bluuebunny/crossref_metadata_2025_split huggingface dataset:
FRACTESUS Project: General Framework of Materials Selection and Testing Processes
Authors: Brynk Tomasz , Lambrecht Marlies , Uytdenhouwen Inge , Arffman Pentti , Altstadt Eberhard , Petit Tom 
Published: October 2021

Abstract
The H2020 project entitled “Fracture mechanics testing of irradiated RPV steels by means of sub-sized specimens (FRACTESUS)” started on the 1st October 2020. The aim of this project is to demonstrate the applicability of miniaturized compact tension specimens in fracture toughness testing of the reactor pressure vessel steels under hot cell conditions. Validation of this method in an industrially-relevant environment will be an important step towards achieving its acceptance by the nuclear authorities, and finally, to induce its prospective usage by the nuclear power plant operators. Successful implementation of a miniaturized specimens testing technique will result, among others, in the optimization of surveillance material usage and savings in irradiated materials testing. The general project overview showing its structure, partners involved and main deliverables was published elsewhere. Here, we focus on some technical aspects being of the utmost importance in the initial stage of the project and which will have a crucial impact on its overall progress. The general consideration on the selection of the best available materials for testing are discussed on the examples of 73W weld, A533B LUS and the WWER-440 base metal 15Kh2MFAA. Moreover, the general scheme of the testing process, which is planned within the project, is briefly presented as well as the basic assumptions about the numerical modeling task aimed for rationalizing experimental findings.
Strength Distribution Characteristics of Elbow Pipes Considering Low Cycle Fatigue Based on Analysis
Authors: Shimazu Ryuya , Sakai Michiya , Ono Yohei , Matsuura Shinichi 
Published: October 2021

Abstract
When designing and evaluating the strength of the piping system of a nuclear power plant, ductile rupture or plastic collapse is often assumed to be the limit state. However, previous studies have reported that the primary failure mode of piping systems is fatigue. Therefore, it is important to conduct a fragility assessment of the piping system under the assumption that the failure mode is fatigue. In this study, a method for calculating the strength distribution characteristics of elbow pipes based on material characteristics is proposed. The linear cumulative damage is considered as a damage assessment index to conduct a fragility assessment by assuming that fatigue occurs. The strain at the elbow pipes was calculated by finite element analysis, while the linear cumulative damage corresponding to the number of cycles was evaluated based on the strain using the Probabilistic ε-N curve. The failure probability of elbow pipes corresponding to the input acceleration was obtained under specific given parameters, and parts of the curves that compose the fragility curve are shown. The analysis results show that the 5% failure probability can be evaluated using the relationship between the failure probability and either the open or closed angle of the elbow pipes, without considering the thickness variations in the fragility assessment of elbow pipes.
Evaluation of an Alloy 52 / Cladded Carbon Steel Repair Weld by Cold Metal Transfer
Authors: Huotilainen Caitlin , Keinänen Heikki , Kuutti Juha , Nevasmaa Pekka , Sirén Henrik , Virkkunen Iikka 
Published: October 2021

Abstract
Extending the lifetime of existing nuclear power reactors is an increasingly important topic. As the existing fleet of nuclear power reactors ages and approaches the end of their design lifetimes or enters periods of lifetime extension, there is an increased probability for defect repairs due to extended exposure to the operating environment (e.g. high temperature, high pressure, corrosion environment, neutron irradiation, etc.). Concerning repair welding, should a critical need for repair arise, qualified and validated solutions must be readily available for rapid deployment. A proposed method using robotized gas metal arc welding-cold metal transfer to repair a “worst-case” scenario, linear crack like defect beneath the cladding, which extended into the reactor pressure vessel steel, was evaluated on laboratory scale in previous works (PVP2020-21233, PVP2020-21236). These previous studies demonstrated that cold metal transfer has the potential to produce high quality welds in the case of a reactor pressure repair.
In the current study, the lessons learned from the previous work were applied to repair a postulated surface crack on a thermally embrittled and cladded low alloy steel plate using a nickel base Alloy 52 filler metal. Two excavations were filled using different weld bead arrangements — a traditional pattern (92 weld beads, Q = 0.6 kJ/min) and a 45°-hatch pattern (184 weld beads, Q = 0.9 kJ/min) — by gas metal arc welding-cold metal transfer. No pre-heating or post-weld heat treatment were applied, to remain in line with what can be expected in a real pressure vessel repair situation. The 0° angle pattern acts as a reference for previous studies, while the 45°-hatch pattern, aims to minimize the residual stresses caused by repair welding. Finite element modeling was used to predict the initial (cladded, embrittled and excavated) condition of the steel plate, followed by simulating the welding using the actual welding conditions and material constants for both bead patterns as input parameters. The resulting deformation, strains and stresses created in the material due to repair welding were predicted and the welding’s effectiveness was estimated. In addition, the post-repair weld mechanical properties and microstructure, specifically focusing on the fusion boundary and heat-affected zone, were evaluated using various microscopy techniques and hardness measurements. The outcomes of the performed simulations, corresponding characterizations and lessons learned are presented in this study.
Advanced Structural Integrity Assessment Tools for Safe Long Term Operation - ATLAS+ Project: Status of the Activities of the WP3 on Modelling in 2021
Authors: Blouin Arnaud , Marie Stéphane , Nicak Tomas , Timperi Antti , Gill Peter 
Published: October 2021

Abstract
The main objective and mission of the ATLAS+ project is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems.
ATLAS+ WP3 focuses mainly on ductile tearing prediction for large defects in piping and associated components: Several approaches have been developed to accurately model the ductile tearing process and to take into account phenomena such as triaxiality effects, or the ability to predict large tearing in industrial components. These advanced models include local approach coupled models or advanced energetic approaches. Unfortunately, the application of these tools is currently rather limited to R&D expertise. However, because of the continuous progress in the performance of calculation tools and accumulated knowledge, in particular by members of the ATLAS+ consortium, these models can now be considered as relevant for application in the context of engineering assessments.
WP3 has been planned to:
• Illustrate the implementation of these models for industrial applications through the interpretation of large scale mock-ups (with cracks in weld joints for some of them),
• Make recommendations for the implementation of the advanced models in engineering assessments,
• Correct data from the conventional engineering approach by developing a methodology to produce J-Δa curve suitable case by case, based on local approach models,
• Improve the tools, guidance and procedures for undertaking leak-before-break (LBB) assessments of piping components, particularly in relation to representing structural representative fracture toughness J-Resistance curves and the influence of weld residual stresses.
To achieve these goals, WP3 is divided into 4 sub-WPs. This paper presents the progress of the work performed in each sub-WP after 48 months of activities.
Effect of Piping System Vibration (FIV, AIV, PIV) on Pipe Support Loads
Authors: Appiah Emmanuel , Wiseman Phillip 
Published: October 2021

Abstract
Integrity of a piping system is a prerequisite for personnel safety and operational reliability in industries where pipelines are critical means of transferring products from one process point to the other, such as power plants, refinery plants, and chemical industries. An essential aspect of designing a reliable piping system is to design supports of suitable load carrying capacity. This also depends on accurate determination of expected support loads including loads due to vibration of the system. Piping design codes such as ASME B31.3 and B31.1 provide a general framework but do not address vibration and its impact from a detailed perspective. In many situations, the potential impact of vibration is overlooked during support load determination. In recent piping system construction, the effect of vibration has increased due to increase in fluid flow rates and use of high strength thin wall materials. Common factors that contribute to vibration include: turbulent flow (flow induced vibration, FIV), relief valve operation (acoustic induced vibration, AIV), rotating and reciprocating equipment (pulsation induced vibrations, PIV). The effect of vibration depends on the strength of excitation and the flexibility of the piping system. As vibration of the piping system increases, loads transfer to the pipe supports also increase. Catastrophic failure of a piping system can occur if its natural frequency lock-in with the frequency of the excitation source. For holistic system integrity, the loads induced due to vibrations need to be accounted for in the support design. In this paper, we investigate the contributions of the various vibration loads in a piping system, the effect of neglecting the various vibration loads on the system integrity, and an empirical method to readily determine the vibration loads to reduce cost and time require in support design processes.
Measurement of Steam-Generator Tube Damping Caused by Anti-Vibration-Bar Supports
Authors: Smith Bruce A. W. , Feenstra Paul , Liu Michael 
Published: October 2021

Abstract
In 2013, Atomic Energy of Canada Limited (now Canadian Nuclear Laboratories) took vibration measurements to examine low-frequency damping associated with tube vibration parallel to anti-vibration bar supports. These tests were performed to better understand and control in-plane fluidelastic instability of tubes in the U-bend region of recirculating steam generators.
In the tests, the damping ratios of a single steam generator tube vibrating parallel to a single pair of anti-vibration bars were measured using both a log-decrement and a power-based method. Non-contacting excitation and position-sensing techniques were employed to improve accuracy. The tests examined the effects of fluid (water or air), natural frequency, gap width, preload, and vibration normal to the bars. Additional tests were also done using a drilled-hole support and without supports to allow validation against previously published work. Most of the tests were performed in water.
This paper describes the tests including the test apparatus, test methods, and analysis techniques. A summary of the results is presented. These results show that the damping ratios measured without any supports and with a drilled-hole support are consistent with previously published data. Contrary to an existing design guideline, the anti-vibration bars resulted in no significant additional viscous or squeeze-film damping for vibration parallel to the bars. The results also showed that anti-vibration bars can introduce significant in-plane Coulomb-type damping because of friction and impacting.
Research on Characteristics of Fluid Exciting Force of Reactor Internals
Authors: Feng Zhipeng , Deng Liwen , Huang Xuan , Shen Pingchuan , Liu Shuai , Cai Fengchun 
Published: October 2021

Abstract
Flow-induced vibration is an important issue related to the safety and reliability of nuclear reactor, which need to be analyzed and evaluated in the design stage. In order to obtain the input loads and key parameters used in the calculation of flow-induced vibration of reactor vessel internals (RVIs) that need to satisfy the engineering requirements. The typical RVIs are selected as the research object, and the fluid exciting force characteristics are studied based on the computational fluid dynamics methods. The results show that the fluid exciting force acting on the RVIs is a wide-band stochastic process. For upper internal, the largest pressure fluctuation occurs at the guide tubes and support columns located near the outlet. Therefore, it is necessary to pay more attention to these guide tubes and support columns in response analysis. As for core barrel, the root mean square value of the pressure fluctuation changes drastically at the inlet and outlet location. For lower internal, the lower flow field of RVIs is relatively disordered, and its pressure fluctuation possesses irregular characteristics. Each component of lower internal need to be considered in analysis and evaluation.
Failure Modes for Acrylic Polymers in Section VIII Pressure Vessels
Authors: Kemper Bart , Richards Guy , Nappi Taylor , Thipparthi Veda , Escobar Ana 
Published: October 2021

Abstract
Section VIII of the Boiler and Pressure Vessel Code is introducing the use of acrylics as a pressure vessel material. The design method is specified in ASME PVHO-1, Safety Standard for Pressure Vessels for Human Occupancy. The current method relies upon an empirical method developed in the 1960–70’s. It does not use “allowable stress” or other mechanical properties traditionally used to calculate design dimensions, but instead uses a fixed range of dimensions for specific shapes and determines the wall thickness using a curve. Understanding the PVHO-1 design assumptions and typical failure modes is important for a non-PVHO pressure vessel designer using acrylics. An ASME Codes & Standards task group is developing a “design by analysis” method (DBA) for acrylics and other glassy polymers for pressure vessel components. The proposed DBA methodology uses Verification and Validation (V&V) techniques and Finite Element Method (FEM) as the design method framework in order to advance the use of glassy polymers as pressure vessel materials.
A Multi-Scale Failure-Probability-and-NDE-Based Fatigue Life Model for Estimating Component Co-Reliability of Uncracked and Cracked Pipes
Authors: Fong Jeffrey T. , Marcal Pedro V. , Rainsberger Robert , Heckert N. Alan , Filliben James J. , Doctor Steven R. , Finney Ned A. 
Published: October 2021

Abstract
The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Committee has recently developed a new Section XI (Nuclear Components Inspection) Division 2 Code [1] named “Reliability and Integrity Management (RIM).” RIM incorporates a new concept known as “System-Based Code (SBC)” originally due to Asada and his colleagues [2, 3], where an integrated approach from design to service inspection is introduced using three new types of statistical quantities: (1) “system reliability index,” or “system co-reliability target” for any system consisting of structures and components, (2) “structural co-reliability,” for any structure, and (3) “component co-reliability” for any component. In a recent paper published in the International Journal of Pressure Vessels and Piping, Fong, Heckert, Filliben, and Freiman [6] developed a new theory of fatigue and creep rupture life modeling for metal alloys at room and elevated temperatures such that the co-reliability of an uncracked component can be estimated from fatigue and creep rupture test data with simple loading histories. In this paper, we extend the theory to include a methodology to estimate the co-reliability of a cracked pipe from fatigue crack growth rate test data, probability of detection (POD) data, and nondestructive evaluation (NDE) of initial crack sizing data for simple loading histories. To illustrate an application of this new modeling approach, we present four numerical example case studies using (a) the fatigue failure data of six AISI 4340 steel specimens at room temperature (Dowling, N. E., 1973) for an uncracked steel pipe, and (b) the fatigue crack growth rate data of 17 specimens of 2024-T3 aluminum (von Euw, Hertzberg, and Roberts, 1972) for a cracked aluminum pipe (see discussion after Eq. (16)). The four cases are: (1) Uncracked and uninspected pipe. (2) Inspected pipe with a crack-found-location-and-size call. (3) Uncracked and inspected pipe with a no-crack-found call and a POD value. (4) Inspected pipe with a crack-found-location-and-size call and a structural health monitoring (SHM) program. Significance and limitations of this new fatigue life modeling approach to the estimation of component co-reliability of uncracked and cracked pipes are presented and discussed.
A Case for New Low Pressure Vessel (LPV) Codes for Design Pressures Below 15 psi (100 kPA)
Authors: Millet Barry , Ebrahimi Kaveh , Lu James , Spencer Donald 
Published: October 2021

Abstract
Today the ASME Boiler and Pressure Vessel Code Section VIII (ASME Code) covers pressure vessels for design pressure above 15 psi (100 kPa) but not for design pressures below 15 psi.
Manufacturers of pressure vessels under 10ft (3048mm) and with design pressure under 15 psi (100 kPa) design to the ASME Code but do not stamp them. The ASME Code is explicit in not allowing this.
Manufacturers of low pressure vessels over 10 ft (3048 mm) in diameter, design and built to “good engineering practice” using Finite Element Analysis, the ASME Code, API and the AISC Manual of Steel Construction.
This paper provides an overview of these existing codes, standards with their methods for design, fabrication and testing then provides an outline of a Code with two classes of low pressure vessels (LPV). The audience for the smaller pressure vessels would be small batch chemical, pharmaceutical, food, and beverage processing facilities who require small near atmospheric pressure vessels. The audience for the larger pressure vessels would be power plants, refineries, chemical plant, steel mills and concrete plants flue gas treatment and CO2 sequestration of exhaust products.
Influence of Finite-Length Plate Proximity on Vortex-Induced Vibrations of Elastic Cable
Authors: Ivanov Oleg , Vedeneev Vasily 
Published: October 2021

Abstract
Vortex-induced vibrations of an elastic cylinder near a finite-length plate are experimentally investigated. A rubber cylinder of diameter D was spanned in the test section of a wind tunnel near a plate of length nearly 6D. The oscillation amplitude peak for a single cylinder was 0.3D, Reynolds number for oscillating peak-amplitude regime based on the cylinder diameter was in the range 180...260. The oscillation amplitude ratio A/D appeared to decrease with the reduction of the gap ratio G/D between the cylinder surface and the plate for the cylinder located sufficiently upstream from the plate trailing edge. However, for the cylinder location at the same level or downstream from the plate trailing edge, there are regions of essentially larger oscillation amplitude compared to the isolated cylinder case. Maximum amplitude increase by 39% was obtained.
Delayed Hydride Cracking Initiation Evaluation of Interacting Flaws in a CANDU Reactor Zr-Nb Pressure Tube
Authors: Scarth Douglas , Doddihal Preeti , Ip Monique 
Published: October 2021

Abstract
In-service evaluation of a flaw that is detected in a Zr-2.5Nb alloy pressure tube in a CANada Deuterium Uranium (CANDU) nuclear reactor includes an analysis to demonstrate that delayed hydride cracking will not initiate from the flaw. The flaw can be assessed either as planar or volumetric with a finite root radius. Two adjacent flaws caused by fretting due to flow-induced vibration were detected by in-service ultrasonic examination in a Zr-2.5Nb alloy pressure tube in a CANDU reactor. The flaws were replicated to obtain details of flaw geometry. The two flaws are volumetric in nature, have irregular shapes, and were determined to intersect. Although both flaws are volumetric, for pragmatic reasons one flaw was treated as planar. The other flaw was treated as blunt. The flaws were evaluated for delayed hydride cracking initiation in accordance with the Canadian Standards Association Standard N285.8. A three-dimensional finite element stress analysis was performed to quantify the interaction between the two flaws, as well as the peak stress at the root of the blunt flaw. Both flaws met the acceptance criteria for prevention of delayed hydride cracking initiation, and the pressure tube was returned to service. This paper provides a summary of the delayed hydride cracking initiation evaluation of these two flaws, and highlights the methods that need to be applied for flaws with irregular shapes and that interact.
Optimization for Nuclear Class 1 Opening Reinforcement
Authors: Reinhardt Wolf 
Published: October 2021

Abstract
An important element in evolving Section III of the ASME Code is the re-examination of Code rules to identify the potential for efficiencies. The present paper looks at the rules for reinforcement of openings. The present Code rules require the area of material removed to create the opening (up to the design thickness) to be added around the opening with certain limits on the distance from the opening. Past studies have suggested that the present rules may add more material than needed to maintain the strength of the vessel in some cases.
The optimum amount of reinforcement is therefore postulated to be determined by the criterion that the limit pressure should not be reduced excessively relative to the limit pressure of the vessel in the absence of the opening.
Limit analysis is performed to derive possible rules on the amount and distribution of around an opening. Options for alternative rules for reinforcement of openings, and restrictions on the present Code rules, are proposed.