Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors

Due to the widespread use of thick-wall Alloy 690 and its corresponding weld metals Alloys 52 and 152 in various replacement, repair, mitigation, and new plant pressurized water reactor (PWR) applications, there is an industry need for an equation or methodology to predict crack growth rates (CGRs) for primary water stress corrosion cracking (PWSCC) of these materials. An international PWSCC CGR Expert Panel was organized by EPRI, with the participation of national laboratories sponsored by the US NRC, to support the development of such PWSCC CGR equations. A database of over 500 Alloy 690 CGR data points and over 130 Alloy 52/152 CGR data points from seven research laboratories was compiled, evaluated and scored for data quality, and assessed to determine the effects of numerous parameters such as temperature, crack-tip stress intensity factor, yield strength, and crack orientation. The process by which these data were evaluated and scored is presented in this paper.

A great deal of PWSCCPrimary Water Stress Corrosion Cracking (PWSCC) testing has been conducted on a range of A690 materials. An expert panel organized by EPRI is working to collect all the available laboratory data for its applicability to actual plant components. This paper will review the considerations that are being used to perform this evaluation, and the results of that evaluation. One of the key variables is cold workCold work, and detailed studies have been conducted to measure the residual strains, so as to determine the amount of cold work that can exist in a heat-affected-zone (HAZ) region of a typical weld. In addition to these measurements on weldments, limitations on bulk cold work on base metal imposed by all vendors will be reviewed. The effect of the orientationOrientation of the specimens tested will be compared with the orientations of typical flaws in operating plants, so as to determine the relevance of the data to those plants.

Extensive SCC growth rate measurements have been performed on Alloy 690 and its weld metals in the past, and this paper focuses on SCC growth rate evaluation of Alloy 52Alloy 52/152 Alloy 152welds with a variety of defects and/or weld repairsWeld repairs, and in the dilution zone. Ductility dip cracking dominated the weld defectsWeld defects, and weld repair mockups were fabricated by EPRI Charlotte to be 20% or 50% excavation and repair, as well as welds with a refuse pass every layer. Only low and very low SCC growth rates were observed in all cases. Studies on weld dilution zone effects of varying Cr content were evaluated using welds created with variable ratios of dual-filler-wire feel, which permits definitive SCC growth rate measurements in a homogenous weld without the ambiguity of having the crack front in undefined composition of an actual weld dilution zone.

High-chromium, nickel-based alloys, including alloy 690 base material and alloys 52 and 152 weld filler metals, are used in the primary system of new pressurized water reactors (PWRs), as well as for replacement and mitigation of components in existing reactors. These materials are thought to be highly resistant to primary water stress corrosion cracking (PWSCC), which is observed in plant service for components fabricated from the low-chromium alloys 600, 82, and 182. For over 10 years, the U.S. Nuclear Regulatory Commission has sponsored a laboratory testing program to measure the PWSCC growth rates of alloys 690, 52, and 152 in environmental conditions representative of PWRs, with the intent to support technical bases for the determination of appropriate in-service inspection requirements. In many tests, the low crack growth rates are confirmed. For certain cases, however, such as in highly cold worked alloy 690 and at dilution zones between high-chromium weld metals and low-chromium base metals, PWSCC growth rates are reported to be similar to those observed in alloys 600, 82, and 182. Challenges arise in the use of these data for predictive models given the relatively few numbers of tests performed for some material conditions and uncertainties about the correlation between conditions of test materials and those found in the field. This paper will present perspectives on factors that may be considered for the application of these data to the analysis of plant components.

Recent stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC) crack growth rate (CGR) testing at Argonne National Laboratory (ANL) of Alloy 52/152 weldments near dissimilar metal weld interfaces have found, on occasion, some rather surprisingly high SCC CGRs. Weld overlays of alloys believed to possess superior SCC resistance due to their higher Cr content are typically applied over welds made with SCC-susceptible alloys with the expectation that they will act as a barrier to SCC. However, testing conducted at ANL revealed that the SCC CGRs near the interface between the two welds was in the 10−10 m/s range. Likewise, SCC CGR data in Alloy 152Alloy 152 weld butter near the interface with Low Alloy Steel, which is a region with some dilution of CrCr dilution, found SCC rates as high as 10−10 m/s. In most cases, SCC propagation occurred in a direction perpendicular of that of the dendritic grains—a direction not usually associated with fast SCC propagation. The objective of this paper is to present and discuss the testing results with a focus on the possible paths for fast IG SCC propagation in these weldments.

Stress corrosion crack (SCC) arrest tests have been conducted on composite material specimens to study the SCC susceptibility of SCC resistant materials in hydrogenated deaerated water. SCC tests were performed at 360 °C at a stress intensity factor of ~35 MPa$$ \surd {\text{m}} $$ with ring loadedRing loaded composite material specimens fabricated from SCC resistant materials of Alloy 690, weld metals EN52 and EN625, low alloy steel and stainless steel. Control specimens of Alloy 600 and EN82H were also tested. SCC results and analytical characterization results are discussed.

The mechanism of primary water stress corrosion cracking (PWSCC) in Alloy TT690 was studied from the viewpoint of the influence of hydrogen behavior and cavity formation. Four parameters were examined: cavity formation at crack tips; deformation in the presence of hydrogen gas; dynamic strain effect on mechanical properties; and nickel plating effect on PWSCC propagation. The correlation that was observed between cavity formation at crack tips and crack growth rates indicates the cavities have an important influence on cracking. Results for creep tests done in the presence of hydrogen gas reveal hydrogen lowers the elasticity and makes the alloy brittle at the stress concentrated region. Results for slow strain rate tensile tests indicate that the dynamic strain which accompanies the hydrogen source increases the apparent yield strength. Additionally, nickel plating, which might change corrosion behavior on the alloy surface, suppresses the crack propagation. These observations provide direct support for hydrogen and cavity related mechanism that account for PWSCC of Alloy TT690.

Crack initiation has often been measured using simple, unmonitored tests such as bolt-loaded U bends, where the time to initiation is only estimated by occasional interruption, and the stress drops by ~12% from the change in modulus along with stress relaxation, which can be substantial in some materials and heats. The objective of this study was to develop and demonstrate improved techniques using actively loaded tensile specimens and continuous on-line monitoring of crack development using reversing DC potential drop. To complement the crack initiationCrack initiation data, the crack growth response of the heatsHeat and conditions was also evaluated.

SCC initiation behavior of 15% CF specimens cut from four different alloy 182Alloy 182 weldments was investigated in 360 °C simulated PWR primary water under constant load at the yield stress using direct current potential drop to perform in situ monitoring of SCC initiation time. Within each weldment, one or more specimens underwent SCC initiation within 24 h of reaching full load while some specimens had much longer initiation times, in a few cases exceeding 2500 h. Detailed examinations were conducted on these specimens with a focus on different microstructural features such as preexisting defects, grain orientation and second phases, highlighting an important role of microstructure in crack initiation of alloy 182.

Interactions between multiple stress corrosion cracks (M-SCC) have a major influence on crack growth but are underestimated in models devoted to the evaluation of the lifetime of industrial components. In this study, the growth and interactions between multiple cracks on a sensitized Alloy 600 in a 0.01 M tetrathionate solution, were studied by digital image correlation Digital image correlation (DIC). Cracks exceeding 55 µm in length and 0.45 µm in opening were successfully detected by DIC. The emergence and intensification of interactions modify the growth of the crack colony which evolves from a mostly surface crack propagation (lack of interactions) to in-depth propagation controlled by crack shielding. A multiphysics phase field model was jointly developed and successfully implemented to simulate intergranular M-SCC. It coupled a robust algorithm based on brittle fracture and a diffusion model. The resulting modeling allowed simulating the interactions between cracks and the shielding effects observed experimentally. Finally, 3-D quantification of crack propagation was performed by micro-tomography and digital volume correlation (DVC).

Experiments in hydrogenated steam were performed on several U-bend specimens extracted from two Alloy 82 welds. Results demonstrated that Alloy 82 is susceptible to SCC in hydrogenated steam at 400 °C and its susceptibility depends on its chemical composition, welding process and thermal treatment. The microstructure was characterized in the apex of the U-bend specimens. Chemical analysis were performed by electron probe microanalyser (EPMA) and secondary ion mass spectrometry (SIMS) on several areas in the weld in order to correlate crack initiation with chemical heterogeneities. It was concluded that there are more cracks in the roots of the weld passes where the impurity content (sulfur, titanium and aluminum) is higher.

Cavitation Peening (CP)Cavitation Peening (CP) is achieved by delivery of Ultra-High Pressure (UHP) water through a nozzle, the pressure drop through the orifice creating cavitation bubbles, and delivering these bubbles to the target metal surfaces. Collapse of the bubbles on the surface generates a shock wave resulting in compressive stresses. Complex or regular geometric surfaces can be treated by “coating” them with cavitation bubbles through this simple, robust, and forgiving process. Laboratory tests have demonstrated that the compressive stresses due to UHP CP preventPrimary Water Stress-Corrosion Cracking (PWSCC) PWSCC crack initiation of Alloy 600. UHP CP additionally does not significantly affect roughness and hardness in depth of the mitigated surfaces nor risk to damage the component. UHP CP can be applied cost effectively to the remaining Alloy 600 primary components in order to prevent expensive repairs or replacements. Process parameter development and a tooling qualification program for the application of the UHP CP process on Alloy 600 RPV head penetration nozzles Top head penetration nozzlewas completed, and this process was successfully implemented in several PWRsPressurized Water Reactor (PWR).

Stress corrosion crackingStress corrosion cracking (SCC) initiation tests have been conducted on Alloy 600 in simulated PWRPWR primary water (PW) with the aim of understanding the effect of surface condition produced by different machining methods on the time to initiation. Surface roughness, residual stress and cold work were characterised using profilometry, X-ray diffraction (XRD), micro-hardness and electron backscatter diffraction (EBSD). Initiation tests used tensile, button-headed specimens manufactured from 15% cold rolled, low temperature annealed Alloy 600 to relate machining parameters to PWSCC initiation susceptibility. Tests were conducted at 360 °C with active loading corresponding to 1% plastic strain. The onset of initiation was detected by in situ direct current potential drop (DCPD) measurement. Results indicate that residual stress and orientation of the crack plane in relation to cold work are critical parameters for Alloy 600 PWSCC initiation susceptibility. This suggests that knowledge of surface roughness cannot be used as the sole acceptance criteria for surface condition with regard to PWSCC susceptibility.

SCC initiation results in simulated PWR primary have been obtained on one mill annealed alloy 600Alloy 600 plate heat in the cold-worked condition. Twelve specimens with similar cold workCold work levels were tested at constant load and three showed much shorter SCC initiation times (<400 h) than the nine others (>1200 h). Post-test examinations revealed that these three specimens all feature an inhomogeneous microstructure where the primary crack always nucleated along the boundary of large elongated grainsLarge elongated grain protruding normally into the gauge. In contrast, such microstructure was either not observed or did not extend deep enough into the gauge in the other specimens exhibiting ~3–7× longer initiation times. In order to better understand the role of this microstructural inhomogeneity in SCC initiation, high-resolution microscopy was performed to compare carbide morphology and strain distribution between the long grains and normal grains, and their potential effects on SCC initiation are discussed in this paper.

Alloy 690 is a nickel-based alloy (60% Ni, 30% Cr, 10% Fe) used in nuclear Pressurized Water Reactors (PWR) for different components and welds (steam generator tubes etc.). They are subjected to thermal ageing up to 60 years which could lead to an order-disorder transition (Ni₂Cr ordered phase formation) by a diffusion-assisted mechanism. This transformation might modify mechanical properties and is suspected to influence the stress corrosion resistance of the affected components. To study ordering kinetics, hardness, thermoelectric power (TEP)Thermoelectric power alongside transmission electron microscope (TEM) observations were conducted on Ni-33%Cr alloys with different iron contentsIron content (0–3 wt%) after various ageing thermal treatments. The ordering activation energies have been determined: they are found to be independent of the iron content. A correlation between macroscopic properties and TEM diffraction results is proposed. Finally, the distribution of iron between matrix and ordered domains was studied.

Mechanical property degradation due to isothermal ageing is of potential concern for alloys based on the Ni–Cr binary system, such as Alloys 625 and 690. The disorder-order phase transformation, which is the primary source of embrittlement, has been studied in Ni–Cr model alloys by experimental approaches. Model alloys with different stoichiometries have been isothermally aged up to 5000 h at three temperatures (373, 418, and 475 °C) and characterized via nano-indentation, atom probe tomography, and transmission electron microscopy. Results show that off-stoichiometry alloys exhibit ordering but at a slower rate than stoichiometric (Ni/Cr = 2.0) alloy.

The objective of this work is to evaluate the effect of hardening on the susceptibility of a Ni-30.7Cr wt% (Ni-33 at.%Cr) model alloy to stress corrosion cracking (SCC)Stress corrosion cracking and to low temperature crack propagation (LTCP)Low temperature crack propagation. Unlike previous studies that employ cold work to induce susceptibility to SCC, this work utilizes isothermal ageing to produce the long range orderedLong range order Ni₂Cr phase. Samples were aged at 475 °C for durations up to 3176 h in order to produce hardness values between 70-100 HRB. After ageing, precracked compact tension specimens were tested for susceptibility to primary water SCC and to LTCP. Samples aged for less than 200 h (70-80 HRB) showed very high resistance to SCC, while intergranular cracking was observed in samples aged for longer times. An activation energy of 129.7 ± 17.6 kJ/mol and a yield strength exponent of 6.4 ± 2.1 were measured for SCC growth at constant K _I and ΔEcP conditions, consistent with observations for Alloy 690Alloy 690. Hardening via long range order had no measureable effect on the toughness of the alloy in air, but degraded the toughness and promoted intergranular fracture in hydrogen deaerated water (i.e. it caused susceptibility to LTCP). The similarity of the yield strength dependence to cold worked Alloy 690 and the common temperature dependence (~130 kJ/mol) to A600, X-750, A690, etc. suggests a common SCC mechanism for all these alloys. Hardening via long range order is a novel method to induce SCC susceptibility in Ni-30 wt%Cr alloys, which avoids some microstructural damage, inhomogeneity, and orientation effects that complicate testing of cold worked material.

Thermally aged nickel based AlloyNickel based alloy 600 was investigated to evaluate the effects of long-term thermal aging and triaxial stress on primary water stress corrosion crackPrimary water stress corrosion cracking initiation behavior. Long-term thermal aging was simulated by heat treatment at 400 °C, a temperature that does not cause excessive formation of second phases that cannot form in nuclear power plant service conditions. Triaxial stress was applied by a round notch in the gauge length of some test specimen; other specimens were smooth. Slow strain rate tests (SSRT) monitored by the direct current potential drop method were conducted to evaluate stress corrosion crack initiation susceptibility of the thermally aged specimens in the primary water environment. For smooth specimens (which experience uniaxial stress), the susceptibility of those thermally aged for the equivalent of 10-years was the highest, while the susceptibility of the as-received specimens was the lowest. However, for the notched specimens (which experience triaxial stress), the specimens thermally aged for the equivalent of 20-years showed the highest susceptibility, while the as-received specimens showed the lowest.

Alloy 718 is an age hardenable, nickel-base alloy used in fuel assembly of Pressurized Water Reactors (PWRs) by virtue of its high strength and resistance to corrosion and stress corrosion cracking (SCC). SCC susceptibility is affected by the microstructure developed during thermal mechanical treatments. The SCC behavior of alloy 718 in three different thermal mechanical treatments (TMTs) and two different heats was studied in PWR primary water environment using constant extension rate tensile (CERT) tests. TMTs have a significant effect on the microstructure and thus the mechanical behavior and the SCC susceptibility of alloy 718. TMTs using a solution anneal at 1093 °C with a two-step ageing treatment (1093 °C/1 h + 718 °C/8 h + 621 °C/8 h) exhibited the best SCC resistance.

A series of Alloy 718 specimens were irradiated in the Halden Reactor under mechanical tensile stresses and in a chemical environment and temperature representative of Pressurized Water Reactor service conditions. The specimens were miniature pin-loaded dogbones, heat treated using either a direct aging cycle or the same aging heat treatment preceded by a solution anneal. Applied stresses ranged between 920 and 1200 MPa. Fracture surfaces examined by SEM displayed a mixture of intergranular regions perpendicular to the applied stress and smoother regions at various angles to the applied stress. It is concluded that intergranular cracking proceeded until the stress on the remaining ligament was sufficient to cause prompt ductile fracture. Fluence at fracture occurred over a range of seven orders of magnitude, with no correlation to applied stress. Time at fracture spanned a much smaller range and was broadly, though weakly, inversely correlated with stress. It appears that time in the environment is a better predictor of failure than is fluence.

Thermal ageing promotes intergranular carbide precipitation and atomic ordering reaction in most commercial nickel-base alloys, and it affects the long-term primary water stress corrosion cracking (PWSCC) resistance of pressurized water reactor components. Alloy 690 with 9.8 wt% Fe was solution annealed and heat-treated at low temperature, then aged between 350 and 550 °C for 10,000 h. No direct observation of ordering was possible, but variations in hardness and lattice parameter suggested the formation of short-range ordering (SRO) with a peak level upon ageing at 420 °C, while a disordering reaction occurred at higher temperatures. Heat treatment induced ordering before thermal ageing was compared to the solution-annealed state. Thermal ageing resulted in the precipitation of Cr-rich M₂₃C₆ carbides at grain boundaries and twin boundaries. Although no link between SRO and an increase in strain localization was observed, the combination of intergranular carbide formation and SRO over longer ageing times was deemed detrimental to the PWSCC resistance of Alloy 690 TT.

Two mechanisms are studied to explain the asymmetrical chromium depletionsChromium depletion observed ahead of SCC crack tips in nickel-base alloys: diffusion-induced grain boundary migration (DIGM)DIGM and plasticity-enhanced diffusionDiffusion. On the one hand, DIGM is evidenced in a model Alloy 600Alloy 600 by focused ion beam (FIB) coupled with scanning electron microscopy (SEM) cross-section imaging and analytical transmission electron microscopy (TEM) after annealing at 500 °C under vacuum and at 340 °C after exposure to primary water. The occurrence of grain boundary migration depends on the grain boundary character and misorientation. On the other hand, the effect of plasticityPlasticity on chromium diffusion in nickel single-crystals is investigated by performing diffusion tests during creep tests at 500 and 350 °C. An enhancement of Cr diffusion is observed and a linear relationship between the diffusion coefficient and strain rate is evidenced. At last, in an attempt to discriminate the two mechanisms, an analytical modeling of the Cr-depleted areas observed at propagating SCC crack tips is proposed.

It is well established that thermally treated (TT) alloy 600 exhibits superior resistance to intergranular stress corrosion cracking (IGSCC) in pressurized water reactor (PWR) primary water environments than do solution annealed (SA) and mill annealed (MA) equivalents of the same material. This improved resistance is nominally ascribed to the prevalence of grain boundary (GB) Cr-carbide precipitates (M₇C₃ or M₂₃C₆). In this study, we perform high-resolution characterization by scanning transmission electron microscopy (STEM) and atom probe tomography (APT) on a heat of alloy 600 with a very low concentration of C (0.01 at.%) and a modest B concentration (46 appm). After SA+TT annealing, this heat exhibits IG Cr-boride precipitates (Cr₅B₃) in the absence of Cr carbides. Despite IG precipitation of Cr₅B₃, the pristine GB shows no measurable Cr depletion and B remains segregated at the GB. During intergranular attack (IGA) in PWR primary water at 360 °C, the Cr₅B₃ precipitates dissolve rapidly. The observed IGA is subtly different from what is typical for alloy 600, specifically with higher concentrations of Li and B within the corrosion oxides while the GB ahead exhibits less depletion of oxidizing species (e.g. Cr, Fe, Si and B) than other alloy 600 heats. Together these observations suggest that IG precipitation of Cr₅B₃ in the absence of Cr carbides has a neutral to slightly positive effect on the IG corrosion resistance of alloy 600 GBs in PWR primary water.

Advanced electron microscopy and surface science techniques were applied to characterize inter- and intragranular oxidation in Ni–Fe–Cr alloys after exposure to 480 °C hydrogenated steam. Intragranular internal Fe and Cr oxidation was observed in all cases while intergranular oxidation, exclusively external or penetrative, varied depending on the Cr content of the alloy. The kinetics and morphology of intragranular internal oxidation and nodule growth were studied through successive short-term exposures with characterization performed between exposures. FIB 3D serial sectioning was used to reconstruct volumes containing oxidized grain boundaries and revealed that diffusion-induced grain boundary migration may play a fundamental role in increasing the outward flux of Cr, Ti, and Al near grain boundaries, depending on the extent of intergranular Cr carbide precipitation. In addition, atom probe tomography was used to study the behaviour of minor impurity elements, Al and Ti, and initial oxidation processes. Further analyses of oxidized samples using three-dimensional ToF-SIMS are also discussed.

Oxidation studies were performed on Alloy 600 in situ in high-temperature H₂/N₂ + H₂O vapor mixture at 400 °C. The initial stages of preferential intergranular oxidation (PIO), shown to be an important precursor phenomenon for primary water stress corrosion cracking, have been successfully identified using the in situ approach. The behaviour of Alloy 600 was observed in real time using the Protochips environmental cell and analysed via analytical electron microscopyAnalytical electron microscopy (AEM). Post in situ AEM analyses were compared with previous ex situ post-exposure characterization results obtained from bulk specimens, demonstrating good agreement. The in situ results confirmed the grain boundary migration and intergranular oxide formation in solution-annealed Alloy 600. The excellent agreement between the in situ and previous studies demonstrates that this approach can be used to investigate the initial stages of PIO relevant to nuclear power systems.

Low pressure superheated H₂-steam system has been extensively used in the past years to accelerate the oxidation kinetics while keeping the conditions representative to PWR primary water. One of the most important requirements of this environment is that needs to replicate the Ni/NiO transition. However, despite several studies have been carried out by different research groups in H₂-steam environment, there is still some level of uncertainty over the thermodynamic of the oxidation process. In this study, the Ni/NiO transition in hydrogenated steam was investigated via electrochemical potential measurements using a Ni/NiO solid state reference electrode. Furthermore, solution annealed Alloy 600 coupons were exposed to H₂-steam at 480 ℃ in order to examine the effect of oxidizing conditions with respect to the Ni/NiO transition on the preferential intergranular oxidation. The effect of the redox potential on the preferential intergranular oxidation is discussed in the context of the precursor stages of stress corrosion cracking for Alloy 600.

The effect of dissolved hydrogen (DH)Dissolved hydrogen on primary water stress corrosion crackingPWSCC of nickel base alloys has been of intense interest for plant operators worldwide. In this study, crack growth rates of Alloy 600Alloy 600 were measured in simulated PWR primary coolant at 330 °C with DH levels of 5, 16, 45 and 75 cc H₂/kg H₂O, respectively. The oxide films formed in the crack tip regions were examined using transmission electron microscopy (TEM). The results show low and similar crack growth rates at all DH levels, without a maximum at 16 cc H₂/kg H₂O. The low DH content favors nickel oxide formation at the crack tip region, whereas the high DH level favors Me₃O₄ type spinel formation. Also, the oxide films were found to grow epitaxially on some metal grain surfaces in the cracks. The possible effects of alloy composition on the oxide films formed, and the effect of DH on the crack growth are briefly discussed.

Stress corrosion cracking (SCC) in Alloy 600 has been studied in simulated pressurized water reactor (PWR) primary water at various temperatures. A clear correlation between temperature and crack growth rate (CGR) was found showing that the CGR increased monotonously within the range of temperatures used in this study (320–360 °C). In order to understand the temperature dependence of CGR, high-resolution characterization was used to study the crack tips. The crack tips obtained from different temperatures were analyzed by high-resolution analytical transmission electron microscopy (TEM) to reveal the crack tip morphology and chemistry, which enable the study of a thermally activated diffusion-based mechanism operating during SCC propagation. Transmission Kikuchi diffraction (TKD) was used to investigate mechanical response-based mechanisms in SCC propagation through quantifying the size and extent of plastic deformation around the crack tips. Results obtained in this study show that the thermally activated diffusion along the grain boundary increased with temperature while the changes of plastic deformation around the crack tip were small and nearly independent of temperature, suggesting that a thermally activated diffusion-based mechanism was dominant.

Long-term grain boundary (GB) damage evolution and stress corrosion crack initiation in alloy 690 are being investigated by constant load tensile testing in high-temperature, simulated PWR primary water. Six commercial alloy 690 heats are being tested in various cold work conditions loaded at their yield stress. This paper reviews the basic test approach and detailed characterizations performed on selected specimens after an exposure time of ~1 year. Intergranular crack nucleation was observed under constant stress in certain highly cold-worked (CW) alloy 690 heats and was found to be associated with the formation of GB cavities. Somewhat surprisingly, the heats most susceptible to cavity formation and crack nucleation were thermally treated materials with most uniform coverage of small GB carbides. Microstructure, % cold work and applied stress comparisons are made among the alloy 690 heats to better understand the factors influencing GB cavity formation and crack initiation.

Long-term constant load stress corrosion cracking (SCC) testing for alloys 690/152/52 at 360 ℃ is ongoing, showing no rupture for more than 105 h, suggesting immunity to primary water (PW) SCC initiation under stress level assumed for primary circuit components in pressurized water reactor (PWR) plants. Since the mechanical plug for steam generators has the largest cold work strain, a mock-up PWSCC test, using a mechanical plug of alloy 690, was also performed for evaluation of time to failure under stress and cold work conditions assumed for operating plan. As a result, it was proven that no crack initiated up to approximately 4 × 104 h at 360 ℃. PWSCC susceptibility was also evaluated in terms of crack growth rate (CGR). The CGR of alloy 690 increased after cold working, and the degree of increment is significantly affected by the nature of carbide precipitate along grain boundaries. It was found that increase in CGR caused by cold working remained relatively low when grain boundary carbides precipitated continuously along grain boundaries and coherently with the matrix. Contrarily, CGR grew higher in the materials with lower coherency. It was also revealed that alloy 690 with no grain boundary carbides (solution annealed alloy) showed a small increase of CGR after cold working.

The effects of local strain distribution on primary water stress corrosion cracking of cold-rolled Alloy 690 with an inhomogeneous microstructure were investigated by measuring dislocation densities using transmission electron microscopy. Many intragranular carbides with a Cr-rich M₂₃C₆ structure were dispersed in the fine grains. The results showed that the dislocation densities near the intragranular carbides were high, regardless of the degree of cold-rolling. Below 20% cold-rolling, the dislocation densities near grain boundaries were higher than that in the grain interior. Meanwhile, the dislocation densities in the grain interior increased to similar value of the grain boundary with increasing degree of cold-rolling up to 40%. The results indicate that the intragranular carbides dispersed in the fine grains play an important role in the local strain distribution of a cold-rolled Alloy 690 with an inhomogeneous microstructure. It is suggested that the high local strain in the grain interior in a severely cold-rolled Alloy 690 induced by interaction between dislocation and intragranular carbides could be responsible for the mixed cracking mode and the high crack growth rate.

The evolution of carbides precipitated on grain boundaries near triple junctions in Inconel Alloy 690 following aging treatments at 715 °C for 2, 15 and 50 h were investigated by SEM and EBSD. The results show that, aging time does not influence the morphology of the carbides precipitated at grain boundaries, however, it does influence the morphology of carbides precipitated at triple junctions. Tiny carbides are precipitated at coherent Σ3 grain boundary, bar like carbides are precipitated at both sides of incoherent Σ3 grain boundary, while on only one side of Σ9 grain boundary, and the morphology of carbides precipitated on Σ27 and random grain boundaries are similar. Therefore, the type of nearby grain boundaries can influence the morphology of carbides precipitated at the other grain boundaries.

The Stress Corrosion Crack Growth Rate (SCCGR) in heavily cold rolled Thermally Treated (TT) Alloy 690 exposed to simulated PWR primary water at 360 °C increases with increasing cold rolling ratio, but the SCCGR in cold rolled Mill Annealed (MA) Alloy 690 remains very low, regardless of cold rolling ratio. Cavities were detected near GB carbides in heavily cold rolled TT Alloy 690 before the SCC tests. There is a good correlation between the existence of cavities near GB carbides and high SCCGRs in heavily cold rolled TT Alloy 690. The number of cavities increases with increasing cold rolling ratio and is affected by heating in air at 400 or 475 °C for ~2000 h and by exposure in simulated PWR primary water at 360 °C. However, the cavities were detected not only in the stressed area but also in the stress-free area of the SCC test specimens of heavily cold rolled TT Alloy 690. By contrast, the effect of Ni content on SCCGRs in Ni base (25–30%)-Cr-Fe alloys is not significant for similar amounts of GB carbide precipitation. The high SCCGRs in heavily cold rolled TT Alloy 690 may be caused by a high density of lattice defects, cavities near GB carbides, cracking of M₂₃C₆ primary GB carbides, and hydrogen absorption, but there is no possibility of creep damage at the test temperature of 360 °C. More detailed tests will be needed to confirm this hypothesis.

This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC)SCC of alloy 690 in simulated pressurized water reactor (PWR)PWR primary water environment. Solution annealed alloy 690Alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidationIntergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migrationBoundary migration. The oxides at the crack tip are mainly composed of NiO and Cr₂O₃ which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr₂O₃ dominates at the oxide/substrate interfaces and the very tip region.

Components of reactor core internals suffer from Irradiation Assisted Stress Corrosion Cracking. Here is studied 08Ch18N10T austenitic stainless steel acquired from decommissioned NPP Nord Unit 1, VVER 440-230 type in Greifswald, where had been irradiated to 5.2 dpa. The material was tensile tested at 20 °C in air and slow strain rate tested at 320 °C in air and in water. SEM observations of the fracture surface found ductile fracture for the air tests, but areas of intergranular fracture typical of IASCC in the water. This paper emphasizes the microscopic examination from three samples to determine the underlying physical damage processes. TEM observations close to the fractured surface focused to the interaction of dislocations with local radiation damage defects and grain boundaries owing to different test conditions. Determination of local chemical composition around the grain boundaries indicated radiation induced segregation; as well presence of helium gas in voids. The observation of tensile tests found the presence of twinning and regions of strained martensite transformation. The nano features of tests at elevated temperature were tangled dislocations, similar in air and water. No effect of the water environment on the deformation structures was observed.

5 meV Ni++ and Fe++ ion irradiations were performed to investigate radiation-induced precipitates evolution in a cold-worked 316 austenitic stainless steel at high doses and temperatures. The irradiation conditions were 23 dpa at 380 °C, 130 dpa at 380 °C, 23 dpa at 500 °C, and 15 dpa at 600 °C. TEM selected electron diffraction (SAED), TEMTEM dark-field imaging and energy dispersive spectroscopy (EDS) Energy Dispersive Spectroscopy (EDS)mapping were used as complementary techniques to determine crystallography, morphology and chemical composition of radiation-induced precipitates. The precipitates were predominantly in form of the Ni–Si rich γ′ phase at all irradiation conditions. The EDS analysis further determined Ni–Si–Mo–P and Ni–Si–Mn rich precipitates after irradiation at 380 and 600 °C, respectively. The precipitates were found close to saturated state between 23 and 130 dpa at 380 °C irradiation conditions. A different effect of higher irradiation temperatures was found between 500 and 600 °C. In case of the irradiation to 23 dpa at 500 °C, the average size of precipitates was similar to irradiations at 380 °C, but the density was lower. However, the precipitates revealed large size and very low density following the irradiation to 15 dpa at 600 °C. The original dislocation network introduced by cold-working was found as dominant sink for intra-granular solute radiation-induced segregation (RIS) and possibly took place as primary nucleation site of radiation-induced precipitates at irradiation temperatures 380 and 500 °C. At the temperature 600 °C, the RIS at dislocation network almost vanished and the main nucleation sites became twin boundaries as more energetically favorable intra-granular sinks.

Understanding the changes in mechanical properties and failure mechanisms as a function of radiation damage is important for long-term operation of structural components in nuclear reactors. Due to the expense, the activation of the sample, and the long duration of neutron irradiation, ion beam irradiations (proton-irradiation and, increasingly, heavy-ion irradiation) are used as surrogates for neutron irradiation. However, the shallow irradiation depths of ion-beam irradiation have restricted mechanical property measurements until the recent advent of small-scale mechanical testing. In previous studies, nano-hardness and yield strength of proton-irradiated 304SS were measured using nanoindentation and in situ microcompression respectively. This study develops an in situ microtensile testing method to provide direct stress-strain curves, including the strain to failure, which previous studies do not provide. In addition, a novel way of quantifying irradiation-induced susceptibility to slip band formation in microscale specimens was demonstrated. Lastly, the paper introduces a new technique for measuring grain boundary strength, demonstrated on an oxidized grain boundary of Alloy 600 exposed to primary water chemistryPrimary water chemistry environment.

The next generation of light water reactors, resource renewable BWR (RBWR), which can be burned trans uranium (TRU) is currently under development at Hitachi. The RBWR requires a high flux of fast neutron for efficient burning of the TRU, which is four times as large as that of the ABWR. Therefore, structural materials require both a high resistance to corrosion and to irradiation. In this study, oxide dispersion strengthened austenitic Stainless steelstainless steelsODS (ODS-ASUS) with high corrosion resistance have been developed. The objective of this research is to evaluate irradiation resistance and SCC Stress Corrosion Cracking (SCC)susceptibility in a simulated reactor water environment for the ODS-ASUS. The materials were irradiated with 6.4 MeV Fe3+ at 673 K up to 8.0 dpa using the DuET facility at Kyoto University. The creviced bent beam (CBB) test is conducted to assess the SCC susceptibility in the hot water, 288 °C, 8 MPa with a dissolved oxygen of 8 ppm.

This paper discusses applications of spherical nanoindentation stress-strainNanoindentation stress-strain curves in characterizing the local mechanical behavior of materials with modified surfaces. Using ion-irradiated tungsten as a specific example, this paper demonstrates that a simple variation of the indenter size (radius) can identify the depth of the radiation-induced-damage zone, as well as quantify the behavior of the damaged zone itself. Using corresponding local structure information from electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) we look at (a) the elastic response, elasto-plastic transition, and onset of plasticity in ion-irradiated tungsten, zirconium and 304 stainless steel under indentation, and compare their relative mechanical behavior to the unirradiated state, (b) correlating these changes to the different grain orientations as a function of (c) irradiation from different sources (such as He, W, and He+W for tungsten samples).

Stress corrosion cracking (SCC) has been found on surfaces of reactor internals which are heavily cold-worked by machining. When repair welding is applied to the cracked region, the cold-worked surface will be included in the welding region. Therefore, the effect of the cold-worked layer on the weldability of irradiated stainless steels (SSs) needs to be clarified. In this study, helium bubblesHelium bubble formed by welding on irradiated SSs were investigated by measurements and comparisons to published results. TIG weldingWelding was performed on an irradiated SS plate with a cold-worked layer generated by wire peening. Re-crystallization was confirmed in the weld after peening at the surface near the weld heat affected zone (HAZ). The helium bubble number density along the grain boundary in the re-crystallized region was the same as in the non-re-crystallized region, whereas the bubble diameter was smaller. Helium bubble growth was suppressed by the re-crystallization behavior.

PWR structural components built from austenitic steels accumulate large amounts of helium and hydrogen during service and it is desired to predict the concentrations of these gases not only for higher neutron exposures, but also for subsequent ion-induced extension of the damage dose. While the neutron-induced sources of helium are well-known, there is often uncertainty in the local thermal neutron fluence, which constitutes the major uncertainty for prediction of helium production. Combining earlier measurements of helium at lower dpa levels with knowledge of the 59Ni behavior we can extrapolate to higher dpa levels since the helium concentration is an excellent retrospective dosimeter of the thermal fluence. Predictions of hydrogen are more difficult, however, due to helium-nucleated cavities storing both transmutant and environmental hydrogen. Subsequent ion irradiation of neutron-irradiated material can then use double-ion and triple-ion injection to maintain the neutron-relevant gas cogeneration rates.

Self-ion irradiation is currently being used to explore the relative void swelling resistance of various candidate advanced alloys for a wide variety of nuclear systems, including light water reactor (LWR) reactors. The credibility of using this surrogate irradiation technique to evaluate potential in-reactor behavior requires that certain facets of neutron-induced behavior be reproduced in the ion simulation. Of particular importance is the ability of ion irradiation to produce the anticipated post-transient swelling rates for fcc and bcc iron-base alloys characteristic of reactor irradiation at ~1.0%/dpa and ~0.2%/dpa, respectively. Using a model duplex Fe-9Cr-C alloy irradiated at 450 ℃ with 8 MeV Fe+ ions it is shown in this study that a post-transient rate of ~0.2%/dpa is observed in the ferrite phase after an incubation period of ~60 dpa. It is also shown that the ferrite phase attains this rate first, while the tempered martensite phase exhibits a longer transient delay prior to the onset of high-rate swelling.

Neutron-induced swelling in austenitic and ferritic steels is sensitive to the carbon level in the steel, as well as its distribution in matrix or precipitates. It has recently become known that ion-irradiation to high dpa levels leads to a progressive ion-beam-induced increase in carbon concentration and precipitation within the ion range, with concurrent reductions in void swelling. This neutron-atypical phenomenon imperils the credibility of ion simulation for light water reactor applications. A series of experiments involving pure iron and a structural alloy HT9, were conducted to identify the source and distribution of injected carbon. It was found that negatively-charged carbon atoms are entrained in the self-ion beam by a Coulomb drag effect, and thereby delivered at low drift energy to the irradiated surface, followed by ion-beam-mixing and diffusion. A technique for filtering out contaminants, especially carbon, oxygen and nitrogen, was developed and resulted in higher, more neutron-relevant swelling levels than achieved without filtering.

Most of the available void swelling (VS) Void Swelling (VS)data are from fast reactors rather than PWRs, but there are indications that the saturation VS rate for PWR-relevant conditions is at least an order of magnitude smaller than that observed in fast reactors. In 2005, VS screening criteria (temperatures ≥320 °C (608 °F) and fluence ≥20 dpa) were calculated. Since publication of these initial screening criteria, a physics-based model has been developed for the prediction of VS in irradiated austenitic stainless steelStainless steel components. Comparisons between experimental data derived from density measurements and transmission electron microscopy (TEM) characterizations suggest that the Cluster DynamicsCluster Dynamics model is capable of predicting the evolution of the irradiated microstructure under PWR conditions.Pressurized Water Reactors (PWR) The Cluster Dynamics model estimates <1.5% VS for solution-annealed or cold-worked austenitic stainless steel at temperatures below 320 °C (608 °F) and doses <20 dpa for all displacement rates. Therefore, the previous conservative screening criteria originally calculated in 2005 are retained.

A mean-field, cluster dynamics model of the microstructure evolution in austenitic steels of light water reactors which reproduces the incubation period of swelling has been developed for the first time. In agreement with observations, it predicts that, although the void nucleation starts from the very beginning of irradiation, their growth beyond certain, relatively small size is delayed until the onset of the transient period of swelling. Such a delayed growth of voids is explained by the frequent interaction of voids randomly distributed over the volume with one-dimensionally migrating clusters of self-interstitial atoms, which are produced in cascades of atomic displacements. The incubation period of swelling is followed by the transient stage, when voids start to grow with increasing rate due to development of the experimentally-observed spatial correlations between voids and extended defects, such as second-phase precipitates and dislocations, which screen voids from the mobile clusters. A critical role of residual gas on void nucleation, which diminishes importance of He atoms from transmutation reactions, is revealed.

The effects of irradiation on Inconel® (Inconel is a registered trademark of Special Metals Corporation and its subsidiaries) X-750, have been shown to lead to embrittlement and intergranular fracture. This is now widely accepted to be a result of intergranular helium bubbles over the fluence range studied. This paper provides a quantitative assessment and a detailed discussion of the radiation-induced defects including; helium bubbles (size and density distribution, and grain boundary area fraction), dislocation loops and stacking fault tetrahedra, and the disordering and dissolution of secondary gamma prime precipitates. The microstructural evolution will be presented and discussed as a function of dose (from ~5.5 to ~80 dpa), helium concentration (~1300 to ~25,000 appm helium), and irradiation temperature (~120–280 to ~300–330 °C).

A novel lift-out, push-to-pull, micro-tensileMicro-tensile, small scale mechanical testing (SSMT)Small Scale Mechanical Testing (SSMT) technique was developed to assess the yield strength, failure strength, and failure mechanisms of activated ex-service Inconel X-750 removed from the CANDU nuclear reactor core after extended service. Neutron irradiated Inconel X-750 components fail in an intergranular manner. Because these ex-service components are less than 1 mm in thickness, conventional tensile specimens cannot be fabricated from them. Thus, large-scale testing is not possible, and specimens on the order of 1 μm × 1 μm × 2.5 μm (thickness × width × gauge length) containing individual boundaries were fabricated in order to assess the grain boundary strengthGrain boundary strength of the material as a function of irradiation temperature and dose. The variability introduced by differences in thermo-mechanical processing during fabrication was also assessed. Application of this new Micromechanicsmicro-tensile testing technique to non-irradiated Inconel X-750 gives good agreement with the bulk yield strength of the nickel superalloy, 1070 MPa. From SSMTs, the measured yield strengths of non-irradiated specimens were 1001 MPa at the outer edge and 1043 MPa at the center of the component. Cold work, introduced by grinding of the outside surface of the component, reduces ductility, as does irradiation. Initial tests indicate that away from the surface in the center, the boundary strength was reduced by ~456 MPa after irradiation to 78 dpa at an average irradiation temperature of 180 °C; the corresponding ductility decreased from 16.6 to ≤2.3% total elongation. Testing is a work in progress and more tests are needed for higher precision with regards to grain boundary strength reduction.

Proton irradiation is a well-known useful experimental technique to study neutron irradiation-induced phenomena in reactor core materials. Type 316 austenitic stainless steel was irradiated with 2 meV protons to doses up to 10 displacement per atom at 360 °C, and the various effects of the proton irradiation on the microstructural changes were characterized with transmission electron microscopy and atom probe tomography. Typical irradiation damage mainly consisted of small dislocation loops, cavities, tiny precipitates and network dislocations. Ni and Si were enriched, whereas Cr, Mn and Mo were depleted on the grain boundaries associated with irradiation-induced segregation. Ni–Si rich clusters were also found in the matrix. A new method to prepare TEM specimens of a proton-irradiated material is suggested, which was shown to be a relatively simple and effective method to chemically eliminate the inherent surface damage induced by a conventional high-energy focused ion beam and subsequent low-energy ion milling treatments.

Although service experience of austenitic stainless steels exposed to PWR primary coolant has been good, stress corrosion crack propagation has been observed in laboratory tests in the presence of ≥15% cold work. Data on crack initiation are much more limited and this study therefore aims to improve the understanding of the conditions under which crack initiation and subsequent development of stress corrosion crackingStress Corrosion Cracking (SCC) might be possible. Testing was performed on two heats of Type 304/304L stainless steelStainless steel under slow strain rate tensile loading. A range of analytical techniques were used to characterize the resultant precursor features and cracking, and digital image correlation before and after testing was also used to evaluate the influence of localized deformation on SCC. The results indicate that crack initiation can occur in austenitic stainless steels exposed to good quality primary coolant under dynamic straining conditions; additional testing underway under more plant-representative conditions will be reported later. Significant influences of steel microstructure on crack initiation susceptibility were observed.

Slow strain rate tensile (SSRT) tests Slow Strain Rate Test (SSRT)were conducted on conventional and tapered samples manufactured from forged Type 304L stainless steel304L Stainless steel to assess the stress corrosion cracking (SCC) Stress Corrosion Cracking (SCC) initiationbehaviour in simulated PWR primary water. Several testing and microstructural parameters were investigated in order to explore the conditions under which crack initiation might occur. Surface preparation appeared to play a very important role on SCC initiation whereby the machined surfaces were the least susceptible to SCC initiation whilst oxide polishing suspension (OPS) polished surfaces were more susceptible. On the machined surfaces the cracks were always transgranular (TG) in nature and associated with the machining marks. Conversely, on fine polished surfaces with oxide polishing suspension the crack morphology was mainly intergranular in nature, although minor transgranular cracking was observed. The regions in the proximity of the δ-ferrite/austenite interface were shown to be very susceptible to SCC initiation especially on the OPS polished surfaces and this was attributed to the strain localization upon dynamic deformation. Furthermore, intragranular inclusions appeared to dissolve and act as initiation sites for transgranular cracking to occur. The roles of strain rate, dynamic deformation and microstructure on the initiation of SCC are also discussed.

Annealed and cold-worked stainless steelStainless steel 316L samples with machined and polished surfaces were tested in simulated pressurized water reactor (PWR) primary water under slow strain rate tensile (SSRT) test conditions to investigate stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC) initiation. Roughness, residual stress and cross-sectional microstructure of the as-machined samples were characterized before SSRT tests. Plan view and cross-sectional examinations were performed after the test. Pre-test characterization indicated that a deformation layer was present on the machined surfaces. This deformation layer consisted of an ultrafine-grainedUltrafine grain layer on the top and deformation bands underneath. The thickness of the deformation layer on the annealed material was greater than that on the cold-worked material. Post-test characterization revealed that the SCC initiation behaviors of the as-machined and polished surfaces were different for both annealed and cold-worked materials. MachiningMachining increased SCC initiation susceptibility of the annealed material as many shallow cracks initiated along the machining marks in the machined surface, and it decreased the SCC initiation susceptibility of the cold-worked material as a reduced number of cracks were identified in the machined surface compared to the polished surface. The factors influencing SCC initiation are also discussed.

Initiation and propagation of cracks under simulated primary water conditions and different slow strain ratesSlow Strain Rate Test (SSRT) have been studied for an austenitic 304-type stainless steel. Two surface finishes were used to better understand the conditions that trigger stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC). The main objective is to identify the mechanism(s) that govern the initiation and propagation of SCC and the influence of microstructure. Crack morphology, stress localisation and local chemical composition were characterized for all samples studied. The characterization methodology includes scanning electron microscopy (SEM), 3D focused ion beam (FIB), Transmission Kikuchi Diffraction (TKD)Transmission Kikuchi Diffraction (TKD), and analytical scanning transmission electron microscopy (STEM).

The field experience of austenitic stainless steels in PWR primary circuits has been generally outstanding. However, a review made by EPRI indicates that significant SCC cases occurred in low flow or stagnant zones where the primary water was probably contaminated by trapped oxygen and/or radiolysis products. In addition, the deleterious effect of off normal surface conditions in the SCC susceptibility is not yet clearly understood. A recent program was launched to get insights regarding the potential deleterious effect of off normal surface and chemistry environments on the SCC susceptibility of cold worked stainless steels. Since the presence of surface treatments due to the various steps of component manufacturing or repairing is sometimes unavoidable, a wide range of experimental techniques is deployed to reproduce some of the various off normal surface conditions. This paper summarizes the initial characterizations as well as the baseline SCC tests in PWR primary environment.

The field experience of austenitic stainless steels in PWR (Pressurized Water Reactor) primary circuits has been generally outstanding. However, the effect of so called “off normal” conditions still plays an important role regarding the SCC (Stress Corrosion Cracking) susceptibility of austenitic stainless steels (SS). Such off normal conditions can either be surface conditions or off normal water chemistry conditions, which is the topic of this paper. This paper summarizes evaluations of the SCC susceptibility of cold worked stainless steels in off normal oxygenated environment including transients. In order to study off normal chemistry conditions, SCC tests were performed in simulated PWR primary water conditions under either hydrogen water chemistry conditions or oxygenated conditions. Since the effect of transients are suspected to play a role under plant conditions, additional tests were performed with continuously changing from oxygenated to hydrogenated conditions and compared to results from tests under purely hydrogenated or oxygenated water chemistry conditions.

Sensitization and deformation have previously been implicated in the stress corrosion cracking (SCC) susceptibility of Type 304 stainless steel (SS) in oxygenated water. However, Type 304L SS, with reduced carbon content, is expected to be resistant to sensitization effects. The current work evaluates the SCC response of two dual certified Type 304/304L SSs after a sensitization heat treatment. It is shown that other material factors, namely boron content and delta ferrite stringers, can lead to sensitization and subsequent SCC even in L-grade materials.

A test program has been completed that re-examined the effect of load ratio, R, on fatigue crack growth rates (FCGRs) of austenitic stainless steel in 482 °F air and deaerated water. Data for the test program were collected at R between 0.1 and 0.95 and ΔK values between 2.5 and 50 ksi√in to ensure an overlapping dataset in R and ∆K. In contrast to the single Paris slope relationship in the ASME code, results from air tests revealed a three-regime curve across many R: a high ΔK regime similar to ASME, an intermediate ΔK regime with a decreased power law exponent and a low ΔK region where FCGR exhibit a steep downturn. Water FCGRs showed two regimes—a single power law regime and a steep low ΔK regime. FCGR sensitivity to R was greatest in the low ΔK regime for both environments.

Fatigue crack closure is widely recognized to promote fatigue crack growth retardation behavior through contacting of fracture surfaces in the crack wake and the resulting reduction in the effective stress intensity factor range (∆K) that promotes crack advance. Experimental measurements of crack closure are typically made using conventional compliance methods, but electrical potential drop has also been used to characterize crack closure behavior. Electrical potential drop measurements have detected electrical shorting across fracture surfaces of stainless steel, nickel base weld, and A508 steel tested in high temperature water environments under cyclic loads. These observations have consistently occurred under loading conditions (low R, following overloads) where closure effects are expected to be prominent and have been shown to correlate with reductions in fatigue crack growth rate. These findings suggest that electrical potential drop measurements may serve as a useful tool in assessing the influence of crack closure on corrosion fatigueCorrosion fatigue retardation behavior.

Injected vacancy, enhanced creep is hypothesized to reduce crack growth rates (CGRs) in deaerated pressurized water (DPW) in austenitic stainless steels with high sulfur levels. CGR reduction is hypothesized to occur by corrosion generated vacancy/dislocation interactions that promote dislocation climb and disrupt planar slip bands. Creep tests using tensile specimens of varying sulfur content were performed in air and DPW at 288 °C. Testing began with a hold at the flow stress, followed by fatigue cycles at room temperature (RT), then holds at flow stress and 105% flow stress. Primary creep was exhibited in the high sulfur material in DPW, after the RT fatigue cycles, and resulted in 0.19 mm of extension. Characterization revealed a corrosion product and a deformed microstructure with extensive planar slip bands in the specimen that crept. Corrosion-generated vacancies are unlikely to be the source of the primary creep. Potential mechanisms for the observed creep behavior will be discussed.

Corrosion fatigue crack growth rate tests were performed at 288 °C in high purity, deuterated water (D₂O), and crack tips were examined for the presence of deuterium (D). A sample from the interior of the specimen (1T-CT) was analyzed for D and total D+hydrogen (H) using time of flight-secondary ion mass spectrometry (ToF-SIMS) and hot vacuum extraction. With SIMS, two regions 500 × 500 µm in size were analyzed. The first region was located immediately in front of the crack tip. The second region was a control, and was 6 mm away from the plastic zone. The deuterium concentration was found to be enhanced by a factor of 8.7–9.3 over the natural abundance in the plastic zone/crack tip while the concentration was enhanced by 5.9 in the bulk material. With no other source of deuterium, the detection of deuterium represents a unique marker. These data also provide evidence that hydrogen species are concentrated at the crack tip in fatigue crack growth processes at elevated temperature.

This paper describes an investigation into the mechanisms influencing environmentally assisted enhancement of fatigue crack growth of 304L stainless steel in PWR primary water. Pre-cracked specimens were tested under loading conditions containing hold periods, either using a trapezoidal waveform, or periods of saw-tooth loading interspersed with long hold periods. Several post-test characterization methods were used to provide insight into the mechanisms influencing crack growth behavior. A correlation was observed between steel sulfur content and reduced environmental enhancement, which was more pronounced under trapezoidal loading than for saw-tooth loading following extended hold periods. Post-test examination linked enhanced crack growth with highly faceted fracture surfaces, whilst lower levels of enhancement showed a less faceted and more heavily oxidized appearance. The observations suggest that, whilst enhanced corrosion due to MnS dissolution from the steel is the cause of retarded crack growth rates, different retardation mechanisms appear to contribute at high and low stress ratios. This programme was sponsored by The Electric Power Research Institute (EPRI) and a full report describing the full programme of work is available as EPRI report#3002007973.

This paper summarizes the results of a research project on environmental effects on the fatigue lifetime of selected low-alloy steels and austenitic stainless steels. Results from investigations on the effect of hold times during fatigue in air, as well as under simulated PWR primary water conditions are presented. Strain controlled fatigue tests with low strain rates and low strain amplitudes have been performed in air using a low-alloy RPV steel as well as the stabilized austenitic stainless steel 347 and the non-stabilized austenitic stainless steel 304L. Similar strain controlled tests under simulated PWR primary coolant conditions were performed using the two above mentioned austenitic stainless steels. The fatigue cycling was performed at 240 °C whereas holds were applied at higher temperature (290 °C). The effect of hold periods is compared to reference tests without hold times and to the prediction based on NUREG/CR 6909, Rev. 1.

Powder bed fusion Direct Metal Laser Melting (DMLM) is an evolving additive manufacturing (AM) fabrication technology that is providing high performance parts to many industries. This technology has significant promise for use in building components for nuclear power plants. Implementation of materials produced using this and similar processes offer a potential step change in efficiency for complex parts production and hence a potential for innovative design as well as cost savings for components in the future. Properties of AM Type 316L have been reported in previous work, showing properties that match wrought properties. The fine grain structure may even lead to better environmental resistance. However, there is a need to confirm the behavior of these innovative materials after exposure to radiation if this innovative technology is to be used in current and future nuclear applications. This paper discusses new efforts being explored via a joint program between GE Hitachi (GEH) and INL (Idaho National laboratory) aimed at developing corresponding un-irradiated and irradiated data for AM materials. This paper will present data for both Type 316L stainless steel, a single-phase alloy, and Ni-base Alloy 718, a precipitation hardened alloy, manufactured using AM. This paper, serving as a progress report, will present the mechanical property and microstructural data for both Type 316L and 718 AM alloys to assess their correspondence to wrought alloy data and establish a baseline for future comparison to irradiated properties. The paper will end by discussing the requirements for using these and other additively manufactured materials in future reactor component applications where irradiated data is not available.

Westinghouse began additively manufacturedAdditively manufactured (AM) materials research activities in 2012 to support development of advanced fuel related components. The initial objective of this work has been the fabrication and delivery of a lead test component to a Westinghouse nuclear utility customer for in-reactor insertion for a limited number of fuel cycles. It is generally recognized that a key criterion for the implementation of AM components would be a thorough understanding of the material response to neutron irradiation. Several alloys were AM fabricated, heat treated, neutron irradiated, and extensively evaluated both in the as-printed condition and in the printed and irradiated condition. Irradiation of AM miniature tensile specimens was performed at the Massachusetts Institute of Technology (MIT) Nuclear Research Reactor to 0.8 dpa at 300 °C (572 °F). Although extensive laboratory testing was performed on these materials, this paper specifically summarizes the results from room and elevated temperature tensile testing of unirradiated and irradiated AM 316L. Testing of the miniature tensile specimens was performed inside a hot cell utilizing in-cell digital image correlationDigital image correlation (DIC) and advanced video extensometryAdvanced video extensometry (AVE). Additional AM alloys are currently being irradiated in the MIT reactor to higher dpa values. The first of these samples will be shipped and subsequently tested in 2017.

Materials in nuclear power system can suffer from thermal/hydrothermal, radiation and chemical degradation due to the high-temperature, high-pressure operation condition along with the presence of water steam and radiation. One particular topic we are addressing is understanding and optimizing materials for fission gas capture. Computational modeling is an efficient tool to investigate materials behaviour in such extreme environment. Westudied a number of materials. One of these is mesoporous silica. We used a combination of Molecular Dynamics (MD) simulation and Monte Carlo (MC) simulation which were validated by detailed experiments. MD simulations reveal the porous structure transformation under high-temperature treatment up to 2885 K, suggesting the pore closure process is kinetically dependent. Based on this mechanism, we predict with the presence of water, the pore closure activation energy will be decreased due to the high reactivity between water and Si-O bond, and the materials become more susceptible to high temperature. A fundamental improvement of the material hydrothermal stability thus lies in bond strengthening. MC simulations then were used to study the the adsorption and selectivity for thermally treated MCM-41, for a variety o f gases in a large pressure range. Relative to pristine MCM-41, we observe that high temperature treated MCM 41 with its surface roughness and decreasing pore size amplifies the selectivity of gases. In particular, we find that adsorption of strongly interacting molecules can be enhanced in the low-pressure region while adsorption of weakly interacting molecules is inhibited. We have also investigated alumina as an example of a ceramic material that can be directly incorporated into the nuclear fuel itself. Unlike uranium oxide fuel, certain phases of alumina have appreciable capacity for gas absorption. The limited diffusion distance of helium and other fission product gases in the fuel may be addressed by coating micron-sized fuel particles with alumina, prior to sintering, using a unique atomic layer deposition process suitable for particles. We have investigated the feasibility of this approach using a combination of helium-focused experiments on fuel surrogate particles, together with analytical calculations of gas production rates and diffusion distances in uranium oxide. Additional studies of nanotubes of carbon and boronitride elucidated fundamental mechanisms of the influence of curvature on gas adsorption.

Martensitic stainless steels, in general, become more susceptible to Environmentally Assisted Cracking (EAC), specifically Hydrogen Assisted Cracking (HAC),Hydrogen Assisted Cracking (HAC) with increasing tensile strength (as reflected by increasing hardness). The aim of this test programme was to determine the susceptibility to HAC of a 12% chromium stainless steel as a function of material hardness, stress and environment. Incremental Step Loading (ISL) tests demonstrate a reduction in failure stress with increasing hardness due to the presence of hydrogen. Relationships between failure stress and hardness/tempering temperature are described. Testing also clearly supports the concept that there is a critical value of nominal stress, at each tempering temperature/hardness, below which HAC does not occur. Constant displacement testing results show that susceptibility to HAC is dependent upon a complex interplay between microstructure (tempering temperature/hardness), stress and environment (availability of hydrogen).

To investigate various methods to mitigate flow accelerated corrosion of carbon steels, we have deposited various metallic and composite coatings on the surface of carbon steels and tested their performance by exploiting flow accelerated corrosion (FAC) simulation experiments. From the results, we found that both Ni-P/TiO₂ composite coating and Fe-based amorphous metallic coating exhibited outstanding FAC resistance thus they are expected to expand the life-time of secondary systems of nuclear power plants. Furthermore, to investigate their life-time in nuclear power plants, we investigated known mechanistic models and commercial models of FAC and imported the parameters of the coated carbon steels into the models.

The structural integrity of the reactor pressure vessel (RPV) of light water reactors (LWR) is of utmost importance regarding operation safety and lifetime. The fracture behaviour of low-alloy RPV steels with different DSA (dynamic strain aging) & EAC (environmental assisted cracking) susceptibilities and microstructures (base metal, simulated weld coarse grain heat affected zone) in simulated LWR environments was evaluated by elastic plastic fracture mechanics (EPFM) tests with different strain rates and by metallo- and fractographic post-test observations. These tests revealed some evidences of high-temperature water and hydrogen effects on the fracture behaviour and potential synergies with DSA and EAC.

Corrosion fatigueCorrosion fatigue testing of a low alloy steelLow alloy steel was undertaken to determine the effect of cycling parameters and low oxygenOxygen effects levels on the crack growth rate at various stress intensity factor ranges. Notched, pre-cracked compact tension specimens were prepared from A516-90 plate material with a sulfur content of 0.020 wt%. These specimens were tested in a water environment with a pH of ~9.0 and a temperature of 177 °C. At each stress intensity factor range, the crack growth rate was compared at three different frequencies with oxygen levels <10 to 150 ppb. At higher stress intensity factor ranges, no effect on crack growth rate from oxygen, rise time, or dwell time was observed. For the lower stress intensity factor ranges, the crack growth rate decreased with oxygen addition and additional dwell times at maximum load. The decrease in crack growth rate at lower stress intensity factors is attributed to crack tip blunting and/or crack closure effects from oxide build-up. At the higher stress intensity factors, the mechanical crack driving force was sufficient to break the oxide and continue growing the crack.

A specifically designed reference electrode was developed for analyzing the electrochemical behaviors of alloy materials in supercritical water (SCW) environments and identifying the associated electrochemical parameters. The internal Zr/ZrO₂ reference electrodes constructed for high-temperature conditions were manufactured and adopted to measure the electrochemical corrosion potential (ECP) of the sample in SCW environments. Before the electrochemical analysis, the oxidation behaviour of zirconium would be investigated in SCW environment. The mass gain of zirconium is assumed due to formation of ZrO₂ and there was only 78% of the original thickness of zirconium existed after the 1300 h immersion test in SCW environments with 8.3 ppm dissolved oxygen. In deaerated SCW environments, the thickness of zirconium is about 88% of the original one. The outcome indicated that the laboratory-made Zr/ZrO₂ reference electrode was able to continuously operate for several months and delivered consistent and steady ECP data of the sample in SCW environments.

The influence has been studied of thermo-mechanical treatment, sensitization conditions, and neutron irradiation on the pitting corrosion resistance of austenitic 316LN stainless steel variants in 10% FeCl₃·6H₂O at 22 °C. Variants of this steel were modified with additions of nitrogen, manganese, copper, and tungsten, as well as testing cast, cold-rolled, grain boundary engineered (GBE), and as-received variants. It was found that the 316LN steel variant with additions of 0.2% N and 2% Mn had the best pitting corrosion resistance of all studied conditions. When irradiated in a light water reactor (LWR) to a maximum fluence of 3 × 1017 n/cm2 (E > 1.1 meV, T_irr < 50 °C), neutron irradiation surprisingly increased the resistance of GBE steels to pitting corrosion. An anisotropy of corrosion resistance of GBE and cold rolled steels was observed.

Solid particle-induced wall thinning of water-cooled nuclear power plant components is an established degradation mechanism that can affect the long term management of plant operations and component reliability. This form of material degradation is identified as erosion-corrosion and is comprised of mechanical removal known as erosion, chemical removal or corrosion, and the enhanced degradation resulting from the combined action of erosion and corrosion known as synergistic wear. Erosion-corrosion’s complicated analysis involves many factors such as fluid velocity, particle size, concentration and shape, pH, and temperature amongst several others. Erosion-corrosion is observed in heat exchanger tubing exposed to raw water (water from ponds, rivers, bays, and lakes), and steam generator blowdown piping. However, quantifying the impact of solid and liquid mixtures on light water reactor component reliability has proven to be difficult, and often underestimated. This paper describes a physics-based and probabilistic erosion model for estimating average wall thinning rates and age-dependent probabilities of exceeding user-defined wall thicknesses.

In Japan, light water reactors are built on the seacoast because they use seawater as the final heatsink. Leakage of seawater from the condenser section of the reactor could lead to contamination of the reactor coolant, and stainless steels can be susceptible to crevice corrosion in chloride-contaminated water. Therefore, it is necessary to develop counter measures for suppressing the initiation of crevice corrosion and for repassivating the corroding crevice to maintain structural reliability. To accomplish this, first the effect of molybdate anion on suppressing the initiation of crevice corrosion on 316L stainless steel in chloride-contaminated water was evaluated by potentiostatic immersion tests. Next, the effect of molybdate anion addition on the repassivation of corroding crevices was also evaluated through potentiostatic immersion tests as a function of the concentration of chloride anion. Based on the results of these examinations, the beneficial effects of the presence of molybdate anion on the suppression of initiation and propagation of crevice corrosion were quantitatively evaluated in terms of critical potentials.

Thermal desorption spectroscopy, secondary ion mass spectroscopy and scanning transmission electron microscopy have been used to investigate the effect of pH on corrosion and hydrogen pick-up behaviour in different samples of Zircaloy-4Zircaloy-4. Samples were autoclave-oxidised in pure water and at an elevated pH (with 50% deuterated water) when compared to commercial reactors. A characteristic desorptionDesorption peak for hydrogenHydrogen has been found at ~650 °C, which occurs when the difference in free energy between hydrogen in the metal and in the gas phase becomes positive. Electron energy loss spectroscopy provided us with a method to detect and measure the thickness of the following layers (from oxide to metal): ZrO₂, a previously reported ZrO suboxide, an oxygen saturated zirconium region and the Zr metal. Overall, samples exposed to a high pH show a longer time to transition and contain far less hydrogen than those oxidised in pure water. A mechanistic explanation will be provided.

The oxidation kineticsOxidation kinetics of supercritical-water-cooled reactor (SCWR) fuel cladding candidate materials, e.g. 15Cr-20Ni stainless steel (1520 SS) in supercritical waterSuper critical water reactor at 650 and 700 °C under 24 MPa has been investigated. Characteristics of oxide layers and its relation to oxidation behaviors are also studied. The applicability of the candidate materials for the fuel cladding of SCWR from oxidation kinetics, spalling susceptibility of oxide layer, and breakdown of Cr₂O₃ layer points of view has been discussed. The results indicate that the threshold condition for spalling of oxide layer is different at 650 and 700 °C. The decrease in oxidation kinetics of 1520 SS with time correspond to the change in rate-limiting process of oxidation from mass transfer through an Fe oxides to mass transfer through a Cr rich oxide layer with time. Based on the oxidation kinetics obtained in this study, 1520 SS is considered suitable for a fuel cladding of SCWR in combination with appropriate CW process. However, detailed evaluation and countermeasures for the degradation due to nodular oxidation are needed before application of tube-shaped 1520 SS in supercritical water at 700 °C. On the other hand, it is estimated that the use of that at 650 °C is acceptable because the weight gain after long-term exposure was considered to be much less than the threshold condition of the spalling.

The primary heat transport system of modern CANDU® (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactors uses A106B piping (i.e., feeder pipes). Feeder cracking has only affected tight-radius bends at outlet feeders (higher temperature), and cracking is limited to regions with high residual stress suffering from wall-thinning by flow accelerated corrosion. To date, the mechanism of feeder cracking has not been identified. This paper includes high-resolution transmission electron microscopy and electron energy loss near edge structure characterization of inside and outside surface intergranular cracks from ex-service CANDU feeders. Prior to this work, no high resolution characterization has been performed for CANDU feeder cracking. All intergranular cracks show evidence of cementite decomposition, leading to decoration of grain boundaries with amorphous carbon, and carbon diffusion along un-cracked boundaries ahead of crack tips. Sulfur has been found on the oxide-metal interface of all intergranular cracks, but is not observed ahead of the crack tips. Sulfur is believed to be from the breakdown of manganese sulfides during service. The cementite decomposition and breakdown of manganese sulfides are believed to be accelerated in the presence of hydrogen produced from the flow accelerated corrosion. Small (<15 nm) voids are also present ahead of some intergranular crack-tips along the ferrite-ferrite boundaries, indicating that hydrogen enhanced, low temperature creep-cracking, may also contribute to intergranular fracture.