Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water

A formulation for the quantitative calculation of the stress corrosion cracking (SCC) growth rate was proposed based on a fundamental-based crack tip strain rate (CTSR) equation that was derived from the time-based mathematical derivation of a continuum mechanics equation. The CTSR equation includes an uncertain parameter r₀, the characteristic distance away from a growing crack tip, at which a representative strain rate should be defined. In this research, slow strain rate tensile tests on sensitized 304L stainless steel in oxygenated high temperature water were performed. By curve fitting the experimental results to the numerically calculated crack growth rate, the parameter r₀ was determined. Then, the theoretical formulation was used to predict the SCC growth rates. The results indicate that r₀ is on the order of several micrometers, and that the application of the theoretical equation in predicting the crack growth rate provides satisfactory agreement with the available data.     

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.     

Studies on weldability of Ti-5Ta-1.8Nb alloy

The welding, qualification and characterization of welds of Ti-5Ta-1.8Nb alloy, which is being developed for high corrosion resistant performance, are reported. Based on the studies performed as per the ASME Section IX standards, welding procedure specification and procedure qualification record have been formulated. The heterogeneous microstructures of the weldment are rationalized, based on phase transformation in the alloy system and differences in the thermal cycles of various microscopic regions.     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.     

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 between microstructure and corrosion behavior of Zr-Nb binary alloy

To investigate the correlation between microstructure and corrosion characteristics of Zr-Nb alloy, the microstructural observation and corrosion test with the change of cooling rate from beta temperature and the variation of Nb content were performed. The oxide characterization was also carried out by synchrotron XRD and TEM. When the Nb is contained less than solid solution limit (0.6 wt%) in Zr matrix, the difference of corrosion rate was not observed in spite of showing the significant changes of microstructures with cooling rate. While, when the Nb content in the alloy is more than 0.6 wt%, the corrosion properties were deteriorated with increasing the supersaturated Nb concentration in matrix and increasing the area fraction of β_Zr. Also it was observed that the supersaturated Nb in matrix was more effective to decrease the corrosion resistance than the β_Zr phase in the same Nb containing alloy, while the equilibrium Nb concentration below solubility limit in the matrix played an important role to enhance the corrosion resistance. During the corrosion testing in steam at 400 °C, the formation of β_Nb phase in water-quenched specimen would result in the reduction of Nb concentration in matrix. Thus, the corrosion resistance is enhanced with the formation of β_Nb phase. It is suggested from this study that the equilibrium Nb concentration below solubility limit in α matrix would be a more dominant factor in the enhancement of corrosion resistance than β phase (β_Nb or β_Zr), supersaturated Nb, precipitate, and internal microstructure such as twin, dislocation and plate.     

The impact of the composition of structural steels on their corrosion stability in liquid Pb-Bi at 500 and 400 °C with different oxygen concentrations

In recent years, heavy liquid metals have found exercise as possible coolants and targets in the conversion of radioactive elements in accelerator driven systems (ADS). Liquid lead-bismuth eutectic alloy is one of candidates for this using tanks to its suitable nuclear and physical properties. Performed examination was aimed at research of compatibility choice materials for parts of ADS with liquid Pb-Bi eutectic alloy, influence of composition choice materials on their corrosion resistance, influence of temperature and oxygen content. We performed corrosion tests of 1000 h each on approximately 20 types of structural steels (austenitic, ferritic and martensitic) in convection loops with flowing Pb-Bi at 500 and 400 °C and using different oxygen concentrations. The impact of Fe, Cr, Ni, Mn, Si, Al and Mo content on the corrosion stability of these steels was measured without and after preliminary passivation through creating thin spinel or oxide layers on their surface.     

Corrosion behaviour of steels in lead-bismuth at 823 K

The corrosion behaviour of the martensitic T91 steel and the austenitic AISI 316L steel was analysed. The steels were immersed in stagnant molten Pb-55.2wt%Bi alloy at 823 K for different exposure times (t = 550-2000 h). The corrosion tests were carried out both under Ar and under Ar-5%H₂ mixture. Under the oxidising conditions (P_O2 = 6 × 10⁻³ Pa), the formation of oxide layers was observed which prevent the penetration of the liquid alloy into the matrix, while under the Ar-5%H₂ mixture (P_O2 = 3.2 × 10⁻²³ Pa), two phenomena occurred: a ‘reactive penetration’ at the liquid alloy/steel interface and the competition between oxidation and penetration.     

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.     

Tracer diffusion of Mn in alloy D9 in presence of sodium

The diffusivity of manganese in vacuum annealed and cold worked alloy D9 in presence of sodium was measured by the standard tracer technique. The lattice diffusion coefficient of manganese in vacuum annealed alloy D9 specimens in the temperature range 773-873 K is given by the expression and that in cold worked alloy D9 specimens is given by where R is in J K⁻¹ mol⁻¹. The activation energy for diffusion in cold worked specimens is less than that in vacuum annealed specimens. The activation energy for diffusion of manganese in presence of sodium is almost four times less than that in various austenitic stainless steels reported in the literature.     

Tritium release from bulk of carbon-based tiles used in JT-60U

Tritium thermal release behavior from the isotropic graphite tile and the CFC tile used as the plasma facing material of JT-60U was experimentally examined. Whole tritium retained in the bulk of tile could not be released by dry gas purge at high temperature in such a period as one day. Utilization of the isotope exchange reaction using purge gas with hydrogen or humid gas was more effective to release the retained tritium. However, approximately 1% of retained tritium was not recovered by the isotope exchange reaction with dry hydrogen even though such high temperature as 1200 °C was applied. Combustion method with oxygen was required to recover all tritium left in the deeper site of the tile. It was observed that combustion of isotropic graphite tile and CFC tile became vigorous at higher temperature than 700 °C though the combustion rate was rather slow at 650 °C.     

Effects of heat input on the microstructure and toughness of the 8 MnMoNi 5 5 shape-welded nuclear steel

A weld metal well proven in the German nuclear industry served as the basis for the certification of a shape-welded steel to be used as base material for manufacture of nuclear primary components. The outstanding properties of this steel are attributed to the extremely fine-grained and stable primary microstructure. Subsequent reheating cycles caused by neighbouring weld beads do neither lead to coarsened brittle structures in the heat-affected zone nor to increase in hardness and decrease in toughness, as is the case with wrought steel materials. One of the largest new reactor vessel design amongst today’s advanced reactor projects is considered to be particularly suitable for the use of shape-welded parts in place of forgings. In addition the need for design and development of new shape-welded steel grades for other new generation reactor projects is emphasized, in which the experience gained with this research could make a contribution.     

Isotope exchange reaction between tritiated water and hydrogen on SiC

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 °C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10⁶ Bq/cm². The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 °C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve fitting method applying the serial reactor model. The reaction rate was observed to be constant as 3.48 × 10⁻⁵ m/s. Additionally tritium release behavior from the surface of SiC in water vapor atmosphere was predicted and compared with that for graphite and stainless steel.     

Influence of δ phase precipitation on the stress corrosion cracking resistance of alloy 718 in PWR primary water

The negative influence of δ phase on the intergranular stress corrosion cracking (IGSCC) resistance of alloy 718 is commonly taken for granted. In addition, δ phase formed at low temperature (about 1023 K) do not present the same characteristics than the one formed at higher temperatures (from 1173 to 1273 K). The aim of the present study is then to understand how δ phase precipitation could enhance crack initiation in alloy 718, whatever the form of δ phase is. For that purpose, several heat treatments leading to δ phase precipitation were realized on two alloy 718 heats, one sensitive to IGSCC and the second not. Specific slow strain rate tensile tests carried out on thin tensile specimens in simulated PWR primary medium at 633 K conclusively prove that δ phase has no effect on the intrinsic sensitivity to intergranular crack initiation of tested heats.     

Corrosion behaviour of stainless steels in flowing LBE at low and high oxygen concentration

The corrosion behaviours of austenitic steel AISI 316L and martensitic steel T91 were investigated in flowing lead-bismuth eutectic (LBE) at 400 °C. The tests were performed in the LECOR and CHEOPE III loops, which stood for the low oxygen concentration and high oxygen concentration in LBE, respectively. The results obtained shows that steels were affected by dissolution at the condition of low oxygen concentration (C_[O2] = 10⁻⁸-10⁻¹⁰ wt%) and were oxidized at the condition of high oxygen concentration (C_[O2] = 10⁻⁵-10⁻⁶ wt%). The oxide layers detected are able to protect the steels from dissolution in LBE. Under the test condition adopted, the austenitic steel behaved more resistant to corrosion induced by LBE than the martensitic steel.     

Surface mediated isotope exchange reactions between water and gaseous deuterium

Maintaining isotopic purity of hydrogen is one of the major tasks in tritium processing systems. The work with multiple isotopes and isotopomers is accompanied by isotope exchanges which is often accelerated by catalysts e.g. surfaces of various materials. In this work, densities of D₂O, HDO produced via isotope exchange reactions in the mixture of D₂, H₂, D₂O, H₂O, HD and HDO contained in a stainless steel (type SS304) vessel were measured as a function of time (40-36 000 s) and pressures near 3.5 × 10² Pa, using mass spectrometry. The derived rates of change of the isotopomers densities are described accurately by a postulated kinetic model.     

Behavior of steels in flowing liquid PbBi eutectic alloy at 420-600 °C after 4000-7200 h

This paper presents the results of steel exposure up to 7200 h in flowing LBE at elevated temperatures and is a follow-up paper of that with results of an exposure of up to 2000 h. The examined AISI 316 L, 1.4970 austenitic and MANET 10Cr martensitic steels are suitable as a structural material in LBE (liquid eutectic Pb₄₅Bi₅₅) up to 550 °C, if 10⁻⁶ wt% of oxygen is dissolved in the LBE. The martensitic steel develops a thick magnetite and spinel layer while the austenites have thin spinel surface layers at 420 °C and thick oxide scales like the martensitic steel at 550 °C. The oxide scales protect the steels from dissolution attack by LBE during the whole test period of 7200 h. Oxide scales that spall off are replaced by new protective ones. At 600 °C severe attack occurs already after 2000 and 4000 h of exposure. Steels with 8-15 wt% Al alloyed into the surface suffer no corrosion attack at all experimental temperatures and exposure times.     

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.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part II. Saturation processes

Sintered pellets of thorium-uranium (IV) phosphate-diphosphate solid solutions (β-Th_4−xU_x(PO₄)₄P₂O₇, β-TUPD) were altered in several acidic media. All the results reported in the first part of this paper confirmed the good chemical durability of the samples. The evolution of the normalized weight loss showed that, in several media, thorium quickly precipitates in a neoformed phosphate-based phase while uranium (IV) is released in the leachate due to its oxidation into the uranyl form. The characterization of neoformed phases was carried out through several techniques involving grazing XRD, infrared and μ-Raman spectroscopies, EPMA, SEM and TEM. SEM micrographies showed that the dissolution mainly occurs at the grain boundaries, leading to the break away of the grains: only the first 15 μm are altered for 2 months in 10⁻¹ M HNO₃. From EPMA and BET measurements, neither the chemical composition nor the specific surface area are significantly modified. Near equilibrium, two neoformed phases were observed and identified by grazing XRD and/or μ-Raman spectroscopy at the surface of the leached pellets: one is found to be amorphous and progressively turns into the crystallized thorium phosphate-hydrogenphosphate hydrate (TPHPH). From the results obtained, a chemical scheme of the dissolution of β-TUPD sintered samples is proposed. The behavior of the actinides in the gelatinous phase appears mainly driven by their oxidation state: thorium remains in the tetrapositive state and is quickly and quantitatively precipitated while uranium (IV) is oxidized into uranyl then released in the leachate. The Th-precipitation as TPHPH first appears scattered then covers the entire surface of the pellet, inducing a delay of the actinides release in the leachate. Both phases act as protective layers and should induce the significant delay of the release of actinides (Th, U) to the biosphere.     

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.