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).
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.
https://www.structint.com/wp-content/uploads/2021/06/News-View-Volume-48-SI-FatiguePro-Version-4.0-Crack-Growth-Module-Application-Case-Study-Complex-Multi-Cycle-Nuclear-Transients.jpg363668Structural Integrityhttps://www.structint.com/wp-content/uploads/2023/05/logo-name-4-930x191-1.pngStructural Integrity2020-09-30 18:00:082021-07-28 18:27:44News & Views, Volume 48 | SI:FatiguePro Version 4.0 Crack Growth Module – Application Case Study Complex Multi-Cycle Nuclear Transients
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.
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.
https://www.structint.com/wp-content/uploads/2021/07/News-View-Volume-47-TRU-Compliance-Expands-into-Physical-Security-How-To-Make-Knowing-A-Good-Thing-Thinning-Handbooks.jpg363668Structural Integrityhttps://www.structint.com/wp-content/uploads/2023/05/logo-name-4-930x191-1.pngStructural Integrity2020-03-03 16:45:142021-07-28 18:27:44News & View, Volume 47 | How To Make Knowing A Good Thing: Thinning Handbooks
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.
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).
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 Universityof 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).
https://www.structint.com/wp-content/uploads/2021/07/News-View-Volume-46-Evaluation-of-Reconfiguration-and-Damage-of-BWR-Spent-Fuel-During-Storage-and-Transportation-Accidents.jpg363668Structural Integrityhttps://www.structint.com/wp-content/uploads/2023/05/logo-name-4-930x191-1.pngStructural Integrity2019-07-01 15:57:032021-07-21 16:00:18News & Views, Volume 46 | Evaluation of Reconfiguration and Damage of BWR Spent Fuel During Storage and Transportation Accidents
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.
Nuclear 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:
As part of the general design criteria for nuclear power plants, the primary structures and systems of the plant must be designed to handle postulated accident events, including the dynamic effects of postulated pipe ruptures. For a Boiling Water Reactor, analyzed events include various accident conditions in the recirculation piping, including a Loss of Coolant Accident (LOCA). One postulated LOCA event is assumed to be an instantaneous double-ended guillotine break of the recirculation line. This event causes several loads to be imparted on the reactor vessel, attached piping, and reactor internal components. [Some loads such as jet impingement, annulus pressurization, and pipe whip impart loads on the outside of the reactor vessel and the attached piping.][ Other loads, including flow-induced drag and acoustic loads, transmit loads inside the vessel on critical components such as jet pumps, core shroud, and the shroud support structure.] Figure 1 shows the pipe and resulting loads.
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