bluuebunny/crossref_metadata_2025_split huggingface dataset:
Identification of Two-Phase Flow Patterns Using Support Vector Classification
Authors: Benito Ines , Mureithi Njuki W. 
Published: October 2017

Two-phase flows are preponderant in industrial components. The present work deals with external two-phase flows across tube banks commonly found in heat exchangers, boilers and steam generators. The flows are generally highly complex and remain theoretically intractable in most cases. The two-phase flow patterns provide a convenient albeit qualitative means for describing and classifying two phase flows. The flow patterns are also closely correlated to fluid-structure interaction dynamics and thus provide a practically useful basis for the study of two-phase flow-induced vibrations.
For internal two-phase flows, maps by Taitel et al. (1980) and others have led to detailed and well defined maps. For transverse flows in tube bundles, there is significantly less agreement on the flow patterns and governing parameters. The complexity of flow in tube arrays is an obvious challenge. A second difficulty is the definition of distinct flow patterns and the identification of parameters uniquely identifying the flow patterns.
The present work addresses the problem of two-phase flow pattern identification in tube arrays. Flow measurements using optical as well as flow visualization via high-speed videos and photography have been conducted. To identify the flow patterns, an artificial intelligence machine learning approach was taken. Pattern classification was achieved by designing a support vector machine (SVM) classifier. The SVM achieves quantitative and non-subjective classification by mapping the flow patterns in a high dimensional mathematical space in which the different flow patterns have unique characteristics.
Details of the flow measurement, parameter definition and SVM design are presented in the paper. Flow patterns identified using the SVM are presented and compared with previously identified flow patterns.
Image Analysis of Indications in Fillet Welding by Phased Array Ultrasonic Technique
Authors: Zhang Youwei , Yang Yuqing , Li Qingfeng , Chen Le 
Published: October 2017

The steel frame for carrying boiler in power generation station is usually connected by fillet welding like corner or tee joints. Traditionally, we inspect this type of welding by Magnetic Particle Testing (MT) and Dye Penetration Testing (PT), by which only the surface or near surface defects can be detected, and the internal defects in butt welding is inspected by Ultrasonic Testing (UT) and Radiographic Testing (RT).
Though ultrasonic Testing (UT) has various advantage in flat butt weld, it requires a better condition on the detection surface avoiding the interference of being rough, irregular in shape, very small or thin, or not homogeneous. Due to the geometric shape or structural constrains in fillet welding, volumetric defects are sometimes unable to be detected by RT. Currently, there are no effective methods of Non-destructive examination (NDE) for the volumetric flaws in this type of welding which connects the steel column and reinforcing plate.
Normally, the volumetric and plannar defects of fillet welding are commonly as: incomplete penetration, slag inclusion, porosity, lack of fusion and cracks, which have a great influence on the welding quality.
This paper aims at using the advantages of Phased Array Ultrasonic Testing (PAUT) to detect the fillet weld:
• Focal depth is variable.
• Scanning range is large, steering beam from −89° to 89°.
• Covering large volume at single position of probe.
By setting different types of artificial flaws, we carried out the reliability analysis of this kind of welding with PAUT.
Through observing the representative images of flaws, we can directly know the location of defects in the weld. According to the data provided by the scan software, we can read its depth and distance from the weld center line, which helps us to locate defects in weld precisely.
Inertia Mass Damper and its Application
Authors: Sunakoda Katsuaki , Yamazaki Issei 
Published: October 2017

From early times adding auxiliary mass to the main mass has been done to mitigate vibration events. And much research of the structure by adding an equivalent mass using fluid or functional fluid has been done.
On the other hand, the research and development of a new damper using rotating inertia mass began in the early 1970s in Japan. The new type damper was termed the “Mechanical Snubber” when it was used for piping and equipment systems in nuclear power plants. Tens of thousands of Mechanical Snubbers are used in Japanese domestic light water reactor and also in foreign countries. As one of the only surviving developers, the author would like to report the development process.
This Mechanical Snubber has a large equivalent inertia mass and it accords with the design criteria of high stiffness of seismic method of nuclear power plant. In recent years, in the field of civil engineering and construction, studies using rotating inertia mass or negative stiffness of mechanism have come into favor. These studies are expected to be applied in structure and bridge engineering.
This paper describes the historical background of inertia mass dampers, the theory of the inertia mass damper (I.M.D.) applied as a product, and the electromagnetic inertia mass damper (E.I.M.D.) developed as a passive and/or semi-active damper. Some experimental studies using shaking table in the National Center for Research on Earthquake Engineering (NCREE) in Taiwan and theoretical studies are introduced.
Study on the Relationship Between Interaction Factors and Stress Intensity Factor for Elliptical Flaws
Authors: Azuma Kisaburo , Li Yinsheng , Hasegawa Kunio 
Published: October 2017

The interaction of multiple flaws in close proximity to one another may increase the stress intensity factor of the flaw in structures and components. This interaction effect is not distributed uniformly along the crack front. For instance, the strongest interaction is generally observed at the point closest to a neighboring flaw. For this reason, the closest point could show a higher value of the stress intensity factor than all other points in some cases, even if the original value at the point of the single flaw is relatively low.
To clarify the condition when the closest point shows the maximum stress intensity factor, we investigated the interaction of two similar elliptical flaws in an infinite model subjected to remote tension loading. The stress intensity factor of the elliptical flaws was obtained by performing finite element analysis of a linear elastic solid. The results indicated that the interaction factors along the crack front can be expressed by a simple empirical formula. Finally, we show the relationship between geometrical features of the flaw and the stress intensity factor at the closest point to a neighboring flaw.
Thermal Aspects of Safety Analysis for Shipment of West Valley Melter
Authors: Laurinat James E. , Kesterson Matthew R. , England Jeffery L. , Ketusky Edward T. , McKeel Charles A. , Hensel Steve J. 
Published: October 2017

The thermal aspects of a safety analysis for shipment of the West Valley melter are presented. The West Valley melter was used from 1996 to 2002 to vitrify regionally sourced high level radioactive waste. The U.S. Department of Energy (DOE) set up the West Valley Demonstration Project to encase this melter and grout it in low density cellular concrete, for disposal. DOE-West Valley requested the Savannah River National Laboratory to prepare a Safety Analysis Report.
The thermal portion of the safety analysis covers Normal Conditions of Transport (NCT) and Hypothetical Accidents Conditions (HAC), as defined in the Code of Federal Regulations. For NCT, it is assumed that the encased melter is stored in either shade or direct sunlight at an ambient temperature of 311 K (100 °F). The defining HAC is exposure to a 1075 K (1475 °F) fire for 30 minutes. Finite element computer models were used to compute temperature profiles for NCT and HAC, given the thermal properties of the melter and its contents and tabulated radiolytic heating source concentrations. The resulting temperature conditions were used to estimate the pressurization due to evaporation of water from the concrete. The maximum calculated gauge pressures were determined to be 81 kPa (12 psig) for NCT and 580 kPa (84 psig) for HAC.
Advanced Applications for HDPE Pipe With New PE-RT Material
Authors: Long Wes 
Published: October 2017

Canfor, a producer of lumber, pulp and paper needed a solution to replace aging 30-inch (760mm) fiberglass reinforced pipe. A new PE-RT product now expands PE into industrial applications requiring resistance to high temperatures and having a Hydrostatic Design Basis (HDB) of 800psi (55 bar) at 180°F (82.2°C).
Through chemical processes, Canfor cooks, washes, and extracts pulp fiber from wood that results in both acidic and caustic effluent with temperatures normally in the 50–60°C range or as high as 70–75°C. Traditional fiberglass pipes have experienced repeated joint failures over time, whereas heat-fused HDPE pipe provides solutions reducing unnecessary maintenance and a longer service life. Standard PE4710 High Density Polyethylene Pipes (HDPE) have pressure ratings limited to 140°F (60°C) and are not normally acceptable for such high temperature acidic and caustic effluent. Additionally, the potential for higher oxidation-reduction potential (ORP) from residual chlorine levels and bleaching also justified turning to a different material based on the potential oxidative attack at high temperatures.
The new PE-RT resin protects against oxidative attacks at high temperatures and the flexible heat-fused HDPE pipe provides considerable cost savings during installation. Compared to fiberglass, up to eight 40-foot lengths of HDPE pipe can be joined by heat fusion per day, whereas only two 6-meter (20-foot) lengths of FRP pipe can be wrapped per day.
The presentation will highlight photos during the installation process and report the advantages of using the new pipe material. This project provides reference for expanding HDPE pipe into new applications using PE-RT materials.
Dynamic Material Test and Analysis for Rupture Study for Pressure Vessel Exposed to Fire in Plant
Authors: Kawai Tetsuya , Mitarai Yasuhiro , Waki Yoshiyuki , Yamabe-Mitarai Yoko , Kimura Kazuhiro , Kasuya Kazuyuki 
Published: October 2017

In case of fire occurring in an Oil and Gas facility, pressurized vessels may be exposed to fire. Though the entire system will be depressurized once a fire is detected, vessels may rupture, leading to risk of flammable, toxic or cryogenic fluid being released into atmosphere. Therefore, pressure vessels should be designed to withstand internal pressure without rupture during exposure to fire, at least until the system pressure can be decreased to a safe level. A pressure vessel rupture study should be conducted in addition to design code calculation in order to ensure a safe design in case of fire. As part of the recent trend for safer plant design, demand for pressure vessel rupture studies is growing and becoming a necessary requirement.
In our previous presentation (PVP2015-45260 [1]), the material data for carbon steel (SA-516 Gr.70) and stainless steel (SA240 type304 and type304L) at high temperature range were obtained through material testing and were presented as our study result. And in the other presentation (PVP2016-63184 [6]) that we’ve made, procedure for pressure vessel rupture study by FEM using the above mentioned material data was developed.
For the present research, material testing in a dynamic condition wherein a more similar condition to an actual fire case were performed and comparison between the test results and FEM analysis was done. In conclusion, recommendation for the application of the pressure vessel rupture study was justified and necessity for further development of the above mentioned study was determined.
FAVOR Version 16.1: A Computer Code for Fracture Mechanics Analyses of Nuclear Reactor Pressure Vessels
Authors: Bass B. Richard , Williams Paul T. , Dickson Terry L. , Klasky Hilda B. 
Published: October 2017

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life.
Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a).
This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.
LEAPOR: A Computer Code for Leakage-Rate Calculations for Cracks in Cooling Water Piping Systems
Authors: Williams Paul T. , Bass B. Richard , Dickson Terry L. , Klasky Hilda B. 
Published: October 2017

This paper describes the development of a new computer code called Leak Analysis of Piping - Oak Ridge (LEAPOR) which calculates estimates for the leakage rate of water escaping from postulated through-wall cracks in a piping segment of a nuclear power plant cooling water system. The ability of nuclear power plant control and safety systems to detect a piping leak prior to breakage is a fundamental requirement of the leak-before-break concept. The design and assessment of leak-detection systems, therefore, requires the determination of through-wall crack leakage rates covering a significant range of operating and flow conditions. For the primary use case of pressurized water reactors, the coolant is subcooled liquid-phase water at high pressures and temperatures, and the leakage flow regimes can range from adiabatic flow boiling (“flashing”) with non-equilibrium vapor generation inside the crack to orifice flow of a subcooled liquid with vapor generation occurring outside of the pipe. The thermohydraulic Henry-Fauske model (with extensions) for non-equilibrium flashing flow through “tight cracks” has been implemented into LEAPOR.
A primary driver in the development of LEAPOR has been that its Software Quality Assurance (SQA) requirements included evaluations for correctness, consistency, completeness, accuracy, source code readability, and testability. The new code should be prepared to successfully meet the criteria of formal SQA audits. The attributes of maintainability, portability, and extensibility also informed LEAPOR’s layered software architectural design.
The paper presents the results of verification and validation studies carried out with LEAPOR where verification by benchmark comparisons to the results of an independently developed leak rate code and validation against experimental data are described.
Development of a Profile Matching Criteria to Model Dents in Pipelines Using Finite Element Analysis
Authors: Woo Janine , Kainat Muntaseer , Adeeb Samer 
Published: October 2017

Current industry standards cite depth and interaction with additional stress risers as the key indicators of pipeline integrity concerns in regards to dents. There have been significant efforts towards the improvement of these benchmarks in recent years. Several dent assessment methods are presented in literature, including research focused on the use of finite element analysis (FEA). The accurate assessment of dents using FEA is heavily reliant on how close the shape produced by the FEA model aligns with the shape of the actual dent. The research presented in this paper has been conducted to evaluate the sensitivity of the stresses and strains to the dent profile shape.
Information regarding the existence, shape, and size of dents is typically provided by in-line inspection (ILI) tools. An FEA model is then built in commercially available software, ABAQUS, to create a dent profile that closely resembles the profile given by the ILI. The study in this paper assesses the effect of different indenter sizes on the stresses and strains within the dent and provides a recommendation to quantify the error between the ILI and FEA profiles. The process of matching a dent profile using FEA is compared to an existing analytical method to calculate strain, the equations proposed in ASME B31.8. The FEA results were found to be more conservative than the strains calculated using ASME B31.8.
Seismic Qualification of Piping Systems by Detailed Inelastic Response Analysis: Part 3 — Variation in Elastic-Plastic Analysis Results on Carbon Steel Pipes From the Benchmark Analyses and the Parametric Analysis
Authors: Nakamura Izumi , Otani Akihito , Morishita Masaki , Shiratori Masaki , Watakabe Tomoyoshi , Shibutani Tadahiro 
Published: October 2017

It is recognized that piping systems used in nuclear power plants have a significant amount of the safety margin, up to the point of boundary failure, even when the input seismic load exceeds the allowable design level. The reason is attributed to the large strength capacity of the piping systems in the plastic region. In order to establish an evaluation procedure, in which the inelastic behavior of piping systems is considered in a rational way, a task group activity under the Japan Society of Mechanical Engineers (JSME) has been conducted. As a deliverable of this activity, a Code Case in the framework of the JSME Nuclear Codes and Standards is now being developed. The Code Case provides the strain-based criteria, an evaluation procedure using the response-spectrum based inelastic analysis, and detailed inelastic response analysis based on a finite element model.
For developing the Code Case, inelastic benchmark and parametric analyses of the tests of a pipe element and piping system made of carbon steel were conducted to investigate the variation of the elastic-plastic analyses results. Based on these analytical results, it is assumed that setting the yield stress has a significant influence on the inelastic analytical results, while the work hardening modulus in the bi-linear approximation of the stress-strain curve has little influence. From the results of the parametric analyses, it is confirmed that the variation in the analytical results among the analysts would be reduced by having a unifying analysis procedure. In this paper, the results of the parametric analyses and the variation in the elastic-plastic analysis are discussed.
Effect of Postweld Heat Treatment Conditions on Mechanical Properties of 9Cr-1Mo-V Steel Welds for Pressure Vessel
Authors: Tanaka Tomohiro , Abe Masamitsu , Nakatani Mitsuyoshi , Terasaki Hidenori 
Published: October 2017

2¼Cr-1Mo steel and 2¼Cr-1Mo-V steel which have high performance in high-temperature and high-hydrogen service are widely used in fabricating process equipment for the oil refining industry. With the demand for plants using low-grade crude oil expected to increase in the future, equipment will after to operate at higher temperatures and higher pressures to meet such demand. In addition, the demand for gasification process equipment for coal using high-temperature strength material, which is to be used as an alternative of oil, is expected to grow also.
Although 2¼Cr-1Mo steel and 2¼Cr-1Mo-V steel are adopted for high-temperature and high-pressure equipment used at temperatures up to 450°C, 9Cr-1Mo-V steel is suited for operation at even higher temperatures and pressures. It is used as structural material for boilers of power plants and elsewhere. The application of 9Cr-1Mo-V steel is anticipated for oil refineries. However, it is necessary for the welds to consider the hardness and the low-temperature toughness in addition to the high-temperature strength.
In this paper, the effect of postweld heat treatment (PWHT) conditions on mechanical properties of 9Cr-1Mo-V steel welds was investigated. The PWHT conditions are organized with reference to the tempering parameter (Larson-Miller Parameter), with the heat treatment conducted in an LMP range of 20.46 to 22.42 × 103. As LMP increased, the hardness and high-temperature strength decreased. In the hardness distribution of the welds subjected to PWHT, the hardness tended to maximize near the bond of the heat affected zone (HAZ); however, applying heat treatment to materials with an LMP equal to or higher than 21.15 × 103 allowed a hardness of 260HV or less to be obtained. In terms of the high-temperature strength, the welds met the specified value as the 0.2 proof stress of ASME SA387-91-2, the base metal, under the PWHT conditions of an LMP equal to 22.42 × 103; the tensile strength was 0.9 times or more of the specified value. The Charpy absorbed energy increased with the LMP, and, when an LMP of 21.99 × 103 or more was used as the PWHT conditions, an absorbed energy of 60 J or more could be obtained at 0°C, as well. From the above results, the effect of PWHT conditions on mechanical properties of the welds is clarified, and it is suggested that 9Cr-1Mo-V steel can be used for the equipment in oil refining plants.