FCGR) usually accompanied a reduced notched tensile strength (NTS) of the specimens, i.e., the faster the FCGR the lower the NTS. It was proposed that the same fracture mechanism could be applied to these two different types of specimens, regardless of the loading conditions. Rapid fatigue crack growth and high NTS loss were found in the H800 (426 °C under-aged) and H900 (482 °C peak-aged) specimens. The HE susceptibility of the steel was reduced by increasing the aging temperature above 593 °C, which was attributed to the increased amount of austenite in the structure. Extensive quasi-cleavage fracture was observed for the specimens that were deteriorated severely by HE.     

Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 °C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 °C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 °C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 °C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same reason, the visible dislocation microstructure and radiation-induced hardening are also not sufficient to cause IASCC alone.     

Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K_Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K_Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing.     

Spectroscopic and microscopic study of the corrosion of iron–silicon steel by lead–bismuth eutectic (LBE) at elevated temperatures

The performance of iron–silica alloys with different silicon composition was evaluated after exposure to an isothermal bath of lead–bismuth eutectic (LBE). Four alloys were evaluated: pure iron, Fe–1.24%Si, Fe–2.55%Si and Fe–3.82%Si. The samples were exposed to LBE in a dynamic corrosion cell for periods from 700 to 1000 h at a temperature of 550 °C. After exposure, the thickness and composition of the oxide layer were examined using optical microscopy, scanning electron microscopy (SEM) and X-ray photoelectron spectrometry (XPS), including sputter depth profiling. Particular attention was paid to the role, spatial distribution, and chemical speciation of silicon. Low-binding-energy silicon (probably silicates or ) was found in the oxide; while elemental silicon (Si) was found in the metal as expected, and silica (SiO₂) was found at the bottom of the oxide layer, consistent with the formation of a layer between the oxide and the metal. Alloys with low concentrations of Si contained only silicate in the oxide. Alloys with higher concentrations of Si contained a layer of silica at the boundary between the oxide and the bulk metal. All of the alloys examined showed signs of oxide failure. This study has implications for the role of silicon in the stability of the oxide layer in the corrosion of steel by LBE.     

Microstructural investigation of the stability under irradiation of oxide dispersion strengthened ferritic steels

Some fuel pin cladding made from a ferritic steel reinforced by titanium and yttrium oxides were irradiated in the French experimental reactor Phénix. Microstructural examination of this alloy indicates that oxides undergo dissolution under irradiation. This irradiation shows the influence of dose and, in a smaller part, of temperature. In order to better understand the mechanisms of dissolution, three ferritic steels reinforced by Y₂O₃ or MgO were irradiated with different charged particles. Inelastic interactions induced by 1 MeV He ion irradiation do not lead to any modification, neither in their chemical composition, nor in their spatial and size distribution. In contrast, isolated Frenkel pairs created by electron irradiation lead to significant oxide dissolution with a radius decrease proportional to the dose. Moreover, the comparison between irradiation with ions (displacements cascades) and electrons (Frenkel pairs only) shows the importance of free point defects in the dissolution phenomena.     

HVEM irradiation damage in heat affected zone of welded SUS304 steel

The heat affected zone (HAZ) of a welded SUS304 steel has been irradiated in an 1250 kV high voltage electron microscope at 673 K and up to 5.4 dpa (displacements per atom) to study the effect of electron irradiation on microstructure. The dislocation loop density of initial irradiation state increased with electron irradiation dose. Void size, void number density and void swelling increased and then saturated gradually with irradiation dose. The depletion of Cr and the enrichment of Ni at the grain boundary were also recognized by EDS analysis in the HAZ of welded SUS304 steel.     

Stability of nanometer-sized oxide clusters in mechanically-alloyed steel under ion-induced displacement cascade damage conditions

Oxide dispersion strengthened ferritic steels are being considered for a number of advanced nuclear reactor applications because of their high strength and potential for high temperature application. Since these properties are attributed to the presence of a high density of very small (nanometer-sized) oxide clusters, there is interest in examining the radiation stability of such clusters. A novel experiment has been carried out to examine oxide nanocluster stability in a mechanically alloyed, oxide dispersion strengthened ferritic steel designated 12YWT. Pre-polished specimens were ion irradiated and the resulting microstructure was examined by atom probe tomography. After ion irradiation to ∼0.7 dpa with 150 keV Fe ions at 300 °C, a high number density of ∼4 nm-diameter nanoclusters was observed in the ferritic matrix. The nanoclusters are enriched in yttrium, titanium and oxygen, depleted in tungsten and chromium, and have a stoichiometry close to (Ti + Y):O. A similar cluster population was observed in the unirradiated materials, indicating that the ultrafine oxide nanoclusters are resistant to coarsening and dissolution under displacement cascade damage for the ion irradiation conditions used.     

Cementation of ILW ion exchange resins: Impact of sulfate ions released by radiolysis on hydrated matrix

Some of the ion exchange resins used during treatment of spent nuclear fuels are intermediate level radioactive wastes which may be damaged by radiolysis process, releasing sulfate ions directly into the cement-based encapsulating material. This work consists in an experimental study of the resulting sulfate attack on the properties of the hydrated matrix: dimensional stability, mineralogy and microstructure of the samples, as well as variations in the chemical composition of the curing solution, were studied during six months. Three sites of delayed ettringite formation were detected: into the cement matrix near the surface exposed to solution, localized in the interfacial transition zone between cement matrix and resins, or progressively replacing the portlandite that initially fulfilled the cracks of anionic resins. During the experiment period, the ettringite precipitation and the expansion detected were moderate, and did not lead to cracking. The material involved was considered as having a good resistance to sulfate attack.     

Study of gel development during SON68 glass alteration using atomic force microscopy. Comparison with two simplified glasses

Literature mentions several physicochemical studies concerning the characterisation of the alteration films that are formed during the dissolution of the nuclear glasses. Up to now, however, no study had been undertaken on the evolution of the alteration film thickness by in situ technique. This study proposes to carry out atomic force microscopy (AFM) in liquid and dry conditions in order to measure the shrinkage or swelling of the alteration film. This work is performed on the glass SON68 and on two glasses with simplified compositions. The results obtained reveal a shrinkage of the alteration film for the simplified glasses and, in situ (underwater), a slight swelling for the SON68 glass caused by the formation of crystalline phase (phyllosilicates) on the surface. In all three cases, when alteration progresses, it increases the density of the gel and the volume occupied by the alteration products tends to be equal to the volume initially occupied by the glass (called isovolumetric alteration). Finally, the drying leads to an important shrinkage. These results could be used to evaluate the potential impact of the internal cracks of an industrial glass block.     

The development of grain-face porosity in irradiated oxide fuel

The predominant mode of fission gas release occurs through atomic diffusion to the grain boundaries. In oxide fuels the fission gases initially precipitate as an array of small lenticular bubbles of circular projection. The arrival of additional gas and vacancies causes these bubbles to grow and coalesce into fewer, larger bubbles. Depending on the irradiation conditions and temperatures, these bubbles may develop either as circular lenticular pores or as extended multi-lobed pores. Eventually the pores may intersect the grain edges where pathways may be formed which enable the gas to migrate to the outer geometry of the fuel and hence to the gap and the pin free volume. Recent extensive PIE campaigns on irradiated fuels have provided a large database of inter-granular porosity development and, from these, models of bubble growth, coalescence, morphological relaxation and venting have been developed.     

Estimation of stress corrosion cracking sensitivity of type 304 stainless steel by magnetic force microscope

An application of a magnetic force microscope (MFM) to the measurement of the chromium depleted regions of type 304 stainless steel is proposed to enable more effective evaluation of the material sensitization to stress corrosion cracking than the conventional methods. The MFM images of sensitized materials show that the magnetizations are induced along grain boundaries by the chromium depletion. The dependence of the magnetization on the sensitization condition conforms to the expected one from the behavior of chromium depletion. Furthermore, the phase identification was performed by electron backscattered pattern technique to reveal the magnetization mechanism due to sensitization. Then, it was found that the magnetization is caused by the transformation from austenite phase to martensite phase. From the discussion on the temperature at which martensitic transformation starts, we see that it seems to be possible to detect regions where the chromium concentration is under 14% by using an MFM.     

The re-crystallization issue in lead-bismuth technology

Numerical and experimental studies were performed to investigate the behaviour of lead-bismuth eutectic (LBE) after solidification. Re-crystallization of LBE is the main phenomenon to consider; it may lead to serious over-stressing of structural materials. The conditions for the target vessel of MEGAwatt PIlot Experiment (MEGAPIE) were especially considered. Some general recommendations were deduced in order to help avoiding dangerous events.     

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.     

Effect of contact conditions on embrittlement of T91 steel by lead-bismuth

The T91 martensitic steel is a candidate structural material for the liquid lead-bismuth eutectic (LBE) MEGAPIE spallation target. This paper first reviews some results on Liquid Metal Embrittlement (LME) of martensitic steels by liquid metals. It appears that LME of steels can occur provided a few criteria are fulfilled simultaneously. Intimate contact between liquid metal and solid metal is the first one. Usually, it is impossible to avoid the oxide film formation on the steel surface even after short exposure to air. This explains the difficulty arising when one would like to determine the susceptibility to LME of T91 steel whilst put into contact with lead-bismuth. Later, we report on different methods of surface preparation in order to remove the oxide layer on the T91 steel (PVD, soft soldering fluxes) and the resulting susceptibility to LME.     

Analysis of water radiolysis in relation to stress corrosion cracking of stainless steel at high temperatures - Effect of water radiolysis on limiting current densities of anodic and cathodic reactions under irradiation

Electrochemical corrosion potential (ECP) is an important measure for environmental factor in relation to stress corrosion cracking (SCC) of metal materials. In the case of SCC for in-core materials in nuclear reactors, radiolysis of coolant water decisively controls ECP of metal materials under irradiation. In the previous models for ECP evaluation of stainless steel, radiolysis of reactor water in bulk was considered to calculate the bulk concentrations of the radiolysis products. In this work, the radiolysis not only in bulk but also in the diffusion layer at the interface between stainless steel and bulk water was taken into account in the evaluation of ECP. The calculation results shows that the radiolysis in the diffusion layer give significant effects on the limiting current densities of the redox reactions of the radiolysis products, H₂O₂ and H₂, depending on dose rate, flow rate and water chemistry, and leads to the significant increase in the ECP values in some cases, especially in hydrogen water chemistry conditions.     

Oxidation behavior of modified SUS316 (PNC316) stainless steel under low oxygen partial pressure

Oxidation behaviors of modified SUS316 (PNC316) and SUS316 stainless steels were investigated under the low oxygen partial pressure of 10⁻³¹−10⁻²² atm at 600-800 °C. Oxygen uptake by these materials parabolically increased with time, and the kinetic rate constants depended on both oxygen partial pressure and temperature. Thus, semi-empirical equations of the parabolic rate constants were obtained to be 2.70×10⁴exp(−109/RT)P_O2^0.279 for PNC316 and 9.23×10⁴exp(−98/RT)P_O2^0.313 for SUS316. For the duplex layer formed under the low oxygen partial pressure, the inner layer consisted of such oxides as Cr₂O₃ and FeCr₂O₄, while the outer layer consisted of non-oxidized α-Fe. Furthermore, oxidation along the grain boundaries was observed for samples oxidized for a long time. From the point of view of fuel cladding chemical interaction evaluation at high burn-up fuel for fast reactors, it is interesting that formation of non-oxidized α-Fe was observed under the low oxygen partial pressure.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part I - Kinetic study

The dissolution of β-TUPD sintered samples was examined in various conditions of pH, temperature, concentrations of anions in the leachate and leaching flow rates. All the normalized dissolution rates were in the range 10⁻⁷ to 10⁻⁴ g m⁻² day⁻¹ even in very aggressive media, showing the good resistance of these ceramics to aqueous alteration. The first part of this paper describes several parameters exhibiting a significant influence on the normalized dissolution rate of the pellets prepared. Both the partial order relative to the proton concentration (n = 0.39-0.41) and the apparent activation energy (E_app = 49 kJ mol⁻¹) were found in good agreement with the data reported for powdered samples showing that the sintering process does not degrade the chemical durability of the ceramics. Moreover, due to the high thermodynamical constant of complexation of phosphate species for tetravalent uranium and thorium, the influence of other ligands such as nitrate, chloride or sulphate on the normalized dissolution rates was limited. Near the equilibrium, the increasing of the leaching time, the temperature or the leachate acidity led to the thorium precipitation at the surface of the pellets either in static or in dynamic conditions. Consequently, the dissolution became clearly incongruent and controlled by saturation processes which are described in the second part of this paper.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.     

Corrosion of steels with surface treatment and Al-alloying by GESA exposed in lead–bismuth

Corrosion tests were performed for T91, E911 and ODS (oxide dispersion strengthened) with surface treatment and Al-alloying by pulsed electron beam (GESA—GepulsteElektronenStrahlAnlage) in flowing lead bismuth eutectic (LBE) with an oxygen content of 10⁻⁶ wt% at 550 °C for 2000 h. The result was that the surface treatment by GESA led to a faster growing multiphase oxide layer which was very homogenous in thickness. After exposure of specimens to LBE, the average oxide layer at the surface was 14–15 μm thick for ODS, 19–20 μm for E911 and 8–22 μm for T91. No dissolution attack occurred. On the surface of the Al-alloyed specimens, thin protective alumina layers were observed at the places where FeAl was formed by the GESA process, otherwise multiphase oxide layers or corrosion attack were observed.