Elastic-Plastic Analysis of Thick-Walled Toroidal Pressure Vessels

Author(s):  
R. Adibi-Asl

Piping systems in process industries and nuclear power plants include straight pipe runs and various fittings such as elbows, miter bends etc. Elbows and bends in piping systems provide additional flexibility to the piping system along with performing the primary function of changing the direction of fluid flow. Distinctive geometry of these toroidal shell components result in a structural behavior different from straight pipe. Hence, it would be useful to predict the behavior of these components with acceptable accuracy for design purposes. Analytical expressions are derived for stresses set up during loading and unloading in a toroidal shell subjected to internal pressure. Residual stresses in the component are also evaluated. The proposed solutions are then compared with three-dimensional finite element analysis at different locations including intrados, extrados and flanks.

Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.


Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.


Author(s):  
Francis H. Ku ◽  
Pete C. Riccardella ◽  
Steven L. McCracken

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.


Author(s):  
Akemi Nishida

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.


Author(s):  
Omid Malekzadeh ◽  
Matthew Monid ◽  
Michael Huang

Abstract Three-Dimensional (3D) CAD models are utilized by many designers; however, they are rarely utilized to their full potential. The current mainstream method of design process and communication is through design documentation. They are limited in depth of information, compartmentalized by discipline, fragmented into various segments, communicated through numerous layers, and finally, printed onto an undersized paper by the stakeholders and end-users. Large nuclear projects, such as refurbishments and decommissioning, suffer from spatial, interface, and interreference challenges, unintentional cost and schedule overruns, and quality concerns that can be rooted to the misalignments between designed and in-situ or previously as-built conditions that tend to stem from inaccessibility and lack of adequate data resolution during the transfer of technical information. This paper will identify the technologies and the methodology used during several piping system modifications of existing nuclear power plants, and shares the lessons learned with respect to the benefits and shortcomings of the approach. Overall, it is beneficial to leverage available multi-dimensional technologies to enhance various engineering and execution phases. The utilization and superposition of various spatial models into 3D and 4D formats, enabled the modification projects to significantly reduce in-person plant walkdown efforts, provide highly accurate as-found data, and enable stakeholders of all disciplines and trades to review the as-found, as-designed, and simulated as-installed modification; including the steps in between without requiring significant plant visits. This approach will therefore reduce the field-initiated changes that tend to result in design/field variations; resulting in less reliance on Appendix T of ASME BPVC Section III, reduction in the design registration reconciliations efforts, and it aligns with the overarching goal of EPRI guideline NCIG-05. Beyond the benefits to design and execution, the multidimensional approach will provide highly accurate inputs to some of the nuclear safety’s Beyond Design Basis Assessments (BDBA) and allowed for the incorporation of actual design values as input and hence removing the unnecessary over-conservatisms within some of the inputs.


Author(s):  
Elodie Gipon

Flow Accelerated Corrosion (FAC) is very effective for nuclear power plant. This generalized corrosion can lead to the rupture of pipe and in some dramatic cases to casualties. During the last 20 years Electricité de France (EDF) has developed software called BRT-CICERO™ for the surveillance of the carbon steel piping system of its Nuclear Power Plants (NPPs). This software enables the operator to calculate the FAC wear rates by taking into account all the influencing parameters such as pipe isometrics, alloy content, chemical conditioning, design and operating parameters of the steam water circuit (temperature, pressure, etc…). This is a major tool to help operators organize their maintenance and inspections plan. The algorithms implemented in BRT-CICERO™ are based on tests conducted by EDF R&D, empirical results (national and international feedback), literature reviews and on permanent adjustments based on the operating feedback, via statistical studies. However, for some piping components, from the turbine’s hall, flow dynamics are not optimized and calculated FAC kinetics may be too conservative. EDF is committed for optimizing and increasing reliability of its maintenance programs to prevent the risk of pipe rupture due to FAC. As in consequence EDF is leading continuous improvement in parameters and calculation algorithms for BRT-CICERO™. Furthermore studies on the geometric characteristics of the pipes were conducted. In BRT-CICERO™ geometric effect of a pipe component (elbow reduction, tees …) is taken into account by considering a factor called “Geo” in the calculation to tune the thickness loss rate according the component type, its characteristics and specific effect on flow mass transfer. EDF implements finite element analysis software to compute the mass transfer coefficient k and so ascertain the “Geo” coefficient. These computed “Geo” coefficients are compared to those used in BRT-CICERO™. If necessary, current “Geo” coefficients used in BRT-CICERO™ will be adjusted and optimized to improve maintenance programs issued from the software. The presentation deals with the calculation method used for these studies and some results will be shown on tube and elbows.


Author(s):  
Kei Kobayashi ◽  
Takashi Satoh ◽  
Nobuyuki Kojima ◽  
Kiyoshi Hattori ◽  
Masaki Nakagawa ◽  
...  

The present design damping constants for nuclear power plant (NPP)’s piping system in Japan were developed through discussion among expert researchers, electric utilities and power plant manufactures. They are standardized in “Technical guidelines for seismic design of Nuclear Power Plants” (JEAG 4601-1991 Supplemental Edition). But some of the damping constants are too conservative because of a lack of experimental data. To improve this excessive conservatism, piping systems supported by U-bolts were chosen and U-bolt support element test and piping model excitation test were performed to obtain proper damping constants. The damping mechanism consists of damping due to piping materials, damping due to fluid interaction, damping due to plastic deformation of piping and supports, and damping due to friction and collision between piping and supports. Because the damping due to friction and collision was considered to be dominant, we focused our effort on formulating these phenomena by a physical model. The validity of damping estimation method was confirmed by comparing data that was obtained from the elemental tests and the actual scale piping model test. New design damping constants were decided from the damping estimations for piping systems in an actual plant. From now on, we will use the new design damping constants for U-bolt support piping systems, which were proposed from this study, as a standard in the Japanese piping seismic design.


Author(s):  
Yukio Takahashi ◽  
Yoshihiko Tanaka

It is essential to predict the behavior of nuclear piping system under seismic loading to evaluate the structural integrity of nuclear power plants. Relatively large stress cycles may be applied to the piping systems under severe seismic loading and plastic deformation may occur cyclically in some portion of the systems. Accurate description of inelastic deformation under cyclic loading is indispensable for the precise estimation of strain cycles and accumulation potentially leading to the failure due to fatigue-ratcheting interaction. Elastic-plastic constitutive models based on the nonlinear kinematic hardening rule proposed by Ohno and Wang were developed for type 316 austenitic stainless steel and carbon steel JIS STPT410 (similar to ASTM A106 Gr.B), both of which are used in piping systems in nuclear power plants. Different deformation characteristics under cyclic loading in terms of memory of prior hardening were observed on these two materials and they were reflected in the modeling. Results of simulations under various loading conditions were compared with the test data to demonstrate the high capability of the constitutive models.


Author(s):  
F. W. Brust ◽  
E. Punch ◽  
E. Kurth

PWR nuclear power plants have dissimilar metal (DM) welds at many junctions between the vessels and the piping. The DM welds are made with Alloy 82 filler materials between carbon steel and stainless steel. These are potentially susceptible to Primary Water Stress Corrosion Cracking (PWSCC). PWSCC is mainly driven by the tensile weld residual stresses (WRS) that develop during fabrication of the piping system. In particular, weld repairs that often occur during the weld fabrication process also play a strong role in the development of the weld residual stress state in and near the DM welds. Most weld residual stress analyses performed to date in order to characterize the weld residual stress state in DM welds for PWSCC crack growth, leakage, and subsequent failure used axis-symmetric assessments. The purpose of this work is to provide direct assessment of the appropriateness of this axis-symmetric assumption on the WRS by comparison with full three dimensional analyses of several nozzles. In particular, weld start stop effects on the original weld will be assessed. In addition, the effect of partial arc weld repairs will be included. Repair cases considered include 15% and 50% deep repairs of length 48-degree and 96-degree of the circumference, along with the baseline case with no repair. The more complex three dimensional WRS state from the three dimensional analyses are compared to the corresponding axis-symmetric solutions and guidelines regarding the appropriateness of 2D solutions are discussed. Finally, some limited calculations of stress intensity factors at locations along the repair are presented.


Author(s):  
Robert A. Robleto

When designing branch connections in low pressure large diameter piping systems as in Figure 1, thicker is not always better. The flexibility factors in ASME B31.3 1 for branch connections do not assist the designer in taking credit for flexibility that may exist in a large diameter intersection. Since the stress intensification factors (SIFs) are relatively high for large diameter piping, many stub-in branch connections will require a pad to meet the code displacement stress limits. In an ASME B31.3 Piping analysis the stiffness of the branch connections is considered to be as stiff as a straight piece of pipe modeled as a beam. This is a simplifying assumption that can lead to expensive conservatism for the component and possibly non-conservatism for nearby equipment especially when large diameter pipe is considered. Branch connection flexibility is often negligible when compared with piping flexibility of straight pipe perpendicular to the deflection and bends which can ovalize under in-plane bending moments. However, studies at KBR show branch connections in large diameter pipe can contribute significant flexibility to a close coupled piping system.


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