A number of large heavy wall reactors showed severe stress corrosion cracking in the high strength low alloy steel shells. The stress corrosion was mainly caused by the fact that no PWHT was performed which resulted in very high hardness of the Heat Affected Zone; this made the HAZ extremely sensitive for stress corrosion cracking in the feed water environment. The extent of the cracking was such that replacement of all reactors was unavoidable. The redesign and fabrication of the new reactors would take over a year. Fitness for Service methodologies were used keep the reactors running safely until the replacements were installed with minimized down time. This resulted in large economic benefits over several years. The paper covers the FFS assessments performed, the special NDT methods required to get the necessary crack size information at operating conditions as well as failure mode and repair options. It demonstrates an application and the benefits of Fitness for Service assessments on key equipment.

A high pressure carbamate condenser (HPCC) is a shell and tube heat exchanger with low pressure steam/condensate on the shell side and high pressure carbamate solution on the tube side. The inlet channel of the exchanger is protected from the internal corrosive environment by means of an austenitic stainless steel liner. During the first internal inspection after commissioning, a bulge was observed in the liner where a pass partition plate is joined to the liner. Surface-breaking cracks were observed at the partition plate to liner welds in the bulged region. The cracks were removed by shallow grinding, but a fitness-for-service assessment was required to evaluate the suitability of the bulged liner for continued service. Due to the complexity of the geometry, a handheld 3D geometry scanner was used to accurately measure the deformed shape of the bulge. Using the measured geometry, a finite element model was developed and used to perform a level 3 fitness for service assessment. This paper will describe the assessment procedure in detail as well as the complex internal repairs that were ultimately implemented in order to restore the integrity of the equipment.

Delayed coker drums are unique in hydrocarbon processing facilities in that estimating their true design and service life has been problematic. Generally, pressure containing equipment in these facilities is designed using the notion of design life based on required pressure thickness and corrosion allowance considerations. Hence, pressure containing equipment is routinely monitored by facility inspectors for wall thickness. Although many analysts have ascribed coke drum failure to “thermal stress cycling”, the difficulty posed by the operation of coke drums results in an inability to measure or calculate the magnitude of the thermo-mechanical “stresses” and the actual number of significant exposures, that is, cycles causing fatigue damage. As well, the use of Code construction practices has been generally misapplied, for this specific equipment, as the practices are intended to define a safe design life rather than a service life. Indirect measures of service life based on shell bulge severity have fallen from favor by being ineffective. A trend to use a strain index method is somewhat more appealing but is based on static load and monotonic material property considerations rather than those properties indicative of thermal cyclic operation. Recent work has shown that thermo-mechanical strain cycling can be characterized quantitatively and used to determine a cyclic service life for both undamaged and damaged coke drums. This paper discusses some of the engineering specifics to generate a high probability estimate of coke drum fatigue service life for a new drum, a damaged-stable drum, drums with weld overlay and for drums exhibiting incremental damage.

The current paper examines the effects of corrosion induced wall thinning on buckling of domed closures onto cylindrical vessels. It is assumed that corrosion is axisymmetric and that the wall is corroded on inside, only. The ratio of corroded wall thickness, t c , to the non-corroded thickness, t, is varied between 0.10 ≤ t c /t ≤ 1.0. Both depth of corrosion and its meridional extend are varied during numerical calculations. Three modelling scenarios for placement of corrosion are considered: (i) corrosion confined to the knuckle, (ii) corrosion spanning evenly the knuckle and spherical parts, and (iii) patchtype area positioned at the apex. Numerical results indicate that the following factors influence buckling performance of the dome: (i) meridional position of corroded area, (ii) depth of corrosion itself, and (iii) meridional span of corroded wall. For example, wall thinning of 10 % over 10 % of meridional length causes almost 20 % drop in buckling strength. The largest drop of load carrying capacity is found when the corroded wall is at the knuckle/crown junction. Here it is shown that assessment of strength based on the collapse mechanism is not only wrong but dangerous. For the case of the corroded dome, the collapse pressure overestimates the load carrying capacity associated with asymmetric bifurcation buckling by 40 %.

This paper describes the failure of a jacketed vessel. The product pressure is 16 barg (inner vessel; D = 2200 mm) and is heated with thermal oil in the jacket (operated at 4barg). The jacket is split up in different zones which can be opened and closed separately. After a shut down, several valves were not opened properly. This resulted in blocking-in of the jacket on the top head of the vessel during the start-up and operation of the vessel. The vessel was heated to the operating temperature (±250°C), causing a pressure increase of the blocked-in thermal oil. The jacket has a wall thickness of 10 mm, and the vessel head has a wall thickness of 22 mm. Because of the pressure increase, failure occurred at a nozzle weld on the inner pressure vessel shell. This resulted in a leakage of thermal oil into the vessel. The jacket itself deformed but did not fail. Based on a detailed FE analysis, it has been concluded that failure occurred as a result of (local) buckling of the 22mm thick elliptical head (diameter of ±2.200mm). This paper describes the failure that occurred and the assessments performed to determine and validate the root cause of the failure. A level 3 assessment according to ASME VIII div 2 Part 5 (1) was used to determine if the vessel is still safe for operation.

Following the flaw indications found in summer 2012 in two Belgian Reactors Pressure Vessels (RPV), WENRA recommended [1] the nuclear safety authorities in Europe to verify the material quality and integrity of the RPV in a 2-step approach: 1) a comprehensive review of the manufacturing and inspection records of the forgings of the RPV, 2) an additional UT examination of the base material of the vessels if needed. In this context, and to consolidate scientific basis on this issue, IRSN, the French technical safety organization, conducted, with CEA support, a test program aiming at studying the consequences of hydrogen flakes in large forgings of primary equipment (RPV, steam generator, pressurizer). Framatome provided the material to be investigated, namely two blocks of a steam generator vessel shell in 18MND5 steel: a block without flake — the reference block — and a block including a high density of hydrogen flakes. This shell — so called VB395 — was rejected because of an incident which occurred during the degassing heat treatment. Fracture toughness has been evaluated from 85 tests in the ductile range and the ductile-to-brittle transition range of the material. The test results on usual 0.5T-CT specimens were compared to those on specimens containing a hydrogen flake replacing the fatigue precrack. The latter were interpreted using 3D elastic-plastic X-FEM simulations allowing the modelling of the irregular flake geometry. Furthermore, large scale bending specimens with multiple flakes have been tested at −100°C. These tests were interpreted thanks to 3D X-FEM simulations allowing the analysis of the hydrogen flake interaction in terms of K J .

During the 2012 outage of the Belgian nuclear power plants (NPP) Doel 3 and Tihange 2 non-destructive testing (NDT) measurements revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels (RPV). A root cause analysis leads to the most likely hypothesis that the indications are hydrogen flakes in segregated zones of the RPV ferritic base material. The laminar and quasi-laminar orientation (0° – 15° inclination to the pressure retaining surface) of the hydrogen flakes, the interaction of several adjacent flakes and the mechanical loading conditions lead to a mixed-mode behavior at the crack tips. In the framework of an ongoing research project, experimental and numerical investigations are conducted with the aim to describe the failure behavior of such complex crack configurations. The experiments are carried out using two ferritic materials. One is a non-irradiated representative RPV steel (SA 508 Class 2) and the second material is a special lower bound melt of a modified 22NiMoCr3-7 steel (FKS test melt KS 07 C) containing hydrogen flakes. A material characterization is done for both materials including tensile specimens, notched round bars, shear-, torsion- and compact-tension-shear (CTS) - specimens to investigate different stress states. Furthermore, flat tensile specimens with eroded artificial crack fields are used to investigate the interaction between the cracks in different arranged crack fields. Numerical simulations are carried out with extended micromechanical based damage mechanics models. For the description of ductile failure an enhanced Rousselier model is used and an enhanced Beremin model to calculate the probability of cleavage fracture. To account the sensitivity for low stress triaxiality damage by shear loading, the Rousselier model was enhanced with a term to account for damage evolution by shear. The Beremin model will be enhanced with a term to account for different levels of triaxiality. For the numerical simulations in the transition region of ductile-to-brittle failure a coupled damage mechanics model (enhanced Rousselier and Beremin) will be used. In this paper, the current status of the ongoing research project and first results are presented.

The loads on the equipment nozzles are generally generated by the piping stress engineer by doing the stress analysis of entire closed loop systems. Subsequently the nozzle loads are passed on to the engineers of the pressure vessel equipment. The value of the loads which have been worked out for the nozzle mostly depends upon the methods/concept by which the piping stress engineer has evaluated the piping loop. Nozzle flexibility/stiffness is the important parameter in evaluation of various components of nozzle loads. The objective of this paper is to explain the effect/influence of flexibility/stiffness generated from three different methods (Anchor, WRC and Finite element method) on nozzle load evaluation and shell/nozzle junction stresses. WRC297 bulletin [6] gives the reference to nozzle flexibility in the appendix A, example no.3. The work presented in this paper is an attempt to compare the nozzle loads calculated by evaluating the flexibilities/stiffness in various methods. Further an attempt has been made to consolidate the results of junction local stresses obtained by the various methods of stiffness/flexibilities which would result in realistic results and overall code acceptable stresses without the results being either overly conservative or unconservative.

Non-destructive testing measurements in the Belgian nuclear power plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels. The most likely explanation is that the indications are hydrogen flakes positioned in segregated zones of the base material of the pressure vessel. These hydrogen flakes have a laminar and quasi-laminar orientation to the pressure retaining surface. Under mechanical loading the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. Due to these specific loading conditions, the assumptions for classical standardized fracture mechanical methods are not met. Currently, there is no verified concept for the evaluation of such kind of crack fields. Therefore the mechanical behavior of components with laminar crack fields and the interaction of cracks in such crack fields are investigated in an ongoing research project. Relevant parameters to describe crack fields in terms of crack size, crack location and crack orientation are derived from literature and own nondestructive measurements. Damage mechanical approaches are used in finite element calculations to investigate the interaction of cracks. Advanced damage mechanical models will be used to investigate crack initiation, crack growth and coalescence of cracks in crack fields. According to the results, representative parameters for crack fields will be derived and critical crack formations determined. The results will be evaluated and compared with state of the art approaches and standards.

We measured the natural frequencies of the fluid-elastic-coupled shell plate vibration excited in a large-sized, flat-bottomed, cylindrical steel tank by observing microtremors at three points of the tank. By comparing the appearance frequencies and the values of the five peaks seen in the observed microtremor spectral ratios of the top or mid-height of the shell plate to the bottom on the tank foundation with the solutions obtained from a fixed-base FEM eigenvalue analysis, we identified the five peaks as belonging to modes (m, n) = (1, 1–5), with m specifying the vertical order and n the circumferential wave number. The measured non-soil-coupled natural frequencies from the spectral ratio agreed fairly well with those obtained from theory using the FEM analysis. The measured frequencies of the fundamental mode (m = n = 1) were also in good agreements with those estimates using a simplified equation assuming a fixed base adopted in seismic codes of the Japanese Fire Service Act. This equation is expected to provide a reliable soil-coupled fundamental-mode natural frequency for a tank resting on firm ground; the storage-soil-coupled effects are presumed to be weak. Without using an FEM analysis, we present a simple approach to determine the non-soil-coupled fundamental-mode natural frequency solely from the observed microtremor spectral ratios. This simplified procedure works very well for the tank examined.

The current seismic design rule on piping assumes elastic analysis without the effect of response reduction due to plasticity, although some degree of plasticity is allowed in its allowable limits. Damping for the seismic design analysis is conservatively determined depending on the number of supports and thermal insulation conditions. These conservative assumptions lead to large amount of design margin. Based on such recognition, to provide a more rational seismic design method, a new Code Case for seismic design of piping is now under development in the framework of JSME Nuclear Codes and Standards as an alternative rule to the current design rule. The Code Case provides detailed inelastic analysis with using shell or solid FEA models as a more rational method. Simplified analysis with an additional damping taking the response reduction due to plasticity into account is now under consideration to incorporate the convenience in design. In this study, a series of analysis was made to see the adequacy of the simplified inelastic analysis. Design margins contained in the current design analysis method composed of response spectrum analysis and stress factors was quantitatively assessed in the view point of additional damping.

Combustion turbines are frequently used because they provide the most power in the smallest footprint and their modular design makes them an economical choice. These machines are used across all land-based industries as well as marine applications. The exhaust system must perform its basic function of conducting exhaust gases, which can be as high as 1250° F (675° C), safely away from the adjacent equipment and workers, and mitigate the exhaust noise under a wide range of requirements and conditions. In addition, new requirements limit shell temperatures and exhaust leakage to prevent fire or explosion of fuel gas that may leak from equipment (ATEX). This paper presents a review of the analytical processes used in the development of a silencer system to achieve optimal performance metrics. These systems are typically comprised of parallel baffles for a wide range of conditions including aero-acoustical performance, system pressure and high flow rates, thermal stresses, environmental conditions (ocean, seismic and wind), flow-induced vibration, corrosion, and fatigue – design life analysis. The specific requirements of the baffle design will be discussed through a specific case study relative to typical rectilinear (parallel) baffles used in many installations, including land-based power generation plants (Figure 1), and offshore platforms (Figure 2). This paper will discuss the analytical methods used to address these challenges via a case study. A combination of static, vibration-pulsation and dynamic structural analysis with specific attention to the seismic analysis of the parallel baffles used in skirt and structural steel supported vessels, as well as acoustical design and flow modeling techniques are used to evaluate the design options.

It is important in the confirmation of the safety of the nuclear power plant to clarify the response behavior of a vertical cylindrical water storage tank under seismic motion. When a vertical cylindrical tank is shaken by a large earthquake, deformation of side shell due to the elephant foot buckling, the oval vibration etc. may occur. The occurrence of those deformations depends on materials, shapes, stored water level and time history of seismic motion. Then, response behavior was obtained for a condensate storage tank (CST) model under large seismic motion such as standard earthquake Ss multiplied by 2 with the elastic-plastic finite element calculation. In this calculation, dynamic water pressure and elastic-plastic characteristics of the material were taken into account. In this case, the elephant foot bulge did not occur but the oval vibration of side shell became dominant. Based on the result, we estimated the structural integrity of the tank.

Evaluating the dynamic response of large, aboveground storage tanks exposed to seismic loading is multifaceted. There are foundation-structure and fluid-structure interaction effects that can influence the overall tank behavior and likely failure modes. Additionally, local stresses at anchor bolt support chair attachments and the shell-to-floor junction can be difficult to quantify without detailed finite element analysis (FEA). Often times, performing explicit dynamic analysis with liquid sloshing effects can be time consuming, expensive, and even impractical. The intent of this paper is to summarize simplified analysis techniques that can be leveraged to evaluate aboveground storage tanks subject to seismic loading. Closed-form calculations to establish a recommended design for a tank, including seismic considerations, are available in storage tank design standards, including API 650 [1] ( Appendix E). Seismic design standards have evolved significantly in recent years. Furthermore, for many vintage, in-service storage tanks, explicit seismic considerations were not incorporated into the original design. In Part I of this study, these design equations and other closed-form solutions are used to evaluate the structural integrity of a large, in-service, mechanically-anchored storage tank. The design equations in API 650 [1] are used to form the basis of simplified, equivalent static analysis, where seismic loads are applied to a three-dimensional FEA model via equivalent lateral body forces. These practical results are then compared to explicit dynamic seismic behavior of the same tank with fluid-structure interaction effects considered (in Part II of this study [2]). These comparisons offer insight into the appropriateness of using simplified hand-calculations and equivalent static analysis (and their relative conservatism) in lieu of more rigorous explicit dynamic and fluid sloshing simulations.

In order to prevent brittle fracture, the pressure and temperature in a reactor pressure vessel (RPV) is controlled by pressure-temperature (P-T) limit curves during the heat-up and cool-down processes. Nuclear power plants should update the P-T limit curves periodically, because of RPV material irradiation embrittlement. Too restricted P-T limit curves may cause difficulty of operating a reactor. The 2007 edition of the RCCM code Annex ZG provides a new method for defining the P-T limit curves. In this paper, two types of the P-T limit curves for a French type RPV are established by different methods, which are the current operation limits based on the 1993 edition of the RCCM code and the new proposed limits according to the 2007 edition of the code. The margins of the current P-T limit curves are evaluated by comparing with the new proposed limit curves. Furthermore, the reasonability of improvements of the new P-T limit curve method is discussed, and their individual effects are investigated, including the conventional defect size, the required material toughness and the stress intensity factor plastic correction. The present results indicate that the current P-T limit curves for the RPV studied are conservative and have about 25∼70 °C margin in the transition temperature range and about 10∼12MPa in the upper shell temperature range, depending on different conditions. The new P-T limit curve method, which not only removes some conservative assumptions in the previous method but also restricts some requirements, is more reasonable and can provide a relaxed operation window. Present results can be a reference for the nuclear power plant owner to release the operation limits and is helpful in enhancing our understanding of the P-T limit curve.

In the aerospace industry, open hole specimens of composite laminates have been used in standardized tests to generate design allowables. Using finite element method (FEM) based tool MicMac/FEA with AB AQUS code interface and statistical design of experiments, Shah, et al. in 2010 [11] studied average-property-based failure envelope with uncertainty estimates of open hole specimen with quasi-isotropic carbon fiber-epoxy laminate. However, their FEM model is deterministic, without uncertainty analysis. In this paper, based on Shah’s FEM model, we developed FEM model of uni-axial strength test of holed composite laminates using ABAQUS with a serious of quadrilateral S4R and trilateral S3R shell element designs. The mesh density ranges from the original 8 × 8 (very coarse) to 48 × 48 (very fine). For each of the meshes, we compute the failure strength from Hasin failure criteria. Then we use a 4-parameter logistic function nonlinear least squares fit algorithm to obtain an estimate of the failure strength at infinite degrees of freedom (d.o.f) as well as its uncertainty at 50,000-d.o.f. and relative error convergence rates. Our results are then compared with Shah’s with the additional advantage that our results have uncertainty quantification that can be compared with experimental data. The significance and limitation of our method on the uncertainty quantification of FEM model of uniaxial strength test of holed composite laminates are discussed.

Stress intensity factor solutions are readily available for flaws found in pipe to pipe welds or shell to shell welds (i.e., circumferential/axial crack in cylinder). In some situations, flaws can be detected in locations where an appropriate crack model is not readily available. For instance, there are no practical stress intensity factor solutions for circular-arc cracks which can form in circular welds (e.g., nozzle to vessel shell welds and storage cask closure welds). In this paper, stress intensity factors for circular-arc cracks in finite plates were calculated using finite element analysis. As a first step, stress intensity factors for circular-arc through-wall crack under uniform tension and crack face pressure were calculated. These results were compared with the analytical solutions which showed reasonable agreement. Then, stress intensity factors were calculated for circular-arc semi-elliptical surface cracks under the lateral and crack face pressure loading conditions. Lastly, to investigate the applicability of straight crack solutions for circular-arc cracks, stress intensity factors for circular-arc and straight cracks (both through-wall and surface cracks) were compared.

The determinant of the Jacobian matrix is frequently used in the Finite Element Method as a measure of mesh quality. A new metric is defined, called the Standard Error, based on the distribution of the determinants of the Jacobian matrices of all elements of a finite element mesh. Where the Jacobian norm can be used to compare the quality of one element to another of the same type, the Standard Error compares the mesh quality of different versions of a finite element model where each version uses a different element type. To motivate this new Standard Error, we investigate the geometric meaning of the Jacobian norm on 3D Finite Elements. This mesh quality metric is applied to 8, 20, and 27 node hexahedra, 6 and 15 node prisms, 4 and 10 node tetrahedra, 5 and 13 node pyramid, and 3, 4, 6, 8, and 9 node shell elements. The shape functions for these 14 element types, or more precisely their first partial derivatives, are used to construct the Jacobian Matrix. The matrix is normalized to compensate for size. The determinant of the Jacobian is calculated at Gaussian points within each element. Statistics are gathered to form the Standard Error of the mesh. To illustrate the applicability of this a priori metric, we present two simple example problems having exact answers, and two industry-type problems, a pipe elbow with a crack and a magnetic resonance imaging (MRI) birdcage RF coil resonance, both having no analytical solution. Significance and limitations of using this a priori metric to assess the accuracy of finite element simulations of different mesh designs are presented and discussed.

Thermal membrane and bending stress fields exist where the thermal expansion of pressure vessel components is constrained. When such stress fields interact with pressure stresses in a shell, ratcheting can occur below the ratchet boundary indicated by the Bree diagram that is implemented in the design Codes. The interaction of primary and thermal membrane stress fields with arbitrary biaxiality is not implemented presently in the thermal stress ratchet rules of the ASME Code, and is examined in this paper. An analytical solution for the ratchet boundary is derived based on a non-cyclic method that uses a generalized static shakedown theorem. The solutions for specific applications in pressure vessels are discussed, and a comparison with the interaction diagrams for specific cases that can be found in the literature is performed.

Post weld heat treatment (PWHT) is often applied to welded pressure vessels and piping to increase the mechanical and metallurgical properties. During the PWHT, a pressure component made of a carbon steel to which PWHT is applied, is heated up to the temperature around 620 deg.C with +/− 25 deg.C tolerances and that elevated temperature is kept for 1 to 2 hours. The local PWHT of the vertical pressure vessel is sometimes planned at a construction site with vertical position, because of some restrictions relating to construction, such as limited access route to the vessel location, and limited space large capacity cranes. In the situation, it is important to assure the structural stability during PWHT. Except wind loads, it is considered that the stress applied in the shell during PWHT is usually not significant, however there have been no published strength criteria corresponds to the PWHT conditions. In this paper, the new elevated temperature tensile tests and constant stress creep strain tests of A516-Gr.70, a typical carbon steel material for pressure vessels, are conducted. Several examples of the numerical analysis using isochronous stress strain curve considering creep effect are presented. From the test result, the importance of the consideration of the creep deformation on the structural integrity assessment of PWHT operation is clarified.