Stress corrosion and thermal fatigue — Experiences and countermeasures in austenitic SS pipings of finnish bwr-plants

A summary of the existence of pipe cracking in Finnish BWR plants is presented covering both thermal fatigue and IGSCC cases. Countermeasures against cracking are evaluated and the measures applied are summarized. Also the results of a research program to monitor ageing of the weld heat affected zones in a pipeline section of a shut-down cooling system are summarized.     

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.     

Separation of Pu-238 and Pu-242 from Irradiated Am-241

Failure analysis was made on samples taken from type 304 stainless steel piping systems (core spray, unloading and feed water pipes) that had cracked in service. In the core spray pipe, large cracks including one penetrating through the wall were found in the upper half of the pipe wall, within the heat-affected zone of the weld joint between the pipe and the nozzle safe end of the reactor pressure vessel. These cracks were of intergranular type, also small transgranular cracks were found beyond the heat-affected zone. Strong correlation was established between the intergranular cracking and severe sensitization in the heat-affected zone.

In the unloading pipe, the region close to the weld joint contained cracks similar to those in the core spray pipe. The feed water pipe, in contrast to the foregoing cases, contained numerous shallow transgranular cracks both within and outside the heat-affected zone.     

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’.     

Stress corrosion cracking of low-alloy, reactor-pressure-vessel steels in oxygenated, high-temperature water

The stress corrosion cracking (SCC) behaviour of low-alloy, reactor-pressure-vessel (RPV) steels in oxygenated, high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both RPV structural integrity and safety, has been a subject of controversial discussions for many years. This paper presents the results of an experimental study on crack growth through SCC in three, nuclear-grade, steels (SA 533 B Cl.1, SA 508 Cl.2, 20 MnMoNi 5 5) under simulated, BWR water-chemistry conditions. Modern, high-temperature water loops, on-line crack-growth monitoring and fractographic analysis in the scanning electron microscope were used to quantify the cracking response of pre-cracked, fracture-mechanics specimens under a variety of mechanical and environmental conditions. Corrosion-assisted crack advance could be only initiated by active loading within the environment. If SCC crack advance at constant load was observed, initiation of crack growth had always occurred while increasing the load to the intended value for subsequent, static-load testing. During the constant load period the rate of SCC crack advance rapidly decayed and crack arrest occurred within a period of <100 h (for tests with K_I60 MPa m^1/2). Supplementary experiments with slowly increasing loading revealed that the initiation of crack growth, and the extent of further crack advance, are crucially dependent upon maintaining both a positive crack-tip strain rate and a high sulphur-anion activity in the crack-tip environment. It is concluded that there is no sustainable susceptibility to SCC crack growth under purely static loading, as long as small-scale-yielding conditions prevail at the crack-tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water-chemistry conditions (>EPRI action level 3) and/or for highly stressed specimens either loaded near to K_IJ or with a high degree of plasticity in the remaining ligament.     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

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.     

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.     

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.     

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K_IC-reference curve and the J_R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.     

Cyclic and transient thermal loading of the HDR reactor pressure vessel with respect to crack initiation and crack propagation

In a series of thermal loading tests at the HDR reactor pressure vessel – thermal stratification, cyclic thermal shock and pressurized thermal shock – the methods applied in safety analysis had to become qualified by a continuous intercomparison of calculated results and experimental data. Above all the complex boundary conditions of the HDR-tests offer a close approximation to the original components, so that they provide a real assessment of the transferability.The results of the thermal mixing tests indicated that during cold water inflow into the RPV longitudinal strains build up in the cylindrical wall which dominate over that in circumferential direction.During the cyclic thermal fatigue tests incipient crack formation in the cladding as well as the behaviour of crack propagation in the cladding and in the base material was analyzed.In the pressurized thermal shock tests, the nozzle region and the cylinder wall in the incipient crack condition were loaded by long cooling streaks. Even in the aggravated loading condition as the result of a routed cold water streak no remarkable indications of crack growth were noticed.In both cases, cyclic and pressurized thermal shock loading, the expected crack propagation was overpredicted by the fracture mechanical methods used.The non-destructive examination methods used were able to locate all of the cracks but they mostly overpredicted the actual crack depth.     

Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP +50 mV_SHE, 0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum K_I value of 25 MPa m^1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate dK_I/dt, but were not dependent on the actual K_I values up to 60 MPa m^1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.     

Fracture mechanical evaluation of an in-vessel melt retention scenario

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.     

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.     

Design improvement of weld design of the reactor internal circulation pump motor casing to the reactor pressure vessel nozzle

The reactor internal recirculation pump (RIP) used in the advanced boiling water reactor (ABWR) design is a glandless wet-motor type pump and is evolved from the pump used in the ABB-A BWRs. On the basis of the proven design viewpoint, the pump nozzle at the bottom head of the ABWR reactor pressure vessel (RPV) to which the RIP is attached is designed as a sleeve-type nozzle as used in the ABB-A BWRs. Several improvements have been made over the ABB-A original nozzle design such as elimination of the weld between the RPV bottom head and nozzle stub (by integral forging) and modification of the weld design (optimization of weld preparation suitable to automatic machine welding and use of insert ring for quality welding). Extensive experimental and analytical studies and the development of machining tools for the penetration bead (the back side of the weld) to inspect the qualification of welding have been performed in Japan with the RIP and the RPV nozzle to confirm the adequacy of the ABWR RIP and the RPV nozzle design.     

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.     

The evaluation of intergranular stress corrosion cracking problems of stainless steel piping in Taiwan BWR-6 nuclear power plant

Taiwan BWR-6 Kuosheng Nuclear Power Plant Unit 1 implemented the inspection of the intergranular stress corrosion cracking (IGSCC)-susceptible weldments of stainless steel piping in the reactor recirculation, reactor water clean-up, residual heat removal, core spray and feedwater systems. The purpose of this paper is to present the status of the fracture problems in the weldments. The crack growth analysis due to IGSCC and the standard weld overlay design based on the ASME Code Section XI and NUREG-0313 Rev. 2 for the fracture weldments are discussed in detail. Then, the contingent programs including the inspection program, fracture evaluation, and the standard weld overlay, are completely established to prevent pipe break during the reactor operation.     

Effect of cladding on the stress intensity factors in the reactor pressure vessel

In general, reactor pressure vessels (RPV) are cladded with stainless steel to prevent corrosion and radiation embrittlement. The ASME Sec. XI specifies that a subclad crack which may be found during the in-service inspection must be considered as a semi-elliptical surface crack when the thickness of cladding is less than 40% of the crack depth. In order to refine the fracture assessment procedures for such subclad cracks, three-dimensional finite element analyses were applied for various subclad cracks embedded in the base metal. A total of 18 crack geometries were analyzed, and the results were compared with those for idealized semi-elliptical surface cracks for two different loading conditions, i.e. internal pressure and pressurized thermal shock. The resulting stress intensity factors for subclad cracks were 50–70% less than those for idealized surface cracks. It has been proven that the condition specified on the ASME Sec. XI is overly conservative for subclad cracks which are assumed to be surface cracks.     

Towards an elastic-plastic fracture mechanics predictive capability for reactor piping

Intergranular stress corrosion cracks have been discovered in the recirculation bypass piping and core spray lines of several boiling water reactor (BWR) plants. These cracks initiate in heat-affected zones of girth welds and grow circumferentially by combined stress corrosion and fatigue. Reactor piping is mainly type 304 stainless steel, a material which exhibits high ductility and toughness. A test program described in this paper demonstrates that catastrophic crack growth in these materials is preceded by considerable amounts of stable crack growth accompanied by large plastic deformation. Thus, conventional linear elastic fracture mechanics, which only applies to the initiation of crack growth in materials behaving in a predominantly linear elastic fashion, is inadequate for a failure analysis of reactor piping.This paper is based upon research initiated by a need to develop a realistic failure prediction and a way to delineate leak-before-break conditions for reactor piping. An effective engineering solution for the type of cracks that have been discovered in BWR plants was first developed. This was based upon a simple net section flow stress criterion. Subsequent work to develop an elastic-plastic fracture mechanics methodology has also been pursued. A survey of progress being made is described in this paper. This work is based on the use of finite element models together with experimental results to identify criteria appropriate for the onset of crack extension and for stable crack growth. A number of criteria have been evaluated. However, the optimum fracture criterion has not yet been determined, even for conditions which do not include all of the complications involved in reactor piping.