Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation

The failure of steam generator tubing is one of the main accidents that seriously affects the availability and safety of a nuclear power plant. In order to estimate the probability of the failure, a probabilistic model was established to predict the whole life-span and residual life of steam generator (SG) tubing. The failure investigated was stress corrosion cracking (SCC) after the generation of one through-wall axial crack. Two failure modes called rupture mode and leak mode based on probabilistic fracture mechanics were considered in this proposed model. It took into account the variance in tube geometry and material properties, and the variance in residual stresses and operating conditions, all of which govern the propagations of cracks. The proposed model was numerically calculated by using Monte Carlo Simulation (MCS). The plugging criteria were first verified and then the whole life-span and residual life of the SG tubing were obtained. Finally, important sensitivity analysis was also carried out to identify the most important parameters affecting the life of SG tubing. The results will be useful in developing optimum strategies for life-cycle management of the feedwater system in nuclear power plants.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Evaluation of the burst characteristics for axial notches on SG tubings

Some events of steam generator tubes have been reported in some nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined (EDM) notches having different lengths were machined on the o.d. of test tubes to study SG tube behavior. Leak rate and ligament rupture pressure as well as the burst pressure were measured for the tubes at room temperature. Rupture pressure of the part through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independent of the flaw types; tubes having 20–60 mm long EDM notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate generated a lower burst pressure than the case of a slow pressurization.     

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.     

Evaluation of SCC crack behavior in zirconium and zircaloy-2 using nonlinear fracture mechanics parameters

Applicability of nonlinear fracture mechanics parameters, i.e. J-integral, crack tip opening displacement (CTOD), and crack tip opening angle (CTOA), to evaluation of stress corrosion crack (SCC) propagation rate was investigated using fully annealed zirconium plates and Zircaloy-2 tubing, both of which produce SCC with comparatively large plastic strain in an iodine environment at high temperatures.Tensile SCC tests were carried out at 300°C for center-notched zirconium plates and internal gas pressurization SCC tests at 350°C, for Zircaloy-2 tubing, to measure the SCC crack propagation rate. The J-integral around semi-elliptical SCC cracks produced in Zircaloy-2 tubing was calculated by a three-dimensional finite element method (FEM) code.The test results revealed that the SCC crack propagation rate dc/dt could be expressed as a function of the J-integral, which is the most frequently used parameter in nonlinear fracture mechanics, by the equation dc/dt = C · Jⁿ, where C and n were experimental constants.Among the other parameters, CTOD and CTOA, the latter appeared to be useful for assessing the crack propagation rate, because it had a tendency to hold a constant value at various crack depths.     

Evaluation of the mitigation effect of hydrogen water chemistry in BWRs on the low-frequency corrosion fatigue crack growth in low-alloy steels

The mitigation effect of hydrogen water chemistry (HWC) on the low-frequency corrosion fatigue crack growth behaviour of low-alloy steels was investigated under those critical boiling water reactor (BWR) system conditions, where fast corrosion fatigue crack growth significantly above the ASME XI ‘wet’ reference fatigue crack growth curves was observed under normal water chemistry conditions (NWC). The experiments were performed under simulated BWR conditions at temperatures of 250, 274 or 288 °C. Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by scanning electron microscope were used to quantify the cracking response. HWC resulted in a significant drop of low-frequency corrosion fatigue crack growth rates by at least one order of magnitude with respect to NWC conditions and is therefore a promising and powerful mitigation method.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

A study of life prediction differences for a nickel-base Alloy 690 using a threshold and a non-threshold model

In this paper we compare and contrast the crack growth rate of a nickel-base superalloy (Alloy 690) in the Pressurized Water Reactor (PWR) environment. Over the last few years, a preponderance of test data has been gathered on both Alloy 690 thick plate and Alloy 690 tubing. The original model, essentially based on a small data set for thick plate, compensated for temperature, load ratio and stress-intensity range but did not compensate for the fatigue threshold of the material. As additional test data on both plate and tube product became available the model was gradually revised to account for threshold properties. Both the original and revised models generated acceptable results for data that were above 1 × 10⁻¹¹ m/s. However, the test data at the lower growth rates were over-predicted by the non-threshold model. Since the original model did not take the fatigue threshold into account, this model predicted no operating stress below which the material would effectively undergo fatigue crack growth. Because of an over-prediction of the growth rate below 1 × 10⁻¹¹ m/s, due to a combination of low stress, small crack size and long rise-time, the model in general leads to an under-prediction of the total available life of the components.     

Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason’s model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.     

Bilinear CTOD/CTOA scheme for characterisation of large range mode I and mixed mode stable crack growth through AISI 4340 steel

An investigation has been reported concerning characterisation of large span stable crack growth (SCG) through AISI 4340 steel in terms of CTOD/CTOA under both mode I and mixed (I and II) mode loadings. The characterisation has been possible through finite element analysis of published experimental results with compact tension type of specimens. As against the earlier observations by many investigators of a bilinear CTOD/CTOA variation, decreasing initially and constant later, characterising the crack growth through other materials, an increasing initially and constant later type of variation is found to be suitable for AISI 4340 steel. The same variation is found to characterise both mode I and mixed mode SCG. The starting value of CTOD/CTOA is 0.035 mm/0.0875 rad; the value at the later stages is 0.08 mm/0.20 rad. The same variation is found to predict accurate enough for engineering applications the initiation (P_i) and maximum (P_max) loads and the variation of load-displacement diagrams over a span of crack growth up to 10 mm.     

Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.     

Modeling of SCC initiation and propagation mechanisms in BWR environments

During operation of mainly BWRs’ (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 °C – corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 °C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with increasing global stresses. In accordance with respective experimental literature data it is shown that decreasing chloride and increasing pH levels of the primary bulk water at 288 °C reduce the total crack propagation rates including anodic path corrosion as well as hydrogen cracking. It is also demonstrated that crack propagation rates can be significantly suppressed by hydrogen water chemistry (HWC) that leads to reduction of bulk surface corrosion potentials. As a conclusion the extended SSC-model for nickel supplies quantitative insight into the frequently controversially discussed high temperature SCC mechanisms of a basic alloying element of BWR components.     

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.     

Creep behaviour of modified 9Cr-1Mo ferritic steel

Creep deformation and fracture behaviour of indigenously developed modified 9Cr-1Mo steel for steam generator (SG) tube application has been examined at 823, 848 and 873 K. Creep tests were performed on flat creep specimens machined from normalised and tempered SG tubes at stresses ranging from 125 to 275 MPa. The stress dependence of minimum creep rate obeyed Norton’s power law. Similarly, the rupture life dependence on stress obeyed a power law. The fracture mode remained transgranular at all test conditions examined. The analysis of creep data indicated that the steel obey Monkman-Grant and modified Monkman-Grant relationships and display high creep damage tolerance factor. The tertiary creep was examined in terms of the variations of time to onset of tertiary creep with rupture life, and a recently proposed concept of time to reach Monkman-Grant ductility, and its relationship with rupture life that depends only on damage tolerance factor. SG tube steel exhibited creep-rupture strength comparable to those reported in literature and specified in the nuclear design code RCC-MR.     

Chemical aspects of iodine-induced stress corrosion cracking failure of zircaloy-4 tubing above 50°C

The chemical environment associated with iodine-induced SCC failure of Zircaloy-4 tubing above 500°C has been characterized. At the critical iodine concentrations which result in SCC initiation and propagation, most of the iodine is present as condensed zirconium subiodides (I/Zr ? 0.4). Only a small part of the iodine remains in the gas phase as ZrI₄. The gaseous ZrI₄ is probably responsible for crack initiation and propagation. The critical ZrI₄ pressures for SCC failure have been estimated in zircaloy/iodine reaction experiments performed with unstressed zircaloy tube specimens. These pressures were confirmed in additional creep rupture tests conducted under controlled ZrI₄ partial pressure conditions. The estimated critical ZrI₄ pressure above which low-ductility SCC failure of the zircaloy tubing always occurs, independent of time-to-failure, varies between 0.005 bar at 550°C and 0.043 bar at 800°C. Below the critical values, however, a rather wide range of ZrI₄ pressures is associated with the onset of the SCC, especially at temperatures below 800°C. A comparison of the experimental results with available thermochemical data in the Zr-I system indicates that the main reaction involved during crack propagation is chemisorption of iodine-containing species on the fresh zircaloy surfaces created by metal straining at the crack tip.     

Fracture mechanics analysis of iodine-induced crack growth in Zircaloy-4 tubing between 500 and 700°C

Low ductility failure of zircaloy tubing due to iodine-induced stress corrosion cracking (SCC) can occur up to about 700°C. The time-to-failure behavior of Zircaloy-4 cladding tubes containing iodine has been described by the elastic-plastic fracture mechanics model CEPFRAME for the temperature region 500 to 700°C. The model includes an empirically-determined computation method for the incubation period of crack formation, as a portion of the time-to-failure, as well as an elastic-plastic model for describing crack growth due to iodine-induced SCC. The total life time of the cladding tube is obtained by adding the crack initiation and crack propagation periods. The incubation period is a temperature-dependent function of both the depth of surface damage (both fabrication pits and machined notches) and the applied load, and is 40 to 90% of the time-to-failure. The elastic-plastic crack growth model is a modified version of the stress intensity K_I-concept of linear-elastic fracture mechanics. The extensions of this concept take into account a plastic strain zone ahead of the crack tip, which effectively increases the crack depth, and in addition, a dynamic correction factor for the crack geometry which is essentially a function of the effective crack depth. Unstable crack growth is predicted to occur when the residual cross section reaches plastic instability.Model results show good agreement with experimental data of tube burst tests at 500, 600, and 700°C. The crack growth velocity at all three temperatures is a power function of stress intensity ahead of the crack tip; the exponent is 4.9. The model can estimate time-to-failure of as-received cladding tubes containing iodine within a factor of 2. Application of the model to temperatures below 500°C is possible in principle. Due to the increasing scatter in experimental data, the structural transformation of the cladding by recrystallization, and the growing importance of creep strain, CEPFRAME has an upper temperature limit of approximately 650°C. The model is suitable for use in computer codes describing LWR fuel rod behavior during reactor transients and accidents.     

Life prediction on SCC of solution annealed stainless steels under laboratory conditions

The stress corrosion cracking (SCC) behavior of austenitic stainless steels, types 304 and 316 has been investigated in acidic solutions to verify whether or not a parameter for prediction of time to failure can be detected as functions of applied stress and environmental factors (temperature, concentration, pH, anion species) by using a constant load. The results show that the steady state elongation rate becomes a parameter for predicting time to failure at a time within 10–20% of time to failure irrespective of the above factors. The steady state elongation rate is also found to become a parameter for the assessment of SCC susceptibility. On the basis of the results obtained, a SCC mechanism is discussed in terms of corrosion current density at crack tips, time to failure and length of crack propagation. The SCC mechanism proposed can be applied to both active path dissolution model and film rupture model.     

Trans-twins stress corrosion cracking behaviors of Alloy 690TT in lead-containing caustic solution at 330 °C

Twin grain boundaries (GBs) are found to be inherently resistant to stress corrosion cracking (SCC), which has become one of the main failure mechanisms of steam generator (SG) tubing since the 1980s and brings huge economic losses to the nuclear power plants. As it is a widely used material for SG tubing, the SCC-resistance of the twins in Alloy 690TT in 10 wt.% sodium hydroxide solution with 100 ppm litharge at 330 °C was studied using C-ring samples. The relationship between the crack paths, twin GBs and the residual strains in the studied areas were analyzed using an environmental scanning electron microscope (ESEM) equipped with electron backscatter diffraction (EBSD) equipment. A continuously stressed C-ring sample without immersion was used to evaluate the effect of residual stress or strain on the microstructure of the twin GBs. The oxides at the crack paths were analyzed by an energy dispersive spectroscopy (EDS). The results show that many twin GBs are cracked during crack propagation. There are more twins with large deviations from the ideal ∑3 twin misorientation in the studied area where the residual strain is high. In situ EBSD analyses verify that higher residual strain causes twins to deviate from the ideal twin microrientation and can even promote twins transiting into random high angle grain boundaries, when the residual strain is high enough. The EDS result illustrates that litharge accelerates the dissolution of the chromium and nickel in the matrix. Overall, the SCC-resistance of the twins in Alloy 690TT in the studied solution is reduced by the destruction of the ideal microrientation of the twin GBs and the preferential dissolution of chromium and nickel at the crack paths. Higher residual strain on the Alloy 690TT and deleterious impurities in the circulating secondary water should be eliminated during the operation of nuclear power plants.     

An aging assessment on ex-service Alloy 800 steam generator tubing

To support SG life extension and plant life management, an aging assessment was performed on a number of ex-service Alloy 800 steam generator (SG) tubes removed from three CANDU^®1 stations with service life spanning from 2 to 27 years. Laboratory tests and examinations were carried out to investigate the potential aging mechanisms of SG tubing. High-temperature electrochemical experiments were performed under simulated SG secondary side crevice chemistry conditions to determine the corrosion susceptibility of the ex-service tubing; metallurgical examinations were carried out to check the chemical compositions, grain size, and hardness of the ex-service tubing materials and secondary ion mass spectrometry (SIMS) analysis was performed to assess the potential surface chromium depletion and grain boundary segregation of the ex-service tubing. Based on the results from the assessment, no increase in the corrosion susceptibility or changes in metallurgical properties of the ex-service tubes resulting from aging were observed. SIMS top-down profiles did not detect any aging-related surface chromium depletion on any of the ex-service tubes. However, SIMS imaging performed on the polished cross-sections of the ex-service tubes observed boron precipitation at the grain boundaries. Since no archived tubing with the same heat number as that of the ex-service tubing is available for comparison, whether the boron precipitation at grain boundaries is attributed to aging through SG operation is not conclusive and needs further clarification. In addition, the impact of this boron precipitation on the integrity of Alloy 800 SG tubing needs further investigation.     

Threshold conditions for iodine-induced stress corrosion cracking of unirradiated zircaloy-4 tubing under internal pressurization

The results are presented of stress corrosion cracking (SCC tests in which nuclear power reactor grade zircaloy-4 tubing specimens were internally pressurized with a mixture of helium and iodine at ($\text{633 ± 5}$) K. Both as-received and artificially preflawed specimens were tested at an initial iodine availability of ~60 g/m² zircaloy surface. It is shown that the failure times in these tests correlate more reliably with hoop stress than with nominal stress intensity or failure strain, and that a threshold hoop stress of ~295 MPa exists for SCC failure within test times up to 605 ks. The origin of this threshold stress is discussed and it is concluded that the observed behavior is consistent with either a critical stress or a critical strain rate being required for the formation of iodine-induced stress corrosion cracks in unirradiated zircaloy tubing.