Caution when applying eddy current inversion to stress corrosion cracking

This study evaluates the applicability of computer-aided eddy current inversion techniques to the profile evaluation of stress corrosion cracking in Inconel welds. Welded plate specimens, which model head penetration welds of pressurized water reactors, are fabricated; notches and stress corrosion cracks are artificially introduced into the weld metal of the specimens. Eddy current inspections are performed using a uniform eddy current probe driven at frequencies of 10 and 40 kHz. Since weld noise is observed uniformly along the weld line, a simple signal processing is applied to eliminate it. First, the artificial notches are reconstructed and good agreements between reconstructed and true profiles are provided, which demonstrates that the computer-aided eddy current inversion technique can deal with defects in welds. Then, numerical simulations are performed to evaluate the profiles of the stress corrosion cracks. In the numerical simulations, the stress corrosion cracks are modeled as a conductive region with a fixed width of 0.3 mm. The cross-sectional profiles of the cracks are reconstructed from measured eddy current signals directly above and along a crack. Although eddy current signals calculated from the reconstructed profiles agree well with measured ones, the true profiles revealed by destructive testing are found to be very different from the reconstructed ones. Whereas the most plausible reason for the difference is the unexpectedly volumetric profile of the stress corrosion cracks, this study has revealed that computer-aided eddy current inversion techniques that have been used to consider cracks in thin structures would not at this point be directly applicable to those in thick structures. It is also important to know in advance those crack features that can adversely impact accurate crack sizing including whether a detected crack is volumetric or not, namely there are many parallel cracks in a cluster or not.     

Halogen stress corrosion cracking of zircaloy-4 tubing

The stress corrosion behavior in the presence of iodine vapor or chlorine gas of unirradiated, recrystallized and stress-relieved Zircaloy-4 as a function of hoop stress, test temperature and contaminant concentration has been investigated using internally pressurized tube specimens. Comparison with similarly tested control specimens (no halogen) shows that at 360°C and 400°C as little as 0.005 to 0.01 mg halogen per cm² Zircaloy surface is sufficient to reduce failure time for hoop stresses above 21 to 24 kpsi. Failure time appears to be essentially independent of iodine up to 0.08 mg/cm² . At equivalent stress ratios (σapplied/σYS) with iodine (i) the higher temperature results in shorter failure times over most of the stress range investigated and (ii) stress-relieved material generally failed before recrystallized material. Failures were typically pinholes except at high stresses where stress rupture occurred. Scanning electron microscopy revealed intergranular failure in recrystallized specimens and the presence of fluted facets in both materials.     

Modelling of iodine-induced stress corrosion cracking in CANDU fuel

Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics® commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.     

Localized deformation as a key precursor to initiation of intergranular stress corrosion cracking of austenitic stainless steels employed in nuclear power plants

Cold-work has been associated with the occurrence of intergranular cracking of stainless steels employed in light water reactors. This study examined the deformation behavior of AISI 304, AISI 347 and a higher stacking fault energy model alloy subjected to bulk cold-work and (for 347) surface deformation. Deformation microstructures of the materials were examined and correlated with their particular mechanical response under different conditions of temperature, strain rate and degree of prior cold-work. Select slow-strain rate tensile tests in autoclaves enabled the role of local strain heterogeneity in crack initiation in pressurized water reactor environments to be considered. The high stacking fault energy material exhibited uniform strain hardening, even at sub-zero temperatures, while the commercial stainless steels showed significant heterogeneity in their strain response. Surface treatments introduced local cold-work, which had a clear effect on the surface roughness and hardness, and on near-surface residual stress profiles. Autoclave tests led to transgranular surface cracking for a circumferentially ground surface, and intergranular crack initiation for a polished surface.     

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.     

Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20–100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ≈50 dpa at ≈370 °C. Slow-strain-rate tensile tests were conducted at 289 °C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at a low ECP, and this susceptibility led to a poor work-hardening capability and low ductility.     

Simulation of stress corrosion cracking behavior in a tube-shaped specimen of nickel-based alloy 600

Stress corrosion cracking (SCC) simulation code has been developed for the evaluation of SCC behavior in specimens in the shape of field components. The code utilizes numerical calculation of stress/strain states at a crack tip using finite element methods and a formula describing the crack tip reaction kinetics containing unknown environmental parameters. The applicability of this simulation code was investigated by applying the code to the evaluation of SCC behavior in a mock-up of a bottom mounted instrumentation tube for a pressurized water reactor subjected to complex stress/strain states. The results indicate that crack growth rate in a component suffering from certain environments can be estimated using the developed SCC simulation code with pre-determined unknown parameters, using the experimental crack growth rate data measured on other specimens in the same environment.     

An analysis of iodine stress corrosion cracking of zircaloy-4 tubing

A fracture mechanics approach to interpreting iodine-vapor stress-corrosion cracking in unirradiated Zircaloy-4 tubing is presented in which crack velocities are related to the fourth power on the stress intensity factor, $\text{K}{}_{\mathrm{I}}$. The crack growth power law on $\text{K}{}_{\mathrm{I}}$ is shown to predict well the time-to-failure in internally pressurized Zircaloy-4 tubing at 360 and 400°C reported by Busby, Tucker and McCauley. The temperature dependency on iodine stress corrosion cracking in Zircaloy can be described by an Arrhenius-type equation in which the activation energy $\text{Q}$ for recrystallized and cold-reduced Zircaloy was determined to be 42.9 and 35.9 kcal/mole, respectively. It is concluded that the geometry of the initial surface flaw, through its attendant elastic stress field, is directly responsible in controlling the SCC time-to-failure, cold working having a relatively small effect on increasing the susceptibility to SCC. The effects of neutron flux on iodine stress corrosion cracking of Zircaloy-4 tubing in-reactor are still unknown.     

Boiling Water Reactor stress corrosion cracking of piping-utility industry research program

Intergranular stress corrosion cracking of Boiling Water Reactor piping has been a prominent problem since 1974 and has reduced plant availability as a result of the requirements (1) for inspection of piping susceptible to cracking and (2) for repair or replacement of piping that has cracked. In 1979, a Boiling Water Reactor Owners Group was formed to expand upon and, where possible, accelerate research into the causes of the problem and its remedies. In this paper, the research program of this international group is described and the program's results to date and expectations are highlighted. Interactions with regulatory groups in the United States of America and reactions of the regulatory groups to the research program are also presented.     

Probabilistic methodology to estimate environmental conditions for localized corrosion and stress corrosion cracking of Alloy 22 in a high-level radioactive waste repository setting

The US Department of Energy (DOE) has indicated that it may use Alloy 22 (Ni-22Cr-13Mo-4Fe-3W) as the waste package outer container material for the potential high-level waste repository at Yucca Mountain, Nevada. This alloy could be susceptible to localized corrosion, in the form of crevice corrosion, and stress corrosion cracking if environmental conditions and material requirements (e.g., existence of crevices or high enough tensile stresses) are met. An approach is proposed to assess the likelihood of environmental conditions capable of inducing crevice corrosion or stress corrosion cracking in Alloy 22. The approach is based on thermodynamic simulations of evaporation of porewaters and published equations to compute corrosion potential and critical potentials for crevice corrosion and stress corrosion cracking as functions of pH, ionic concentration, temperature, and metallurgical states from fabrication processes. Examples are presented to show how the approach can be used in system-level assessment of repository performance.     

Primary water stress corrosion cracking inspection ranking scheme for alloy 600 components

The Palisades nuclear plant has developed a comprehensive inspection program to support safe, reliable, and cost-effective operation of all Alloy 600 nozzles and safe ends in the primary coolant system (PCS). As a part of the Palisades Alloy 600 Project, an inspection prioritization scheme was developed to help the plant focus its resources on high-risk components and plan appropriate inspection activities for the other components. The inspection prioritization scheme is based on the susceptibility of the components to primary water stress corrosion cracking (PWSCC), component failure consequences, component leak detectability and component radiation exposure. The scheme provides a simple, systematic and technical base for selecting Alloy 600 components for inspection. The scheme, however, could be used to develop an inspection schedule or to select the highest priority components for mitigation or replacement.     

Quantitative examination of stress corrosion cracking of alloy 600 in high temperature water

Stress corrosion cracking (SCC) examination of Inconel 600 steam generator tubing has continued at Brookhaven National Laboratory, using U-bends, constant load and slow extension rate tests, leading to Arrhenius plots of failure times versus inverse temperature for crack initiation and propagation. Effect of applied load can be expressed as log-log curves for failure times versus stress. Variations in environment and cold work are included in all the experiments. Microstructure and composition of oxide films on Inconel 600 surfaces were examined after exposure to pure water at 365°C, and stripping with the bromine-methanol method. Results are consistent with a mechanism of transient creep, film rupture and a mass-transport-limited anodic process.     

Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals

The stress corrosion cracking (SCC) rate of reactor internals of boiling water reactors (BWR) continues to increase with on-line operating years. The recent occurrences of cracking in the weld heat affected zones of high carbon stainless steel core shrouds correlate with the years of operation and the water chemistry history. Recently, cracking has also been found in shrouds that were constructed of low carbon or stabilized stainless steels. While these steels are more resistant to intergranular stress corrosion cracking (IGSCC) in the as-fabricated condition, this field experience establishes that the conditions under which the materials will crack in core structures are attributable to the combined effects of high residual stresses, associated with the shroud construction, the presence of a more aggressive, oxidizing environment in the core and to microstructural changes in the material. These changes result from the manufacturing process as well as thermal exposure during operation. Studies of materials that have cracked in the field, as well as high temperature material studies in the laboratory, are being performed to understand the mechanisms. The use of highly oxidizing, high purity water environments is integral to reproducing the conditions for cracking. The status of the laboratory efforts to gain understanding and to verify the mechanisms are presented. Modeling of IGSCC is also a key tool used to understand the cracking behavior of the low carbon stainless steels. The PLEDGE (Plant Life Extension Diagnosis by GE) model is used to support the hypotheses that tie together the role of the different contributing elements: residual stress, core water chemistry and microstructural features. The crack growth modeling is also used to evaluate the benefits of different strategies to manage and mitigate cracking of reactor internals such as hydrogen water chemistry.     

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.     

The effect of Nb additive on Te-induced stress corrosion cracking in Ni alloy: a first-principles calculation

Nb can improve the resistance of Ni-based Hastelloy N alloy to Te-induced intergranular embrittlement. First-principles calculations are performed to research this mechanism by simulating the Ni（111） surface and the ∑ 5（012） grain boundary. The calculated adsorption energy suggests that Te atoms prefer diffusing along the grain boundary to forming the surface-reaction layer with Nb on surface of the Ni alloy. First-principles ten- sile tests show that the Nb segregation can enhance the cohesion of grain boundary. The strong Nb-Ni bonding can prevent the Te migration into the inside of the alloy. According to the Rice-Wang model, the strengthen- ing/embrittling energies of Nb and Te are calculated, along with their mechanical and chemical components. The chemical bonds and electronic structures are analyzed to uncover the physical origin of the different effects of Te and Nb. Our work sheds lights on the effect of Nb additive on the Te-induced intergranular embrittlement in Hastelloy N alloy on the atomic and electronic level.     

Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600

In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.     

Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water reactor (BWR). Therefore, the evaluation of SCC propagation is important for the safe maintenance of the core shroud. This paper attempts to predict the remaining life of the core shroud due to SCC failures in BWR conditions via SCC propagation time calculations. First, a two-dimensional finite element method model containing H6a girth weld in the core shroud was constructed, and the weld processing was simulated to determine the weld's residual stress distribution. Second, using a basic weld residual stress field, the SCC propagation was simulated using a node release option and the stress redistribution was calculated. Combined with the J-integral method, the stress intensity factors were calculated at depths of 2, 3, 4, 8, 12, 16, 19, 22, 25 and 30 mm in the crack setting inside the core shroud; then, the SCC propagation rates were determined using the relation between the SCC propagation rate and the stress intensity factor. The calculations show that the core shroud could safely remain in service after 9.29 years even when a 1-mm-deep SCC has been detected.     

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.