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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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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 | Baffle-Former Bolt Management- Cost:Benefit Studies

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

By:  Tim Griesbach and News & View, Volume 46 | Baffle-Former Bolt Management- Cost:Benefit Studies

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). 

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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

By:  Bill Lyon

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

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News & View, Volume 46 | Delivering Value- Modernization of Plant Automation Controls

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

By: Gerry Davina

News & View, Volume 46 | Delivering Value- Modernization of Plant Automation Controls

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.

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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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News & View, Volume 46 | Acoustic and Blowdown Load Calculations for Reactor Internals

News & Views, Volume 46 | Acoustic and Blowdown Load Calculations for Reactor Internals

By: Matthew Walter

News & View, Volume 46 | Acoustic and Blowdown Load Calculations for Reactor InternalsAs 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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