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.     

Japanese efforts in relation to BWR pipe cracking

The stress corrosion cracking was found in a Japanese Commercial BWR power plant first in 1974, then also in the others. To overcome this problem, much effort has been made to improve the stress condition, material and environment which are known to be the three causative factors of IGSCC. Also several studies have been carried out from the fracture mechanics point of view on the stress corrosion cracking behaviors in the BWR primary piping. Our experiences of pipe cracking, development of remedies and/or countermeasures examples of their application to our units and the fracture mechanistic research on piping LBB are shown in the following.     

Formation and growth of cracking in feed water pipes and RPV nozzles

The observation of numerous small and large cracks in ferritic feed water pipes of boiling (BWR) and pressurized water reactors (PWR) in the last few years has led to basic research into the causes of cracking and the crack growth mechanisms.In horizontal feed water pipe sections connected to nozzles of reactor pressure vessels (RPV) of BWR's as well as of steam generators (SG) of PWR's, circumferential macro and micro cracks were detected. These cracking phenomena could be observed in base material of pipes as well as in weld seam regions. The examination of the stress state displayed that the cracked pipe regions have been exposed to a number of cyclic thermal transients (thermal shock, flow stratification) during start-up (hot stand-by) and shut-down periods of the plants. During thermal transient periods, local and global cyclic stresses in the referred pipe cross sections have been induced which in interaction with the influence from environment (in operation as well as in shut-down periods) and local geometrical imperfections led to the initiation and formation of macro and micro cracks.In the reactor water clean-up system of BWR through which reactor water is fed from the RPV to the main feed water line, two longitudinally welded elbows have been detected to be severely cracked. Both elbows have been subjected to an internal pressure corresponding to RPV and additionally to a relevant “in-plane” bending moment. These longitudinal cracks were found to be started from the inner elbow surface. In one case the longitudinal crack was situated in the base material and was enlarged to leakage. In the second elbow the longitudinal crack was located in the heat affected zone (HAZ) of a longitudinal weld. In both cases the macro cracks started either from corrosion pits located in defective areas of the magnetic protection layer or from geometrical notches (weld root). The semi-elliptic small cracks got linked to more extended shallow cracks.Formation and growth mechanism of these cracks have been studied at the MPA Stuttgart in laboratory under simulated operation conditions which were held as realistic as possible compared with those in nuclear power plants.The results of experimental studies in laboratory as well as conclusions based on the above mentioned cracking phenomena in piping have been used as basic information for a realistic design of large scale (RPV) thermal shock experiments under operation conditions. The formation and growth mechanism of these cracks and their detection by means of NDE during thermal transients at the inner surface of RPV nozzle and at the adjacent cylindrical areas of RPV shell will be described.     

Application of fracture mechanics methods in safety analysis of piping components in subcreep and creep behavior

A survey and review program for the application of fracture mechanics methods in elevated temperature design analysis and safety evaluation was initiated in December 1976. The first report [1] surveyed and assembled the material for a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction and safety analysis of piping components. The second report [2] provided the basic concepts and a review of the problem areas associated with the development of analytical and experimental programs for a systematic evaluation and comparison of the currently available fracture mechanics theories. The basis for such an evaluation was described in terms of a series of benchmark problems which accurately specify conditions of geometry, loading and environment characteristic of large diameter piping systems in nuclear service.The objective of this third report is to establish a data base and detail the additional analytical techniques needed to confirm the validity of existing analytical methods and improve the state of the art in current problematic areas effecting the interpretation and extension of safety evaluation methods. The need for such a program in the elevated temperature field has been demonstrated by a number of independent surveys on various safety aspects of LMFBR related structural analysis methods and matetials problems. The results of this program, however, will be applicable not only to reactor plants operating at elevated temperatures, but will also lead to improvements of light water reactor evaluation methods for operating and accident conditions.The current state of elevated temperature reactor design technology is embodied in the standards and codes which provide guidance and minimum requirements for systematic design and evaluation procedures. These, however, do not necessarily provide specific absolute values which, if satisfied in the course of design, will guarantee thirty to forty years of uninterrupted life. There are numerous assumptions and approximations embodied in these standards concerning materials behavior, damage mechanisms, and failure modes at elevated temperature. There are also numerous areas of uncertainty and conflicting opinion in the interpretation of the existing test data and in the analysis and evaluation methods. Furthermore, the standards and codes leave some areas to the judgement of the designer, some of which require explicit justifications, but no standards or rules are provided.The overall safety therefore lies, at the present time, in the combination of rigorous enforcement of current standards, judicious application of experience with high temperature equipment even if not in nuclear service, and the surveillance of actual operating conditions. In the past, one criterion proposed for elevated temperature design has been that the time for crack initiation should exceed the design life. However, due to the complexities of the piping structures and the nature of the stress history during service, the evaluation of initiation times is difficult and often leads to uneconomical designs. In addition defects may exist in the component before it enters service. Hence, the knowledge of the growth rates of cracks and the residual strength of the components containing cracks is important in a realistic design evaluation. For more brittle materials and lower temperature applications where plasticity is restricted, linear elastic fracture mechanics methods have been developed. For more ductile materials where the plastic zones near the cracks are larger, linear fracture mechanics methods are not directly applicable, but in these nonelastic cases the opening displacement and J integral methods of assessment have been proposed. In the complex situation encountered in nuclear power plant design, the analysis must also account for cyclic thermal strains, time dependent creep, and the effect of harmful environments which are not explicitly treated in the above-mentioned methods. In this report an in-depth review is presented in sufficient detail to illustrate the degree of agreement between the theoretical and empirical methods available in the literature and indicate the scope of the additional analyses and experimental work needed for the development of reliable safety evaluation methodology.For pure cylindrical bending, cracks perpendicular to the load start to grow when reaches a critical value which is generally larger than the corresponding critical uniaxial tension value. There appears to be a thickness effect in the bending case which is probably due to interference from the compressive sides of the crack.For a circular plate with lateral pressure and small lateral displacements, results agree with the bending data when using the nominal bending stress . For larger displacements when bulging occurs, the results agree with the tensile data when the nominal tensile stress is used.For curves surfaces, such as a cylinder under internal pressure, the data agree with the expression developed by Folias both for axial cracks under hoop stress σ and for circumferential cracks under axial stress σ Generally, the expressions were accurate up to , showing a tendency to be lower than the experimental data at higher values of the parameter. The parameter is a promising one.To study the influence of cracks at different angles to the applied load, analysis and data are available including the stress component parallel to the crack in the stress field around a crack tip. This, together with the concept of a critical circumferential stress at a critical distance (α = 0.1) ahead of the crack provides improved correlation with fracture predictions for both the angle of fracture and the critical stress intensity factor for the angled cracks in flat plates.For a hollow cylinder under torsion with angled cracks, the best correlation was given by the same analysis although the results were not as conclusive as for the flat plate. From elastic theory useful curves for the variation of K₁, K₂, and K₃ around the border of an elliptically shaped crack are available.In a plane stress fracture the addition of a biaxial stress produces an increase in the apparent fracture toughness compared with the uniaxial case. However, there is as yet no evidence to show that there would be the same increase in a plane strain situation. Hence, in the absence of biaxial information the uniaxial fracture data may be the most conservative for flat plates. However, for shells there will also be a curvature effect.In an analogous manner, fatigue crack propagation rates appear to be less rapid under biaxial stresses than under uniaxial stress. However, this shift is not great and generallly will be masked by other effects such as environment and temperature service situations.The analysis of cracks in weldments with residual stress effects are also available. In the case of a crack in a weld the estimated residual stress distribution agreed reasonably well with some experimental data for elastic conditions. Results indicate that there can be a tensile stress intensity factor even when the original residual stress distribution has changed to compressive. A point to remember is that residual stresses near welds can be beyond yield.An analysis based on Lagrangean mechanics is useful for indicating the different effects of liquids and gases as pressurizing media in hollow pipes. The results show that whereas gases maintain their pressure as a crack begins to propagate, the pressure in the liquid can quickly decrease so that subsequent catastrophic failure is less likely even in large diameter piping.     

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’.     

A critical survey on the application of plastic fracture mechanics to nuclear pressure vessels and piping

Plastic fracture mechanics techniques have been developed to treat the regime where extensive plastic deformation and stable crack growth occur prior to fracture instability in the tough ductile materials used in nuclear systems. As described in this paper, a large number of crack tip parameters can be used in a plastic fracture resistance curve approach. However, applications using the J-integral currently predominate. This parameter has significant advantages. It offers computational ease and can provide a lower bound estimate of the fracture condition. But, J also has a disadvantage in that only a limited amount of stable crack growth can be accommodated. The crack tip opening angle parameter, in contrast, can be valid for extensive stable crack growth. But, with it and most other realistic alternatives, the computational convenience associated with the J-integral is lost and finite element or other numerical methods must be employed. Other possibilities such as the two-criterion approach and the critical net section stress are also described in the paper. In addition, current research work focussed upon improving the theoretical basis for the subject is reviewed together with related areas such as dynamic plastic analyses for unstable crack propagation/arrest and creep crack growth at high temperatures. Finally, an application of plastic fracture mechanics to stress corrosion cracking of nuclear piping is made which indicates the possible anti-conservative nature of the current linear elastic assessments.     

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel, which is excellent in terms of strength, corrosion resistance and weldability, has conventionally been used. Cast duplex stainless steel contains the ferrite phase in the austenite matrix, and thermal aging after long-term service is known to decrease fracture toughness. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secure, even when such through-wall crack length is assumed to be as large as the fatigue crack length grown for a service period of up to 60 years.     

Estimation of crack opening area for leak before break analysis of nuclear reactor system

The crack opening area has been estimated by various models for leak before break analysis of nuclear reactor systems. A number of linear elastic fracture mechanics models have been employed. A model has been derived based on Forman et al. [Forman, R.G., Hickman, J.C., Shivakumar, V., 1985. Stress intensity factors for circumferential through cracks in hollow cylinders subjected to combined tension and bending loads. Eng. Frac. Mech. 21, 563-571] model for stress intensity factors. The linear elastic fracture models entail the use of plasticity corrections. The convergence behaviour of the iterative implementation for plasticity correction has been studied. An elasto-plastic fracture mechanics model has also been adopted for comparative study undertaken here. The range of applicability of the models has been reviewed.     

Fracture mechanics evaluation of the structural material for fast breeder reactors

Results of fracture mechanics investigations on austenitic steels used for LMFBRs (Liquid Metal Fast Breeder Reactors) are presented. A summary of reported tests on straight piping and elbows with through wall flaws is given which agree well with predictions made by using a plastic instability model. Crack growth experiments and calculations indicate that initial flaws will not extend significantly during service. Even if considerable crack growth is postulated cracks will penetrate the piping wall with a high safety margin to unstable crack configurations. Theoretical investigations of flawed structures under high strains show that the effect of crack size can be discussed similarly to the elastic range. The information demonstrate that with respect to the design requirements and operating conditions of LMFBRs a sudden rupture of the piping can be excluded. The integrity of the coolant boundary is given also in case of initial flaws.     

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.     

Static and cyclic crack growth of low alloy steels in high-oxygen water simulating HDR test conditions

Steel samples of reactor pressure vessel and piping steels from the German HDR programme have been tested in high oxygen water at different temperatures simulating HDR test conditions. The specimens have been exposed to sequences of static and cyclic loading or to purely cyclic loading. During the tests, threshold stress intensity values for stress corrosion cracking and crack growth rates with various cyclic loading parameters were determined. Extensive fracture surface and oxide layer investigations were also performed. Water chemistry parameters such as dissolved oxygen content, pH and conductivity were continuously monitored during the tests. Finally, the measured laboratory water chemistry parameters were compared to those measured in the HDR plant during full scale testing of components and the relevance of the results for normally operating plants is discussed.     

Implications of recent developments in the plastic fracture mechanics field to the pci stress corrosion problem

Fractographic observations on irradiated Zircaloy cladding stress corrosion fracture surfaces are considered against the background of recent developments in the plastic fracture mechanics field. Dimples have been observed on the fracture surfaces of failed cladding, even though the cracks in metallogiaphic sections are tight, i.e., crack propagation is associated with a low crack tip opening angle. This result is interpreted as providing evidence for an environmentally assisted ductile mode of fracture. The presence of this fracture mode forms the basis of an argument, which adds further support for the view that power ramp stress corrosion cladding failures are caused by stress concentrations that produce stress gradients in the cladding.     

Fracture mechanics issues for irradiated blanket structures

This paper reviews some of the factors that will affect fracture behavior of fusion reactor structures and summarizes some component life predictions based on linear elastic fracture mechanics analysis. The review includes discussion of the environments to which the components will be subjected, the response of materials to these environments, the time dependent nature of the structural response, and the fracture related failure mechanisms.Radiation environments and complex loading conditions in a fusion reactor cause a variety of material phenomena. These phenomena include irradiation swelling and creep, strength changes due to matrix hardening, helium embrittlement, and surface effects such as sputtering and blistering.The interaction of thermal creep, irradiation creep, and swelling results in complex time, temperature, and neutron fluence dependent stress histories in first wall and blanket structures. These effects reduce compressive thermal stresses during the burn portion of a reactor operating cycle and result in residual tensile stress during the non-burn portion of the cycle. The cyclic nature of these stresses, particularly in a tokamak reactor, and the presence of undetected flaws provide a basis for the application of fracture mechanics. Linear elastic fracture mechanics analysis techniques have been applied to predict component life for several conceptual tokamak fusion reactor designs. These analyses show that the structural life may be limited by growth of initial flaws to a coolant leakage. Results indicate that for neutron wall loadings below 2 to 3 Mw/m², life is likely to be controlled by stresses during the burn period and, at higher wall loadings, by residual stresses during the non-burn period.Fracture toughness properties tend to be reduced by irradiation. Therefore, brittle fracture will be a potentially critical failure mode. Fatigue crack growth and fracture characteristics of the design will affect the operating mode of a reactor and influence the performance of different types of reactors. Tests are currently planned to develop material crack growth and fracture toughness data [1] for candidate alloys because these properties have been shown to be important.     

Crack growth under cyclic thermal shock loading

Experimental and theoretical results of fatique crack growth under cyclic thermal shock loading are presented. The experiments were done by cooling hot steel plates by a water jet. Linear elastic fracture mechanics is applied to predict fatigue crack growth theoretically. For this stress intensity factors of semi-elliptical surface cracks were calculated by use of the weight function method. Measured fatigue crack growth is compared with predictions from the theory.     

LBB technology application to the primary piping system of the NPP V1 Jaslovske Bohunice

Due to several deficiences of the WWER Model 230 type reactor a leak before break demonstration of this reactor is of primary importance. The complex project for NPP V1 Jaslovske Bohunice includes a static and dynamic stress analysis of the primary piping, a fatique damage analysis, leak rate assessments and an analysis of the stability of the heavy components supports. The material database includes data on fracture mechanics, on assessment of corrosion properties, and on the influence of 100000 hr service exposure on base metal and welds including disimilar welds. The program was supported by large scale experiments on RPV safe-end, pressurizer safe-end, elbow welds with through-wall cracks and leak rate measurements. The results and applications are discussed.     

Tests on large scale LMFR piping Part I: crack resistance properties of through-cracked straight 700 mm nominal diameter pipes tested under bending at ambient temperature and 550°C

Knowing the crack resistance properties of a structure is essential for fracture mechanics safety analyses. Considerable attention has to be given in many cases to the through-wall case, since this is generally believed to be the controlling case with regard to complete pipe failure. Within a cooperative fracture mechanics programme of Electricite de France (EdF), Novatome and Siemens/KWU, bending tests with monotonously increasing load on circumferentially cracked straight pipes of typical liquid metal fast reactor (LMFR) main piping dimensions were performed. In this paper a summary report is given on crack resistance curves based on the crack tip parameters S, J and JM. The data are compared with those of C(T) specimens. The experiments have demonstrated an enormous potential for stable crack extension under global bending which is a typical loading for LMFR piping structures. The results of checking the transferability of laboratory specimen crack growth characteristics to the cracked pipes on the austenitic stainless steel 316 L demonstrate that the fracture mechanics concept for a reliable transfer of crack resistance data from small specimen geometries to large structures needs further qualification for high toughness materials.     

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.     

Cyclic crack growth evaluation of a 20MnMoNi55 piping steel in high-oxygen reactor water

Samples of a low alloy steel piping material taken from the full scale corrosion fatigue test loop of the Heissdampfreaktor (HDR) plant have been tested at 240°C in high oxygen reactor water. The small-scale specimens (CT25) were exposed to a similar loading spectrum to that which has been used in the full-scale corrosion fatigue tests at the HDR-plant. During the autoclave tests cyclic crack growth rates were determined. Fracture surface investigations were performed not only for the laboratory test specimens but also for the fracture surface of a sample taken from the HDR test loop piping after the full scale test. In this paper the autoclave testing results and fracture surface observations are presented and compared to those obtained in the HDR piping tests.     

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.     

Opening and extension of circumferential cracks in a pipe subject to dynamic loads

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.