News & Views, Volume 49 | Inspection Optimization- Probabilistic Fracture Mechanics

News & Views, Volume 49 | Inspection Optimization: Probabilistic Fracture Mechanics

By:  Scott Chesworth (SI) and Bob Grizzi (EPRI)

News & Views, Volume 49 | Inspection Optimization- Probabilistic Fracture Mechanics

The goal was to determine whether the frequency of current inspection requirements was justified or could be optimized (i.e., increase the interval of certain inspections to devote more attention to higher-value inspections and thereby maximize overall plant safety).

Executive Summary
Welds and similar components in nuclear power plants are subjected to periodic examination under ASME Code, Section XI.  Typically, examinations are performed during every ten-year inspection interval using volumetric examination techniques, or a combination of volumetric and surface examination techniques.  Nuclear plants worldwide have performed numerous such inspections over plant history with few service induced flaws identified.

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News & Views, Volume 49 - PEGASUS- Advanced Tool for Assessing Pellet-Cladding Interaction

News & Views, Volume 49 | PEGASUS: Advanced Tool for Assessing Pellet-Cladding Interaction

By:  Bill Lyon, PE and Michael Kennard

News & Views, Volume 49 - PEGASUS- Advanced Tool for Assessing Pellet-Cladding Interaction

PEGASUS provides a fully capable computational environment to solve the unique, detailed 3D analyses required for the evaluation of PCI.

In the current economic environment in which nuclear units compete with less costly energy sources, a quicker return to full power correlates to more power generated and increased operating efficiency.  This may be achieved with shorter startup post-refueling or a quicker return-to-power following any number of plant evolutions including load follow, control blade repositioning, equipment outage or maintenance, testing, extended low power operation, scram, etc.  Such strategies to increase operating efficiency may enhance the risk of pellet-cladding interaction (PCI), a failure mechanism that occurs under conditions of high local cladding stress in conjunction with the presence of aggressive chemical fission product species present at the cladding inner surface.  These conditions can occur during rapid and extensive local power changes and can be further enhanced by the presence of fuel pellet defects (e.g., missing pellet surface, MPS).  Several commercial reactor fuel failure events in the last eight years, as recently as early 2019, suggest a PCI-type failure cause.  To safely manage changes in core operation, the margin to conditions leading to PCI-type failures must be determined prior to implementation of such operating changes.

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News & Views, Volume 49 | Autobook- Nuclear Physics Automation Code

News & Views, Volume 49 | Autobook: Nuclear Physics Automation Code

News & Views, Volume 49 | Autobook- Nuclear Physics Automation CodeBy:  Sasan Etemadi, P.E. and Mark Drucker, P.E.

The AUTOBOOK code reduces human errors, increases efficiency, and streamlines the reload analysis process

AUTOBOOK facilitates plant operation by providing nuclear power plant Reactor Engineers and Reactor Operators with cycle-specific information about the physics characteristics of the reactor core in a core data book document. Structural Integrity has created the AUTOBOOK computer code to automate the creation of this document.

AUTOBOOK is a Quality Assured code developed under a licensee’s software quality assurance (SQA) program. SI provides a full complement of SQA documents, including a Software Requirement Specification (SRS), a Software Design Description (SDD), Verification and Validation (V&V) Plan and Test Report, a User Manual, and Software Installation Instructions (SII).

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Structural Integrity Associates Achieves Milestone with Pegasus Code Development

Structural Integrity Associates Achieves Milestone with Pegasus Code Development

Structural Integrity Associates | Structural Integrity Associates Achieves Milestone with Pegasus Code Development

Structural Integrity Associates | Structural Integrity Associates Achieves Milestone with Pegasus Code DevelopmentOn October 28th, the Structural Integrity (SI) Nuclear Fuel Technology Team achieved a major milestone in completing the first Verification & Validation phase in the development of its nuclear fuel performance and behavior code Pegasus©.  “This is a significant step by the SI Team” commented Vick Nazareth, SI Fuel Director.  “We have been developing Pegasus© since 2017 to incorporate cutting edge computational technology and four decades of fuel behavior modeling and analysis expertise into a software program”.  The code addresses a need for deeper fuel integrity insights within the nuclear industry to achieve next level fuel performance and licensing.  Dr. Joe Rashid, Scientist and Senior Technology Developer of the code added “this code will analyze fuel behavior through the entire fuel cycle from initial startup to used-fuel storage”.

SI announced the development of Pegasus© in the SI newsletter in 2019, Introducing Pegasus: State-of-the-Art Nuclear Fuel Behavior with the objective of enhancing the fidelity of fuel behavior and performance in support of advanced fuel technologies.  The Pegasus© code will go through additional validation testing over the next several months to meet a production roll-out in early 2021 in support of fuel performance behavior analysis across a broad spectrum of light water reactor and advanced reactor fuel designs.

” I am proud of the SI Fuel Team”, said Mark Marano, SI CEO.” This milestone exemplifies our ability to provide innovative structural integrity solutions for clients across structures, systems, components, water chemistry and nuclear fuel.”

Structural Integrity is an employee-owned specia­­lty engineering and services company providing innovative engineering solutions and services to achieve asset management excellence across multiple industries including Nuclear, Fossil, Oil & Gas, Renewables, and Critical Infrastructure.

News & View, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

News & Views, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

By:  Dick Mattson and Minghao QinNews & View, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

A U.S. BWR utility contracted with Structural Integrity (SI) to review their current reinspection guidance documents relative to those contained in the BWRVIP inspection guidelines, the purpose of which was two-fold:

  1. ­Are current reinspection guidelines compliant with industry requirements?
  2. ­Are there components where reinspection intervals could possibly be extended?

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News & View, Volume 48 | Plant Materials Aging and Degradation

News & Views, Volume 48 | Plant Materials Aging and Degradation – Nuclear IGSCC Mitigation Optimization and Equipment Advances

By:  Erica Libra-Sharkey

INDUSTRY CHALLENGE

News & View, Volume 48 | Plant Materials Aging and Degradation

From the US Department of Energy, Office of Nuclear Energy, “The demanding environments of an operating nuclear reactor may impact the ability of a broad range of materials to perform their intended function over extended service periods. Routine surveillance and repair/replacement activities can mitigate the impact of this degradation; however, failures still occur. With reactors being licensed to operate for periods up to 60 years, with further extensions under consideration, and power uprates being planned, many of the plant systems, structures, and components will be expected to tolerate more demanding environments for longer periods. The longer plant operating lifetimes may increase the susceptibility of different systems, structures, and components to degradation and may introduce new degradation modes.

While all components potentially can be replaced, decisions to simply replace components may not be practical or the most economically favorable option. Therefore, understanding, controlling, and mitigating materials degradation processes and establishing a sound technical basis for long-range planning of necessary replacements are key priorities for extended nuclear power plant operations and power uprate considerations. https://www.energy.gov/ne/materials-aging-and-degradation.

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News & View, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

News & Views, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

By:  Bill WeitzeNews & View, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

100% of thermal stress was treated as nonlinear gradient stress and linear bending stress was about 12% of the moment stress. Structural Integrity’s (SI’s) review of the stress terms used in piping analysis show that pressure stress does create bending stress in components…

EPRI Report 3002014121 “Development of Fatigue Usage Life and Gradient Factors” has developed fatigue usage adjustment factors that account for: 1) increased cyclic life associated with the growth of potential engineering size fatigue cracks in thicker components (thickness factor, TF; also called life factor, LF), and 2) the presence of through-thickness stress gradients (gradient factor, GF). (TF is used in the issued Code Case.)  These factors are applied to cumulative usage factor, U, in air.

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News & View, Volume 48 | Examination Optimization for PWR and BWR Components

News & Views, Volume 48 | Examination Optimization for PWR and BWR Components

By:  Scott Chesworth, Bob Grizzi, and Dilip Dedhia

Optimizing the inspection interval for high-reliability components whose examinations have a significant outage impact.News & View, Volume 48 | Examination Optimization for PWR and BWR Components

Welds and similar components in nuclear power plants are subject to periodic examination under ASME Code, Section XI.  Typically, examinations are performed during every ten-year inspection interval using volumetric examination techniques, or a combination of volumetric and surface examination techniques.  Nuclear plants worldwide have performed numerous such inspections over the plant history with few service induced flaws identified.  Since personnel health and safety, radiation exposure, and overall outage costs associated with these inspections can be significant, Structural Integrity (SI) was contracted by the Electric Power Research Institute (EPRI) to review the technical bases for the inspection intervals for select components.  The goal was to determine whether the frequency of current inspection requirements was justified or could be optimized (i.e., reduced in order to devote more attention to higher-value inspections and thereby maximize overall plant safety).  Special priority was given to components demonstrating an exceptional history of reliability and whose examinations have a significant outage impact.

News & Views, Volume 48 | Environmentally-Assisted Fatigue Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

News & Views, Volume 48 | Environmentally-Assisted Fatigue – Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

By: Dave Gerber and Terry HerrmannNews & Views, Volume 48 | Environmentally-Assisted Fatigue Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

Environmentally-Assisted Fatigue (EAF) screening is used to systematically identify limiting locations for managing EAF effects on Class 1 reactor coolant pressure boundary components wetted by primary coolant.  This article provides an overview of the methods developed and used by Structural Integrity (SI) for Class 1 components having explicit fatigue analyses performed using ANSI/ASME B31.7(1) and ASME Section III(2).  A future article will discuss how this is performed for Class 1 piping designed and analyzed to ASME/ANSI B31.1(3).

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News & View, Volume 48 | SI-FatiguePro Version 4.0 Crack Growth Module Application Case Study; Complex Multi-Cycle Nuclear Transients

News & Views, Volume 48 | SI:FatiguePro Version 4.0 Crack Growth Module – Application Case Study Complex Multi-Cycle Nuclear Transients

News & View, Volume 48 | SI-FatiguePro Version 4.0 Crack Growth Module Application Case Study; Complex Multi-Cycle Nuclear Transients

By: Curt Carney

As plants enter their initial or subsequent license renewal period one of the requirements is to show that fatigue (including environmental effects) is adequately managed.  For some locations in pressurized water reactors (PWRs), it can be difficult to demonstrate an environmental fatigue usage factor less than the code allowable value of 1.0.  Therefore, plants are increasingly turning to flaw tolerance evaluations using the rules of the ASME Code, Section XI, Appendix L.  Appendix L analytically determines an inspection interval based on the time it takes for a postulated flaw (axial or circumferential) to grow to the allowable flaw size.  For surge line locations, this evaluation can be very complex, as the crack growth assessment must consider many loadings, such as: insurge/outsurge effects, thermal stratification in the horizontal section of the line, thermal expansion of the piping (including anchor movements), and internal pressure.  Trying to envelope all of these loads using traditional tools can lead to excess conservatism in the evaluation, and short inspection intervals that reduce the value of an Appendix L evaluation.