Cracking in stabilized austenitic stainless steel piping of German boiling water reactors—characteristic features and root cause

Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur.     

The IGSCC, sensitization, and microstructure study of Alloys 600 and 690

Both Alloys 600 and 690 were studied to understand the effect of heat treatment on the sensitization and SCC behavior of these alloys. The microstructural evolution and chromium depletion near the grain boundaries were carefully studied using analytical electron microscopy. The majority of the precipitates formed in Alloy 600 was found to be M₇C₃ with a hexagonal structure (a₀ = 1.398 nm, c₀ = 0.45 nm); whereas the carbides found in Alloy 690 were identified as M₂₃C₆ with an fcc structure (a₀ = 1.06 nm). Modified Huey test performed in boiling 40% HNO₃ was used to study the effect of heat treatment and degree of sensitization. Constant load tests and constant extension rate tests were performed in the solution containing sodium thiosulfate to study the SCC resistance of these alloys. The results of the constant load tests for Alloy 600 indicated that the susceptibility to SCC is sensitive to the chromium depletion depth at grain boundary, and the minimum value to prevent SCC failure is approximately 8 wt%. No SCC was observed for Alloy 690 tested using constand load and CERT in the same environments. All tests showed that Alloy 690 has a far better resistance to intergranular attack and SCC than Alloy 600, which is believed due to its high chromium content. It is therefore anticipated that Alloy 690 now a better substitute to Alloy 600 as a steam generator tubing material for pressurized water reactor will also offer a superior corrosion resistance when “sensitized” and in particular if exposed to sulfur containing media such as thiosulfate solutions.     

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.     

The estimation of lifetime distribution of Alloy 800 steam generator tubing

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors worldwide and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under appropriate operating conditions. In planning refurbishment of nuclear plants stations, a key concern is the longevity of existing SGs up to the 60-year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and refines its estimation based on data specific to CANDU operating conditions. The paper presents a more advanced Bayesian probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of potential occurrences of degradation on reliability of SG tubing.     

PWSCC of thermally treated alloy 600 pulled from a Korean plant

Three tubes of alloy 600 were pulled out from a Korean nuclear power plant. The microstructure was analyzed using an optical microscope and TEM. Information on the crack length and depth was obtained by metallography, and crack detection and evolution were evaluated by analyzing the eddy current data obtained from each in-service-inspection (ISI). The carbon content in the pulled tubes was higher (around 0.03 wt.%) than that (around 0.015 wt.%) of Alloy 600 tubing used in other operating nuclear power plants. Most carbides in the pulled tubing were distributed in the grains rather than along the grain boundaries. The poor microstructure might come from high carbon contents, low temperature annealing, or high residual stresses during tube straightening. Mill annealing temperature should be high enough to dissolve all carbon in order to decorate the grain boundaries with semi-continuous carbide precipitation during 700 °C thermal treatment. Shot peening seemed to suppress the growth of the axial cracks, while it was analyzed to play a role in increasing crack growth in the wall thickness direction.     

Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation

The Canadian Nuclear Standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation of mechanical properties of Monel 400 tubes after 7-18 effective full power years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.     

The effect of grain boundary misorientation on the intergranular M23C6 carbide precipitation in thermally treated Alloy 690

The precipitation characteristics of chromium carbides on various types of grain boundaries in Alloy 690 thermally treated at 720 °C for 10 h were studied through transmission electron microscopy. Precipitation of the intergranular chromium carbides, identified as Cr-rich M₂₃C₆, was retarded on the low angle grain boundaries, compared to that on the random high angle grain boundaries on which coarse and discrete ones were found. They were rarely found on the coherent twin boundaries, however, needle-like ones were evolved on the incoherent twin and twin related Σ9 boundaries. Precipitation of the chromium carbides was also suppressed on the nearly exact coincidence site lattice boundaries such as Σ11 and Σ15, for which the Brandon criterion was fulfilled. The results of the intergranular M₂₃C₆ carbide precipitation were explained in terms of the influence of the grain boundary energy.     

Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment

Advanced transmission electron microscopy techniques were carried out in order to investigate stress corrosion cracking in Alloy 600 U-bend samples exposed in simulated PWR primary water at 330 °C. Using high-resolution imaging and fine-probe chemical analysis methods, ultrafine size oxides present inside cracks and intergranular attacks were nanoscale characterized. Results revealed predominance of Cr₂O₃ oxide and Ni-rich metal zones at the majority of encountered crack tip areas and at leading edge of intergranular attacks. However, NiO-structure oxide was predominant far from crack tip zones and within cracks propagating along twin boundaries and inside grains. These observations permit to suggest a mechanism for intergranular stress corrosion cracking of Alloy 600 in PWR primary water. Indeed, the results suggest that stress corrosion cracking is depending on chromium oxide growth in the grain boundary. Oxide growth seems to be dependent on oxygen diffusion in porous oxide and chromium diffusion in strained alloy and in grain boundary beyond crack tip. Strain could promote transport kinetic and oxide formation by increasing defaults rate like dislocations.     

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.     

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

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

Appling grain boundary engineering to Alloy 690 tube for enhancing intergranular corrosion resistance

The feasibility of applying the grain boundary engineering (GBE) processing to Alloy 690 tube manufacturing for improving the intergranular corrosion resistance was studied. Through small amount of deformation by cold drawing using a draw-bench on a production line and subsequent short time annealing at high temperature, the proportion of low Σ coincidence site lattice (CSL) grain boundaries of the Alloy 690 tube can be enhanced to about 75% which mainly were of Σ3ⁿ (n = 1, 2, 3, …) type. In this case, the grain boundary network (GBN) was featured by the formation of highly twinned large size grain-clusters produced by multiple twinning during recrystallization. All of the grains inside this kind of cluster had Σ3ⁿ mutual misorientations, and hence all the boundaries inside the cluster were of Σ3ⁿ type and formed many interconnected Σ3ⁿ type triple junctions. The weight losses due to grain dropping during intergranular corrosion for the samples with the modified GBN were much less than that with conventional microstructure. Based on the characterization by scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD) technique, it was shown that the highly twinned large size grain-cluster microstructure played a key role in enhancing the intergranular corrosion resistance: (1) the large grain-cluster can arrest the penetration of intergranular corrosion; (2) the large grain-cluster can protect the underlying microstructure.     

Root causes of intergranular attack in an operating nuclear steam generator tube

This paper reports the secondary side intergranular attack of an Alloy 600 tube, which was located within sludge piles in the hot-leg side of an operating nuclear steam generator. Carbide distribution along the grain boundaries and chromium depletion were analyzed using optical microscopy and transmission electron microscopy. Local crevice chemistry in contact with the defect was also assessed from the hideout return test data and oxide film analysis results using energy dispersive spectroscopy. The main causes of this defect are discussed based on the microstructure, local chemistry and operation temperature.     

Cs–Te corrosion depth dependence on distribution of chromium carbide precipitation in high chromium steel

In an attempt to investigate Cs–Te corrosion depth dependence on distribution of chromium carbide precipitation in high chromium steel, Cs–Te corrosion out-pile tests of two 9Cr steels with different distributions of chromium carbide were carried out at 975 K for 100 h, and their corrosion depths were compared. The corrosion is obviously more advanced in a specimen which has grain boundary carbide than in the one that does not. A considerable reason of the result is that the carbide distributed at grain boundaries promoted the corrosion reaction and the corrosion extended along the grain boundary. This is the first case in which the Cs–Te corrosion depth dependence on distribution of chromium carbide precipitation in Fe–Cr steel is clarified experimentally.     

Stress corrosion cracking susceptibility of steam generator tube materials in AVT (all volatile treatment) chemistry contaminated with lead

Alloy 600 steam generator tubing has shown a high susceptibility to stress corrosion degradation at the operation conditions of pressurized water reactors. Several contaminants, such as lead, have been postulated as being responsible for producing the secondary side stress corrosion cracking that has occurred mainly at the location where these contaminants can concentrate. An extensive experimental work has been carried out in order to better understand the effects of lead on the stress corrosion cracking susceptibility of steam generator tube materials, namely Alloys 600, 690 and 800. This paper presents the experimental work conducted with a view to determining the influence of lead oxide concentration in AVT (all volatile treatment) conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing.     

Helium release in uranium dioxide in relation to grain boundaries and free surfaces

Nuclear reaction analyses (NRA) based on the ³He(²H,⁴He)¹H reaction were previously performed to follow the evolution of implanted ³He in polycrystalline UO₂ samples. Experimental results pointed to an enhancement above 800 °C of the diffusion coefficient of helium over several microns in the vicinity of the grain boundaries, with respect to the diffusion coefficient within the grain. This was ascribed to the fact that grain boundaries are probably defect sinks which locally modify the defect concentrations.This study aims at demonstrating the particular effect of grain boundaries on helium migration. To this end, ³He implanted polycrystalline UO₂ samples were cracked then annealed at 900 °C. Helium migration in the vicinity of the grain boundaries and near the crack was investigated by means of NRA microanalyses. Helium depletion extends over far larger distances in the vicinity of the grain boundaries than near the crack. Experimental evidence has been collected of the particular effect of grain boundaries on helium migration, which do not act as free surfaces at which helium atoms are simply released.     

Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

The environmentally assisted cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and sulfur content of the base metal on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates for the DM weld heat-treated at 621 °C for 24 h were observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. Sulfur could diffuse out of the base metal and segregate along the grain boundaries of the dilution zone, leading to weakening the grain boundary strength and the SCC resistance of the Alloy 52-A508 weld.     

Development of a single loop EPR test method and its relation to grain boundary microchemistry for alloy 600

A single loop electrochemical potentiokinetic reactivation test method has been developed for alloy 600 that produces good passivation on all the surfaces, good etching during the reactivation scan and no appreciable pitting. It is able to quantify and discriminate between samples with a wide range of degree of sensitization. The Pa value correlates well with the minimum level of chromium in the depletion regions at the grain boundaries. It has been shown that the width of the attacked regions is much larger than the width of chromium depletion regions and it does not show any direct correlation with either depth or width or with a volume parameter of chromium depletion regions. It has been shown that the chromium carbides are not attacked during the test and that the intragranular regions attacked during the test are the sites of chromium carbides in the grain matrix. A modified Pa parameter is shown to be sensitive down to 7.5 wt% chromium in the depletion regions and indicates that the intragranular carbides have shallower depletion profiles than those at grain boundaries. Comparison of the results of the single loop and the double loop tests showed a good correlation.     

Materials aspects of BWR plant life extension

Considerable experience with plant equipment performance in nuclear power stations has indicated that the principal factors limiting the life of BWRs and PWRs are materials related. Specifically, for LWRs it is known that these materials issues generally include parameters related to stress corrosion cracking, corrosion fatigue, wear and radiation embrittlement. Not only do these parameters affect and limit the actual useful design life of plant components but also affect the plant's operating availability. In all these cases, the elimination or control of one or more of these critical parameters should improve the plants availability and significantly extend the useful service life.In the present paper, research performed to address the intergranular stress corrosion cracking (IGSCC) area is described. Specific emphasis is placed on Type 304 stainless steel which has suffered IGSCC in piping in the heat-affected-zone (HAZs) adjacent to the welds in the BWR primary system. Research has developed and qualified a number of techniques which address the three necessary conditions for IGSCC in BWRs: (1) sensitized microstructure, i.e., chromium depletion at the grain boundaries during welding; (2) over yield tensile stress; and (3) oxygenated (200 ppb) high temperature (288Another potential life-limiting IGSCC phenomenon for certain components, irradiation assisted stress corrosion cracking (IASCC) of stainless steel exposed to a high neutron flux, is also discussed. Unlike the IGSCC, IASCC results in intergranular cracking of annealed material at low stress. Fortunately, preliminary research has indicated that some of the techniques utilized for IGSCC control in piping as well as new controlled impurity level stainless steel alloys may reduce the future potential IASCC concern to an insignificant level.     

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.     

Alloy 800: new stress rupture and creep data for pressurized components in high temperature reactors

In an HTR plant high temperature components of Alloy 800 materials are subjected to temperatures from about 550 to 850°C in the long term; higher temperatures may occur in the short term. Thus the creep and stress-rupture parameters govern the design of these components. Since in recent years the scope of experimental data available for Alloy 800 materials has considerably increased, a new evaluation of the creep and creep-rupture properties was performed using a data bank computer. The relationships between the characteristics of the creep and creep-rupture behaviour and the metallurgical parameters were investigated by multilinear regression analyses. On the basis of the results of these analyses and after discussion in material expert committees new material specifications were determined for different types of Alloy 800. They were included into the draft standards DIN 17459 and DIN 17460 under the material standard nos. 1.4958, 1.4958 Rk and 1.4959. Besides the new values for the 1% total plastic strain limit and the creep-rupture strength for the types of Alloy 800 under consideration, isochronous stress-strain relations were derived on the basis of creep curves of reference heats.