bluuebunny/crossref_metadata_2025 huggingface dataset:
Reconstitution of Reactor Pressure Boundary Components Design Stress Report for License Renewal
Authors: Woods Kirby , Thomas Kenneth
Published: January 2011

The majority of United States Commercial Nuclear Power Plants (CNPP) within the next 10 years will reach the end of their license to operate and can be renewed per the “Atomic Energy Act” of 1954. This act allowed the commission to issue commercial electric power nuclear plants a license to operate (“licensed, but not to exceed 40 years, and maybe renewed upon the expiration of such period, (Chapter 10, Sec. 103(c))).” These CNPP licenses are also governed by the NSSS vendor specification requirements and by the American Society of Mechanical Engineers (ASME) design code standards. This connection is in the form of a stress report that defines the “cyclic life” adequacy criteria for this operational limit of 40 years. The license extension subsequently requires a reconstitution of the initial design stress report input parameters per ASME IWA-4120, 4223 & 4311 (e) for the renewal period extension. This requirement can entail an analysis of the operating conditions and cycles to demonstrate the material elasticity is maintained. The proposed approach for this reconstitution effort was a reanalysis in the form of a study of the Nuclear Steam System Supplier (NSSS) vendors’ original methodology to determine the NSSS vendor specification requirement for ASME code compliance and “cyclic life” adequacy. The information acquired from this evaluation has demonstrated the application to be a complex and simplistic approach. This effort to unravel the composite loading (thermal and pressure transients) condition in relation to specific plant and incident cycles provides both an understanding of component end-of-life limits and supports a comprehensive template for future fatigue life management programs. This paper summarizes this reconstituting effort that utilizes the original vessel stress analysis report to support the license renewal effort, provides a template for future fatigue management programs, demonstrates the conservatism of design, and offers an insight into the design philosophy revealing an elegant process that assures against failures.
The Stress Corrosion Cracking of Austenitic Stainless Steel Heat Exchange Tubes: Three Cases Study
Authors: Xu Shugen , Wang Weiqiang , Liu Huadong
Published: January 2011

In this paper, three leakage failure cases of heat exchange tubes have been introduced. The reasons of the leakage for austenitic stainless steel tubes and overlay welding layer on the tube sheet have been analyzed. Through the investigation of the operation process and histories of the equipment, and after chemical compositions analysis of tube material and corrosion products, metallographic test of specimens with cracks, and fracture surface scan with Scanning Electron Microscope (SEM), the cracking reason and mode are described as the Stress Corrosion Cracking (SCC) of austenitic stainless steel. This kind of cracking in three cases was induced by the micro chloride in the high temperature water (or steam). Moreover, sulfide and dissolved oxygen also reduced the threshold value of chloride concentration and enhanced the corrosion rate for SCC. The cracking mode of Case A and B are transgranular; and Case C is intergranular. It indicates that for this kind of in-service heat exchanger, the operators should not only control the chloride concentration in feed water, but also the sulfide and dissolved oxygen in the future. The austenitic stainless steel tubes (China steel types-1Cr18Ni9Ti and 0Cr18Ni10Ti, equal to Type 304 and Type 321 according to ASME code) used in this cases are not fit to this condition. Thus, for the new heat exchanger design, the tube material should be changed into austenitic-ferritic (duplex phase) steel, such as 2205 Series, which has an excellent performance for SCC resistance in the high temperature water (or steam) with chloride.
Guide for Applying Codes and Standards for Safe Operation and Maintenance of Pressure Equipment in Refinery and Petrochemical Services
Authors: Reynolds John
Published: January 2011

After pressure equipment (aka fixed or static equipment) is designed, fabricated, and constructed to ASME new construction codes and standards (C/S), it is delivered and placed in-service. After that the In-service Inspection (ISI) and Post-Construction C/S begin to govern. Within the ASME, the Post Construction Committee (PCC) produces and maintains the standards that govern equipment after it has been placed inservice. Within the API Standards Organization, the Subcommittee on Inspection (SCI) and the Corrosion and Materials Subcommittee (CMSC) produce and maintain most of the ISI standards and recommended practices that govern pressure equipment in the refining and chemical process industry. This paper shows how many of those ISI and PCC C/S are intended to work together to maintain the safety and reliability of pressure equipment and piping after it has been placed in service. This paper also highlights what’s new with many of these C/S that have been recently updated or newly published. Both the API and ASME use the rigorous, standardized consensus building process outlined by the American National Standards Institute (ANSI) for formulating and publishing their respective C/S. This paper will show how users of these ISI/PCC codes and standards are benefited by the application of the ANSI consensus process.
Verification of the Output Signal Change of Control Rod Position Indicator at the End of Its Life Time
Authors: Yu Je-Yong , Kim Ji-Ho , Huh Hyung , Choi Myong-Hwan , Sohn Dong-Seong
Published: January 2011

The system-integrated reactor currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron-free operation and the use of a nuclear heating for the reactor start-up. The Control Element Drive Mechanism (CEDM) for the integral reactor is designed to raise and lower the control rod in steps of 2mm in order to satisfy the design features of the integral reactor which are the soluble boron-free operation and the use of a nuclear heating for the reactor start-up. The actual position of the control rod could be achieved to sense the magnet connected to the control rod by the position indicator around the upper pressure housing of CEDM. It is sufficient that the actual position information of the control rod at a 20mm interval from the position indicator is used for the core safety analysis. As the magnet moves upward along the position indicator assembly from the bottom to the top in the upper pressure housing, the output voltage increases linearly step-wise at 0.2VDC increments. Between every step, there are transient areas which occur by a contact closing of three reed switches, which is the 2-3-2 contact closing sequence. In this paper the output voltage signal corresponding to the position of the control rod is expected as the 2-1-2 contact closing sequence due to the aging of the magnet at the end of its life time. We verify that the output signal change at the end of the life time of the position indicator could not affect the operational performance of the position indicator. Also it is found that the voltage divider network for the output signal of the position indicator is designed to be robust and reliable during the aging process.
Experience of Xenon Oscillation During an Initial Physics Test of UCN Unit 5 in Korea
Authors: Kim Yong Rae , Choi Tae Young , Shin Sun Ho , Seong Ki Bong
Published: January 2011

Initial core of Ulchin Nuclear Unit 3 (UCN3), which is one of earlier OPR1000 model, was 4 batches and designed as annual cycle after second cycle. The utility requested that UCN Unit 5 (UCN5), which is another of OPR1000 model, had capability of a longer cycle operation from second cycle. Therefore, KNF modified the number of batches from 4 to 3 for OPR1000 initial core, as well as, the number of burnable absorber, and the cutback length of the absorber. However, due to these changes, Xenon oscillation was slightly increased at 100% power during the physics test of UCN5, while that oscillation at 100% power in UCN3 had been gone down without any control rod motion. The xenon oscillation direction is related to axial stability index. The index of UCN3 increased from a slightly negative at BOC to positive at EOC, the index of UCN5 was positive even at BOC, which meant that the core does not go to be stable without the control rod motion. The core of UCN5 became the steady state by the insertion of control rods into the core. To meet the physics test condition, the oscillation was controlled by control rods immediately. After the happening, KNF optimized the cutback length of burnable absorber rods and applied to APR1400, which will keep being stable in xenon oscillation during physics test at the initial cycle.
Effects of Several Factors on the Solution of Failure Probability of Pipes in Nuclear Power Plants
Authors: Park Jai Hak , Lee Jae Bong , Choi Young Hwan
Published: January 2011

A probabilistic integrity assessment program for pipes is developed based on the existing PRAISE program. In the program crack growth due to fatigue loading and stress corrosion can be considered and the probability of failure or leakage of pipes in nuclear power plants can be calculated. Crack growth simulation is performed based on stress intensity factors and failure of a pipe is determined based on J integral or net section yielding. The effects of several important parameters, such as crack depth distribution parameter, operating temperature, and concentration of O2, on the failure and leak probabilities of pipes are examined. And the effects of stress intensity factor and J integral solutions are also examined. It is noted that the failure probability can be changed according to stress intensity factor and J integral solutions.
Establishment of the Advanced Startup Turnover Process for New Nuclear Power Plants
Authors: Kim Il Hwan , Park Sang Chul
Published: January 2011

This paper provides the management and technical issues of a successful turnover process from construction to the nuclear startup organization. The management issues include the development of effective organizations of construction completion group, turnover organization, enhancement of communication between the related groups, the systematic training programs for nurturing startup experts, and also the focused management of essential system turnover for each key event. The paper will also discuss other management issues required for successful systems turnover. Technical issues which will be addressed in this paper for successful turnovers will include the development of computerized Nuclear Power Construction Management System (NPCMS), three dimensional (3-D) CAD program, how to sub-scope the systems and partial turnover to expedite the system turnover, and the issues of maintaining the cleanliness of main control room and electrical equipment rooms.
Improvement of Mechanical Rebar Splices for N.P.P. Construction
Authors: Lee Jin-Gyu , Ju Young-Tae , Choi Jang-Kyu , Kim Sang-Gu
Published: January 2011

All mechanical rebar splices used in domestic nuclear power plant construction sites have been made by foreign manufacturers. Among them, ERICO’s Cadweld (sleeve with ferrous filler metal splices type) has been most commonly used in preceding nuclear power plants owing to the wide field application, but as it is foreign-made product, there is a long lead time (approximately 3∼4 months) to deliver to the designated construction site. Furthermore, it is costly to store surplus products. Despite these demerits, there wasn’t good alternatives besides ERICO’s Cadweld. But, Cold Roll Formed Parallel Threaded type has been developed recently and used in Shin-Wolsong #1, 2.
Further Studies to Evaluate Crack Opening Areas for Through-Wall Cracks in the Vicinity of Welded Attachments
Authors: Sharples J. K. , Madew C. J. , Charles R. , Budden P. J.
Published: January 2011

A paper was presented at the 2009 ASME PVP Conference on evaluating, by finite element techniques, crack opening area (COA) and stress intensity factor, KI, values for through-wall cracks located in the region where an attachment is welded to a plate geometry. Both membrane and bend loads were considered. In addition, based on the stress profile in the un-cracked complex geometry over the region where the cracks would be introduced, COA and KI values were evaluated for the same crack sizes located in a simple plate geometry. This enabled information to be established on the conservatism, or otherwise, of using simple plate solutions to evaluate COA and KI for cracks in the complex geometry. The present paper reports on further studies that have been undertaken to investigate the effect on the previous COA and KI results of considering (i) large displacement theory which may be important for combined membrane and bend loading, and (ii) contact elements in the finite element models since in the previous studies, the mesh was allowed to “overlap on itself” when crack closure was evident due to compressive stresses during bend loading.
A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Pipe Systems
Authors: Leishear Robert A.
Published: January 2011

Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions exist. Pipe ruptures at nuclear facilities were attributed to hydrogen explosions inside pipelines, in nuclear facilities, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents an ignition source for hydrogen was questionable, but these accidents, demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein.
Higher Mode Frequency Effects on Resonance in Machinery Structures
Authors: Leishear Robert A.
Published: January 2011

The complexities of resonance in multi-degree of freedom systems (multi-DOF) may be clarified using graphic presentations. Multi-DOF systems represent actual systems, such as beams or springs, where multiple, higher order, natural frequencies occur. Resonance occurs when a cyclic load is applied to a structure, and the frequency of the applied load equals one of the natural frequencies. Both equations and graphic presentations are available in the literature for single degree of freedom (SDOF) systems, which describe the response of spring-mass-damper systems to harmonically applied, or cyclic, loads. Loads may be forces, moments, or forced displacements applied to one end of a structure. Multi-DOF systems are typically described only by equations in the literature, and while equations certainly permit a case by case analysis for specific conditions, graphs provide an overall comprehension not gleaned from single equations. In fact, this collection of graphed equations provides novel results, which describe the interactions between multiple natural frequencies, as well as a comprehensive description of increased vibrations near resonance.
Prediction of Transient Creep Response Under Combined Primary and Secondary Loading
Authors: Yazdani Nezhad Hamed , O’Dowd Noel P.
Published: January 2011

Residual stress effects on creep deformation and fracture play a significant role in structural integrity assessments of engineering components. The focus of the current work is to investigate creep behaviour of mechanically loaded cracked structures in the presence of residual stress fields. Finite-element analyses have been carried out on single edge notch bend, SEN(B), and tension, SEN(T), specimens at different residual stress and mechanical stress levels. The redistribution time and associated stress relaxation for combined primary and secondary (residual) stresses have been determined (from the finite-element analysis) and interpreted using the transient fracture mechanics parameter, C(t). The observed trends are consistent with earlier studies involving combined thermal and mechanical stress.
Simulation of Fluid Flow Using Reduced-Order Modeling by POD Approach Applied to Academic Cases
Authors: Pomarede Marie , Hamdouni Aziz , Liberge Erwan , Longatte Elisabeth , Sigrist Jean-Franc¸ois
Published: January 2011

Tube bundles in steam boilers of nuclear power plants and nuclear on-board stokehold are known to be exposed to high levels of vibrations. This coupled fluid-structure problem is very complex to numerically set up, because of its three-dimensional characteristics and because of the large number of degrees of freedom involved. A complete numerical resolution of such a problem is currently not viable, all the more so as a precise understanding of this system behaviour needs a large amount of data, obtained by very expensive calculations. We propose here to apply the now classical reduced order method called Proper Orthogonal Decomposition to this case. This choice could lead to reduced calculation times and allow parametrical investigations thanks to a low data quantity. But, it implies several challenges inherent to the fluid-structure characteristic of the problem. Previous works on the dynamic analysis of steam generator tube bundles already provided interesting results in the case of non flowing fluid — i.e. quiescent fluid [J.F. Sigrist, D. Broc; Dynamic Analysis of a Steam Generator Tube Bundle with Fluid-Structure Interaction; Pressure Vessel and Piping, July 27–31, 2008, Chicago]. A first step on the implementation of POD in academic cases (one-dimensional equations, 2D-single tube configuration) is presented. Future work will consist in working on the tube bundle configuration, first in the fixed case and then with structure motion allowed. Present study shows the efficiency of the reduced model to reproduce similar problems from a unique data set for various configurations as well as the efficiency of the reduction for simple cases.
Limit Load Estimate Using Reference Volume Approach
Authors: Gudimetla P. S. Reddy , Seshadri R. , Katna Munaswamy
Published: January 2011

In this paper two novel methods (elastic reference volume method and plastic reference volume method) for reference volume correction while finding out limit loads in the components or structures are presented. These reference volume correction concepts are used in combination with mα-Tangent method to obtain the lower bound limit load of general component or structure.
9975 Shipping Package: Performance of Alternate Materials for Long-Term Storage Application
Authors: Skidmore T. E. , Daugherty W. L. , Hoffman E. N.
Published: January 2011

The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton® GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton® GLT-S O-rings). The shipping packages are part of a long-term storage configuration at the Savannah River Site (SRS). Therefore, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.
Preliminary Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion With Validation From Mechanical Testing
Authors: Crawford Susan L. , Doctor Steven R. , Cinson Anthony D. , Watts Michael W. , Moran Traci L. , Anderson Michael T.
Published: January 2011

Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, are being conducted to evaluate nondestructive examinations (NDE) coupled with mechanical testing of butt fusion joints in high density polyethylene (HDPE) pipe for assessing lack of fusion. The work provides information to the United States Nuclear Regulatory Commission (NRC) on the effectiveness of volumetric inspection techniques of HDPE butt fusion joints in Section III, Division 1, Class 3, buried piping systems in nuclear power plants. This paper describes results from preliminary assessments using ultrasonic and microwave nondestructive techniques and mechanical testing with the high speed tensile impact test and the bend test for determining joint integrity. A series of butt joints were fabricated in 3408, 12 inch (30.5 cm) IPS DR-11 HDPE material by varying the fusion parameters to create good joints and joints containing a range of lack of fusion conditions. Six of these butt joints were volumetrically examined with time of flight diffraction (TOFD), phased array (PA) ultrasound, and the Evisive microwave system. The outer diameter (OD) weld beads were removed for microwave evaluation and the pipes ultrasonically re-evaluated. In two of the six pipes both the outer and inner diameter (ID) weld beads were removed and the pipe joints re-evaluated. Several of the pipes were sectioned and the joints destructively evaluated with the following techniques: high speed tensile test, bend test, and focused immersion ultrasound on a joint section removed from the pipe coupled with slicing through the joint and examining the revealed surfaces. The fusion parameters, nondestructive, and destructive evaluation results will be correlated to validate the effectiveness of what each NDE technology detects and what each does not detect. This is an initial limited study which will aid in identifying key future work.
Potential Difference Around Semi-Elliptical Crack on the Outer Surface of Pipe for Non-Destructive Evaluation of Crack by DC-PDM
Authors: Tada Naoya , Uchida Makoto , Maeda Hiroki
Published: January 2011

Remaining life evaluation has been attracting much attention to secure the safety of aged power generating plants. A crack on the outer surface of pipe is a typical kind of damage observed in aged plants. Accurate information of crack largely helps the remaining life evaluation based on fracture mechanics. The direct-current potential difference method (DC-PDM) is known as one of the suitable non-destructive techniques to monitor the initiation and growth of crack in the pipes. Usually, numerical analyses are necessary to evaluate the geometry of crack from the potential difference because no analytical solution was given in most cases. In this study, a solution of the potential difference around a semi-elliptical crack on the outer surface of pipe is presented and the result is compared with that obtained by finite element analysis.
Piezoelectric Wafer Active Sensors for Structural Health Monitoring: State of the Art and Future Directions
Authors: Giurgiutiu Victor
Published: January 2011

Piezoelectric wafer active sensors (PWAS) are lightweight and inexpensive enablers for a large class of structural health monitoring (SHM) applications. The presentation will start with a brief review of PWAS physical principles and basic modeling. Then, the presentation will consider the several ways in which PWAS can be used for damage detection: (a) embedded guided-wave ultrasonics, i.e., pitch-catch, pulse-echo, phased arrays, thickness mode; (b) high-frequency modal sensing, i.e., the electro-mechanical (E/M) impedance method; (c) passive detection, i.e., acoustic emission and impact detection. Emphasis will be placed on recent developments. Special attention will be given to the mechatronics predictive modeling of the complete cycle from electrical excitation through piezoelectric transduction, ultrasonic acoustic waves, and finally reverse piezoelectric transduction to generate the received electric signal. Power and energy aspects of this process will be considered and discussed. The presentation will end with conclusions and suggestions for further work.
IASCC Evaluation Method for Irradiated Core Internal Structures in BWR Power Plants
Authors: Takakura Kenichi , Tanaka Shigeaki , Nakamura Tomomi , Chatani Kazuhiro , Kaji Yoshiyuki
Published: January 2011

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research & development study for the aging management & maintenance of the nuclear power plants. Based on the JNES project results, JNES proposed “IASCC evaluation guide for BWR core internals”. The purpose of this paper is to describe the background of the guide, especially crack growth rate (CGR) tests for irradiated stainless steels. The CGR tests had been carrying out with neutron irradiated compact tension (CT) specimens under constant load and crack growth was measured using the reversing dc potential drop (DCPD). Irradiation had been conducting in the core region of the Japan Material Testing Reactor (JMTR) in simulated BWR water environments. The specimens were irradiated to fluence ranging from 5×1024 n/m2 (0.7 dpa) to 1 × 1026n/m2 (E>1MeV)(14 dpa). CGR of SUS304L and SUS316L were formulated as a function of fluence and stress intensity factor. Outline of IASCC evaluation guide for irradiated core internals was described in this paper. The method requires time to time evaluation, that is, residual stress and CGR are calculated for the neutron fluence of the evaluating time, and then, structural integrity of core internal is evaluated by fracture mechanics analysis. The guide is tentative one and going to be modified by reflecting new knowledge and discussion.
A Study on Improvement Through Analysis of Cold Hydrostatic Test Cases in Korea Nuclear Power Plants
Authors: Song Ki Sang , Chae Kyeong Sik
Published: January 2011

The objective of this study is to analyze the effectiveness of the Cold Hydrostatic Test (CHT) process and determine the optimum method of completing a CHT through case studies in the Korea nuclear power plants. In this study, all the 9 CHT cases, performed for the past sixteen years (1993 to 2009) in Korea nuclear power plants, will be examined and evaluated. There are twenty (20) operating Units and eight (8) Units under construction at 3 nuclear facility sites in Korea. Among the 20 Units, only 4 Units at the Wolsong site are pressurized heavy water reactors (PHWR), the others are pressurized light water reactors (PWR). CHT is based on the requirements of ASME NB-6200 & NC-6220. CHT is a mandatory test to verify integrity of weld points and interfaces associated with the equipment and pipes of the Reactor Coolant System (RCS) pressure boundary. The design pressure of the RCS is 2,500psia (175.8 kg/cm2 a). The major steps of test sequence of a CHT is RCS filling, venting, heat up, pressurization and inspection. Reactor Coolant Pump (RCP) operation is utilized as thermal input to raise RCS temperature over 120 °C. The Chemical and Volume Control System (CVCS) Charging Pump, or temporary hydro pump is used to pressurize the RCS. CHT requires pressure to be raised and maintained more than 10 minutes at 1.25 times of design pressure, and then be depressurized and inspected at the design pressure of 2,500psia (175.8 kg/cm2 a). According to the analyzed results of the CHT cases, all CHTs were successfully conducted but there are several items which need to revised and modified for increased effectiveness of the CHT. These items include pressurizer manway gasket leakage, improper process of the procedure and others. In conclusion, the results of this study will be used to prevent similar errors and improve the effectiveness of the CHT for future nuclear power plants projects in Korea.
Safety Classification Informed Structural Integrity Assessment of Nuclear Plant Components
Authors: Bell C. D. , Watson C. T. , Howson J. , Walker A. E.
Published: January 2011

A methodology has been developed for establishing the structural integrity demonstration requirements for nuclear plant components. The methodology determines the safety classification of components based upon the consequences of loss of safety functions, which together with an assessment of the potential for failure enables risk categories to be assigned to the components. The appropriate structural integrity assessment and manufacturing inspection requirements are then determined based upon the component risk categories. Adherence to the new methodology, which is supplemental to the requirements of Section III of the ASME Boiler and Pressure Vessel Code, provides, where appropriate, additional assurance of component structural integrity that meets with the UK safety targets. This paper provides the background and an overview of the methodology. Examples of the application of the methodology to nuclear plant components are also included in the appendices of the paper.
Measurement of Residual Stresses in Thick Section Steel Electron Beam Welds
Authors: Kingston E. , Smith D. J. , Zheng G. , Gill C. , Hurrell P.
Published: January 2011

Residual stresses were measured and predicted in steel electron beam welded samples, one supplied by Rolls-Royce, UK, and a second by CEA, France. In the case of the Rolls-Royce sample, measurements were made before and after post-weld heat treatment. The deep hole drilling technique was applied to obtain measurements across and through the EB welds. It is shown that the tensile residual stresses are confined to a region either side of the EB weld that corresponds to about 20% of the wall thickness. Post-weld heat treatment is shown to substantially reduce the magnitudes of the peak residual stresses. Comparisons between measurements and predicted residual stresses in the Rolls-Royce sample show that predicted hoop stresses are substantially greater than measured although there is a fair comparison between measured and predicted radial stresses.
Investigation of Transformation Induced Plasticity and Residual Stress Analysis in Stainless Steel Welds
Authors: Dai H. , Moat R. , Mark A. F. , Withers P. J.
Published: January 2011

The aim of this paper is to investigate the implications for weld residual stresses of martensitic transformation induced plasticity (TRIP) in stainless steel filler metal. The TRIP strains occurring during cooling under different uniaxial load levels have been obtained using digital image correlation (DIC) for a residual stress relieving low transformation temperature weld filler known to show little variant selection on cooling as a function of stress. In order to investigate the efficacy of current FE transformation plasticity models of different levels of sophistication in simulating TRIP strains, a finite element model, incorporating the so-called Greenwood-Johnson effect was used to simulate these constrained dilatometry measurements. To assess the implications of the different approaches to modelling TRIP for weld residual stresses, the TRIP coefficients determined from the above experiments were incorporated into an FE model simulating the residual stresses that are generated when a single weld bead is deposited on to a stainless steel base plate. It was found that including TRIP had a significant influence on the weld stresses, while the differences between the models were much smaller.
Challenges Associated With License Renewal Fatigue Evaluations for an Older-Vintage BWR Plant
Authors: Novotny Tyler D. , Herrmann Terry J. , Mileris George V. , Pace Raymond M.
Published: January 2011

As plants prepare for license renewal, extending their original operating licenses from 40 years to 60 years of plant operation, various tasks are required to be performed. Some of the required tasks include validation of the current fatigue evaluations for ASME Class 1 components, evaluation of environmental affects for a subset of those components, and calculation of revised component life to demonstrate acceptability for the entire license renewal period. Validation of the current fatigue evaluations is a significant challenge for older-vintage plants because the documentation of transient parameters and event severity was often not rigorous, especially during the early years of plant operation. This is primarily because historical data gathered in the early years was difficult to obtain before the implementation of modern computer systems and was therefore only collected at limiting times and locations. As a result, transients were typically considered to be full design basis severity events. Also, the design basis transient definitions, as defined in the original Final Safety Analysis Report (FSAR), were limited to a simpler, bounding set to help simplify the process. Whereas these simplifications are typically conservative and make it easier for the staff implementing plant procedures, they can challenge the ability of utilities to be able to demonstrate long-term acceptability for license renewal where additional margins are required. The purpose of this paper is to provide an example of the efforts expended on this issue for the Pilgrim Nuclear Power Station (PNPS). The intent is to educate the reader on the efforts involved, the vast array of input that is needed, and the possible hurdles that may be encountered for the validation and re-evaluation of Class 1 component fatigue for license renewal. Prior to starting work on the fatigue analysis, PNPS requested that SI develop a design specification for the fatigue reevaluation project. The importance of this document is that it maintained consistency and documented methods to be used along with design inputs.
Development of Burn Test Specifications for Fire Protection Materials in RAM Packages
Authors: Smith Allen C. , Gupta Narendra K. , Blanton Paul S.
Published: January 2011

The regulations in 10 CFR Part 71 require that radioactive material (RAM) packages must be able to withstand specific fire conditions, given in 10 CFR 71.73, during Hypothetical Accident Conditions (HAC). This requirement is normally satisfied by extensive testing of full scale test specimens under required test conditions. Since fire test planning and execution is expensive and only provides a single snapshot into a package performance, every effort is made to minimize testing and supplement the tests with results from computational thermal models. However, the accuracy of such thermal models depends heavily on the thermal properties of the fire insulating materials, properties that are rarely available at the regulatory fire temperatures. To the best of the authors’ knowledge, no test standards exist that could be used to test the insulating materials and derive their thermal properties for the RAM package design. This paper presents a review of the existing industry fire testing standards and proposes testing methods that could serve as a standardized specification for testing fire insulating materials for use in RAM packages.
A Process to Consider Practical Applications of Alternative Assembly Procedures of ASME PCC-1
Authors: Bausman Anita R. , Waterland A. Fitzgerald Jerry
Published: January 2011

This paper explores a process to consider whether or not to use alternative assembly procedures such as those contained in Appendix F of the latest ASME Post Construction Code-1, Guidelines for Pressure Boundary Bolted Flange Joint Assembly [1]. The new edition of ASME PCC-1-2010 is greatly expanded, and Appendix F contains five assembly methods that vary from the traditional “Legacy Pattern” of incremental torque application in a “star” pattern up to a pre-selected target torque which is then followed by a circular pass (or passes) at a target torque. What reasons drive an end-user to use something besides the “tried and true” Legacy Pattern? What issues does the user need to consider? Which of the alternative assembly methods might work best for an end-user, and why? Safety, performance, and economics all must be balanced when evaluating relative risk. Given that risks are considered and managed appropriately — significant benefit is available to the end-user who uses a more stream-lined assembly method on selected applications.
Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant
Authors: Jung Se-Kwon , Goodman Adam , Harrold Joe , Alchaar Nawar
Published: January 2011

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.
Development of ASME Section X Code Rules for High Pressure Composite Hydrogen Pressure Vessels With Non-Load Sharing Liners
Authors: Newhouse Norman L. , Rawls George B. , Rana Mahendra D. , Shelley Bernard F. , Gorman Michael R.
Published: January 2011

The Boiler and Pressure Vessel Project Team on Hydrogen Tanks was formed in 2004 to develop Code rules to address the various needs that had been identified for the design and construction of up to 15,000 psi hydrogen storage vessel. One of these needs was the development of Code rules for high pressure composite vessels with non-load sharing liners for stationary applications. In 2009, ASME approved new Appendix 8, for Section X Code which contains the rules for these vessels. These vessels are designated as Class III vessels with design pressure ranging from 20.7 MPa (3,000 psi) to 103.4 MPa (15,000 psi) and maximum allowable outside liner diameter of 2.54 m (100 inches). The maximum design life of these vessels is limited to 20 years. Design, fabrication, and examination requirements have been specified, included Acoustic Emission testing at time of manufacture. The Code rules include the design qualification testing of prototype vessels. Qualification includes proof, expansion, burst, cyclic fatigue, creep, flaw, permeability, torque, penetration, and environmental testing.
A Hybrid Method to Develop SSI Parameters for Rigid Embedded Foundations of Arbitrary Shape
Authors: Johnson James J. , Maslenikov Oleg R. , Nakaki David K. , Hashimoto Philip S. , Bayraktarli Yahya , Zuchuat Olivier
Published: January 2011

CLASSI and SASSI are two sub-structuring approaches to analyze the phenomena of soil-structure interaction (SSI) with extensive applications in the nuclear power industry. Many similarities exist in the two methods; both approaches treat the SSI problem as a linear problem (soil and structure are modeled as behaving linearly), solve the problem in the frequency domain, and treat complex free-field wave propagation mechanisms, including incoherence of ground motion. SASSI has distinct advantages in modeling embedded foundations of irregular shapes, including foundation flexibility. Standard versions of CLASSI treat the foundation of the structures as behaving rigidly with respect to earthquake motions and the calculation of overall response. Structures are modeled in CLASSI using their fixed-base eigen-system. This permits very detailed structure models of 100,000s degrees of freedom represented by 1000s of fixed-base modes to be incorporated directly into the SSI analysis. Other advantages of CLASSI are the ability to interrogate sub-structure elements of the problem efficiently, and the ability to cost-effectively perform probabilistic response analyses of soil-structure systems. The limitation of CLASSI is its inability to treat embedded foundations of complex geometry and the assumption of rigid foundation behavior. This paper summarizes the general CLASSI approach to the SSI problem. Next, a Hybrid approach to developing SSI parameters is presented, whereby the advantages of SASSI for foundation modeling are used in the development of foundation impedances and scattering functions for embedded foundations and analysis in CLASSI. The computer program RIGID is validated and verified against published results. An application to the modeling of the SSI parameters for the Mu¨hleberg Nuclear Power Plant Reactor and SUSAN Buildings is presented.
Vibration Tests of 3-Story Benchmark Structure With Three Semiactive Dampers
Authors: Matsuoka Taichi , Sunakoda Katsuaki , Hiramoto Kazuhiko , Yamazaki Issei , Fukukita Akira , Loh Chin-Hsiung , Lin Pei-Yang
Published: January 2011

In a previous paper the authors developed a semiactive damper that generates electrical power, and carried out vibration tests using a 3-story benchmark structure at NCREE, Taiwan in 2006. At that time, the dampers were installed at 1st and 2nd floors. The damper has a large inertia mass by flywheel and controllable damping force by generator, and a load capacity of 30 kN. In the test, the damper at 1st floor was only controlled by Bang-bang control that was based on Lyapunov function. In this paper as the next step, in order to demonstrate more effectiveness of vibration suppression, the dampers are installed at all floors of the 3-story structure, and vibration tests using the same structure are carried out again in 2008. The control law which is proposed here is based on Lyapunov function or predictive switching control for all of damper at each floor. The structure has 9 m high, 3 m wide, 2 m span, 18 tons total weight. Several earthquake waves normalized to be 150 gal are inputted horizontally to the base by a shaking table, and the seismic responses of each floor are estimated quantitatively. It is obviously from the experimental results that the seismic reductions for acceleration and displacement in case of large inertia mass are much better than the case of small one, but control effect is decreasing. Because seismic reduction can be depended on a balance between inertial force and controllable damping force, and plus the one of the reason is time delay of the damper. At last, we can summarize that the semiactive damper is available for seismic isolation in practical use.
Advanced Construction Methods for New Nuclear Power Plants
Authors: Jung Dae-Yul , Kang Yoon-Kee , You Chang-Hyung
Published: January 2011

This paper covers advanced construction technologies that are generally used for nuclear power plants and presents advanced construction methods that can contribute to an efficient and short construction schedule. The construction schedule is driven by the activities of the critical path. The advanced construction methods consist of Modularization, Improvement of Mechanical Rebar splices, Application of 3D-CAD system for information and control, and the installation of RVI and RCL at the same time. Some methods have been applied in actual Nuclear power projects and others have been developing under the research and development program. It incorporates the experiences and insights from recent nuclear construction projects all over the world.
Fast Neutron Reduction Techniques for Plant Lifetime Extension
Authors: Yoo Choon Sung , Kim Byoung Chul , Kwon Tae Je
Published: January 2011

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.
An Estimation of C(t) of Transient Creep Under Combined Mechanical and Thermal Stresses
Authors: Oh Chang-Young , Song Tae-Kwang , Kim Yun-Jae
Published: January 2011

In this paper, a prediction of C(t), a crack tip parameter for the transient creep condition, for the circumferential cracked pipe under combined mechanical and thermal stresses are presented. The estimation formulae for C(t)-integral of the cracked component operating under mechanical load alone have been provided for decades [1–6]. However, high temperature structures usually work under combined mechanical and thermal load. And the interactions between mechanical and thermal loads make the relaxation rate different from those produced under mechanical load alone. In this study, 3-dimensional finite element analyses are conducted to calculate the C(t)-integral under combined mechanical and thermal load. As a result, redistribution time for the crack under combined mechanical and thermal load is re-defined to quantify the C(t)-integral. The estimation of C(t)-integral using this proposed redistribution time agrees well with FE results.
Effective Mass of Fluid for Rocking-Bulging Interaction of Rigid Rectangular Tank Whose Bottom Plate Rectilinearly Uplifts
Authors: Taniguchi Tomoyo
Published: January 2011

In experimental and analytical studies of the rocking response of a circular cylindrical tank under the action of the purely horizontal and translational ground motion, the author analogically quantified the mass of fluid contributing to both bulging and rocking motion of the tank. It was called “the effective mass of fluid for the rocking-bulging interaction.” Its dynamical role in the rocking motion of the tank was thoroughly investigated. However, applying it to design process requires us to use its rigorous definition. To date, the fluid pressure on the tank induced by the impulsive (= bulging) motion and the rocking motion and their effective masses of fluid for each motion were mathematically defined, respectively. Therefore, this paper tries to define the effective mass of fluid for the rocking-bulging interaction based on the fluid pressure on the tank mathematically. The effective mass of fluid for the rocking-bulging interaction is understood as a part of the effective mass of fluid for the bulging motion that is also under the action of the rotational inertia. The influence of the rotational inertia on the effective mass of fluid for the bulging motion is measured by a ratio of the apparent density of fluid contributing to the rocking motion to the original density of fluid. The distribution of the apparent density of fluid contributing to the rocking-bulging interaction is drawn for the various aspects of tanks. The ratio of the effective mass of fluid for the rocking-bulging interaction to the total mass of fluid of the tank is given as the function of the aspect ratio of the tank and the ratio of the uplift width of the tank bottom.