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 | 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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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 | 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.

News & View, Volume 47 | TRU Compliance Expands into Physical Security | How To Make Knowing A Good Thing - Thinning Handbooks

News & View, Volume 47 | How To Make Knowing A Good Thing: Thinning Handbooks

By:  Stephen Parker and Eric HoustonNews & View, Volume 47 | TRU Compliance Expands into Physical Security | How To Make Knowing A Good Thing - Thinning Handbooks

SI has developed a process to mitigate the negative outcomes of piping examination.  One part of that process is Thinning Handbooks, which have resulted in direct savings in excess of $10 Million for one nuclear plant.

Examination of Safety Related Service Water piping is driven by a number of factors, all of which tend to converge on the objective of finding localized thinning prior to the thinning becoming a problem.  In other words, examinations are performed to eliminate the risk of a leak and ensure that the wall thickness remains greater than tmin (the minimum required uniform wall thickness).  However, the rules, regulations, and economic realities mean that only bad things happen from an exam regardless of what is found.

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News & View, Volume 46 | Application of Probabilistic Flaw Tolerance Evaluation Optimizing NDE Inspection Requirements

News & Views, Volume 46 | Application of Probabilistic Flaw Tolerance Evaluation Optimizing NDE Inspection Requirements

By:  Christopher Lohse

News & View, Volume 46 | Application of Probabilistic Flaw Tolerance Evaluation Optimizing NDE Inspection RequirementsThere have been several industry initiatives to support optimization of examination requirements for various items/components (both Class 1 and Class 2 components) in lieu of the requirements in the ASME Code, Section XI.  The ultimate objective of these initiatives is to optimize the examination requirements (through examination frequency reduction, examination scope reduction, or both) while maintaining safe and reliable plant operation.  There are various examples of examination optimization for both boiling water reactors (BWRs) and pressurized water reactors (PWRs).  Each of these technical bases for examination optimization relies on a combination of items.  The prior technical bases have relied on: (1) operating experience and prior examination results as well as (2) some form of deterministic and/or probabilistic fracture mechanics.   For BWRs, the two main technical bases that are used are BWRVIP-05 and BWRVIP-108.  These technical bases provide the justification for scope reduction for RPV circumferential welds, nozzle-to-shell welds, and nozzle inner radius sections.  For PWRs, the main technical basis for RPV welds is WCAP-16168.  These technical bases are for the RPV welds of BWRs and PWRs which represent just a small subset of the examinations required by the ASME Code, Section XI.  Therefore, the industry is evaluating whether technical bases can be optimized for other components requiring examinations. 

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News & View, Volume 46 | Assessing Prestress Losses in a Nuclear Containment Structure for License Renewal

News & Views, Volume 46 | Assessing Prestress Losses in a Nuclear Containment Structure for License Renewal

By: Eric Kjolsing

News & View, Volume 46 | Assessing Prestress Losses in a Nuclear Containment Structure for License RenewalNuclear power plants around the world are approaching the end of their original 40-year design life.  Efforts are underway to extend the operating license for these plants to 60 years or beyond.  As part of the license extension, it must be demonstrated that the reactor containment building remains able to safely perform its intended functions for the extended duration of operation.  Many of these containment buildings utilize a post-tensioned concrete design where the tendons are grouted after tensioning.  Since these grouted tendons cannot be re-tensioned, an assessment for the loss in prestress beyond the original design life must be performed.

This article describes a methodology to assess the structural performance of a containment structure over time as a function of confidence in the tendon losses and is split into three parts:

  1. A description of the methodology
  2. A representative probabilistic assessment
  3. Representative analysis results

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