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.     

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.     

Erratum

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.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

Flaw evaluation of Japanese carbon steel piping by load curve approach

New simplified flaw evaluation method, ‘load curve approach’ was developed to evaluate the fracture load of circumferentially surface-cracked pipe. This approach has the same functions with the current two-criteria approach. Fracture stress and fracture criteria are easily estimated by two load curves based on elastic–plastic fracture mechanics and plastic collapse. Fracture analysis was conducted for Japanese carbon steel piping using this approach. The approach showed the dependency of flaw geometry and pipe diameter on pipe fracture. Z-factors were calculated from this approach and compared with Z-factors by ASME Boiler & Pressure Vessel Code, Section XI (ASME-XI) and Japanese Code. It is shown that Z-factors by the load curve approach can improve the conservativeness in the estimation of pipe fracture load.     

Piping response testing associated with pipe rupture

EPRI has sponsored an experimental program in the pipe whip impact and pipe rupture and depressurization areas. Sixteen pipe whip tests were performed with 3 in Schedule 80 (or 10) carbon steel pipes impacting on rigid target or concrete slab. The major testing parameters include distance, impact location, pipe rupture location, and concrete slab thickness and strength. The piping crushing at impact correlates with impact force and target response behavior. Conservatism was established by comparing measured and calculated impact forces. The pipe rupture and depressurization tests were carried out using 6 in stainless steel and carbon steel pipes under either PWR or BWR fluid conditions. These tests are of axial crack with initial machined-in surface flaw. It was found that pipe rupture would occur only if a long unstable through-wall crack was embedded in a sufficiently long unstable part-through crack (in the pipe wall). All other flaw configuration tested led to pipe leakage only. Reaction forces were measured which show conservatism of simplified method for fully ruptured condition. No good crack propagation information was obtained.     

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software.     

Survey of extreme load design regulatory agency licensing requirements for nuclear power plants

Since 1965, when extreme load requirements began to be considered explicitly in nuclear power plant design, there has been a gradual divergence in requirements imposed by national regulatory agencies. However, nuclear plant safety is an international problem because of the potential international effects of any postulated plant failure. For this reason this paper has been prepared in an attempt to highlight the differences in national criteria currently used in the extreme load design of nuclear plant facilities. No attempt has been made to evaluate the relative merit of the criteria established by the various national regulatory agencies. This paper presents the results of a recent survey made of national atomic energy regulatory agencies and major nuclear steam supply design agencies, which requested a summary of current licensing criteria associated with earthquake, extreme wind (tornado), flood, airplane crash and accident (pipe break) loads applicable within the various national jurisdictions. Also presented are a number of comparisons which are meant to illustrate the differences in national regulatory criteria. One example of this difference is the combination of pipe break and earthquake requirements for reactor coolant system design in the US while they are considered separately in Japan. Another example is the difference by which design earthquakes are defined in the US and Japan. In the US, typically a safe shutdown earthquake (SSE) is defined which is considered to be largest earthquake that could rationally be expected to occur at the site, based on regional tectonics or active fault characteristics. A second earthquake term the operational basis earthquake (OBE) which has a required intensity equal to at least half the SSE is defined as the earthquake expected during the 40 yr life of the nuclear plant facility. In Japan a design basis earthquake (DBE) is usually defined based on regional fault characteristics and historical data which could reasonably be expected at the site during the life of the facility. A maximum hypothetical earthquake is then defined for critical structure equal to 1.5 times the DBE for which safe shutdown of the plant must be assured.Other differences highlighted include the fact that the vertical acceleration due to earthquakes have been considered a constant in Japan, while in the US a vertical response spectra has been specified. In W. Germany the effects of aircraft crashes are considered for all plant sites, while in other countries they are considered only when plant sites are in close proximity for commercial airports. In Scandinavia until now there has been no specific requirement for earthquake-resistant design. The US appears to be the only country that routinely considers tornadoes and tornado-borne missiles as a design requirement. In addition to reporting current requirements there is an attempt to project or determine trends in future licensing requirements such as the anticipated reduction of the OBE minimum requirement in the US from one-half to one-third of the SSE.     

Case studies of risk-informed inservice inspection of nuclear piping systems

In the present paper, a probabilistic failure analysis is used to find failure probabilities of piping segments, and a probabilistic risk assessment model is employed to obtain risks to a nuclear power plant should these failures occur. The multiplication of the piping failure probability and the consequence for that particular failure results in the risk contribution of the pipe. The degrees of risk for different piping segments can then be ranked, and their results can be used as the basis for planning a risk-informed inservice inspection program. Numerical studies are offered with special emphases on: (1) the status and experience with RI-ISI applications in Taiwan; (2) the comparison of risk-rankings performed with three different methods developed in the US; (3) aspects of the probabilistic fracture mechanics calculation including the flaw size distributions and stress corrosion cracking model. The results indicate the proposed method can indeed be adopted for planning a cost effective inservice inspection program.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

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.     

Influence of mis-match of weld and base material creep properties on elevated temperature design of pressure vessels and piping

The stress and strain concentrations developed at the weldments during the long time operation of pressure vessels and piping at high temperature due to the mis-match in the creep properties of weldment constituents (weld, heat affected zone and base metal) are estimated using detailed finite element analysis. Three materials, viz. 2.25Cr 1Mo, SS 316 LN and modified 9Cr 1 Mo which are the most commonly used materials in the nuclear and thermal power plants are considered. A longitudinal seam weld with single and double V (X) configurations are analysed. Parametric studies have been done on weld angle and stresses. Based on the analysis, critical locations and the maximum stress concentration factors in the weldments for the above materials are identified. The weld design procedures of the currently used pressure vessel and piping codes are commented. The importance of ductility based failure criteria is emphasised.     

高通量工程试验堆出口管道焊缝破口频率计算

近年来,国内外进行多项研究堆概率安全分析,其中管道破口导致的失水事故是堆芯损坏的重要风险来源。本文参考管道破口计算程序PRAISE(Piping Reliability Analysis Including Seismic Events)方法,选取压力壳型研究堆——高通量工程试验堆(High Flux Engineering Test Reactor,HFETR)的运行工况,对其反应堆冷却剂出口管道的焊缝进行分析,得到运行中该处焊缝发生各类破口的频率。     

Research for the rationalization of nuclear power plant pipe break criteria

The need for a new design basis for pipe break criteria is demonstrated by noting the potential deleterious effect of present criteria in piping during normal operation. Recent advances in fracture mechanics and stress analysis permit development of rational, realistic and conservative criteria that will make possible significant improvements in piping system design. Research needed to form the basis for new criteria is suggested and the nuclear industry is encouraged to work towards this goal.     

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.     

Progress in development of acceptance criteria for local thinned areas in pipe and piping components

Over the last 30 years there has been a considerable amount of research conducted on the effect of corrosion on the burst strength of buried gas and oil transmission pipelines. The results of numerous burst tests on artificial flaws and corroded pipe removed from service were used to validate an empirical analysis that was essentially the limit–load solution for an axial crack in a pipe under pressure loading. This basic concept led to acceptance standards in ANSI B31G, and a more recent modified B31G criterion using the RSTRENG computer program developed at Battelle. This program takes into account variable flaw depths rather than the parabolic flaw shape assumed in the original B31G criterion. Since that time, more fundamental research has been conducted to develop a more accurate and theoretically based failure criterion. The Battelle/Pipeline Research Committee International PCORR computer program is an example of a special purpose shell-element based, finite element, PC criterion for the evaluation of local thinned area (LTA) flaws. This program has evolved with time from linear-elastic to elastic-plastic stress with provisions for axial as well as hoop stresses. The development and new insights into blunt flaw behavior resulting from this program will be one aspect covered in this paper. In the nuclear industry erosion-corrosion, or flow-accelerated corrosion, in single-phase liquid lines has become a major problem. Computer programs, such as the EPRI Checworks program, have been developed to assist the plant operators with deciding where to focus their inspections. However, to date no generally validated acceptance criteria have been developed for the plant piping. Plant piping, whether in nuclear power plants, fossil power plants, or petrochemical plants, have several differences from buried pipelines which need to be considered. The buried pipelines typically have low longitudinal stresses that frequently are compressive, and have no pipe fittings such as tees, elbows, and reducers except at compressor stations. Plant piping needs to consider hoop stresses and axial tension loads from the pressure, as well as, bending stresses from dead-weight loads, thermal expansion stresses, and seismic loads. In an effort to develop flaw acceptance criteria for Section XI of the ASME Boiler and Pressure Vessel Code, the criteria in Code Case N-480 have been revised and implemented into a new code case (the number has not yet been assigned). These criteria essentially use either the ANSI B31G approach for axial flaws, or the ANSI B31.1 or ASME Section III stress analysis rules to show that the residual strength of the thinned region meets the initial design stress limits. This paper presents some of the validation efforts recently undertaken to determine the inherent margins in the design stress equation approach compared with the applied safety factors in the axial and circumferential flaw limit–load solutions in: (i) the gas and oil pipeline industries; (ii) the proposed criteria in Belgium for the nuclear industry and other criteria, and (iii) the preliminary criteria from a recently proposed ASME Code Case on erosion/corrosion acceptance criteria and the ASME Appendix H criteria for flawed ferritic nuclear pipe.     

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the “force equivalent area” (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques.     

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.     

The IPIRG dynamic pipe loop test facility

The International Piping Integrity Research Group (IPIRG) Program was a group program conducted at Battelle, managed by the U.S. Nuclear Regulatory Commission, and funded by a consortium of organizations from nine nations. A unique pipe loop test facility was designed and constructed for the program to evaluate the behavior of nuclear piping containing flaws and subjected to high rate loading typical of high amplitude seismic events. The facility was carefully designed with rigid anchors and special support bearings to provide well-controlled boundary conditions that can be accurately modelled in numerical analyses. Extensive instrumentation provided pipe system response data and pipe fracture data that serve as a test bed to evaluate various structural and fracture analyses.