Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Würgassen Nuclear Power Station need not be feared. Although the material and corrosion–chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert's agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept.     

Primary water stress corrosion cracking behaviors in the shot-peened Alloy 600 TT steam generator tubings

The effect of compressive residual stress on the primary water stress corrosion cracking behavior was investigated, based on the J-1 and J-2 nuclear power plant data. The following analyses were performed such as: (i) Weibull slope; (ii) crack growth rate; (iii) average crack length; (iv) crack length distribution. Alloy 600 TT exhibits strong heat to heat variations in its sensitivity to PWSCC. Crack growth rate was retarded after shot-peening. The compressive residual stress induced by shot-peening was more effective on new, short cracks, than on existing, long cracks. However, whether the ‘new’ cracks were initiated after peening is an unresolved issue, due to the present ECT sensitivity limit.     

Modeling of weld residual stresses in core shroud structures

This paper presents a computational model to predict residual stresses in a girth weld (H4) of a BWR core shroud. The H4 weld is a multi-pass submerged-arc weld that joins two type 304 austenitic stainless steel cylinders. An axisymmetric solid element model was used to characterize the detailed evolution of residual stresses in the H4 weld. In the analysis, a series of advanced weld modeling techniques were used to address some specific welding-related issues, such as material melting/re-melting and history annihilation. In addition, a 3-D shell element analysis was performed to quantify specimen removal effects on residual stress measurements based on a sub-structural specimen from a core shroud. The predicted residual stresses in the H4 weld were used as the crack driving force for the subsequent analysis of stress corrosion cracking in the H4 weld. The crack growth behavior was investigated using an advanced finite element alternating method (FEAM). Stress intensity factors were calculated for both axisymmetric circumferential (360°) and circumferential surface cracks. The analysis results obtained from these studies shed light on the residual stress characteristics in core shroud weldments and the effects of residual stresses on stress corrosion cracking behavior.     

Evaluation of stainless steel pipe cracking: Causes and fixes

This paper discusses (1) studies of impurity effects on susceptibility to intergranular stress corrosion cracking (IGSCC), (2) intergranular crack growth rate measurements, (3) finite-element studies of the residual stresses produced by induction heating stress improvement (IHSI) and the addition of weld overlays to flawed piping, (4) leak-before-break analyses of piping with 360° part-through cracks, and (5) parametric studies on the effect of through-wall residual stresses on intergranular crack growth behavior in large diameter piping weldments. The studies on the effect of impurities on IGSCC of Type 304 stainless steel show a strong synergistic interaction between dissolved oxygen and impurity concentration of the water. Low carbon stainless steel (Type 316NG) appear resistant to IGSCC even in impurity environments. However, they can become susceptible to transgranular SCC with low levels of sulfate or chloride present in the environment. The finite-element calculations show that IHSI and the weld overlay produce compressive residual stresses on the inner surface, and that the stresses at the crack tip remain compressive under design loads at least for shallow cracks.     

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.     

Probability of crack-induced failure in BWR recirculation piping

The Lawrence Livermore National Laboratory (LLNL) has estimated the probability of double-ended guillotine break (DEGB) in the reactor coolant piping of Mark I boiling water reactor (BWR) plants. Two causes of pipe break are considered: crack growth at welded joints and the seismically-induced failure of component supports. For the former a probabilistic fracture mechanics model is used, for the latter a probabilistic support reliability model. This paper describes a probabilistic model developed to account for effects of intergranular stress corrosion cracking (IGSCC). The IGSCC model, based on experimental and field data compiled from several sources, correlates times to crack initiation and crack growth rates for Types 304 and 316NG stainless steel against material-specific ‘damage parameters’ which consilidate the separate effects of coolant environment (temperature, dissolved oxygen content, level of impurities), stress (including residual stress), and degree of sensitization. Application of this model to actual BWR recirculation piping shows that IGSCC clearly dominates the probability of failure in 304SS piping, mainly due to cracks that initiate within a few years after plant operation has begun. Replacing Type 304 piping with 316NG reduces failure probabilities by several orders of magnitude.     

An elastoplastic phase-field model for the evolution of hydride precipitation in zirconium. Part II: Specimen with flaws

An elastoplastic phase-field model, described in Part I, was applied to bulk materials containing flaws such as sharp cracks and blunt notches. An additional set of long range order parameters, namely, stress-free strains for flaws, was introduced. The nucleation and growth of hydrides near a void or a crack were simulated by the proposed elastoplastic phase-field model. The effects of notch root radius, hydrogen concentration in solid solution, yield stress of the matrix and the level of externally applied stress on hydride morphology around flaws were studied. It is demonstrated that parameters such as the distribution of the tensile stress component perpendicular to the hydride platelet normal may be closely monitored during hydride growth near a flaw with or without externally applied stresses. Combined with a fracture criterion and real experimental data, the model is capable of predicting the rate and morphology of hydride precipitation, and crack initiation near flaws.     

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.     

BWR pipe crack and weld clad overlay studies

Several topics pertaining to the problem of stress corrosion cracking (SCC) of piping in boiling water reactors are addressed in this paper: (1) the effects of impurities, dissolved oxygen content, and strain rate on susceptibility of SCC of “Nuclear Grade” Type 316NG and sensitized Type 304 stainless steel, (2) finite-element analyses and experimental measurement of residual stresses in weldments with weld overlays, and (3) analysis of field components to assess effectiveness of in-service inspection techniques and the in-reactor performance of weld overlays. Several anion impurities including sulfates, chlorides, nitrates, borates, and carbonates were studied under both near neutral and slightly acidic conditions. At the low impurity concentrations expected in reactor coolant systems (<0.1 ppm), the sulfur species appear to be the most deleterious. They promote intergranular SCC in sensitized stainless steel and transgranular SCC in the low-carbon “Nuclear Grade” stainless steel. Correlations between experimental data and a phenomenological model that describes the effect of strain rate on SCC are presented. Measurements of the residual stresses produced by weld overlays confirm that the process is very effective in producing compressive stresses on the inner surface of the weldment. Examination of a weld overlay removed from service suggests that no additional throughwall crack growth had occurred after application of the overlay.     

Numerical weld modeling — a method for calculating weld-induced residual stresses

In the past, weld-induced residual stresses caused damage to numerous (power) plant parts, components and systems (Erve, M., Wesseling, U., Kilian, R., Hardt, R., Brümmer, G., Maier, V., Ilg, U., 1994. Cracking in Stabilized Austenitic Stainless Steel Piping of German Boiling Water Reactors — Characteristic Features and Root Causes. 20. MPA-Seminar 1994, vol. 2, paper 29, pp.29.1–29.21). In the case of BWR nuclear power plants, this damage can be caused by the mechanism of intergranular stress corrosion cracking in austenitic piping or the core shroud in the reactor pressure vessel and is triggered chiefly by weld-induced residual stresses. One solution of this problem that has been used in the past involves experimental measurements of residual stresses in conjunction with weld optimization testing. However, the experimental analysis of all relevant parameters is an extremely tedious process. Numerical simulation using the finite element method (FEM) not only supplements this method but, in view of modern computer capacities, is also an equally valid alternative in its own right. This paper will demonstrate that the technique developed for numerical simulation of the welding process has not only been properly verified and validated on austenitic pipe welds, but that it also permits making selective statements on improvements to the welding process. For instance, numerical simulation can provide information on the starting point of welding for every weld bead, the effect of interpass cooling as far as a possible sensitization of the heat affected zone (HAZ) is concerned, the effect of gap width on the resultant weld residual stresses, or the effect of the ‘last pass heat sink welding’ (welding of the final passes while simultaneously cooling the inner surface with water) producing compressive stresses in the root area of a circumferential weld in an austenitic pipe. The computer program (finite element residual stress analysis) was based on a commercially available code (Hibbitt, Karlsson, Sorensen, Inc, 1997. user's manual, version 5.6), and can be used as a 2-D or 3-D FEM analysis; depending on task definition it can provide a starting point for a fracture mechanics safety analysis with acceptable computing times.     

Seismic analysis and fracture evaluation of stabilized core shroud

This paper deals with the seismic analysis and fracture evaluation of a stabilized core shroud in a boiling water reactor of nuclear power plant. To study the adequacy of original seismic loadings, the dynamic behaviors of core shrouds with cracks, without cracks and with stabilizers are analyzed. Seismic analysis of a lumped-mass model of reactor internals is then performed to obtain the seismic loadings around various weldments of the repaired core shroud. The interaction between the core internals and this repaired core shroud is thus taken into account in this study. Further, fracture analysis of the stabilized core shroud is performed to obtain the stress intensity factors along the crack front of horizontal welds based on these seismic loadings. The computed results show that the influence of existing cracks in the stabilized core shroud is insignificant on the overall structural integrity.     

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.     

Applications of a fracture mechanics model of structural reliability to the effects of seismic events on reactor piping

A fracture mechanics model of structural reliability is described. The model assumes that failure occurs due to the subcritical and catastrophic growth of as-fabricated defects. The material properties, stress history, number and dimensions of the initial cracks are treated as random variables. Crack growth is calculated using fracture mechanics principles. The capability of modeling two-dimensional cracks and thickness gradients of the applied stresses represents a significant advance over previous work in this field.The model has been used to estimate the influence of earthquakes on the integrity of circumferential girth butt welds in the large (diameter greater than 30 in.) primary coolant system pipes of a commercial pressurized water reactor. In the absence of earthquakes, the probability of leaks and catastrophic double-ended guillotine breaks is estimated to be 10^?6 and 10^?12 per plant lifetime, respectively. These probabilities were only slightly increased by the occurrence of earthquakes. The cyclic stresses in the heatup-cooldown cycle had the greatest effect on the crack growth. Radial gradient thermal stresses due to temperature fluctuation of the coolant during transients have only a small effect on the amount of crack growth. Sensitivity studies show that significant changes in modeling assumptions are needed before the calculated failure probabilities are raised to the level of current estimates. This suggests that perhaps factors such as design and construction errors or stress corrosion cracking may be significant contributors to the probabilities.     

Current research on environmentally assisted cracking in light water reactor environments

The effect of dissolved oxygen level on fatigue life of austenitic stainless steels is discussed and the results of a detailed study of the effect of the environment on the growth of cracks during fatigue initiation are presented. Initial test results are given for specimens irradiated in the Halden reactor. Impurities introduced by shielded metal arc welding that may affect susceptibility to stress corrosion cracking are described. Results of calculations of residual stresses in core shroud weldments are summarized. Crack growth rates of high-nickel alloys under cyclic loading with R ratios from 0.2 to 0.95 in high-purity water that contains <5 and 300 ppb dissolved oxygen at 240, 289, and 320°C, are summarized.     

Surface-crack aspect ratio development during corrosion-fatigue crack growth in low-alloy steels

The development of aspect ratios (crack depth/half-crack length), was studied for semi-elliptical surface cracks in low-alloy steel undergoing corrosion-fatigue in an elevated temperature aqueous environment. Water-enhanced crack growth behavior is influenced by the concentration of hydrogen sulfide at the crack tip, and the sulfide concentration is in turn influenced by mass-transport considerations. The mass-transport characteristics of surface cracks may be different from those of more common test specimens; e.g. compact tension specimens. It is also shown that the method of preparing surface-cracked specimens can have an influence upon the crack growth behavior; surface cracks emanating from crack-starter notches may behave differently than ‘natural’ surface cracks because of differences in the mass-transport paths. It is also shown in that the rate of water flow along the length of a surface crack can affect the resulting crack aspect ratio and crack growth rates.     

Mechanical integrity analysis of TMI-1 OTSG unplugged tubes

Small I.D. circumferential defects have been identified in many steam generator tubes. The origin of the cracks is known to be chemical, not mechanical. A fracture mechanics evaluation has been conducted to ascertain the stability of tube cracks under steady-state and anticipated transient conditions. A spectrum of hypothetical crack sizes was interacted with tube stresses derived from the load evaluation using the methods of linear elastic fracture mechanics (LEFM). Stress intensities were calculated for part-through wall cracks in cylinders combining components due to membrane stress, bending stress, and stresses due to internal pressure acting on the parting crack faces as the loads are cycled.The LEFM computational code, “BIGIF”, developed for EPRI, was used to integrate over a range of stress intensities following the model to describe crack growth in INCO 600 at operating temperature using the equation (ΔK)^3.5.The code was modified by applying ΔK_Th, the threshold stress intensity range. Below ΔK_Th small cracks will not propagate at all. Appropriate R ratio values were employed when calculating crack propagation due to high cycle or low cycle loading.Cracks that may have escaped detection by ECT will not jeopardize tube integrity during normal cooldown unless these cracks are greater than 180° in extent. Large non-through-wall cracks that would jeopardize tube integrity are not expected to evolve because in axi-symmetric tensile stress fields, cracks propagate preferentially through the tube wall rather than around the circumference. Tube integrity can be demonstrated for mid-span tube regions and for the transition region as well.The as-repaired transition geometry is a design no less adequate than the original. The as-repaired condition represents an improvement in the state of stress due to mechanical and thermal loads as compared to the original.     

An overview of current methods used for assessing surface flaws in nuclear reactor vessels

During the life of nuclear reactor vessels several inspections are performed on the pressure retaining components. After those inspections are performed, significant indications must be evaluated to determine if repair is needed. Section XI of the ASME Boiler and Pressure Vessel Code gives guidelines for evaluation of the flaws found in the inspections. In this paper the primary steps in the evaluation procedure are assessed in light of current research. Several inadequacies are found in the procedure, especially in the shape assumed for fatigue crack growth and crack propagation-arrest events. The material properties specified for use in the procedure for initiation and arrest of cracks are shown to be overly conservative and in need of a statistical base. The stress intensity factor solution specified in this procedure is also shown to be overly conservative. On the basis of these inadequacies and over-conservatism, recommendations are made for changes in Section XI and future research.     

Stress corrosion cracking of Alloy 600 in high temperature water

Slow strain rate stress corrosion tests have been performed on specimens cut from four separate heats of Alloy 600 steam generator tubing. The material was tested in the mill-annealed and thermally-stabilised conditions and after various low temperature ageing treatments. Only limited cracking was observed, even for tests at 340°C, but the initiation of intergranular cracking was easier on the inner than on the outer surfaces on the tubing. Polarization data has been obtained in high-temperature water and in saturated boric acid and saturated lithium hydroxide at the atmospheric boiling points, and slow strain tests were performed at controlled potentials in these environments. Again, only very short cracks formed during the slow strain rate tests which were performed at a strain rate of about 10⁻⁶ s⁻¹. The data is discussed in terms of the probable crack tip strain rates that would exist in these tests and at other strain rates. It is argued that if cracking occurs, the main role of very slow strain rate tests is to provide time for initiation and crack growth, so that cyclic loading or intermittent loading long term tests are likely to be more successful in sustaining crack growth in this alloy.     

2D simulation of the initiation and propagation of crack array under thermal fatigue

Various components of nuclear reactors are submitted to various thermo-mechanical loadings. Thermal fatigue cracking has been clearly detected in reactor heat removal system (RHRS) of pressurized water reactors (PWRs). The present study focuses on AISI 304 L stainless steel used in PWRs. The thermal fatigue behavior of this steel has been investigated using a specific thermal fatigue facility called “SPLASH”. This test equipment allows the reproduction of multiple crack networks similar to those detected during component inspections. The present study deals with the modeling of crack networks initiation and propagation. It is structured in two parts: (i) experimental details and main characteristics of the cracks networks, and (ii) numerical simulation of multiple cracks initiation and growth problem, using an elastic–plastic thermal–mechanical computation and a generalized Paris’ law. The model presented in this study gives predictions in a good agreement with observations, as far as the evolution of the mean and deepest cracks during cycling is concerned.