bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on 3-D Measurement Method Using Image Processing Technique for Dynamic Displacement in Shake Table Tests
Authors: Fujita Satoshi , Furuya Osamu , Niitsu Yasushi , Mikoshiba Tadashi , Nakano Daisuke
Published: August 2008

This study has been doing research and development of 3-dimensional measurement method using an image processing technique for a measurement of dynamic displacement in shake table test. This measurement system is a very convenient system because it can measure the 3-D dynamic displacement in shake table test in which the several makers are only attached to the surface of an experimental structure. The system therefore is the most suitable measurement system for an evaluation of complex 3-dimensional behaviour of test model. Fundamental hardware and software for the measurement system has been constructed until now. The fundamental dynamic measurement accuracy and effectiveness of the measurement system has been also confirmed from several shake table tests. In this paper, the dynamic measurement accuracy in three measurement units on the assumption of practical use is described from the results of shake table tests in National Research Institute for Earth Science and Disaster Prevention using the collapse building and the inverted pendulum model.
Dynamic Characteristics of a Sliding Bearing Using Lubricant Material for Base Isolation of Light Structures
Authors: Masaki Nobuo , Takeuchi Sadamitsu , Hirata Hisashi
Published: August 2008

The authors have been developed a system for base isolation of a wooden house and a prefabricate house. It consists of sliding bearings and laminated rubber bearings. There are two types of sliding bearings, one has the coefficient of friction of 0.046, the other has 0.06. Each of sliding bearings supports 30 ∼ 40kN of vertical load, the surface pressure of the sliding shoe is 10 ∼ 12 N/mm2 (10∼12MPa). The horizontal stiffness of the laminated rubber bearing is 38.5N/mm. For base isolation of an ordinary wooden house, 12 ∼ 16 sliding bearings and 4 ∼ 6 laminated rubber bearings are used. If the mass of a wooden house is 40ton and it is supported by 16 sliding bearings of the coefficient of friction of 0.046 and attached by 4 laminated rubber bearings with the horizontal stiffness of 38.5 N/mm, therefore it has 0.249 of the equivalent damping factor and 3.2 second of the natural period. As for the base isolated houses which are used this system, 500 houses have been constructed in Japan currently by the August of 2002 (Takahashi, 2002). As an advanced system of this base isolation system, we have developed a new type of the base isolation system of sliding bearing using silicone lubrication materials and rubber spring devices.
Construction of Generalized Neural Network System for Recognizing Several Non-Linear Behaviors
Authors: Mazda Taiji , Otsuka Hisanori , Yabuki Wataru , Iwamoto Kensuke
Published: August 2008

Generally, in formulating a spring-mass model for hysteric behavior of materials and members with inelastic characteristic, a mathematical model based on load-deformation experimental results is considered. The model must approximate the inelastic hysteresis of the material. However, assumption of material’s behavior using mathematical models is crucial, since it may cause serious errors if inappropriate model is applied for a particular situation. This paper describes multiple layered neural network to simulate the non-linear hysteretic behavior like Ramberg-Osgood model, modified bilinear model and Takeda model. In this study, based on the pattern recognition ability of neural network, non-linear hysteretic behavior was modeled by the network directly without replacing it with a mathematical model. The effectiveness and applicability of the network in numerical analysis were evaluated. Generalized multiple layered neural network to evaluate non-linear hysteretic curve was constructed. The network can recognize well the three types of hysteretic curve. The network is available as a subroutine of non-linear spring in dynamic response analysis.
An Advanced Cap Model for Concrete Materials
Authors: Tsai C. S. , Chen Ching-Shyang , Lin Yong-Zhang , Chen Bo-Jen , Chen J. C.
Published: August 2008

Concrete is a pressure-sensitive material. To ensure the safety of structures, it is required to understand the mechanical behavior of concrete. The stress-strain relation of concrete is highly nonlinear and pressure dependent. Therefore, design problems involving concrete materials need an appropriate constitutive model to quantify its stress-strain relation. Presented in this paper is a new cap-type constitutive law for concrete. The model is prominent in the sense that it satisfies the compatibility between the failure surface and the yield cap. It has modified the classical cap model to obtain smooth yield surfaces. In addition, the model effectively describes strength variations along various directions in the octahedral plane. The model has shown to realistically predict concrete responses in experiments by 7 parameters. The proposed concept can also be extended to include as many previous models for describing various observed material behavior as it is required.
Finite Element Formulations and Theoretical Study for VCFPS
Authors: Tsai C. S. , Chiang Tsu-Cheng , Chen Bo-Jen
Published: August 2008

The friction pendulum system (FPS), a type of base isolation technology, has been recognized as a very efficient tool for controlling the seismic response of a structure during an earthquake. However, previous studies have focused mainly on the seismic behavior of base-isolated structures far from active earthquake faults. In recent years, there have been significant studies on the efficiency of the base isolator when subjected to near-fault ground motions. It is suggested from these studies that the long-duration pulse of near-fault ground motions results in significant response of a base-isolated structure. In view of this, an advanced base isolator called the variable curvature friction pendulum system (VCFPS) is proposed in this study. The radius of the curvature of VCFPS is lengthened with an increasing of the isolator displacement. Therefore, the fundamental period of the base-isolated structure can be shifted further away from the predominant period of near-fault ground motions. Finite element formulations for VCFPS have also been proposed in this study. The numerical results show that the base shear force and story drift of the superstructure during near-fault ground motion can be controlled within a desirable range with the installation of VCFPS. Therefore, the VCFPS can be adopted for upgrading the seismic resistance of the structures adjacent to an active fault.
Error Propagation for Explicit Pseudodynamic Testing
Authors: Chang Shuenn-Yih
Published: August 2008

Error propagation characteristics of the Newmark explicit method, the modified Newmark explicit method and the α-function dissipative explicit method in performing a pseudodynamic test are studied herein. It is shown that the Newmark explicit method is non-dissipative while the modified Newmark explicit method and the α-function dissipative explicit method are dissipative and can eliminate the spurious participation of high frequency responses. Furthermore, analytical results of the error propagation analysis reveal that the modified Newmark explicit method and the α-function dissipative explicit method have much better error propagation properties when compared to the Newmark explicit method. The major disadvantages of the modified Newmark explicit method are the positive lower stability limit and undesired numerical dissipation. This implies that the α-function dissipative explicit method might be the most appropriate explicit pseudodynamic algorithm among the three integration methods.
Development of a New Method of Baseline Correction on Earthquake Strong Motions and Its Application to Long Period Sloshing Responses of Liquid Storage Tanks During Strong Earthquakes
Authors: Minowa Chikahiro
Published: August 2008

In this paper, a new method of baseline correction on strong motion acceleration records is presented and the fundamental concept for baseline corrections on the earthquake strong motions is described. Considering the filtering effect, the earthquakes ground motion displacements of 1995 JMA KOBE, 1999 Kocaeli YPT and 1999 Chi-Chi TCU068 are discussed. Also, the linear sloshing responses of large liquid tanks subjected to these motions were discussed. Since liquid storage tanks show the low frequency (long period) sloshing characteristics and the strong motion characteristics of 1999 Kocaeli and Chi-Chi earthquakes are also low frequencies and large permanent displacements, the sloshing responses in large liquid tanks, especially in long natural periods, were significantly affected by the low frequency motions (large permanent displacements) of these devastating earthquakes. It is very important to use suitable ground motion characterized low frequency content for earthquake resistant design of liquid storage tanks. The baseline correction method presented in the paper may be adequately used to correct strong motion records for large liquid storage design.
Investigation on Buckling Behavior of Cylindrical Liquid Storage Tanks Under Seismic Excitation: 2nd Report — Investigation on the Nonlinear Ovaling Vibration at the Upper Wall
Authors: Morita Hideyuki , Ito Tomohiro , Hamada Koji , Sugiyama Akihisa , Kawamoto Yoji , Ogo Hideyasu , Shirai Eiji
Published: August 2008

When a thin walled cylindrical liquid storage tank suffers a large seismic base excitation, buckling phenomena such as elephant foot bulge at the bottom portion and nonlinear ovaling vibration at the upper portion shows nonlinearity between the input and response level and suddenly occurs for the excessive input level, thus will be called as “nonlinear ovaling vibration” hereafter in this paper, may be caused. In the 1st report, the elephant foot bulge phenomena and the liquid pressure effects were investigated. In this 2nd report of the series of studies, the effect of nonlinear ovaling vibration phenomena were investigated based on the dynamic buckling tests using scaled models of thin walled cylindrical liquid storage tanks for nuclear power plants. The mechanism and the effect of vertical excitation and liquid sloshing were also studied and discussed.
Investigation on Buckling Behavior of Cylindrical Liquid Storage Tanks Under Seismic Loading: 3rd Report — Proposed Design Procedure Considering Dynamic Response Reduction
Authors: Sugiyama Akihisa , Setta Koji , Kawamoto Yoji , Hamada Koji , Morita Hideyuki , Ito Tomohito , Ogo Hideyasu , Shirai Eiji
Published: August 2008

As for thin walled cylindrical liquid storage tanks in nuclear power plants, the current elastic design guideline against seismic loading might result in too conservative component design as compared with elasto-plastic design in general industries. Therefore, it is thought possible to make the design guideline more reasonable by taking dynamic response reduction into account. In this series of study, experiments using scaled models were carried out, and seismic behavior of thin walled cylindrical liquid storage tanks was simulated to investigate energy absorption capacity and seismic resistance of those tanks. In this 3rd report of series of studies, seismic behavior of tanks was simulated to estimate a dynamic response reduction factor. This factor is based on the energy absorption capacity of structures. Through experiments and numerical study, a response reduction factor to design thin walled cylindrical liquid storage tanks has been proposed.
Seismic Responses of Transmission Tower-Pile-Soil Dynamic Interaction
Authors: Li Hong-Nan , Xiao Shi-Yun , Wang Su-Yan
Published: August 2008

The mechanical model of Soil-Pile-Structure interaction of transmission tower-cable system is presented in this paper, and the corresponding equations of motion are derived, in which the nonlinear characteristics of soil is included in the dynamic time history analysis. The computer program for the system is complied and used to calculate the earthquake response of an actual transmission tower. The results of numerical calculation are compared with and without considering soil-pile-structure interactions, and show that the dynamic interaction should be taken into account in the soft foundation.
Design Competition for Development of a General Purpose Fissile Package
Authors: Houghtaling T. K. , Abramczyk G. A. , Hoang D. S. , Leduc D. R. , Lutz R. N. , Watkins R. W. , Wu T. , Smith A. C. , Blanton P. S. , England J. L. , Gelder L. F. , Gupta N. K. , Ketusky E. T. , McKeel C. A. , Bellamy J. S.
Published: August 2008

In general, the design process involves envisioning and developing concepts for a component or system, combining these fractions into an integrated whole and evaluating the final design against functional requirements. A major challenge is developing components or systems to a level of maturity that permits feasibility evaluation of the integrated whole while optimizing opposing performance functions (e.g., thick for strength, but thin for heat transfer). Economic pressure often drives design concepts to conservative bases early in the process. The approach presented in this paper is a highly cost-effective means of developing alternative design solutions for given set of design requirements — in this case a radioactive materials transportation package. In response to planned phase-out of the Department of Transportation 6M specification package, an internal competition was held at the Savannah River Site to develop design proposals for the 6M replacement. Two teams were provided with design requirements, design evaluation criteria, a limited budget and short time frame to develop proposals for a General Purpose Fissile Package (GPFP). Resulting GPFP designs, while quite different, met the performance requirements, were economical, operator friendly, light weight, certifiable, and able to accommodate a wide range of radioactive materials. The methodology of an in-house design competition proved to be a very cost effective way of advancing design alternatives from design requirements to practical design proposals. Estimated fabrication costs for each of the designs are significantly less than other packages currently available for the proposed service.
Gas Analysis of Plutonium Materials Tested in a Bell Jar
Authors: Hensel Steve J. , Kies Jonathan R.
Published: August 2008

Seventy cans containing plutonium bearing materials have been tested for gas generation in bell jars at the Savannah River Site (SRS). Gas samples from five of the tests were taken and analyzed for hydrogen, oxygen, nitrogen, methane, carbon dioxide, carbon monoxide, and nitrous oxide. Gas samples from all five cans showed significant oxygen consumption. Hydrogen and carbon dioxide were found in non-negligible quantities in all gas samples. Hydrogen and carbon dioxide generation rates are presented along with oxygen depletion rates. A description of the plutonium bearing materials is also provided.
Repackaging of Two Bulged Crimp Sealed Cans Containing Plutonium Bearing Materials
Authors: Watkins R. W. , Hill R. D.
Published: August 2008

Two cans containing plutonium bearing materials were found during radiography surveillance activities to be bulged. The cans had been stored in DOT 6M shipping containers at the Savannah River Site. The material in the first can (Item CZA96-179) was packaged can/bag/can configuration with the inner and outer cans being crimp sealed. The crimp sealed innermost can was clearly deformed from the radiography picture taken for surveillance purposes. This material had been stored in the shipping container since the mid 1970’s. The second can (Item 50014440) contained plutonium bearing material of a different origin. The material had been repackaged at the Savannah River Site in the mid 1990’s, and the repackaged can was stored in a 6M shipping drum. A special puncturing tool, which secured the can and allowed for a very controlled puncture of both outer and inner cans was used in a glovebox. The glovebox has a dry air system and an argon supply. The puncturing tool utilized a non-sparking punch and an argon purge. The cans were repackaged into filtered outer cans. A description of the puncturing tool, repackaging activities, and of the materials will be provided.
Modeling of Stress Corrosion Cracking for High Level Radioactive-Waste Packages
Authors: Lu S. C. , Gordon G. M. , Andresen P. L. , Herrera M. L.
Published: August 2008

A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain radioactive-waste repository. SCC is one form of environmentally assisted cracking resulting from the presence of three factors: metallurgical susceptibility, critical environment, and tensile stresses. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022, the environment is represented by the water film present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the stress is principally the weld induced residual stress. SCC has historically been separated into “initiation” and “propagation” phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding). To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulas for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, the time to through-wall penetration for the waste package can be calculated. The SDFR model relates the advance (or propagation) of cracks, subsequent to the crack initiation from bare metal surface, to the metal oxidation transients that occur when the protective film at the crack tip is continually ruptured and repassivated. There exists a threshold stress intensity factor, which provides a criterion for determining if an initiated crack or pre-existing manufacturing flaw arrest. This paper presents the research results that quantify the threshold stress, threshold stress intensity factor, and the parameters in the crack growth rate equation based on experimental results developed specifically for Alloy UNS-N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. Stress mitigation by the laser peening technique is also discussed.
Corrosion Characteristics of Titanium Alloys in Multi-Ionic Environments
Authors: Wong Lana L. , Estill John C. , Fix David V. , Rebak Rau´l B.
Published: August 2008

Yucca Mountain (Nevada) is designated as a high-level nuclear waste repository. The nuclear waste will be isolated by a series of engineered barriers. The metallic engineered barriers will consist of a double-wall container with a detached drip shield. The material for the external wall of the container is Alloy 22, a corrosion-resistant Ni-Cr-Mo alloy. Titanium grade 7 has been proposed for the drip shield. Ti alloys are highly resistant to all forms of corrosion due to the formation of a stable, protective and strongly adherent oxide film. The aim of this research was to characterize the general and localized corrosion behavior of Ti Gr 7, 16 and 12 in simulated concentrated ground waters. Welded and non-welded coupons were exposed for up to 5 years to the vapor and liquid phases of acidic and alkaline multi-ionic solutions at 60°C and 90°C. This paper describes the results obtained after approximately 2-1/2- to 5-1/2-year exposure to the testing electrolyte solutions. In general, the highest corrosion rate was obtained for Ti Gr 12; however, in all of the tested conditions, the corrosion rate was generally lower than 100 nm/yr. For all alloys, the highest corrosion rate was obtained in the concentrated alkaline solution.
Failure Pressure Simulations for 3013 Inner Can Under Room Fire Conditions
Authors: Gupta Narendra K.
Published: August 2008

The 3013 container is widely used for long-term storage and transporting of the fissile material (PuO2) in the DOE complex. The design of 3013 container is governed by the DOE Standard 3013. Thermal and structural analyses were performed to assess the integrity of a closed 3013 inner can assembly during a postulated room fire transient. The inner can assembly is engulfed in a 932°F (500°C) 30-minute long room fire. The fire transient results in a can temperature of about 900°F and an upper bound internal pressure of 641 psi due to the evaporation of moisture and generation of gases in the stored PuO2. The structural analyses show that the inner can assembly can withstand a pressure of 641 psi and meet the ASME B&PV Code stress acceptance criteria for one time accident loading. The failure pressures of the 3013 inner can at can metal temperatures of 900°F, 1200°F, and 1400°F are estimated to be 1435 psi, 1051 psi, and 705 psi respectively.
Thermal Analysis and Test Results for the Overpack of a Typical Radioactive Materials Package
Authors: Gupta Narendra K. , Smith Allen C.
Published: August 2008

In the course of the development and certification of the 9975 Package, extensive thermal analyses were performed and the package subjected to the regulatory HAC thermal test. The results of the thermal analysis and materials tests of the cane fiberboard overpack material were evaluated in comparison with the package HAC thermal test results. The evaluation confirmed that the thermal analysis correctly predicted the performance of the 9975 in the HAC fire test. The post test examination revealed that the heat affected region of the Celotex® overpack correlated well with the calculated temperature distribution.
Analysis of the Impact of a Tunnel Fire Environment on a Spent Nuclear Fuel Transportation Cask
Authors: Bajwa Christopher S.
Published: August 2008

On July 18, 2001, a train carrying hazardous materials derailed and caught fire in the Howard Street railroad tunnel in Baltimore, Maryland. Due to this accident, questions were raised about the performance of spent nuclear fuel transportation casks under severe fire conditions, similar to those experienced in the Baltimore tunnel fire. The U.S. Nuclear Regulatory Commission (NRC) evaluates the performance of spent fuel transportation casks under accident conditions. Title 10 of the Code of Federal Regulations Part 71 section 73(c)(4), (10 CFR 71.73(c)(4)) requires that transportation packages used to ship radioactive material must be designed to resist an engulfing fire of a 30 minute duration and prevent release of radioactive material to the environment. The staff of the NRC, in cooperation with the National Transportation Safety Board, the National Institute of Standards and Technology, Pacific Northwest National Labs and the Center for Nuclear Waste Regulatory Analysis, have undertaken an analysis to determine the thermal conditions present in the Howard Street tunnel fire, as well as analyze the effects that such a fire would have on a spent fuel transportation cask. This paper describes the analytic models used in the assessment and presents a discussion of the results.
Development of Prediction Equations on Charpy Upper Shelf Energy for Japanese RPV Steels
Authors: Sakamoto Kazunobu , Hirota Takatoshi , Osaki Toru , Tomimatsu Minoru
Published: August 2008

It is well known that as the embrittlement due to neutron irradiation on reactor pressure vessel (RPV) steels, there is the tendency of the reduction of Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 1024n/m2 in OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 years plant operation for the integrity assessment of the RPVs. The USE prediction equations have been developed through the regression analyses of the test reactor data combined with Japanese surveillance test data. This research was entrusted by the Ministry of Economy, Trade and Industry in Japan.
French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects
Authors: Moinereau Dominique , Frund Jean-Michel , Churier-Bossennec Henriette , Bezdikian Georges , Martin Alain
Published: August 2008

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....
The European Network AMES for Reactor Component Ageing in the 6th EURATOM Framework Programme
Authors: Sevini F. , Debarberis L. , Gerard R. , Brumovsky M.
Published: August 2008

The AMES (Ageing Materials European Strategy) European network started its activity in 1993 with the aim of studying ageing mechanisms and remedial procedures for structural materials used for nuclear reactor components. Operated by JRC-IE, it has been supporting the co-ordination of the project cluster throughout the 4th and 5th EURATOM Framework Programs, carrying out projects on with plant life management implications. A short summary is given of the main outcome. Main frame of the network in the 5th Framework Programme is the ATHENA project, which is aimed at summarizing the obtained achievements and edit guidelines on important issues like the Master Curve, Effect of chemical composition on embritllement rate in RPV steels, Re-embrittlement models validation after VVER-440 annealing and open issues in embrittlement of VVER type reactors. The enforcement of the ERA (European Research Area) concept arisen with start of the 6th EURATOM Framework in 2003 will produce some changes and reorganisation. The network will be part of a broader initiative on PLIM including in a more integrated way NESC, ENIQ, NET and AMALIA networks. The paper shows an overview of the concluded projects, achievements of the running ones and open issues tackled in the 6th EURATOM FWP and a summary of the plans for a new broader network on NPP Plant Life management (SAFELIFE).
A Decision Advisor to Manage the Implications of Alloy 600 Issues
Authors: Perdue R. K. , Mycoff C. W. , Woodcock J. , Meyer T. A. , Llovet R.
Published: August 2008

A Decision Advisor process has been designed to efficiently support plant level decisions, and the process has been specialized to support decisions pertaining to plant locations that have alloy 600 degradation issues. The process guides the decision maker through phases of framing the problem, developing options, selecting alternatives, and planning. The decision making process is supported by a computer software model that quantifies the net present value costs of each identified strategy. The Alloy 600 Decision Advisor model calculates crack initiation and growth and the impact on operations and revenue for a given inspection and repair strategy. The three major modules in a component Decision Advisor are: engineering issues of crack propagation and detection, plant operations, and financial impacts. The crack propagation and detection module captures the current state-of-the-art knowledge regarding alloy 600 (and alloy 182/82 for welds) degradation, combining statistical analysis of historical data and expert engineering judgment to produce a forecast of cracks and leaks over the relevant planning period. This is input to the plant operational impact module, which calculates the resulting impacts on plant revenue and cost drivers. The model can be readily updated as new information (such as inspection results) comes available. Results are provided in a report describing the goals and objectives together with the technical and financial input, and providing the plant-level net present value cost of each alternative. The net present value costs include typical costs of implementation (such as inspections and repairs) as well as the effects on power production and plant revenue.
Implicit Fatigue Design of Piping Systems for License Renewal of B31.1 Plants
Authors: Chang K. C.
Published: August 2008

The fatigue analysis of older vintage nuclear power plants was normally performed following the implicit method defined in the USAS B31.1, Power Piping Code. The USAS B31.1 Code applies reduction factors to allowable stresses for loadings with specific numbers of cycles. This effectively reduces stress amplitudes due to thermal bending and prevents fatigue damage from occurring. However, it does not provide sufficient detail to address the environmental impacts on fatigue when license renewal application is prepared for a specific plant. Explicit fatigue analyses were performed for many components in plants whose piping was designed by the implicit fatigue analysis methods. Over the years, modified or improved plant operating procedures may have been implemented. More sophisticated calculations became available for use on several components for which some explicit fatigue analyses is required to address the environmental effects for 60 years of operation. This paper addresses the type of information required to evaluate the effect of environmental factors on fatigue in a license renewal application for plants with reactor coolant system pressure boundary piping designed to B31.1 Code.
License Renewal in the U.S. and Plant Life Management (PLIM)
Authors: Young Garry G. , Rinckel Mark A.
Published: August 2008

License renewal of operating nuclear power plants in the United States has become one of the most successful nuclear industry activities in the past few years. It is anticipated that over 90% of the 103 operating nuclear power plants in the United States will pursue license renewal and seek an additional 20 years of operation. Some plants may pursue operation to 80 years or longer since the license renewal rule does not limit the operating life of a nuclear power plant. The requirements for renewing the operating license of a nuclear reactor in the United States are contained in Nuclear Regulatory Commission (NRC) Regulation 10 CFR Part 54, which addresses general, technical, technical specification, and environmental requirements. The most labor intensive element of the requirements are the technical requirements, which include addressing an integrated plant assessment (IPA) and time limited aging analyses (TLAA). The cost of performing the needed reviews and obtaining a renewed license ranges between $10M to $15M. The license renewal rule focuses on aging of passive long-lived components and aging management programs that manage those structures and components. The aging management programs credited to manage aging include both existing programs (e.g., ASME Section XI) and a few new programs (e.g., Reactor Vessel Internals Aging Management Program). Commitments to aging management programs for license renewal may be implemented and tracked through a comprehensive plant life management (PLIM) program. PLIM is the process to integrated equipment aging management with other plant activities to maximize plant value. PLIM can save the operating plant a significant amount of money by effectively planning and implementing component refurbishment and replacement. The ultimate decision to seek license renewal and continue operation is based on PLIM, which considers aging, safety, and economics.
Significant Corrosion of the Davis-Besse Nuclear Reactor Pressure Vessel Head
Authors: Lang Theodore A.
Published: August 2008

In March of 2002, significant corrosion of the Davis-Besse reactor head was discovered. The Davis-Besse reactor head is of standard construction, composed of low alloy steel and clad with stainless steel. Alloy 600 control rod nozzles penetrate the reactor head, attached with J-groove welds. During an ultrasonic inspection, three of these nozzles were found to have through-wall cracks induced by Primary Water Stress Corrosion Cracking (PWSCC). Undiscovered leakage of borated water over the course of several operating cycles from one of these nozzles led to localized cooling and wastage of the reactor head near the nozzle. This leakage, less than 0.2 gpm (0.8 l/min), was small in comparison to allowable unidentified leakage, but larger than typical PWSCC leakage. The greatest damage to the low alloy steel reactor pressure vessel head was an oblong cavity, approximately 7 × 5 inches (18 × 13 cm), penetrating to the stainless steel cladding. The cracks in this nozzle were axially oriented, which would previously have been considered low risk because they would not have caused control rod ejection. However, the damage led to an increase in risk of a loss of coolant accident, prolonged loss of generation, and replacement of the reactor pressure vessel head. In addition to the industry wide regulatory impact of this event, the Nuclear Regulatory Commission has indicated that there may be a need to revise the inservice inspection requirements in Section XI of the ASME Code. This paper provides a brief synopsis of PWSCC in Control Rod Drive Mechanism nozzles, describes the inspection activities that led to the discovery of both the cracking and the corrosion, and describes the extent and technical cause of the damage. Management and human performance issues that allowed the damage to progress to an advanced state are discussed, since this event would not have been noteworthy if administrative controls and programs had been properly implemented.
Effects of Flow Accelerated Corrosion on Piping Steady-State Vibration Evaluations
Authors: Voll Brian J.
Published: August 2008

Piping steady-state vibration monitoring programs were implemented during preoperational testing and initial plant startup at most nuclear power plants. Evaluations of piping steady-state vibrations are also performed as piping and component failures attributable to excessive vibration are detected or other potential vibration problems are detected during plant operation. Additionally, as a result of increased flow rates in some piping systems due to extended power uprate (EPU) programs at several plants, new piping steady-state vibration monitoring programs are in various stages of implementation. As plants have aged, pipe wall thinning resulting from flow accelerated corrosion (FAC) has become a recognized industry problem and programs have been established to detect, evaluate and monitor pipe wall thinning. Typically, the piping vibration monitoring and FAC programs have existed separately without interaction. Thus, the potential impact of wall thinning due to FAC on piping vibration evaluations may not be recognized. The potential effects of wall thinning due to FAC on piping vibration evaluations are reviewed. Piping susceptible to FAC and piping susceptible to significant steady-state vibrations, based on industry experience, are identified and compared. Possible methods for establishing links between the FAC and vibration monitoring programs and for accounting for the effects of FAC on both historical and future piping vibration evaluations are discussed.
Fitness-for-Service of Pressure Relieving Systems
Authors: Short W. E.
Published: August 2008

Fitness-for-service assessments most commonly have been applied to tanks and pressure vessels. However, fitness-for-service evaluations are applicable to virtually all equipment, piping and hardware in an operating facility that, upon failure, would result in economic loss of equipment or production, exposure of plant personnel to hazardous conditions, toxic or hazardous releases to the environment, and/or risk to public safety. Federal and state regulations require operating industrial facilities to have risk management programs in place that include the design basis for safety-relief systems installed to protect pressurized equipment from overpressure. For process facilities that have been in operation for many years, the original design basis and calculations for the safety-relief devices often are no longer available. A fitness-for-service assessment of an existing pressure relieving system includes verification of the relief device size and specification, and review and substantiation of required documentation. This paper presents the required elements of a pressure relieving system size verification and documentation program for existing process facilities. The required technical information to perform a fitness-for-service assessment is described. Some restrictions and limitations of the codes and standards that are applied for design guidance of pressure relieving systems are challenged [5–13].
An Investigation on Mechanical Energy Supply in Rotating Flow Passage of Centrifugal Pump
Authors: Tanaka Takaharu
Published: August 2008

Centrifugal pump is a typical turbomachinery, which transfers mechanical energy to hydraulic energy through the rotational motion of impeller blades. It is commonly used and generally operated at a very high efficiency. Therefore, it would seem that theoretical discussion of performance and experimental observations of internal flow conditions inside the pump should be fully understood by now. However, it appears that neither the basic expressions nor the theoretical design methods are that clear. For example, the most fundamental definition of pump head, which is the most important equation in pump textbooks, is not often well explained. The purpose of this oral presentation is to share preliminary results of on-going studies on the energy transfer in centrifugal pumps.
TOFD Automatic Ultrasonic Testing for Condition Monitoring of Coke Drums
Authors: Penso Jorge A. , Owen Robert , Oka Masaaki
Published: August 2008

Cracking and bulging in welded and internally lined pressure vessels that work in thermal-mechanical cycling service have been well known problems in the petrochemical, power and nuclear industries. In spite of this problem has been studied during the last fifty years, published literature and industry surveys show that similar problems still occur nowadays. Typical examples of this problem are the coke drums in the delayed coking units refinery process. Delayed coking units are among the refinery units that have higher economical yields. To shut down these units represents a high negative economical impact in refinery operations. Also, the maintenance costs associated with repairs are commonly very high. Cracking and bulging occurrences in the coke drums, most often at the weld areas, characterize the history of the operation of delayed coking units. To anticipate through wall cracking in these coke drums, AUT (automatic ultrasonic testing); Dual TOFD (time of flight diffraction) and the Phased Array technique simultaneous inspection system was selected among other inspection techniques as a condition monitoring tool during an unit turnaround. The inspection methodology in combination with fracture mechanics was used to classify discontinuities as acceptable and non-acceptable. This indicated approach helped to optimize the workscope during the turnaround and establish guidelines for inspection and repair of the delayed coker unit. This work presents the different steps followed during the inspection and fitness for service evaluation. Also, this study shows advantages and disadvantages of the AUT-Phased Array technique.
Study and Application of Health Monitoring by Fiber Optic Sensors in Civil Engineering
Authors: Li Hong-Nan , Li Dong-Sheng , Wang Su-Yan
Published: August 2008

In civil engineering, the smart health monitoring method by use of fiber optic sensor is a new approach that evaluates the structural health situation. The current status in applications of fibre optic structural health monitoring in civil engineering structures with a brief introduction of the advantages, basic principles of fibre optic sensors is described in this paper. Leakage detection and potential damage to pipelines are emphasized. Finally, existing problems for packing and implementing fibre optic sensors in structures are discussed.
Finite Element Study on Piezoelectric Sensors for Blade Vibration Measurements
Authors: Sunar M. , Al-Bedoor B. O.
Published: August 2008

The output of piezoelectric sensors is studied numerically as a method for monitoring the vibrations of blades used in turbomachinery. Two thick piezoelectric layers are bonded on the bottom and top surfaces of a blade at the root between the blade and the disk. The placement of piezoelectric sensors at the root allows measuring and/or computing the blade vibrations from the root without exposing the sensors to fast rotation and corrosive environment. As a preliminary investigation, the blade is modeled as a cantilever beam mounted with two piezoelectric layers at its root. The beam and piezoelectric layers are analyzed with the finite element method The beam is excited by step and sinusoidal forces acting upward at its free end. The numerical results indicate the effectiveness of placing piezoelectric sensors at a blade root for measuring the blade vibrations.
Effect of Fluid Mass on Non-Linear Natural Frequencies of a Rotating Beam
Authors: Al-Qaisia Ahmad A.
Published: August 2008

The non-linear natural frequencies of the first three modes of a beam clamped to a rigid rotating hub and carrying a distributed fluid along its span are investigated. The mathematical model is derived using the Lagrangian method and the continuous system is discretized using the assumed mode method. The resulted unimodal nonlinear equation of motion was solved using two methods; harmonic balance (HB) and time transformation (TT), to obtain approximate analytical expressions for the nonlinear natural frequencies. Results have shown that the two terms harmonic balance method (2THB) is more accurate than the time TT method. Results for the effect and type of distribution, i.e. uniform or linearly distributed, on the variation of the nonlinear natural frequency with the rotational speed of the system and how they affect the stability are studied and presented in non-dimensional form.
Investigation of Stress Indices and Directional Loading of Eccentric Reducers
Authors: Wais Edward A. , Rodabaugh E. C. , Carter R.
Published: August 2008

Stress indices and stress intensification factors are used in the design of piping systems that must meet the requirements of ASME Section III for Class 1 and Class 2 systems. This study reviews the present values for eccentric reducers and provides new test data for comparison, which takes into account the directionality of the loading. Suggestions are presented which significantly improve the evaluation of reducers.
Relaxed Hanger Spans for Non-Critical Piping
Authors: Basavaraju Chakrapani , Fox Ronald C.
Published: August 2008

Piping systems are normally used to transport air, gases, steam, water, and other fluids. Piping systems in power plants, petro-chemical, and other industrial facilities that carry non-hazardous fluids like air, water, etc. with no significant pressures at moderate temperatures can be considered as non-critical. Codes such as ASME B31.1 provide suggested dead weight spans for the placement of supports. In this paper, relaxed hanger spans are computed for non-safety related piping systems taking in to consideration the maximum bending stress, maximum deflection or pipe sag, and bearing stresses due to dead weight for pipe sizes ranging from 1” to 42” diameter. For situations in which large diameter pipes are directly sitting on steel with a line contact, bearing stresses are computed using finite element analysis as well as simple formulas. A comparison of relaxed hanger spans with B31.1 suggested spans is presented. Significant benefits in lowering the cost due to reduced number of dead weight supports, and the associated savings due to reduction in materials, fabrication, and installation can be derived by using the relaxed spans. These relaxed hanger spans were utilized over the past nine years on a variety of non-safety related piping systems. Experience with more than a dozen power projects of Pulverized coal (PC), Cogen, Simple cycle, and Combined cycle types indicates that the piping systems, which used relaxed hanger spans, are operating normally and satisfactorily without any problems.
Effects of Inline Rotating Component on Piping System Dynamic Characteristics
Authors: Ibrahim Zakaria N.
Published: August 2008

Piping system dynamic characteristics are determined utilizing the classical Eigen-Value problem solution. Systems with integrated inline rotating components would experience some changes in these characteristics. The intensity of these changes depends on the relative magnitudes of their rotational momentums. Pseudo single degree of freedom ‘SDOF’, systems that include gyroscopic momentum of a rotating shaft are formulated and are subjected to seismic base excitations. The comparative responses of these SDOF are investigated. The gyroscopic inertia significantly alters their predicted seismic responses. For multi-degree of freedom system, simplified methodology is formulated to provide quick assessment of including the gyroscopic effects of the rotating components. The lumped parameters of the inline-rotating component within the piping system require the velocity degrees of freedom to be included in the analysis. This results in the non-classical Eigen-Value problem type of formulation. The latter is simplified utilizing the classical Eigen-Value solution conventionally termed as normal modes extraction. The normal modes synthesis reduces the number of degrees of freedom for the non-classical Eigen-Value problem to a manageable level. The concept of mode shape utilization coefficient ‘MSUC’, is introduced to provide quick assessment of the system. The gyroscopic inertia does not dominate the system dynamic characteristics. Low to moderate rotational momentums, typically installed by the industry, slightly affect the piping system dynamic characteristics. This justifies the industry practice of ignoring the rotational momentums within the piping systems routine analysis. It is also found that very high rotational momentums are artificially required to dramatically change the system dynamic characteristics.