bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Residual Stress Profiles for Defect Tolerance Assessments of Thick Section Electron Beam Welds
Authors: Hurrell P. R. , Pellereau B. M. E. , Gill C. M. , Kingston E. , Smith D. , Bouchard P. J. 
Published: November 2014

This paper describes the results of weld model analysis and deep hole-drilling measurements undertaken to evaluate residual stress distributions in austenitic and ferritic steel thick section electron beam welds. The work was undertaken in support of a Rolls-Royce and TWI development programme in the UK, for a Reduced Pressure Electron Beam (RPEB, 0.1 to 1mbar) welding process using a mobile local vacuum seal for the manufacture of thick section pressure vessel and pipe welds for nuclear power plant applications. Measurements were undertaken on representative mock-ups including a 160mm thick SA508-3 forging circumferential seam weld, in both the as-welded and furnace post weld heat treated condition. A 316L stainless steel plate butt weld and a 304L pipe girth weld of 80mm and 36mm thickness respectively were also analysed.
There is now considered to be sufficient understanding of the residual stress fields generated by thick Electron Beam (EB) welds, to propose through thickness ‘upper bound’ R6 Level 2 stress profiles for use in defect tolerance assessments. The intention is to incorporate residual stress profiles of this type into the R6 structural integrity assessment procedure, following peer review. This would represent a significant step forward in demonstrating technology readiness for plant applications. It is also anticipated that an ASME Code Case will be drafted and proposed for the RPEB welding process.
EB welding is a relatively low heat input process, compared with a multi-pass arc weld, such that the fusion zone and heat affected zone are narrow. The centre of an EB weld is the last region to solidify and cool-down, so typically there is a high degree of restraint to weld metal contraction, thereby generating a highly tri-axial yield magnitude tensile stress state at the mid-thickness location. The stress components acting in the longitudinal welding direction and through-thickness orientation tend to be large in the centre of EB welds of high aspect ratio (depth / width). By contrast, lower stress levels are produced on the surfaces acting transverse to the weld plane compared to conventional multi-pass metal arc welds. The transverse stress component is most likely to be required for the assessment of any postulated EB welding defects. The residual stress field decays rapidly with distance from the EB joint into the adjacent parent metal. Symmetrical stress distributions are typically generated in a 1-pass EB plate weld and stress fields are characteristically sinusoidal of wavelength between 1 and 4 times the section thickness.
Development of the RCC-MR Creep Deformation Model for the Prediction of Creep and Stress Relaxation in Type 321 Stainless Steel
Authors: Moffat A. J. , Douglas J. P. , White M. , Spindler M. W. , Austin C. , Jacques S. 
Published: November 2014

In this paper a creep deformation model has been developed for Type 321 stainless steel which has been based on a modified version of the creep model that is used in the French fast reactor design code RCC-MR.
The model has been evaluated using: 1) constant load creep data covering the temperature range from 550°C to 650°C and 2) constant displacement, stress relaxation data obtained from creep-fatigue tests at 650°C. Samples in the heat-treatment conditions of solution-treated, aged, and simulated ‘heat affected zone’ have been assessed.
The standard RCC-MR model was fitted to the constant load data and provided good predictions of forward creep. However, when this model was used to predict stress relaxation it was observed that the model significantly over predicted creep strain rates and therefore the level of stress drop during each cycle.
During constant load tests the stress remains relatively constant (noting that true stress does increase a small amount prior to rupture). However, in relaxation tests the stress varies significantly over the dwell. Due to the poor predictions of stress relaxation it was hypothesised that the fitted model did not capture the stress dependence of creep appropriately. The RCC-MR model was therefore modified to include a primary and secondary threshold stress term that is a function of the accumulated creep strain.
This work indicates that the RCC-MR model, modified to include threshold stresses, can be used to provide good predictions of both forward creep and stress relaxation in Type 321 stainless steel. Further work is required to validate this model on stress relaxation data at additional temperatures and lower start of dwell stresses.
Physical and Numerical Methods for the Dynamic Behavior of the Fast Reactor Cores
Authors: Broc Daniel , Cardolaccia Jérome , Martin Laurent 
Published: November 2014

In the frame of the GEN IV Forum and of the ASTRID Project, a program is in progress in the CEA (France) for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts.
The cores are constituted of Fuel Assemblies (of FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fluid Assemblies. The FA and the NS are slender structures, which may be considered as beams, form a mechanical point of view. The dynamic behavior of this system has to be understood, for design and safety studies. Two main movements have to be considered: global horizontal movements under a seismic excitation, and opening of the core. The dynamic behavior of the core is strongly influenced by contacts between the beams and by the sodium. The contacts between the beams limit the relative displacements. The fluid leads to complex interactions between the structures in the whole core.
The paper presents the physical and numerical methods and tools used to describe and simulate the phenomena. A key point is the Fluid Structure Interaction (or FSI): the interactions between the beams and the liquid sodium. The fluid movement is assumed to be described by the equations of a perfect fluid. Simple and efficient homogenization methods may be used to reduce the size of the problem. These methods are integrated in a general computer code, CAST3M developed at the CEA Saclay. This computer code allows to take into account the impacts between the beams. Some applications are presented.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 4 — Failure Behavior of Crossover Piping for Seismic Isolation System
Authors: Otoyo Teruyoshi , Otani Akihito , Fukushima Shunsuke , Jimbo Masakazu , Yamamoto Tomofumi , Sakakida Takaaki , Onishi Shigenobu 
Published: November 2014

This paper provides a part of the series titled “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. This part shows the failure behavior of crossover piping installed in a seismic isolated plant. The considered crossover piping is supported on one side by an isolated building and by a non-isolated building on the other side. During an earthquake, the piping structure is deformed due to the large relative displacements between the two buildings and at the same time excited by the different building seismic responses. Therefore, the high-pressure crossover piping structure requires both flexibility and strength.
In this study, 1/10 scaled shaking tests and FEM analyses have been performed to investigate the failure behavior of the crossover piping, where both seismic motions and excitations have been taken into account. It was confirmed that the failure occurs at the piping elbow through low cycle fatigue. Moreover, the results of the elastic-plastic response analysis, which simulates an extreme level of excitation corresponding to more than three times the design level, are in good agreement with the test results. The simulation also succeeded in predicting the experimental failure location.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 3 — Seismic Response of Crossover Piping for Seismic Isolation System
Authors: Otani Akihito , Otoyo Teruyoshi , Hirai Hideo , Iiizumi Hirohide , Shimizu Hiroshi , Saitoh Mitsuru , Onishi Shigenobu 
Published: November 2014

This paper, which is part of the series entitled “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”, shows the linear seismic response of crossover piping installed in a seismically isolated plant. The crossover piping, supported by both isolated and non-isolated buildings, deforms with large relative displacement between the two buildings and the seismic response of the crossover piping is caused by two different seismic excitations from the buildings. A flexible and robust structure is needed for the high-pressure crossover piping.
In this study, shaking tests on a 1/10 scale piping model and FEM analyses were performed to investigate the seismic response of the crossover piping which was excited and deformed by two different seismic motions under isolated and non-isolated conditions. Specifically, as linear response analysis of the crossover piping, modal time-history analysis and response spectrum analysis with multiple excitations were carried out and the applicability of the analyses was confirmed. Moreover, the seismic response of actual crossover piping was estimated and the feasibility was evaluated.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 5 — Fatigue Test of the Crossover Piping
Authors: Mizuno Kotoyo , Shimizu Hiroshi , Jimbo Masakazu , Oritani Naohiko , Onishi Shigenobu 
Published: November 2014

This paper provides a part of the series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”.
It is assumed the main steam crossover piping is damaged by the ratcheting deformation based on the relative displacement and the inertia load by the earthquake between the buildings and the internal pressure.
This part shows a low cycle ratcheting fatigue test using the scaling model under the combined loadings based on the relative displacement and the inertia load by the earthquake between the buildings and analyses were performed to confirm the failure modes and the fatigue life of the pipe elbow for the fatigue damage of the long-period ground motion.
As a result, the fatigue life under combined loads was sufficiently higher than the design criteria and analyses are good match with the test results. So, it confirmed the structural integrity of the crossover piping.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 1 — The Work Schedule of Project and a Seismic Design of Crossover Piping System
Authors: Suzuki Yutaka , Sato Kunihiko , Iiizumi Hirohide , Hisatsune Masakazu , Onishi Shigenobu 
Published: November 2014

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities” [1]–[4]. This part describes the work schedule of this project and the summary of a seismic design for crossover piping system.
Since the Southern Hyogo Prefecture Earthquake in 1995, a seismic isolated design has been widely adopted for Japanese typical buildings. The Japanese government accepted utilizing seismic isolation technology for nuclear power facilities with the 2006 revision of the “Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities”. Under these backgrounds, the Japan national project with the participation of all electric power companies and reactor vendors has been started from 2008 to develop seismic isolation systems of nuclear power facilities under the support of the Ministry of Economy, Trade and Industry.
In the design of seismic isolated plant, the crossover piping systems, such as Main Steam line and other lines related to the safety system have the important roles for overall plant safety. Therefore, the design of multiply supported piping systems between isolated and non-isolated buildings is one of the major key issues.
This paper focuses on the seismic response analysis of Main Steam crossover piping between seismic isolated Reactor Building and non-isolated Turbine Building. Multiple input response spectra and time history analyses of the crossover piping have been performed and the structural integrity of piping and the validity of the multiple input analysis method have been verified based on comparisons with the results obtained by conventional response spectrum analysis using enveloped floor response spectrum.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 10 — Evaluation of Seismic Isolator Design
Authors: Asano Hiroyuki , Hirotani Tsutomu , Nakayama Takashi , Norimono Takemi , Aikawa Yuji , Sato Kunihiko , Asakura Shinji , Jimbo Masakazu , Umeki Yoshito 
Published: November 2014

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. This part shows an evaluation of seismic isolator design established in this project where several methods are newly developed. The major four accomplishments are as follows.
One: establishment of design earthquake specially considered for seismically isolated nuclear power facilities. The design earthquakes are made to fit multiple target spectra with different damping factors considering a building, equipment and seismic isolators for more precise response analyses.
Two: design and development of a high-performance seismic isolator. Against the large design earthquakes, a seismic isolator is newly developed which has a large diameter lead plug for more damping; the isolators were actually manufactured and tested.
Three: seismic response analyses for seismically isolated nuclear power plants. Light water reactors are designed where the structural characteristics of the seismic isolation system is reflected.
Four: evaluation of thermal effects on seismic isolators by a long-duration earthquake. Considering a long-duration earthquake, the heat generation phenomenon in the lead plug is analytically evaluated to ensure the lead plug’s damping performance.
By introducing these accomplishments, the realistic design of a seismically isolated nuclear power plant is achieved.
Study on Stability Analysis on Acoustic Resonance in Heat Exchanger Tube Bundles
Authors: Nishida Eiichi , Hamakawa Hiromitsu 
Published: November 2014

Acoustic resonance may occur in heat exchangers such as gas heaters or boilers which contain tube bundles. This resonance is classified in self-excited oscillation, and feedback effect between vortex shedding and sound field plays important role. The final goal of our study is to develop a method by which to predict the resonance attack critical gas flow velocity and maximum resonance amplitude at the design stage. In order to reach this goal, it is essential to formulate the feedback effect between vortex shedding and a resonance mode concerned, and to execute a stability analysis of the resonance mode. There are two mechanisms in the feedback process: as the acoustic resonance grows, vortex strength is increased and vortex shedding synchronization grows.
This paper is concerned with the proposal of phenomenological model suitable to explain this mechanism and the formulation of these kinds of feedback mechanism with the use of this model. The model adopts vortex shedding wake oscillator model which is effective for tube vibration problems. Tube vibration movement is replaced by acoustic particle movement. Another improvement of our study is the introduction of statistical modeling of the wake oscillator to express vortex shedding synchronization effect. Here, the randomness of vortex shedding is explicitly modeled by a probability density function of the phase of the oscillator, and this function depends on the level of acoustic resonance. Based on these ideas to express the vortex/acoustics interaction, the formulas of stability analysis were derived.
Reliability of Water Jet Peening as Residual Stress Mitigation
Authors: Kono Yusuke , Okimura Koji , Tajima Masakazu , Noda Masaki 
Published: November 2014

As a countermeasure against high residual stress, some residual stress improvement methods have been developed; Water Jet Peening (WJP) [1] [2] for components installed in water, Shot Peening by Ultrasonic-wave vibration (USP) for components installed in air, and outer surface irradiated Laser Stress Improvement Process (L-SIP)[3] for components that can only be accessed from the outer surface.
WJP is applied to Reactor Vessel (RV) outlet/inlet nozzle safe-end joints (Alloy600 weld metal), RV Bottom Mounted Instrument (BMI) inner surface and J-weld. Especially, it is difficult to apply the technology to BMI because BMI inner surface is a very narrow space (inner diameter; approximately 10–15mm) and BMI J-weld configuration is a complicated 3-dimensional form.
On the occasion of actual application, the verification tests have been carried out and checked that a stress improvement was effective as one of PWSCC mitigations. And the compressive stress induced by the WJP was verified to continue to exist under actual plant operation conditions.
Thus, in addition to replacing the material with Alloy 690, converting the residual stress to compressive one can prevent the occurrence of PWSCC.
Validation of a Microwave Based Inspection System for HDPE Butt Fusions
Authors: Stakenborghs Robert J. 
Published: November 2014

The use of High Density Polyethylene (HDPE) is growing rapidly due to its corrosion resistance characteristics and ease of construction when compared to metal piping systems. As a result of its widespread use, a method of inspecting the piping joints is becoming essential to ensure safe, sound piping systems. This is particularly true for butt fusion joints, which can contain micro-structural defects that impact long term integrity of the piping system. These defects have proven to be difficult to detect using traditional means of inspection, such as ultrasound, because these micro-structural defects do not result in an acoustic reflector or impedance difference.
This paper presents the results of a rigorous validation program for a microwave non-destructive inspection technique for HDPE butt fusions. The program includes inspection of a cross section of defect types that comprise geometric type defects, such as back and side drilled holes, and butt fusion inclusions. It also includes defects of types commonly found in field applications, such as grass, oil, fine particulate and other contaminants and procedural deviations, such as low and high pressure and longer than typical dwell times (wait time between heating the pipe ends and fusing them). The success of the microwave inspection protocol and equipment in detecting the many types of defects that comprise the set of samples has been reduced to Probabilities of Detection (POD) and Probabilities of False Alarm (POFA). These statistics have been calculated in a traditional fashion and are presented in this paper.
Residual Stress Evaluation of Small Diameter Stainless Steel Piping by Experimental Measurements and Finite Element Analysis
Authors: Hicks Trevor G. , Mabe William R. , Miller Jason R. , Mullen John V. 
Published: November 2014

Residual stresses within stainless steel pipe welds may impact both component inspections and in-service cracking. Various investigators have examined piping residual stresses in large diameter piping using both finite element modeling and experimental techniques, but limited information is available for small diameter piping. This investigation uses both experimental methods and analytical modeling to evaluate the impact of welding on the residual stresses along the inner diameter (ID) of two small diameter pipe specimens.
Results of the investigation showed that tensile axial residual stresses were observed in the heat affected zone (HAZ) along the ID of the thin-wall pipe specimen with distinct regions of tensile and compressive axial stress which correlate well with the location where the last weld segments of the final weld pass were deposited. Higher stresses were also observed in the HAZ on the side where the final weld pass was deposited. By contrast, testing of the thick-wall pipe specimen showed significantly lower levels of tensile stresses along the pipe ID with the higher stress regions being biased toward the pipe outer diameter (OD).
Numerical Analysis of a RHR Piping System Subjected to Seismic Loading
Authors: Shen Ming-Yi , Lai Zih-Yu , Chai Juin-Fu , Lin Fan-Ru , Huang Yin-Nan , Yu Ching-Ching 
Published: November 2014

The objective of this study is to build a credible numerical model using SAP2000 for fragility analysis of RHR piping system, and to establish the load pattern of a cyclic loading test to identify the seismic vulnerability of the system. The RHR piping system selected for test and numerical analysis is duplicated from a part of the RHR system in a nuclear power plant (NPP) in Taiwan, and this part is distributed between the floor of the reactor building (RB) to the wall of the reinforced concrete containment vessel (RCCV) of the sample NPP.
The numerical model for the sample RHR piping system was developed, and the nonlinear response-history analysis was conducted using input motions compatible with the floor response spectrum at the anchor points of the sample piping system subjected to Safe Shutdown Earthquake (SSE). Based on the distribution of resultant inertial forces and the responses at critical locations of the piping system under SSE, the magnitudes and locations of the equivalent concentrated static loads were determined and used in the pushover analysis to estimate the capacity of the RHR piping system. The numerical results will be verified by the aforementioned cyclic loading test. More studies are on the way including shaking table test and fragility analysis for the sample piping system to further identify the seismic performance and risk of the system.
Models and methods for analyzing DCE‐MRI: A review
Authors: Khalifa Fahmi , Soliman Ahmed , El‐Baz Ayman , Abou El‐Ghar Mohamed , El‐Diasty Tarek , Gimel'farb Georgy , Ouseph Rosemary , Dwyer Amy C. 
Published: November 2014

Purpose:To present a review of most commonly used techniques to analyze dynamic contrast‐enhanced magnetic resonance imaging (DCE‐MRI), discusses their strengths and weaknesses, and outlines recent clinical applications of findings from these approaches.Methods:DCE‐MRI allows for noninvasive quantitative analysis of contrast agent (CA) transient in soft tissues. Thus, it is an important and well‐established tool to reveal microvasculature and perfusion in various clinical applications. In the last three decades, a host of nonparametric and parametric models and methods have been developed in order to quantify the CA's perfusion into tissue and estimate perfusion‐related parameters (indexes) from signal‐ or concentration–time curves. These indexes are widely used in various clinical applications for the detection, characterization, and therapy monitoring of different diseases.Results:Promising theoretical findings and experimental results for the reviewed models and techniques in a variety of clinical applications suggest that DCE‐MRI is a clinically relevant imaging modality, which can be used for early diagnosis of different diseases, such as breast and prostate cancer, renal rejection, and liver tumors.Conclusions:Both nonparametric and parametric approaches for DCE‐MRI analysis possess the ability to quantify tissue perfusion.