bluuebunny/crossref_metadata_2025 huggingface dataset:
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.
Proactive In-Situ Welding Stress Control for Laser Repair Welding of Irradiated Austenitic Materials
Authors: Chen Jian , Tatman Jonathan , Chen Zongyao , Feng Zhili , Frederick Greg
Published: October 2017

Substantial research has been performed in recent years to determine the effects and feasibility of welding on highly irradiated austenitic materials. This research has been driven by the need to preemptively develop welding techniques capable of repairing highly irradiated light water reactor (LWR) components susceptible to detrimental corrosion and cracking. However, the materials used to fabricate internal LWR components become increasingly difficult to weld with in-service age due to irradiation-induced generation of helium in the material matrix over time. This paper introduces a patent-pending technology that proactively manages the stresses during laser repair welding of highly irradiated reactor internals to avoid the occurrence of intergranular helium-induced cracking. The technology development relied on numerical simulations that made it possible to refine and optimize the innovative welding concept and to identify specific process conditions achieving significant reduction of tensile stress (or even formation of compressive stress) near the weld pool in the heat-affected zone on cooling. The candidate welding process conditions identified by the numerical simulations were experimentally tested on stainless steel plates (Type 304L) with a laser welding system purposely designed and engineered to incorporate the proactive stress management concept. In-situ temperature and strain measurement technique based on digital image correlation were applied to validate the numerical simulations.
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.
High Pressure Tubing Material Performance Review
Authors: Warren Kraig D.
Published: October 2017

The purpose of the study was to find an alternative supply of material for high pressure tubing used in the manufacture of plastic in Low Density Polyethylene (LDPE) plants. The material selected by A&A Machine & Fabrication, LLC was HS220-27Ca4, which is manufactured by TimkenSteel. The HS220-27Ca4 material is a fine-grain, forged rolled alloy with optimal chemistry of medium carbon and a balance of chromium, nickel and molybdenum to achieve ultra-high strength and toughness. The material was laboratory tested to fatigue standards defined by ASME Section VIII Division 3 Article KD-3. In addition, the material was processed by A&A Machine & Fabrication, LLC to autofrettage requirements set forth in ASME Section VIII Division 3 Article KD-5. To satisfy industry standards and customer specifications, dimensional and non-destructive test data was gathered throughout the fabrication process. This information and comparative chemistry data is presented and reviewed as a means to demonstrate acceptance of HS220-27Ca4 for use in LDPE process.
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.
Probabilistic Fracture Mechanics Analysis of Boiling Water Reactor Vessel on Relatively Low Failure Probability Problem Using PROFAS-RV PFM Analysis Code
Authors: Kim Jongmin , Lee Bongsang
Published: October 2017

The PFM approach has been widely used to evaluate the integrity of reactor pressure vessel (RPV) in nuclear power plant. Since the 1980s, a number of probabilistic fracture mechanics (PFM) analysis codes have been developed to perform the probabilistic analysis for RPV, and these codes are continuously updated by reflecting recent irradiation shift model, database of fracture toughness and compendia of stress intensity factors. The author developed a PFM analysis program for RPV, PROFAS-RV (PRObabilistic Failure Analysis System for Reactor Vessel), recently, which can evaluate failure probability of RPV using recent RTNDT shift model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx A16 code as well as required basic functions of PFM program. In this paper, the failure probabilities of boiling water reactor (BWR) for cool-down and low temperature over pressurization (LTOP) transient are calculated by using the own PFM analysis code, PROFAS-RV. This work was conducted as part of an international collaborative study. The effects of key parameters such as transient, fluence level, Cu and Ni content, initial RTNDT and RTNDT shift model on the failure probability are systematically compared and reviewed. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial RTNDT. However, the effect of Cu and Ni content is negligible for the very low fluence of 0.02×1019 n/cm2 because there is no additional irradiation embrittlement. The effect of initial RTNDT on the failure probability is more significant for the lower fluence region in both transients. The failure probability of LTOP transient is lower than that of cool-down transient, and the probability of failure with irradiation shift model of 10CFR50.61a is larger than that of R.G.-1.99 rev. 2 at the fluence ranges 0.2×1019 n/cm2 to 0.5×1019 n/cm2.
Expression of a Generic Full-Range True Stress-True Strain Model for Pipeline Steels Using the Product-Log (Omega) Function
Authors: Ndubuaku Onyekachi , Martens Michael , Cheng J. J. Roger , Adeeb Samer
Published: October 2017

Steel pipelines are subjected to a variety of complex, and sometimes difficult to predict, loading schemes during the fabrication, installation and operation phases of their lifecycles. Consequently, the mechanical behavior of steel pipelines is not only influenced by the steel grade but also by the loading history of the pipe segments. Due to the resultant intricacies of the nonlinear load-deformation behavior of pipelines, adequate numerical analysis techniques are usually required for simulation of pipelines under different loading schemes. The validity of such numerical simulations is largely influenced by the accuracy of the true stress-true strain characterization of the pipeline steels. However, existing stress-strain mathematical expressions, developed for the characterization of metallic materials over the full-range of the stress-strain relationship, have been observed to either loose predictive accuracy beyond a limited strain range or, for the more accurate full-range models, are cumbersome due to their requirement of a large number of constituent parameters. This paper presents a relatively accurate and simple true stress-true strain model which is capable of accurately predicting the stress-strain behavior of pipeline steels over the full range of strains. The proposed stress-strain model is characteristically unlike existing stress-strain models as it is essentially defined by a Product-Log function using two proposed parameters, and is capable of capturing a reasonable approximation of the yield plateau in the stress-strain curve. To validate the proposed model, curve-fitting techniques are employed for comparison to experimental data of the stress-strain behavior of different pipeline steel grades (X52 – X100). Excellent agreements are observed between the proposed model and the different pipeline steels over the full-range of the true stress-true strain relationship. Furthermore, the applicability of the proposed model is validated by means of a proposed parametric procedure for predicting the ultimate compressive strength of shell elements.
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.
Hydrogen Embrittlement Susceptibility of Clad Steel Pipes
Authors: Jemblie Lise , Olden Vigdis , Nyhus Bård , Akselsen Odd Magne
Published: October 2017

The objective of the present work has been to evaluate the hydrogen embrittlement susceptibility of 316L - X60/X65 carbon steel hot rolled bonded clad pipes through experimental investigations and cohesive zone numerical simulations. Fracture mechanical testing in air and under in situ electrochemical hydrogen charging revealed little hydrogen influence on the samples with a Ni-interlayer between the clad and the base material, while significant hydrogen influence on the sample without a Ni-interlayer. Cohseive zone simulations were able to predict the detrimental influence of hydrogen on the fracture initiation toughness. The simulations suggest that hydrogen trapped at dislocations are mainly responsible for the embrittlement of the clad pipe base material.
A Study on the Ball Indentation Test for Linear Hardening Metals
Authors: Zhang Tairui , Wang Weiqiang , Li Aiju
Published: October 2017

In this work, a numerical analysis based on the linear hardening model and ball indentation test was proposed to determine the properties of materials such as austenitic stainless steels that obey linear hardening behavior. Experiments on 304 and 321 austenitic stainless steel showed that the numerical method provides yield and tensile strength values with a maximum error of less than 5% and the hardening behavior identical with uniaxial tensile tests, while the average error for yield strength was more than 30% with the empirical method. The reasons why the empirical method produces larger errors for non-power law materials were analyzed in detail. The elastic modulus calculation method was also re-investigated, and the model based on re-loading showed much bigger equivalent contact diameters and provided more accurate elastic modulus values in a simple way.
The Effects of Low Temperature Carbon Diffusion Treated Fasteners on Thread Galling Resistance
Authors: Clark Karson P.
Published: October 2017

Kolsterising® is a low temperature carbon diffusion treatment, referred to as carburization, in which carbon is forced into the surface of metal. This results in a significant increase of surface hardness without the formation of crystalline solids or chemical elements.1 Carburization is not a coating process therefore the possibility of delamination is eliminated with the process. 2 Low temperature carbon diffusion treatment is being considered by a chemical company to possibly aid in the disassembly of higher alloy fasteners in attempts to reduce maintenance and equipment overhaul time during plant outages. This paper will explore the effects surface carburization, or carbon diffused surface hardening, has on the resistance of thread galling for B8M Class 1, B8M Class 2, and Hastelloy® C-276 fasteners.3 With the data provided from this initial study, a better understanding of the benefits low temperature carbon diffusion treated fasteners provides will be obtained. This paper has a limited scope that will utilize hand torqueing at various percentages of bolt yield to analyze the resultant clamping force of carburized fasteners in comparison to that of standard fasteners. This study is a milestone to further clarify the thread galling resistance that carburized fasteners provides.
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.
A Novel Metamodeling Approach for Time-Temperature Parameter Models
Authors: Haque Mohammad Shafinul , Ramirez Christopher , Stewart Calvin M.
Published: October 2017

The classic approach to the prediction of stress-rupture involves using time-temperature parameters. Since 1952 with the introduction of the Larson-Miller parameter, many different time-temperature parameters have been introduced. There have been several attempts to generalize the many time-temperature models into a single “metamodel”. A common approach is to use a pre-defined stress-parameter function that is force fit to the submodels leading to biased predictions. In this study, a novel metamodeling approach is applied to derive a single metamodel that incorporates twelve time-temperature parameters (eight existing and four new time-temperature parameter models are exploited). This metamodel allows the instantaneous and efficient evaluation of all twelve submodels simultaneously. The metamodel and twelve submodels are evaluated against four isotherms of alloy P91 stress-rupture data in the form of time-temperature parameterization isostress lines, points of convergence, and master curves for each model. A statistical analysis of the uncertainty of stress-rupture predictions with respect to the experimental uncertainty of the experimental database is performed. A key characteristic of time-temperature parameters is the regression analysis of stress versus the calculated parameter. An analysis of stress-parameter functions is performed to determine the optimal function that produces physically realistic predictions. Finally, a guideline to the selection of the most physically realistic time-temperature parameter and stress-parameter function for a given material is presented.
Resilience of Masonry Systems in Nuclear Power Plants Under Blast Risk
Authors: Campidelli Manuel , El-Dakhakhni Wael W. , Tait Michael J. , Mekky Waleed
Published: October 2017

The resilience of the built environment to high explosives poses a significant challenge to the professionals tasked with the design of blast resistant facilities. Current standards — including the ASCE 59-11 and CSA S850-12 — fail to address this challenge in design provisions targeting a single parameter of structural performance, while neglecting other key indicators of performance recovery that define the very concept of resilience. In order to investigate their significance in the design process, two resilience parameters known as robustness and rapidity are evaluated for an archetype blast scenario — a nuclear power plant (NPP) featuring reinforced concrete block masonry walls exposed to a blast hazard, namely, the detonation of an explosive charge within an open (outdoor) area of the industrial complex. The adopted methodology integrates resilience–based analysis and probabilistic risk assessment, in order to account for the uncertainties associated with threat (attack likelihood); hazard (attacker’s success likelihood); load input variables — including location, mass, and type of explosive; resistance variables — including material properties and wall geometry; and loss variables — including the costs of repair and replacement. Based on the current analysis, recommendations are made to incorporate resilience metrics in standards for blast protection, so as to foster more resilient industrial facilities.
Multisensor Degradation Data Fusion and Remaining Life Prediction
Authors: Wang Changxi , Elsayed E. A. , Li Kang , Cabrera Javier
Published: October 2017

Multiple sensors are commonly used for degradation monitoring. Since different sensors may be sensitive at different stages of the degradation process and each sensor data contain only partial information of the degraded unit, data fusion approaches that integrate degradation data from multiple sensors can effectively improve degradation modeling and life prediction accuracy. We present a non-parametric approach that assigns weights to each sensor based on dynamic clustering of the sensors observations. A case study that involves a fatigue-crack-growth dataset is implemented in order evaluate the prognostic performance of the unit. Results show that the fused path obtained with the proposed approach outperforms any individual sensor data and other paths obtained with an adaptive threshold clustering algorithm in terms of life prediction accuracy.
Application of SDS Method to Predict Fracture Toughness Temperature Dependency of A533B Steel
Authors: Meshii Toshiyuki , Ishihara Kenichi , Nakano Hiroki
Published: October 2017

Recently, the authors have proposed a new method for scaling the crack tip stress distribution under small scale yielding condition and named it as T-scaling method [1, 2]. This method identifies the different stress loads for materials with different tensile properties but identical in terms of K or J. Then by accepting the knowledge “fracture stress for slip induced cleavage fracture is temperature independent [3],” a framework to predict the fracture load Pc and fracture toughness Jc at an arbitrary temperature from the already known Pcr and Jcr at a reference temperature Tr was proposed and validated for 0.55% carbon steel JIS S55C [1, 2]. This framework was named as SDS method. This paper presents that the SDS method was valid to predict 1TCT Jc temperature dependency of A533B steel [4] in the range of −9 ≤ T-T0 ≤ 27 °C, where T0 is the master curve reference temperature. The SDS method seems to have a possibility to solve the problem the master curve is facing in the relatively higher temperature region, by requiring only tensile tests.
Screening Criteria of Fracture Assessment Methods for Pipes Having a Circumferential Surface Flaw
Authors: Machida Hideo , Kato Takehiko
Published: October 2017

This paper describes examinations of the screening criteria of fracture assessment methods for pipes having a circumferential surface flaw. The failure mode screening criteria for ferritic steel pipes are defined in Appendix E-11 of the Rules on Fitness-for-service for Nuclear Power Plants [1] (hereinafter, the FFS Codes) of the Japan Society of Mechanical Engineers (JSME) and ARTICLE C-4000 of ASME Section XI [2] (hereinafter, ASME Sec. XI). However, these screening criteria were assessed under limited conditions; they are not intended for detailed evaluation. Essentially, the fracture mode depends on a stress-strain curve, a J-R curve, flaw shape (length and depth), pipe shape (R/t), etc. First, among these parameters, these that have a large effect on fracture strength evaluation were selected. The important parameters for fracture strength are a stress-strain curve, a J-R curve, flaw depth. Next, failure modes according to these parameters were classified. Finally, a screening method to determine whether the failure mode is Limit Load (LL) or Elastic-Plastic Fracture Mechanics (EPFM) for all pipe materials was proposed. The proposed method is a general method not limiting the applicable pipe materials.
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.
Benchmarking of Explicit Dynamic Finite Element Analysis for Storage and Transportation Casks Under Impact Loading Conditions
Authors: Yoshida Shinji , Okamoto Toshiki , Takagi Shunsuke
Published: October 2017

Impact analyses for storage and transportation cask drop tests were performed using the explicit dynamic finite element analysis code to demonstrate analytical accuracy and benchmark the analytical code. The test series for this benchmarking were provided by Sandia National Laboratories (SNL) as the Structural Evaluation Unit Benchmark Problem Statement. 30 foot and 120 foot drop test cases in which the test unit was dropped from 30 feet and 120 feet onto an essentially rigid target were selected as the benchmarking problems. A comparison of deformation and acceleration between the actual cask drop tests and those predicted by explicit dynamic finite element analysis performed using LS-DYNA was performed in this paper. The simulation results showed reasonably good agreement with respect to deformation and acceleration between the calculated and the measured result by adjusting the material properties of the impact limiter to match the amount of crush.
Validation of the New Post-RCCM Option From Code_Aster Through Benchmark Comparisons With Other Industrial Codes
Authors: Métais Thomas , Plessis Sarah , Miralles Julien
Published: October 2017

Fatigue is identified as a significant degradation mode that affects nuclear power plants world-wide. The various international codes and standards (ASME, RCC-M, JSME, etc...) offer rules to predict its damaging effect on the locations of the various components of an NPP. These rules, which ensure conservatism and safe operation, have grown in complexity over the years because they have integrated R&D results showing aggravating effects (such as PWR environment degradation) that were not explicitly included in the original analyses (such as Environmental Assisted Fatigue [1] [2]) but also because an economically viable design of components has required optimization and refinement of mechanical assessment methods.

EDF has been developing since 1989 its own in-house FEA code baptised code_aster that is included in the Salome-Meca mechanical package. Salome-Meca is open-access and can be used freely by international users. It is continuously improved with a release at a rate of one new software version per year and it integrates the most recent results obtained by the EDF R&D, in fields as broad as fracture mechanics, XFEM and fatigue.

The fatigue post-processing in code_aster offers a span of criteria (Dang Van, Stress Intensity, etc...) to pick and choose from and even offers the possibility to make up owns one fatigue criteria. It also offers the possibility to post-process fatigue according to the RCC-M rules: this is implemented in code_aster through the so-called POST_RCCM operator.

Recent developments have enabled EDF to update the POST_RCCM operator in order to bridge the gap with modern fatigue industrial codes. The update effort has included in particular the development of a new option to evaluate fatigue for nozzles according to the RCC-M Annex ZE-200 as well as the clarification of the methodologies used for combining time points and transients. In addition, POST_RCCM includes the possibility to perform calculations with integration of environmental effects, with significant freedom to design one’s own corrections. The option is even left open to integrate a correction per the methodology proposed in [2]. The overall work scope and progress as of early 2016 was reported in [3] and is now in its validation phase.

The validation of a fatigue computer code is not an easy task as the full validation using hand calculations would be a highly tedious effort, given the technicality and the multiple choices to make along the various steps of the fatigue analysis. As a result, an accepted way of validating the results obtained with a given code is to perform a benchmark analysis against another industrial fatigue code. This strategy was adopted for the validation of the new POST_RCCM functionalities and this paper presents the work performed and the results obtained.
Shakedown Limits for Hillside Nozzles in Cylindrical Vessels
Authors: Bakry Ahmed K. , Saleh Chahinaz A. , Megahed Mohammad M.
Published: October 2017

This research paper is concerned with the mechanical behavior of the cylindrical vessels with hillside nozzles when subjected to both pressure and nozzle bending loads in cyclic forms. The influence of hillside angle on shakedown (SD) limits of the connection under cyclic pressure and combined steady pressure with cyclic nozzle bending is investigated. A shell FEA model is built for the assembly using five different hillside angles ranging from 0° to 40°. Shakedown limits are determined by a direct technique known as the Nonlinear Superposition Method (NSM). Bree diagrams for cyclic out of plane opening (OPO) / in plane (IP) nozzle moments combined with steady internal pressure are determined. The results show an increase in both OPO and IP shakedown moments as the hillside angle is increased. In addition, the OPO shakedown limit moments for all hillside angles was found to be insensitive to the level of internal pressure in contrary to IP shakedown moment which starts to reduce with pressure for the high pressure range.
Evaluation of Fatigue Crack Propagation Behavior Crossing Interface in Cladded Plates
Authors: Nagai Masaki , Murai Kiminori , Nagashima Toshio , Miura Naoki
Published: October 2017

In nuclear power plants, inner surface of reactor pressure vessel, which is made of low-alloy steel, is cladded by austenitic stainless steel to improve corrosion resistance of vessel. In an evaluation of fatigue crack propagation behavior at J-welding portions of bottom mounted instrumentation in reactor pressure vessel, crack propagation crossing dissimilar materials interface between base metal of low-alloy steel and stainless steel cladding is assumed. Fatigue crack growth rate of the cladding is predicted to be captured by that of the base metal because thickness of the cladding is very thin compared that of the base metal. In this study, fatigue crack propagation analyses of cladded plates under cyclic loading were performed using XFEM, which can model crack independently of finite elements, as a preliminary study for crack propagation analyses considering dissimilar materials interface.
Numerical Experiment on Two-Phase Flow Behaviors in Tube Bundle Geometry for Different Mixtures
Authors: Benguigui William , Deri Enrico , Lavieville Jerome , Mimouni Stephane , Longatte Elisabeth
Published: October 2017

Since steam generators are used, many experiments have been performed to understand the different phenomena appearing in them. One of the main issues to reproduce a similar flow on a reduced-scale experiment is the choice of the two-phase mixture. Air/water and boiling freon are among the most used two-phase flow mixtures. In the present document, a finite-volume CFD code dedicated to multi-phase flows based on a two-fluid approach (extended to n) is used to compare two mixtures numerically. Thanks to two experiments, an air/water horizontal channel and a freon/freon inclined tube bundle, a first part highlights the validity of the two-phase flow modeling. Then, based on inlet superficial velocities from both experiments, a numerical experiment is performed by using each mixture on the other experiment. Void fraction profiles and two-phase flow regimes are compared in order to exhibit the behaviors of both mixtures in a steam generator like tube bundle. Results turn out to provide useful information (void fraction profiles mainly) about mixture properties, and near-wall loads in the vicinity of cylinders are investigated.
Core Seismic Experiment and Analysis of a Large Number of Element Models for Fast Reactor
Authors: Iwasaki Akihisa , Matsubara Shinichiro , Yamamoto Tomohiko , Kitamura Seiji , Okamura Shigeki
Published: October 2017

To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Generally, the fast reactor core is made of several hundred core elements in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement, vertical displacement (rising) and impact force of each core element may cause a trouble for control rod insertability, reactivity insertion and core element intensity. Therefore, a seismic analysis method of a fast reactor core considering three-dimensional nonlinear behavior, such as bouncing, impact, fluid-structure interaction, etc. was developed. Validation of the core element vibration analysis code in three dimensions (REVIAN-3D) was conducted by 1/1.5 scale 32 core element mock-ups one-row test and 1/2.5 scale 313 core element mock-ups hexagonal-matrix test. In this verification, the applicability of the result obtained on a single model test or a small number of scale tests is verified when the number of core components increases. The vertical behavior (rising displacement) and horizontal behavior (Impact force, horizontal response) as a single core element of the analysis result agreed very well with the experiments.