Nuclear Energy Archives

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

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

September 30, 2020/in Bill Weitze, Fracture Mechanics, News and Views, Nuclear Asset Integrity, Nuclear Energy/by Structural Integrity

By: Bill Weitze

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.

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

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

September 30, 2020/in Bob Grizzi, Dilip Dedhia, News and Views, Nuclear Asset Integrity, Nuclear Energy, Scott Chesworth/by Structural Integrity

By: Scott Chesworth, Bob Grizzi, and Dilip Dedhia

Optimizing the inspection interval for high-reliability components whose examinations have a significant outage impact.

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

September 30, 2020/in Dave Gerber, News and Views, Nuclear Asset Integrity, Nuclear Energy, Terry Herrmann/by Structural Integrity

By: Dave Gerber and Terry Herrmann

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⁾.

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

September 30, 2020/in Curt Carney, News and Views, Nuclear Asset Integrity, Nuclear Energy, SI:FatiguePro/by Structural Integrity

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 & Views, Volume 48 | Optimizing Cathodic Protection Commitments Aging Management Program (AMP XI.M42)

September 30, 2020/in Mark Jaeger, Nuclear Energy, Shane McManus/by Structural Integrity

By: Shane McManus and Mark Jaeger

License renewal applications (LRAs) often involve commitments to future actions. These can be classified into one of three categories: appropriate, overcommitment, and ambiguous implementation. Appropriate commitments include those actions that are expected by the NRC (such as those explicitly identified in the GALL⁽¹⁾and GALL-SLR⁽²⁾) as well as some less restrictive actions that are technically justified by engineering evaluation. These commitments can generally be implemented within one operating cycle using existing technology, are cost-effective, and are consistent with the GALL and GALL-SLR.

Overcommitments and those commitments with ambiguous implementations can be avoided and cost-effectiveness optimized by obtaining independent third party reviews (ITPR) of the LRA.

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

March 3, 2020/in News and Views, Nuclear Asset Integrity, Nuclear Energy, Stephen Parker/by Structural Integrity

By: Stephen Parker and Eric Houston

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 t_min (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.

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

July 1, 2019/in Chris Lohse, News and Views, Non-Destructive Evaluation, Nuclear Asset Integrity, Nuclear Energy, Reactor/by Structural Integrity

By: Christopher Lohse

There 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.

News & Views, Volume 46 | Baffle-Former Bolt Management: Cost/Benefit Studies

July 1, 2019/in Chris Lohse, News and Views, Nuclear Energy, Tim Griesbach/by Structural Integrity

By: Tim Griesbach and

For the past several years baffle-former bolt (BFB) cracking in pressurized water reactors has become a significant concern for of PWR plants. In 2016, three similar Westinghouse designed plants (Indian Point 2, Salem 1, and D. C. Cook Unit 2) experienced significant numbers of cracked BFBs, attributed to irradiation-assisted stress corrosion cracking (IASCC). These plants had common characteristics that included the 4-loop plant design, downflow configuration, and Type 347 stainless steel bolting material. BFB cracking is not an entirely new phenomenon as it was initially detected in the French PWR fleet in the 1990s. However, the extent of cracking found in some of the US plants has greatly exceeded prior cracking. Extensive industry programs have identified and categorized by tier group the most susceptible plants, and the EPRI Materials Research Program (MRP) has published guidance regarding baffle-former bolt UT inspections for PWR plants for detection of degraded and cracked bolts in the baffle-former assembly (MRP-2017-009).

News & View, Volume 46 | Evaluation of Reconfiguration and Damage of BWR Spent Fuel During Storage and Transportation Accidents

News & Views, Volume 46 | Evaluation of Reconfiguration and Damage of BWR Spent Fuel During Storage and Transportation Accidents

July 1, 2019/in Bill Lyon, News and Views, Nuclear Energy, Nuclear Fuel/by Structural Integrity

By: Bill Lyon

Structural Integrity Associates is participating in a Department of Energy (DOE) Integrated Research Projects (IRP) program focused on storage and transportation of used nuclear fuel (UNF). The project, entitled Cask Mis-Loads Evaluation Techniques, was awarded to a university-based research team in 2016 under the DOE Nuclear Fuels Storage and Transportation (NFST) project. The team is led by the University of Houston (U of H) and includes representatives from the University of Illinois at Urbana-Champaign, the University of Southern California, the University of Minnesota, Pacific Northwest National Laboratory, and staff members from the Nuclear Fuel Technology and Critical Structures and Facilities divisions of SI. The primary objectives of NFST are to 1) implement interim storage, 2) improve integration of storage into an overall waste management system, and 3) prepare for large-scale transportation of UNF and high-level waste. The goal of the cask mis-load project is to develop a probabilistically informed methodology, utilizing innovative non-destructive evaluation (NDE) techniques, determining the extent of potential damage or degradation of internal components of UNF canisters/casks during normal conditions of transport (NCT) and hypothetical accident conditions (HAC).

News & Views, Volume 46 | Delivering Value: Modernization of Plant Automation Controls

July 1, 2019/in Automation Controls, Chemistry, Gerry Davina, News and Views, Nuclear Energy/by Structural Integrity

By: Gerry Davina

The modernization of plant automation controls represents a step change in performance that optimizes Operations and Maintenance resources, shifting their focus to performance maintenance and plant monitoring and away from inefficient corrective maintenance and troubleshooting.

According to the U.S. Energy Information Administration, the average age of the U.S.-based nuclear power plant is approximately 38 years old. Three of the “youngest” plants (Watts Bar, Nine Mile Point 2 and River Bend) all began construction in the mid-1970’s with their designs approved years earlier. In terms of industrial control systems, this means that most, if not all, of the plants in the U.S. nuclear fleet, continue to operate with 1970s in automation equipment and technology. Although it can be argued that the equipment and technology have proven to stand the test of time, the reality of the digital age, with low cost and high-powered processors, is that relay-based control systems are long-obsolete and no longer practical for any automation system that requires more than a handful of relays and switches. In an industry that has publicly advocated a commitment to improved reliability and efficiency, ironically, the most evident impact for any plant with the continued use of 40-year-old automation equipment and technology is poor system reliability and inefficiency burdening both Operations and Maintenance resources.