Investigations on the integrity of fast breeder reactor piping and components

Results of investigations made with respect to the integrity of LMFBR (Liquid Metal Fast Breeder Reactor) piping and components are presented. The classification of sodium systems as Moderate Energy Fluid Systems is shown to be an important principal element to determine the failure mechanisms which are relevant. Based upon the selection of materials, design features and high-quality engineering standards the evaluation of the crack growth morphology of surface flaws contribute to the ensurance of the structural integrity. The crack shape development for bending stress distribution over the wall-thickness, which is a typical loading of FBR structures is discussed. It has been shown that even for such unfavourable loading conditions the through crack lengths are bounded. There is a considerable distance from critical crack configurations calculated by tearing modulus concept. Results from a large scale elbow test at operating temperature are reported. They contribute to the crack shape development under fatigue loading with bending type stress distribution over the wall thickness and are in good accordance with calculations. Acceptance criteria for flaws in structures are proposed showing that the structural integrity for coolant boundary and components of FBR can be assessed with a high degree of reliability.     

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates da/dt demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K_I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.     

Fracture behavior of carbon steel pipe with local wall thinning subjected to bending load

To evaluate the structural integrity of power plant piping, monotonic bending tests are conducted on 4- and 3.5-in. diameter full-scale carbon steel pipe specimens with local wall thinning. The local wall thinning is simulated as erosion/corrosion metal loss. The eroded area of the wall thinning is subjected to tensile or compressive stress by applied bending moment. The deformations or fracture behaviors at maximum moments are found to be classified into three types. When the eroded area is subjected to tensile stress, ovalization or crack initiation/growth occurs at the maximum moment. When an eroded area is subjected to compressive stress, ovalization or local buckling occurs. The occurrence of ovalization, crack initiation/growth, or local buckling depends on the initial size of local wall thinning. From the relationships among ovalization, crack growth and local buckling, allowable sizes for local wall thinning are proposed.     

Fatigue and fracture behavior of straight pipe with flaws in inner surface

Fatigue and fracture tests of piping models with flaws in the inner surface were carried out to investigate the fatigue crack growth, coalescence of multiple cracks and fracture behavior.Two straight test pipes with and without weldment in the test section of AISI type 304L stainless steel were tested under almost the same test conditions by imposing moment loads. Three artificial defects were machined in the inner surface of the test section of the test pipes and the fatigue test was performed until the cracks coalesced and grew through the thickness. Subsequently, a static load was imposed on the test pipe which contained a large crack in the test section.The fatigue test results are compared with an analytical crack growth behavior predicted by the method described in the Section XI of ASME Code, and show slower crack growth than that of the prediction. From the fracture test results, it is found that the test pipes can endure considerably high load.     

Fracture mechanics evaluation of small diameter piping considering the latest experimental results

The results obtained from investigations carried out on austenitic piping of small nominal diameter (DN80 and DN50) are introduced and discussed together with their assessment using fracture mechanics methods. Essential results are summarised as following. The pipes with flaws (fatigue crack) down to a depth to a_max/t=0.51 (DN80) as well as a_max/t=0.62 (DN50) and a circumferential extension of results 2α=120° reached bending angles up to 26°. The ASME collapse load (test collapse load) was exceeded considerably and the experimental maximum load could not be reached. Failure due to a leakage or rupture did not occur in any test. The maximum crack extension was 0.69 mm (DN80, a_max/t=0.51) resp. 0.3 mm (DN50, a_max/t=0.62). The experimental maximum load can approximately be assessed by the limit analysis. The fracture mechanics approximation methods GE/EPRI and LBB/NRC calculated a/t=0.4 and 2α=120° initiation loads above the experimental maximum load for pipes containing flaws. These results confirmed the procedures for the proof of integrity of small diameter piping by updating information on load, deformation and failure behaviour of austenitic piping damaged with circumferential flaws. Using these results may formulate a final safety concept for the proof of integrity of small diameter piping by completing the current concepts.     

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.     

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.     

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.     

Fatigue growth and fracture risk from postulated cracks in pressurized-water reactor piping welds

The necessity of taking quick corrective action when a crack indication is discovered in a nuclear piping weld, has led Framatome to evaluate beforehand the potential risk of such a situation by investigating postulated cracks. Considering pessimistic loading conditions to act on a postulated crack of a given shape and orientation enables the determination of the critical size of such a defect. Introduction of fatigue crack growth then yields the maximum crack size that can be tolerated, given the remaining lifetime of the unit. Additionally, detailed analysis of the scenario that leads to these results contributes to the understanding of the potential risk and helps in alleviating it. In this paper, a review of the basic principles and the application to the case of a branch connection weld are presented.     

Cessation of environmentally-assisted cracking in a low-alloy steel: experimental results

The presence of dissolved metallurgical sulfides in pressure vessel and piping steels has been linked to environmentally-assisted cracking (EAC), a phenomenon observed in laboratory tests that results in fatigue crack growth rates as high as 100 times that in air. Previous experimental and analytical work based on diffusion as the mass transport process has shown that surface cracks that are initially clean of sulfides will not initiate EAC in most applications. This is because the average crack tip velocity would not be sufficiently high to expose enough metallurgical sulfides per unit time and produce the sulfide concentration required for EAC. However, there is a potential concern for the case of a relatively large embedded crack breaking through to the wetted surface. Such a crack would not be initially clean of sulfides, and EAC could initiate. Previous experiments have suggested that under some conditions, EAC could be persistent. This paper presents the results of a series of experiments conducted on two heats of an EAC susceptible, high-sulfur, low-alloy steel in 243°C low-oxygen water to further study the phenomenon of EAC persistence at low crack tip velocities. A load cycle profile that incorporated a significant load dwell period at minimum load was used. Experiments using compact tension specimens with various initial precrack depths were employed to simulate the breakthrough of embedded cracks. The results showed that EAC ceased after several hundred hours of cycling. This indicates that significant dwell periods can allow sufficient time for sulfur diffusion to turn off EAC provided that the initial crack tip velocities are not unusually high.     

OPDE—The international pipe failure data exchange project

Certain member countries of the Organization for Economic Cooperation and development (OECD) in 2002 established the OECD pipe failure data exchange project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large through-wall flow rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of design code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection.Currently the OPDE database includes approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-safety-related (non-Code) piping. This paper presents the motivations and objectives behind the establishment of the OPDE project. The paper also summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. An overview of the database content is included to place it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance. Finally, a brief summary is given of current database application studies.     

Probabilistic fracture assessment of piping systems based on FITNET FFS procedure

A probabilistic method based on the fracture module of the FITNET FFS procedure is developed to perform the structural integrity analysis for piping systems. Monte Carlo simulation is used to calculate the failure probability of the whole piping system, as well as that of separate defects by considering the random variables in the method. Both the sensitivity of uncertainties of variables and the model sensitivity are analyzed to identify the most important parameters that affect the failure probability of piping systems, thereby providing an insight into the countermeasures against the failure risk. The results show that the outer diameter of the pipe has the strongest influence on the failure probability of a piping system having a circumferential crack of 0.757 rad, followed by the bending moment, the piping wall thickness, the fracture toughness, the crack angle, the axial force, the ultimate tensile strength and the yield stress.     

Application of the J-integral concept to thermal shock loadings

In the frame of our analytical work the applicability of ductile fracture mechanical J-integral concept on mechanical and thermal shock loaded structures with flaws is investigated. By that the behaviour of possible flaws in components of power plants during accidents can be described (e.g. reactor pressure vessel and piping during emergency cooling).The analyses presented in this paper have been performed with a version of the finite element code ADINA [1] extended by fracture mechanical options. The postanalyses of the first series of pressurized thermal shock experiments (PTSE-1A, B, C) performed at ORNL show stress intensity factors (K_I) calculated from J-integrals which are about 10% lower than values of OCA programs [2] based on the linear elastic K-concept usually used for brittle materials. The discrepancy may be referred to different treatment of the influence of plasticity. The results assessed in the frame of the cleavage fracture concept coincide well with the measured times respectively crack tip temperatures at crack initiation and arrest.In the first thermal shock experiment (NKS-1) performed at the MPA-Stuttgart a circumferentially deep cracked test cylinder with overall upper shelf material conditions has been investigated. The postcalculations based on the J-integral with J_R-controlled crack growth show good coincidence between analytical determined and measured structure and fracture mechanical quantities but they are accompanied with numerical problems due to unloading and large plasticity effects.     

The importance of the triaxiality modified J concept for the assessment of pressurized components with surface flaws

With the progress of stable crack growth of surface flaws observed in panels or pressure vessels a canoe-shaped crack front is formed. The crack propagation in the longitudinal direction is more pronounced that in the wall thickness direction. Therefore, the canoe effect is important with respect to a leak-before-break assessment because the actual through crack length is influenced by this effect. Based on the J integral concept crack initiation and crack propagation in ductile materials are described by J resistance curves which were found to be dependent on the constraint effect of the specimen geometry. Prediction of local crack growth by taking a conservative (flat) J_R-curve into account results in a nonconservative estimate of the axial extension of the surface crack [W. Brocks, H. Veith and K. Wobst, in K. Kussmaul (ed.), Fracture Mechanics Verification by Large Scale Testing, Mech. Eng. Publication Limited, London, 1991]. This means that the influence of local constraint effects on crack resistance has to be considered.Ductile crack growth of semi-elliptical surface cracks in side-grooved specimens F(SCTsg) under tension made from German standard steel StE 460 will be reported on. The development of the canoe effect of an SCTsg specimen was also analysed by a finite element simulation of ductile crack growth which was modelled by using the node shift and node release technique and controlled by crack mouth opening displacement versus crack growth curves from the experiment. The simulation allows the determination of local J_R-curves in dependence on the local multiaxility of the stress state to verify the constraint modified J concept. It is demonstrate that the slope of the J_R-curves decreases with increasing multiaxiality of the stress state near by the crack front.     

Development of modified piping evaluation diagram for leak-before-break application to Korean next generation reactor

Recently, the piping evaluation diagram (PED) is accepted in nuclear industry for an efficient application of leak-before-break (LBB) concept to piping system at an initial piping design stage. The objective of this paper is to develop the modified PED, which can account for the variation of the material properties of the PED development stage and those of the assembly stage. For this purpose, a parametric study was performed to investigate the effect of stress–strain curve on the detectable leakage crack length and the effect of fracture resistance curve on the LBB allowable load. Finite element analyses were also performed to investigate the effect of stress–strain curve on the LBB allowable load. Finally, a modified PED was proposed as a function of crack length and the allowable safe shutdown earthquake load. The LBB analyses based on the modified PED are in good agreement with those based on the traditional PED. By adopting the modified PED, the variation of material properties can be considered in the LBB analysis and the computing times required for the application of LBB during the design process can be considerably reduced.     

承压热冲击下压力容器断裂力学分析

依据美国核管会（NRC）最新法规要求和研究进展,阐述了压水堆核电厂反应堆压力容器（RPV）承压热冲击（PTS）最新评估方法。基于热工水力系统程序RELAP5和有限元分析软件ANSYS,针对某传统二代压水堆核电厂模拟在PTS典型瞬态过程下热工响应行为及压力容器模型断裂力学分析,并评估不同瞬态的危险性及其随压力容器材料脆性的变化。分析表明：表面裂纹和靠近内壁面的埋藏裂纹比深埋裂纹更易发生开裂;同等条件下轴向裂纹较环向裂纹更易开裂,且大中破口事故下轴向裂纹远较环向裂纹更易贯穿壁厚。     

Stress-corrosion cracking experience in piping of light water reactor power plants

Cracking has been observed in the heat-affected zones of welds that join small diamter austenitic steel piping and associated components in boiling water reactors. It was concluded that much of this was caused by intergranular stress-corrosion cracking. In 1975 the US Nuclear Regulatory Commission established a pipe crack study group to investigate this cracking in order to minimize and curtail this phenomenon. In 1978 intergranular stress-corrosion cracking was observed for the first time in large diameter piping. A second pipe crack study group was formed, with an expanded charter, to continue investigations and answer specific questions concerning pipe cracking. This paper summarizes the results of the two study group investigations and presents the major conclusions and recommendations regarding the causes, detection, and control of such pipe cracking. Also discussed is the history of the observed cracking, metallurgy associated with intergranular stress-corrosion cracking, the effects of the primary coolant chemistry, developed stress levels in the heat-affected zones of piping, methods of crack detection, and the importance of leak detection.     

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.     

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.     

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.