BWR SCC topic (incl. CFD considerations)

Webinar Date: October 26, 2010
Instructors: Daniel Sommerville, Jay Gillis
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Current Status of Rick Based Programs

Webinar Date: November 17, 2010
Instructor: Scott Chesworth
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About the Instructor

Scott Chesworth has a MS in Mechanical Engineering from the University of California, Davis. He has extensive experience in risk-informed in-service inspection applications, including FatiguePro™ application experience and experience performing structural, stress, fatigue, fracture mechanics, and dynamic analysis. Chesworth also has solid working knowledge of ANSYS® finite element software.

License Renewal/Fatigue/TLAA Updates

Webinar Date: December 8, 2010
Instructors: Dave Gerber, Tim Griesbach
Registration: Click here to register for this live Webinar

About the Instructors

Tim Griesbach has a BS and MS in Materials Engineering from Case Western Reserve University, and over 30 years of experience in materials behavior and structural integrity of major nuclear components. His specialty is technical consulting utilizing state-of-the-art technologies for mitigating and resolving material degradation concerns in nuclear reactor vessels, internals, piping, and other major components. Griesbach has led numerous EPRI® projects to develop tools for managing such aging effects as fatigue and reactor vessel embrittlement in nuclear pressure vessels, and he has worked with utilities, NEI, and NRC to resolve such critical industry issues as pressurized thermal shock in PWRs.

ASME Code Section XI Flaw Evaluation for Nuclear Components

ASME Code Section XI stipulates flaw evaluation for nuclear components in both PWRs and BWRs, resulting in either acceptance with no further action, acceptance with supplemental re-examination, or repair/replacement. This SI Webinar provides a comprehensive overview of the elements of flaw evaluation and disposition, including flaw characterization and acceptance standards; vessel and piping flaw evaluations, and an introduction to weld overlays. The presentation includes detailed descriptions of planar, laminar, and linear flaws; discussion of linear-elastic and elastic-plastic fracture mechanics, as well as limit load analysis; and information on weld overlay experiences at PWRs and BWRs, experimental weld overlay programs, and verification of weld overlay experience.

Presented on: August 1, 2007
Instructor: Peter Riccardella
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About the Instructor

Peter Riccardella has a PhD in Mechanical Engineering from Carnegie Mellon University, and over 30 years of experience in design and analysis of large structural components. He is an authority on application of fracture mechanics to pressure vessel and piping problems, notably feedwater system thermal fatigue cracking, stress corrosion cracking in power plant piping, and pressure vessel embrittlement and pressurized thermal shock. Riccardella pioneered the development of flaw evaluation procedures and acceptance standards for ASME's Nuclear In-Service Inspection Code.

Advancements in Finite Element Analyses for the Nuclear Industry

Finite element analyses have been used more and more extensively to evaluate and qualify structures and components for ASME Code compliance in the nuclear industry. Improvements in FEA capabilities and computational speed make sophisticated FEA more practical. The webinar showcases analyses performed by SI that utilize advanced FEA techniques to solve industry problems. The techniques include 3D modeling, elastic-plastic material behavior, large strain and plastic deformation, weld residual stress prediction, and FEA based fracture mechanics.

Presented on: March 24, 2010
Instructor: Francis Ku
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About the Instructor

Francis Ku holds a BS in Mechanical Engineering from University of California - Berkeley and an MS in Mechanical Engineering from San Jose State University. He is a senior finite element analyst with expertise in finite element weld residual stress simulations and a recent focus on automation in finite element analyses.

Buried Piping Integrity Initiative

On December 9, 2009, the Nuclear Energy Institute (NEI) announced the industry's unanimous approval of a voluntary policy to better manage issues related to the integrity of underground piping at nuclear power plants. The implementation schedule is aggressive and will require sites to begin work during the 1st Quarter of 2010 in order to meet the deadline.

Structural Integrity Associates, Inc. (SI) has been supporting nuclear power plants with buried piping issues since 2002. As part of an EPRI Balance of Plant Corrosion project, SI has also been developing the new industry database and interface software for managing all nuclear buried piping data - including design, maintenance and inspection information. This database will become part of the new BPWorks Version 2 risk assessment solution to be released to EPRI's Buried Piping Issues Group (BPIG) member companies later in 2010.

Meeting the requirements of the new NEI initiative by December 2010 will require a complete understanding of the activities necessary to support a meaningful risk ranking, which must be capable of differentiating the unique levels of risk of individual buried piping sections around the plant. Risk ranking identifies the magnitude, timing, locations, and tools required to characterize the condition of the buried piping - the second phase of the initiative (to be completed in 2011). To this end, SI is planning a short webinar specifically targeting the issues necessary for the planning and execution of activities necessary to insure that utilities meet their end of 2010 objectives. <br>

The webinar will be a brief (20 minute) management level overview of the issues to consider for 2010 and the project timelines involved with gathering and populating data to support the new BPWorks Version 2 risk algorithms. The webinar may be repeated or supplemented with additional webinars addressing subjects such as developing written programs that address roles and activities, risk acceptance criteria, inspection options, understanding risk threats, etc.

Presented on: January 19, 2010
Instructor: Steve Biagiotti
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About the Instructor

Steve Biagiotti, PE has a MS in Metallurgical Engineering from the Colorado School of Mines, and has 20 years of experience in corrosion control at pipeline, production, and refinery operations in the oil and gas industry. Biagiotti is a pioneer in the implementation of Integrity Management written procedures, data integration, HCA identification, code interpretation, and risk minimization practices and algorithms. Other areas of expertise: in-line inspection, direct assessment (ECDA, ICDA, SCCDA), failure and root-cause analysis, and material selection.

BWR Reactor Pressure Vessel Internals

BWR reactor pressure vessel internals began cracking at many units early in plant life. The industry responded with an industry-wide BWR Vessel and Internals Project (BWRVIP) in 1994, which addressed all BWR internals and helped avoid shutdowns by instituting and executing plans to resolve related issues. This SI Webinar discusses the history of cracking in BWR reactor pressure vessel internals, current status of this cracking, and challenges in evaluating BWR internals during outages. The presentation also deals with the impact of BWR internals on license renewal, specifically: how reactor internals can be effectively managed through existing BWRVIP inspection and evaluation guidelines.

Presented on: February 13, 2008
Instructor: Marcos Herrera
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About the Instructor

Marcos Herrera, PE holds both a MS in Engineering from the University of California, Berkeley, and a MS in Engineering Management from Santa Clara University. He has over 25 years of experience in structural mechanics evaluations of nuclear power plant components, including finite element analysis, fracture mechanics, and fatigue analysis. Herrera has also been heavily involved in development of disposition methodologies for cracking in nuclear plant components.

Code Case Evaluation of Pipe Flaws & Wall Thinning

This SI Webinar discusses two ASME Code cases of great importance to the nuclear industry: Code Case N-513-2 (Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping) and Code Case N-597-2 (Requirements for Analytical Evaluation of Pipe Wall Thinning). N-513-2 provides requirements that may be used to accept pipe and tube flaws, including through-wall flaws, in certain conditions, without repair/replacement activity for a limited time, thereby avoiding unscheduled shutdowns. N-597-2 provides requirements that may be used for analytical evaluation of certain pipes and fittings subjected to internal or external wall thinning, thereby providing an alternative to repair/replacement. The presentation covers both Code cases and their scope, why they are important, key definitions, Code case evaluation approach, inspection requirements, and status of NRC acceptance.

Presented on: December 5, 2007
Instructor: Bob McGill
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About the Instructor

Bob McGill, PE has a MS in Mechanical Engineering from the University of Illinois. He is proficient in development of dynamic thermal hydraulic computer models, applying his experience in heat transfer, thermodynamics, and fluid mechanics. McGill is knowledgeable in Fortran, True Basic, and Visual Basic programming languages, as well as in finite difference numerical methods. He is also experienced in laboratory and field data acquisition.

DLL (Distributed Ligament Length)

DLL (Distributed Ligament Length) is a tool to help engineers evaluate the significance of cracks observed in the BWR core shroud and internal piping.
DLL calculates structural margins based on Limit Load Analysis, Linear Elastic Fracture Mechanics (LEFM) Analysis, Elastic Plastic Fracture Mechanics (EPFM) Analysis and Crack Growth Analysis. The program can be used to evaluate the as-found results from shroud and internal piping inspections, or it can be used as a tool to help determine the required amount of inspection a priority.

Presented on: August 19, 2010
Instructors: Angah Miessi, Dilip Dedhia, Anitha Gubbi and Guest co-presenter from EPRI, Bob Carter
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Failure & Root-Cause Analysis

Root-cause analysis (RCA) is critical to understanding why a failure may have occurred and, most importantly, how to avoid repeat or related failures. This recent SI Webinar discusses the fundamental objectives of RCA, the important differences between failure analysis and RCA, the general approach for performing RCA, and why RCA is critical to reliable plant performance. Included are key training elements for root cause investigators and examples of common RCA methods. The Webinar also delves into why RCA requires a multi-faceted investigation of equipment failures, related personnel actions, process deficiencies, and organizational weaknesses, with insight into why many RCAs stop short of the goal. Importance of third-party reviews, such as those offered by SI, is also addressed.

Presented on: September 17, 2008
Instructor: Terry Herrmann
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About the Instructor

Terry Herrmann, PE holds a MS in Engineering Management from Syracuse University, and has 30 years of experience in design, construction, testing, and failure analysis. He is particularly expert in root-cause analysis, and has extensive experience in areas related to equipment reliability and risk, including PRA applications, systems engineering, Maintenance Rule, surveillance testing, and preventive maintenance. Herrmann is SI's lead engineer for license renewal.

Financial Risk Optimization for Run/Repair/Replace Timing

Plant component run, repair, or replace decisions are often tough calls. Plant owners need unit reliability, but are pressured by aging equipment and available repair/replacement budgets. This SI Webinar offers a different approach to financial risk optimization (FRO), using financial/decision analysis methods already employed by corporate financial planners. The presentation includes a comprehensive overview of FRO; how to determine high-risk components using failure history data and plant personnel interviews; how to estimate probability of failure v.s. time from history data and interviews; incorporation of probabilistic remaining life analysis results after component inspection; the structure of a risk-based methodology that optimizes run/repair/replace timing to minimize risk; and determination of highest-risk-driving components and how they affect run/repair/replace decisions.

Presented on: September 24, 2008
Instructor: David Mauney
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Finite Element Concrete Analysis Fundamentals & Applications

This SI Webinar provides an introduction to finite element analysis of concrete structures and components using ANSYS® and LS-DYNA® codes. Fundamentals of concrete behavior are presented as they relate to concrete material models available in both finite element codes. The Webinar includes modeling of concrete cracking, modeling reinforcement, how to use the ANSYS concrete element in non-linear analysis, new ANSYS features for modeling concrete, and an overview of the material models for concrete in LS-DYNA.

Presented on: December 3, 2008
Instructor: Shari Day
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About the Instructor

Shari Day, PE has a MS in Civil Engineering from the University of Colorado, and over 20 years of experience in civil and structural engineering, with emphasis in finite element analysis. She is an expert in stress and structural analysis, including transient dynamic analysis, elastic-plastic material modeling, non-linear behavior, and seismic analysis, with specialized experience in non-linear concrete modeling, drop-and-impact analysis, and structural stability. Day is proficient in ANSYS®, LS-DYNA®, and other finite element codes.

ANSYS is a registered trademark of ANSYS, Inc. LS-DYNA is a registered trademark of Livermore Software Technology Corp.

Fundamentals of Welding, Part 1, Overview

This webinar introduces the fundamental terminology of welding and welding metallurgy to those people working with welders and welding programs. It provides with an overview of welding processes, materials, discontinuities, welding parameters and Code considerations.

Presented on: February 9, 2010
Instructor: Luis Yepez
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Fundamentals of Welding, Part 2, Nuclear Applications

This webinar builds on the overview of welding (Fundamentals of Welding – Part 1 Overview) to introduce some of the concepts necessary to successfully implement a nuclear welding program. It focuses on the similarities and differences of welding programs related to new fabrication, repair & maintenance as well as component replacement.

Presented on: February 10, 2010
Instructor: Luis Yepez
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Generator Rotor Inspection & Life Assessment

Many generator rotor components—including rotor bores, retaining rings, rotor tooth-top dovetails, and shafts—require close inspection and life assessment. This SI Webinar covers inspection techniques, including linear phased-array (LPA) ultrasonic dovetail inspections, as well as LPA inspections of rotor shafts for under-coupling cracking initiated by fretting fatigue. The presentation also discusses how to analyze inspection data for each of the above components.

Presented on: March 6, 2008
Instructors: Larry Nottingham, Scott Rau
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About the Instructors

Larry Nottingham holds a BS in Mechanical Engineering from the University of Pittsburgh, and has over 35 years of experience in design, maintenance, and nondestructive evaluation of turbines, generators, and other power plant equipment. He also has extensive experience in development and delivery of advanced nondestructive evaluation systems and procedures for numerous power plant applications, with emphasis on turbine and generator components and high-energy piping.

Scott Rau, PE has a MS in Structural Engineering from the University of California, Los Angeles. He has extensive experience in stress analysis, including finite element analysis and modeling of structures and components with linear or non-linear behavior; dynamic system response and time history analysis; and transient thermal elastic-plastic analysis. Rau has performed remaining life evaluations of over 50 high-energy piping systems, and of similar numbers of steam turbine and generator rotors.

Heat Exchanger/Condenser Asset Management

Heat exchangers and condensers, critical components in nuclear plant cooling systems, are necessarily exposed to a variety of water chemistries and operating conditions, and heat exchanger life in particular is limited by tube leaks, cracking, and ruptures, as well as heat transfer end-of-life when tube plugging limits are reached. This new SI Webinar discusses heat exchanger and condenser life limits; mandated inspections and controls; current approaches to extend tube life such as tube plugging, retubing, sleeves, modular replacements, and mid-span staking; and such improved approaches as heat exchanger-specific tube plugging criteria, pit propagation control, financial risk-based repair/replacement decisions, corrosion monitoring, and biofilm monitoring.

Presented on: July 15, 2009
Instructor: George Licina
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About the Instructor

George Licina has a BS in Metallurgical Engineering from the University of Illinois. He has over 30 years of experience in evaluating environmental degradation of materials in power plant and other industrial environments, including all forms of corrosion and stress corrosion cracking in aqueous environments, irradiation embrittlement, and compatibility with liquid sodium. Licina is a recognized expert in microbiologically influenced corrosion.

High-Density Polyethylene Pipe Nondestructive Examination

In this SI Webinar, you will learn past industry issues with metal piping in raw water applications, the benefits of using HDPE piping in place of metal piping,
volumetric examination of HDPE butt-fusion joints, the advantages of Phased Array UT vs. other NDE methods, and SI's volumetric examination capabilities using Phased Array.

Presented on: December 16, 2009
Instructor: Caleb Frederick
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Introduction to Flow Induced Vibration

Flow Induced Vibration (FIV) phenomena arising out of fluid-structure interactions are quite common in both BWRs and PWRs. Piping systems, valves, steam generator tubes, heat exchanger tubes and other critical components encounter FIV loads during the course of their normal operations. Recent implementations of the Extended Power Upgrade (EPU) in BWRs have brought upon unique challenges in identifying FIV phenomena and finding solutions to mitigate or eliminate them. This webinar introduces basic FIV terminology, well known mechanisms, empiricisms and practical solutions, particularly useful to a novice to the subject.

Presented on: June 4, 2008
Instructor: Raju Ananth
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About the Instructor

Raju Ananth, P.E., holds a Ph. D. in Mechanical Engineering from the University of Rochester. He has over 30 years of industrial experience in solving practical problems in the areas of vibration, stress analysis, heat transfer and reliability. He was recently involved in the team effort that identified acoustic resonance in BWR steam piping systems and implemented mitigating systems. has a BS in Mechanical Engineering from the Colorado School of Mines, and is experienced in implementing and managing client Integrity Management projects. Among his achievements: he has conducted direct assessment projects for multiple clients, and served as lead engineer in all phases of Internal Corrosion Direct Assessment for a large natural gas transmission pipeline operator. Gardner is especially knowledgeable in Integrity Management Plan regulations and in direct assessment documentation and reporting.

Introduction to Nuclear Plant Services

This Webinar highlights SI's capabilities and products for the nuclear industry. SI is known for stress, fracture and materials engineering, but there are many other areas of expertise within SI's doors. A summary of SI's history and significant contributions to solving technical issues and providing solutions to industry issues will be provided. SI's diverse capabilities and offerings will be discussed, including many capabilities that might be new to some of our clients. Also, SI's organization and technical resources will be shown.

Presented on: October 21, 2009
Instructor: Marcos Herrera
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About the Instructor

Marcos Herrera, PE holds both a MS in Engineering from the University of California, Berkeley, and a MS in Engineering Management from Santa Clara University. He has over 25 years of experience in structural mechanics evaluations of nuclear power plant components, including finite element analysis, fracture mechanics, and fatigue analysis. Herrera has also been heavily involved in development of disposition methodologies for cracking in nuclear plant components.

Introduction to Phased-Array Ultrasonic Technology

Phased-array ultrasonic technology has been used only recently for nuclear NDE applications. This SI Webinar introduces basic phased-array technology, operations, and several current nuclear applications, including examination of weld overlays and dissimilar metal welds. The Webinar addresses linear phased-array probes, sound field formation, focal law inputs, and touches on future specialty phased-array examinations in the nuclear industry.

Presented on: May 27, 2009
Instructor: John Hayden
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About the Instructor

John Hayden has over 25 years of NDE Level III experience in PWR and BWR nuclear power plants. He has received advanced EPRI® training in IGSCC detection/automated ultrasonic qualification; flaw sizing/automated and manual ultrasonic qualification; weld overlays/manual ultrasonic qualification; and performance demonstration initiative/automated and manual ultrasonic qualification, in addition to EPRI NDE instructor training. He is familiar with industry codes and publications — including those of ASME, NRC, BWRVIP — and their impact on nuclear NDE.

Introduction to Stress Corrosion Cracking

This new SI Webinar discusses the fundamentals of stress corrosion cracking (SCC), exploring ductile v.s. SCC failures, SCC initiation v.s. growth, SCC specimens and testing, and SCC in LWRs.

Presented on: September 30, 2009
Instructor: Barry Gordon
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About the Instructor

Barry Gordon, PE has a MS in Metallurgy and Material Science from Carnegie Mellon University, and is also a Corrosion Specialist (NACE). Gordon has over 35 years of experience in materials corrosion behavior in nuclear power plant environments, and in fact developed and qualified hydrogen water chemistry (HWC) and co-patented zinc injection for stress corrosion cracking mitigation in BWRs. He is also an adjunct professor at the Colorado School of Mines.

Leak-Before-Break Methodologies for Reactor Coolant Systems

Leak-before-break (LBB) approaches may be used to eliminate consideration of the dynamic effects of pipe rupture in reactor coolant piping systems, justifying removal or non-installation of pipe-whip restraints and jet impingement barriers. NRC LBB criteria require demonstrating a low probability of pipe rupture, then determining maximum critical flaw size via fracture mechanics, followed by thermal-hydraulic analysis to determine leakage for one half the critical size for normal plant operation, and demonstrating that there is adequate margin between predicted leakage and plant leakage detection capability. This informative SI Webinar discusses LBB methodology, NRC requirements, and current developments, as well as LBB evaluations performed by SI - and accepted by the NRC - at several plants.

Presented on: May 6, 2009
Instructor: Art Deardorff
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About the Instructor

Art Deardorff, PE has a MS in Mechanical Engineering from the University of Arizona, and a BSME from Oregon State University. He has over 35 years of experience in design, analysis, testing, and failure assessment of structures, systems, and components. Deardorff's achievements include development of innovative approaches for assessing remaining life of nuclear and fossil power plant components, and participation in several industry programs to address thermal fatigue and leak-before-break for reactor coolant systems. He has taught ASME Code seminars worldwide.

License Renewal Part 1: Fatigue Management

License renewal has elevated the importance of a more formal fatigue monitoring program at nuclear units. This SI Webinar, the first of two parts (see License Renewal Part 2, Environmental Fatigue, below), details the license renewal process, which must demonstrate that a unit's fatigue design basis is maintained during the period of extended operation. Discussed are fatigue monitoring, fatigue considerations in design, and fatigue management in the license renewal world, including current NRC fatigue regulations, NRC's Generic Aging Lessons Learned Report, the issue of whether counting cycles is sufficient, and how to establish an optimum fatigue management program. Plant examples are included.

Presented on: October 3, 2007
Instructor: David Gerber
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About the Instructor

David Gerber, PE has a BS in Mechanical Engineering from the University of California, Davis, and a MBA from Santa Clara University. An expert in fatigue behavior of structural components, Gerber has over 30 years of experience in nuclear power plant component and systems engineering.

License Renewal Part 2: Environmental Fatigue

Environmental fatigue assessment is now part of the license renewal process at nuclear plants. This SI Webinar, the second of two parts (see License Renewal Part 1, Fatigue Management, above), provides background on the issues, including NRC studies under the Fatigue Action Plan and more-recent laboratory data; current NRC requirements for environmental fatigue evaluations for license renewal, as detailed in NRC's Generic Aging Lessons Learned Report; the current methodology for environmental fatigue assessment; and what ASME is doing to address environmental fatigue issues.

Presented on: October 10, 2007
Instructor: Gary Stevens
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Managing Aging PWR Internals

Concerns about aging degradation of PWR internals have become a high priority as nuclear units approach license renewal: renewal applications must consider the effects of aging to meet conditions defined in the NRC's Generic Aging Lessons Learned Report. In response, EPRI® MRP Reactor Internals Focus Group (RI-FG) has prepared draft inspection and evaluation guidelines for managing the effects of aging degradation in PWR internals. The guidelines, to be eventually issued under the NEI 03-08 Materials Initiative, will include developing an Aging Management Program (AMP) and inspection plan for PWR internals. SI, working closely with RI-FG, has beta-tested the draft guidelines for several units. This SI Webinar discusses management of aging PWR internals, RI-FG draft guidelines, and SI technical support for plants considering aging management reviews for license renewal, as well as those considering development of AMPs.

Presented on: April 1, 2009
Instructor: Tim Griesbach
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About the Instructor

Tim Griesbach has a BS and MS in Materials Engineering from Case Western Reserve University, and over 30 years of experience in materials behavior and structural integrity of major nuclear components. His specialty is technical consulting utilizing state-of-the-art technologies for mitigating and resolving material degradation concerns in nuclear reactor vessels, internals, piping, and other major components. Griesbach has led numerous EPRI® projects to develop tools for managing such aging effects as fatigue and reactor vessel embrittlement in nuclear pressure vessels, and he has worked with utilities, NEI, and NRC to resolve such critical industry issues as pressurized thermal shock in PWRs.

NEI 03-08 & Materials Degradation Management Programs

The nuclear industry's Materials Initiative NEI 03-08, developed in response to instances of materials degradation in nuclear plants, requires a Materials Degradation Management Program at all PWR and BWR units, in addition to INPO audit of these programs. This new SI Webinar discusses NEI 03-08, INPO audits, and the potentially costly implications for the industry. The Webinar will also describe how SI can support nuclear plants in Materials Degradation Management Program development, implementation, third-party reviews, utility training, preparation for INPO audits, and related services.

Presented on: September 2, 2009
Instructor: Tim Griesbach
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About the Instructor

Tim Griesbach has a BS and MS in Materials Engineering from Case Western Reserve University, and over 30 years of experience in materials behavior and structural integrity of major nuclear components. His specialty is technical consulting utilizing state-of-the-art technologies for mitigating and resolving material degradation concerns in nuclear reactor vessels, internals, piping, and other major components. Griesbach has led numerous EPRI® projects to develop tools for managing such aging effects as fatigue and reactor vessel embrittlement in nuclear pressure vessels, and he has worked with utilities, NEI, and NRC to resolve such critical industry issues as pressurized thermal shock in PWRs.

pc-CRACK™

In this webinar, the following topics will be discussed:
• Introduction to pc-CRACK 4.0
• Critical crack size calculation
• Remaining life calculation of a component subjected to fatigue crack growth
• Remaining life calculation of a component subjected to PWSCC
• Allowable crack size using ASME Section XI IWB-3640

Who should attend:
• Fracture mechanics analysts who deal with remaining life of components subjected to fatigue crack growth, SCC, PWSCC, and allowable crack size computations using the ASME code.

Presented on: May 6, 2010
Instructor: Dilip Dedhia
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About the Instructor

Dilip Dedhia has a PhD in Materials Science from the Oregon Graduate Center and over 30 years of experience in deterministic and probabilistic fracture mechanics analyses. Dedhia's particular expertise is in fracture mechanics methods, including high-temperature creep fatigue crack growth. He was co-developer of EPRI® BLESS™ code for headers and pipes as well as co-developer of EPRI's TULIP™ for high-temperature tubing. Dedhia also developed SI's pc-CRACK™ 4.0 for general-purpose fracture mechanics analysis.

Pressure-Temperature Limits

Reliable pressure-temperature limits must be established to provide adequate margins of safety against brittle fracture of nuclear reactor pressure vessels, particularly in the beltline region, where irradiation causes material to become more brittle. This SI Webinar provides background on pressure-temperature limits as they relate to brittle fracture of RPVs, and includes practical information on pressure-temperature limit methodologies, calculations, NRC/ASME/ASTM governing documents, and methodology approvals.

Presented on: June 27, 2007
Instructor: Gary Stevens
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Probabilistic Fracture Mechanics and its Applications to the Nuclear Industry

Deterministic fracture mechanics analysis often involves computing critical crack size or remaining life of a component subjected to cyclic or steady state stresses. Since many of the inputs needed to carry out the analysis have considerable scatter, conservative bounds are employed to estimate the critical crack size or the remaining life. The final results that are obtained using such methods may be overly conservative. Probabilistic fracture mechanics (PFM) overcomes this difficulty by considering the variables with scatter as distributed random variables. Rather than pass/fail, it provides probability of certain event occurring, for example, the probability of the critical crack size being reached. Monte Carlo simulation is the most commonly used technique for computing the probabilities.
In this webinar, the basic principles of PFM will be reviewed with examples from WinPRAISE, a probabilistic fracture mechanics software for computing probabilities of leaks and breaks in nuclear power plant cooling piping subjected to fatigue, IGSCC or PWSCC. Application of PFM to RPV subjected to an overcooling event will also be discussed.

Presented on: July 28, 2010
Instructor: Dilip Dedhia
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About the Instructor

Dilip Dedhia has a PhD in Materials Science from the Oregon Graduate Center and over 30 years of experience in deterministic and probabilistic fracture mechanics analyses. Dedhia's particular expertise is in fracture mechanics methods, including high-temperature creep fatigue crack growth. He was co-developer of EPRI® BLESS™ code for headers and pipes as well as co-developer of EPRI's TULIP™ for high-temperature tubing. Dedhia also developed SI's pc-CRACK™ 4.0 for general-purpose fracture mechanics analysis.

Recent & Pending Changes to ASME Section III Piping Codes

Changes are coming to ASME Section III Piping Design Codes, including redefinition of the elastic-plastic strain amplification factor; guidance for treatment of thermal stratification stresses; changes to stainless steel and carbon steel fatigue design curves; new stress indices; and new requirements for socket welds. This SI Webinar provides a technical description of the upcoming ASME code changes, plus latest word on the status of these changes in ASME's approval process.

Presented on: February 25, 2009
Instructor: Paul Hirschberg
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About the Instructor

Paul Hirschberg, PE holds a MS in Mechanical Engineering from the University of California, Berkeley, and a B.Eng from Cooper Union. His expertise is in piping stress analysis, vibration, thermal stratification, fatigue management, seismic analysis, and piping systems design, backed by over 30 years of experience in the design and analysis of nuclear power plant piping systems and components. Hirschberg also has extensive experience in fatigue monitoring, ASME and B31.1 Codes, vibration analysis, PWR systems, flaw evaluations, and erosion-corrosion.

Service Water System Issues

Nuclear plant service water systems are invariably large, and complicated by non-standard designs, safety-related components, varied water chemistries and materials, diverse operating conditions, wet layups, and other factors. These differences in design, materials, chemistry, and operation necessitate plant-specific approaches for inspection, degradation management, corrosion mitigation, repairs, and replacements. This comprehensive SI Webinar provides an overview of all of the above, including mandated inspections and controls; reactive and preventive approaches; service water system design parameters; implications of sizing and intermittent flow conditions; degradation types and modes; and other issues.

Presented on: April 2, 2008
Instructor: George Licina
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About the Instructor

George Licina has a BS in Metallurgical Engineering from the University of Illinois. He has over 30 years of experience in evaluating environmental degradation of materials in power plant and other industrial environments, including all forms of corrosion and stress corrosion cracking in aqueous environments, irradiation embrittlement, and compatibility with liquid sodium. Licina is a recognized expert in microbiologically influenced corrosion.

Turbine & Generator Rotor Bore Inspection & Life Assessment

Overly conservative assessments of the probability of turbine and generator rotor bore catastrophic failure can lead to unnecessary repairs or replacements. This SI Webinar discusses a balanced approach to inspection and life assessment to maximize the life of rotors, covering inspection, identification of damage mechanisms, operating condition assessment, material properties, acceptance criteria, and repair/replace/reinspect scheduling. SI's Webinar also includes information on new NDE techniques and new analysis tools for rotor bore applications.

Presented on: February 6, 2008
Instructors: Larry Nottingham, Scott Rau
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About the Instructors

Larry Nottingham holds a BS in Mechanical Engineering from the University of Pittsburgh, and has over 35 years of experience in design, maintenance, and nondestructive evaluation of turbines, generators, and other power plant equipment. He also has extensive experience in development and delivery of advanced nondestructive evaluation systems and procedures for numerous power plant applications, with emphasis on turbine and generator components and high-energy piping.

Scott Rau, PE has a MS in Structural Engineering from the University of California, Los Angeles. He has extensive experience in stress analysis, including finite element analysis and modeling of structures and components with linear or non-linear behavior; dynamic system response and time history analysis; and transient thermal elastic-plastic analysis. Rau has performed remaining life evaluations of over 50 high-energy piping systems, and of similar numbers of steam turbine and generator rotors.

Weld Overlays in the Nuclear Industry

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in BWRs since the 1980s. Overlays have also been recently applied in PWRs where primary water stress corrosion cracking (PWSCC) has developed. This SI Webinar offers a history of IGSCC in BWRs and PWSCC in PWRs; weld overlay background and status; and the industry's BWR and PWR experience with weld overlays. The presentation also covers weld overlay repair objectives, design, analysis, inspection requirements, and planning for weld overlay implementation. Licensing issues involving weld overlays at BWRs and PWRs are also discussed.

Presented on: July 9, 2008
Instructor: Marcos Herrera
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About the Instructor

Marcos Herrera, PE holds both a MS in Engineering from the University of California, Berkeley, and a MS in Engineering Management from Santa Clara University. He has over 25 years of experience in structural mechanics evaluations of nuclear power plant components, including finite element analysis, fracture mechanics, and fatigue analysis. Herrera has also been heavily involved in development of disposition methodologies for cracking in nuclear plant components.

Welding of Alloy 52M Overlays over Dissimilar Metal Welds (Part 1)

In this webinar, you will learn the successful plant implementation of Alloy 52M filler
material overlays to mitigate SCC in dissimilar metal welds has presented several welding challenges. Participants will learn about these challenges and the testing performed to develop effective solutions.

Presented on: December 2, 2009
Instructors: Dick Smith of Structural Integrity, Greg Frederick of EPRI
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