Improvement of FAC Maintenance Program Issued From BRT-CICERO™ via CFD Calculations

2016 ◽  
Author(s):  
Elodie Gipon
Keyword(s):  
Mass Transfer ◽  
Nuclear Power ◽  
Power Plants ◽  
Specific Effect ◽  
Piping System ◽  
Element Analysis ◽  
Accelerated Corrosion ◽  
Wear Rates ◽  
Maintenance Program ◽  
Calculation Algorithms

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.

Effects of Local Wall Thinning With Crack on Stress Intensity Factor for Pipes Subject to Combined Pressure and Bending

2019 ◽  
Author(s):  
Joy (Xiaoya) Tao ◽  
Lei Zhu
Keyword(s):  
Stress Intensity Factor ◽  
Finite Element ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Limit Load ◽  
Minimum Thickness ◽  
Element Analysis ◽  
Accelerated Corrosion

Abstract At ageing power plants, local thinning of pipework or vessel is unavoidable due to erosion/corrosion or other reasons such as flow accelerated corrosion (FAC) — one of the common degradation mechanisms in pipework of nuclear power plant. Local thinning reduces the structure strength, resulting in crack initiation from the corrosion pit or welding defect when subject to cyclic loading. General practice is to use the minimum thickness of the thinned area to calculate both limit load and stress intensity factor (SIF) in performing Engineering Critical Assessment (ECA) using Failure Assessment Diagram (FAD). Using the minimum thickness is normally overly conservative as it assumes that thinning occurs grossly instead of locally, leading to unnecessary early repair/replacement and cost. Performing cracked body finite element analysis (FEA) can provide accurate values of limit load and SIF, but it is time consuming and impractical for daily maintenance and emergent support. To minimise the conservatisms and provide a guidance for the assessment of locally thinned pipework or vessel using existing handbook solutions, a study was carried out by the authors on the effect of local thinning on limit loads. The study demonstrates that local thinning has significant effect on limit load if the thinning ratio of thinning depth to original thickness is larger than 25%. It concluded that the limit load solutions given in handbooks (such as R6 or the net section method) are overly conservative if using the minimum local thickness and non-conservative if using the nominal thickness. This paper discusses the effect of local thinning on SIFs of internal/external defects using cracked body finite element method (FEM). The results are compared with R6 weight function SIF solutions for a cylinder. A modified R6 SIF solution is proposed to count for the effect of local thinning profile. Along with the previous published paper on limit load it provides comprehensive understanding and guidance for fracture assessment of the local thinned pipework and vessel.

Wall Thinning Analyses for Secondary Side Piping of Korean NPPS Using CHECWORKS Code

2002 ◽  
Author(s):  
K. M. Hwang ◽  
T. E. Jin ◽  
S. H. Lee ◽  
S. C. Jeon
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Time ◽  
Accelerated Corrosion ◽  
Wear Rates ◽  
Wall Thinning ◽  
Management Technology ◽  
Reactor Types ◽  
Secondary Side

Since the mid-1990s, nuclear power plants in Korea have experienced wall thinning, leaks, and ruptures of secondary side piping caused by flow-accelerated corrosion (FAC). The pipe failures have increased as operating time progresses. In order to prevent FAC-induced pipe failures and to develop an effective FAC management strategy, KEPRI and KOPEC, along with KHNP’s support, have conducted a study for developing a systematic FAC management technology for all domestic nuclear power plants. As part of the study, FAC analyses were performed using CHECWORKS code. The analysis results were used to select components for inspection on each nuclear power plant. The site application feasibility of the analysis results was proven by comparisons of predicted and measured wear rates. This paper focuses on the introduction of the FAC analysis results for secondary side piping associated with three types of domestic nuclear power plants and the comparisons of predicted and measured wear rates. This paper also represents the comparisons of analysis results according to reactor types, power rates, and systems to facilitate the development of FAC management technology.

Elastic-Plastic Analysis of Thick-Walled Toroidal Pressure Vessels

2008 ◽  
Author(s):  
R. Adibi-Asl
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Toroidal Shell ◽  
Piping System ◽  
Straight Pipe ◽  
Element Analysis ◽  
Process Industries ◽  
Piping Systems

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.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

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.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

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.

A Multi-Dimensional Approach to Design and Execution of Modifications, Refurbishments, and Beyond for Existing Nuclear Power Plants

2020 ◽  
Author(s):  
Omid Malekzadeh ◽  
Matthew Monid ◽  
Michael Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Three Dimensional ◽  
Spatial Models ◽  
Technical Information ◽  
Lessons Learned ◽  
Piping System ◽  
Full Potential ◽  
Cad Models

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.

Evaluation of Flange Calculations Using Strain-Based Acceptance Criteria

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elastic Stress ◽  
Calculation Methods ◽  
Acceptance Criteria ◽  
Class 1 ◽  
Torque Moment ◽  
Calculation Algorithms ◽  
Calculation Code

Abstract The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint different y and m values for different kinds of gasket are invented in ASME BPVC Section III [1]. The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled. For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other. In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.

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