bluuebunny/crossref_metadata_2025 huggingface dataset:
Design of Skirt to Cone Joint in Coke Drum: A Parametric Approach Based on Fatigue Analysis
Authors: Panwala M. Sohel M. , Jaiswal S. L.
Published: May 2012

Coke Drums are critical equipment in refineries due to variable temperature and pressure. The temperature is highly fluctuating which causes the thermal fatigue in coke drums. The phenomenon of this thermal fatigue plays a very crucial role in failure of coke drums. In the present study, a coke drum is subjected to pressure - temperature reversal with each cycle of 48 hours duration. Temperature and pressure varies from 338 to 738 K and 0 to 0.404 MPa, respectively. The material of construction is 1.25Cr-0.5Mo (SA-387 Gr. 11 Cl. 1). This coke drum is at higher temperature of 738 K for 24 hours in one full cycle. The skirt to shell and cone junction is a critical portion of coke drum because it is highly susceptible to fatigue at higher temperature. In addition to this, the skirt is provided with slots at specific pitch all around circumference to induce flexibility. This has great influence on fatigue life due to localized stress at higher temperature. API 579-1/ASME FFS-1 2007 is used for the fatigue life evaluation in the present case. Nonlinear transient thermal analysis is coupled with the elastic-plastic structural analysis for calculation of stresses and strains. These stresses are further used for deducing permissible cycles for fatigue as per the approach based on API 579-1. A Parametric study has been carried out in Finite Element Analysis tool ‘ANSYS-12.1’ for different configuration of skirt to cone junction specifically in terms of inside crotch radius. The effect on fatigue life has been studied with variation in inside crotch radius. It has been found that fatigue life at skirt to cone junction is increasing with increase in crotch radius. But, this increase in crotch radius is adversely effects on slot tip where the life is drastically reducing. Such parametric study can be considered for selection of inside crotch radius for given skirt thickness and diameter.
Internal Stress Generation in Austentic Stainless Steels During Creep Deformation
Authors: Rao Ashwin , Chong Kok Boon , Bouchard P. John , Fitzpatrick Michael E.
Published: May 2012

It has been well established in materials such as austenitic steels and aluminium alloys that plastic strain leads to generation of internal stresses. A number of intergranular factors such as the anisotropic stiffness and yield behaviour of a single crystal, orientation of grain families to the loading direction, and the constraint each grain places on its neighbours are responsible for creating these stresses. The presence of these accumulated internal stresses in power plant components is important because they interact with the applied external stress and play a critical role in the initiation and development of material degradation that may lead to eventual failure. This study focuses on measuring the generation of internal (intergranular) strains and stresses in austenitic stainless steels subjected to creep deformation. Creep processes increase the overall inelastic strain in a material and this correspondingly alters the internal strain state. A combination of in-situ and static neutron diffraction measurements was conducted to assess the internal strain generation through a material’s creep life. These experiments have revealed similarities between internal strain generation during primary creep deformation and that during monotonic tensile deformation leading to the conclusion that common mechanisms may be responsible. Such studies are important in the current context when a number of power plants are being life extended. Components in high temperature service applications undergo a number of creep-fatigue cycles during their operation. It is vital to have accurate and robust life assessment procedures that can take account of long-term internal strain evolution effects to maintain economic but safe operation and avoid costly repairs or replacement.
Relationship Between Cross Section of Test Piece and Input Energy for Failure
Authors: Minagawa Keisuke , Fujita Satoshi , Kitamura Seiji , Watakabe Tomoyoshi
Published: May 2012

The 2007 Niigataken Chuetsu-oki Earthquake damaged a nuclear power plant. Some of the plant’s structure sustained plastic damage although fracture was avoided. Since accumulation of this damage induces fracture, cumulative damage and a margin for fracture should be clarified. Therefore a method which is able to evaluate cumulative damage is required. The authors have dealt with this issue from the viewpoint of the energy balance equation. The energy balance equation is one of the valid methods for structural calculation, and it indicates cumulative information of motion. The authors have already confirmed a qualitative relationship between the energy and fatigue failure, and agreement between the energy based method and conventional fatigue methods. However, clarifying the quantitative relationship is required for application to general structures. Previously, we investigated the influence of test piece width to energy for failure. In this paper, influence of cross section of test piece to energy for failure is investigated by vibration experiments. Authors focused on second moment of area and section modulus as parameter of cross section. Five types of test piece were made for vibration experiment. One is a basic model. The others are a model which has same second moment of area as the basic model and a model which has same section modulus as the basic model. Then, vibration experiments which caused fatigue failures were conducted on the experiment models. After that, input energy for failure of each test piece was compared. As a result of the experiment, it was confirmed that input energy for failure was same if section modulus was same.
Design by Finite Element Analysis on Tubesheet of Heat Exchanger With a Central Hole
Authors: Li Guo , Zhao Jianping
Published: May 2012

As one of major components of heat exchanger, tubesheet is of paramount importance to enhance the safety of heat exchanger whether its design is reasonable. As the diversity of the heat exchangers’ operating condition, structures of tubesheets also become very special. A fixed tubesheet heat exchanger with a central pipe whose diameter is greater than other heat exchanger tubes is presented in this paper. A central hole in tubesheet will weaken the stiffness and intensity of the tubesheet and cause local stress concentration along the central hole’s edge. However, there is no design method for this kind of tubesheet with a central hole and it is unable to be calculated using various international standards available. In this paper, finite element analysis code ANSYS is used to simulate the real complex structure, real loads and boundary conditions of the tubesheet so that the design problem can be visualized. The three-dimensional finite element model of the tubesheet is built considering the influence of tube channel, partial shell and heat transfer tube bundle. The different pressure in shell-side and tube-side will cause the mechanical stress. The temperature gradient exists widely and the tubesheet, shell as well as heat transfer tubes can’t transform freely. Therefore, there may exists high thermal stress due to the high temperature difference in shell-side and tube-side. The thermal stress has great impact on the total stress distribution. So the simulation of the temperature field is very important. By means of thermal analysis coupling with structure analysis, the distribution of temperature, stress and deformation is obtained. Through evaluating the stress intensity of the tubesheet, it is found that the dangerous region is located at the edge of the tube distribution region and local stress concentration along the edge of the central hole is not obvious. The result shows that the tubesheet is appropriately designed and the design by finite element analysis method is feasible. This paper provides a solution for this kind of structure’s analysis design in engineering application.
The Stability of Fine, Sub-Grain Microstructures Within Carbon Depleted Regions of Dissimilar Metal, Ferritic, Creep Resistant Welds
Authors: Dawson Karl , Tatlock Gordon J.
Published: May 2012

The duration of post weld heat treatments (PWHT) applied to thick section multi-pass dissimilar metal welds (DMW), involving ferritic creep resistant steels of differing chromium content, are shown to have a considerable impact on the performance of the welded joint. Welding consumables of alloy types P22 and P24 have been used to form joints with P91 base alloy which were subsequently post weld heat treated for varying durations. High resolution transmission electron microscopy (TEM) has been exploited in the characterisation of precipitation in the weld material and the heat affected zone. It has been shown that uphill diffusion of carbon from the low to the higher alloy material during PWHT and creep test conditions occurs in all specimens. Selected area diffraction (SAD) and convergent beam electron diffraction (CBED) studies of carbon extraction replicas reveal extensive dissolution of M23C6 and M7C3 carbides in the decarburised zone of the weld alloy subsequent to post weld heat treatments. However, welds completed using Nb and V containing consumables retain a fine distribution of MX precipitation in the carbon depleted regions after PWHT. The retention of these microstructure stabilising carbonitrides facilitates the preservation of an ultra fine sub-grain microstructure, thus avoiding recrytallisation which is invariably observed in post weld heat treated P22:P91 DMWs. Cell size comparisons of the sub-grain microstructures have been investigated utilising channelling contrast back scattered scanning electron images of as welded and post weld heat treated material.
Long-Term and High Throughput Operation of a Detonation Chamber by Utilizing Shakedown Phenomenon Under Initial Overload in Plastic Range
Authors: Asahina Joseph K. , Rodriguez Edward A. , Nickell Robert E. , Shirakura Takao
Published: May 2012

ASME published Code Case 2564 [1] in January 2008 on impulsively loaded vessels, which is currently the first design guidance of this kind worldwide. The authors participated in the Task Group on Impulsively Loaded Vessels reporting to Sec. VIII, Div. 3 [2] and were also intimately engaged in the development, design and operation of the DAVINCH® detonation chambers. DAVINCH® is an acronym for Detonation of Ammunition in a Vacuum INtegrated CHamber, utilized for destruction of old non-stockpile chemical weapons, which were recovered from underground and sea in several areas of Japan, China, and Belgium. The quantity destroyed to-date has reached as high as 7,000 munitions.
Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping
Authors: Kalyanam S. , Shim D.-J. , Krishnaswamy P. , Hioe Y.
Published: May 2012

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.
Experimental Study on the Flaw Characterization of Multiple Flaws for Ductile Fracture
Authors: Iwamatsu Fuminori , Miyazaki Katsumasa , Saito Koichi , Hamanaka Tetsuya , Takahashi Yoshiaki
Published: May 2012

Flaw characterization of multiple flaws for ductile fractures was verified by comparing results of flat plate fracture tests and flaw combination rules in typical fitness-for-service codes. Single or double coplanar surface flaws were introduced into flat plate specimens made of Type 304 stainless steel using an electric discharge machine. Fracture tests of sixteen flat plates with flaws were conducted using a tensile testing machine. According in results of the fracture tests, collapse behavior for flat plates with double coplanar flaws was divided into two broad categories. One was coalescence behavior, in which the double flaws combined before the wall was penetrated. The other was penetration behavior, in which the wall was penetrated before coalescence. Rules for characterizing multiple flaws in fitness-for-service codes in Japan, the United States, and Europe were surveyed in order to compare them with test results. When the multiple flaws satisfy the criteria of rules in fitness-for service codes, a single combined flaw is assumed to conservatively estimate the limit load analysis for a ductile fracture. Flaw combination rules in typical fitness-for-service codes were found to conservatively estimate ductile fractures.
Estimation of Maximum Load for Pipes With Multiple Circumferential Flaws by Limit Load Analysis
Authors: Iwamatsu Fuminori , Miyazaki Katsumasa , Saito Koichi , Hamanaka Tetsuya , Takahashi Yoshiaki
Published: May 2012

Fully plastic failure stress for a single circumferential flaw in a pipe is evaluated by the limit load criteria in accordance with Appendix C in the ASME Code Section XI and Appendix E-8 in the JSME S NA-1-2004. However, multiple flaws such as stress corrosion cracking are frequently detected in the same circumferential cross section in a pipe. Limit load analysis has been validated for pipes with multiple circumferential flaws. Quasi-static four-point bending tests were performed on Type 304 stainless steel pipes with single, double, or triple circumferential flaws. Maximum loads measured in these tests were estimated by limit load analysis for pipes with multiple circumferential flaws. All estimation results using flow stress defined by the JSME S NA-1-2008 are conservative compared to the test results. Estimation results using flow stresses obtained from tests for the pipe with a single flaw quantitatively agree with test results.
Durability Methodologies for Material Systems
Authors: Reifsnider Kenneth , Raihan Rassel , Majumdar Prasun
Published: May 2012

In the present work, we are concerned with heterogeneous material systems that have multiple distinct materials and void phases, and associated interfaces, for which the geometric scale and specific morphology at the micro/nano-level play an essential role in the global properties, functional behavior, and material system performance. Such heterogeneous combinations of materials that are engineered to function together to create unique properties and performance that is a function of their interaction are called material systems. Engineered heterogeneous material systems are at the heart of revolutionary advances in devices that convert and store energy (e.g., batteries, fuel cells, solar cells, capacitors, and many electro-optical devices), and other membrane-based devices used in chemical and fuel processing, sequestration, and extraction. In recent years, advances in a variety of additive manufacturing methods and techniques have made it possible to design, control, and fabricate specific micro- or nano-structures to achieve prescriptive functional performance of the material systems and devices in which they appear. However, systematic multiphysics analysis methods properly set on field equations that represent the local details are not available, so that first-principles understandings and designs of those materials are not properly founded. Recently, the DoE established an Energy Frontiers Research Center for Physics Based Nanostructure Design and Fabrication of Heterogeneous Functional Materials, called the HeteroFoaM Center, to address this and related questions. The present paper presents some initial findings of part of that effort related specifically to durability. For the present study, our approach to the question of durability of engineered material systems will be to construct a damage model, wherein damage is defined as (and by) changes in material state as a function of some generalized time variable that defines the intensity and history of the applied conditions that drive those changes. To bring the reality of engineering practice to our discussion, the durability concepts will be related to the damage and functional degradation observed in solid oxide fuel cells (SOFCs) during service. SOFCs convert the chemical energy of fuel to electrical power. This focus will serve to define the scope of our discussion, which will be the durability of complex, heterogeneous material systems as measured by degradation of their functional performance defined by mechanical, thermal, and electrical behavior.
A Numerical Shape Optimisation Method for Blowing Glass Bottles
Authors: Groot J. A. W. M. , Giannopapa C. G. , Mattheij R. M. M.
Published: May 2012

Industrial glass blowing is an essential stage of manufacturing hollow glass containers, e.g. bottles, jars. A glass preform is brought into a mould and inflated with compressed air until it reaches the mould shape. A simulation model for blowing glass containers based on finite element methods, which adopts a level set method to track the glass-air interfaces, has previously been developed [1–3]. A considerable challenge in glass blowing is the inverse problem: to determine an optimal preform from the desired container shape. In previous work of the authors [4, 5] a numerical method was introduced for optimising the shape of the preform. The optimisation method described the shape of the preform by parametric curves, e.g. Bezier-curves or splines, and employed a modified Levenberg-Marquardt algorithm to find the optimal positions of the control points of the curves. A combined finite difference and Broyden method was used to compute the Jacobian of the residual with respect to changes in the positions of the control points. The objective of this paper is to perform an error analysis of the optimisation method previously introduced and to improve its accuracy and performance. The improved optimisation method is applied to modelled containers of industrial relevance, which shows its usefulness for practical applications.
Steady-State Positions of a Pop Action Safety Valve
Authors: Rit Jean-Franc¸ois , Moussou Pierre , Teygeman Christophe
Published: May 2012

Pressure relief valves in water pipes are known to sometimes chatter when the inlet pressure slightly exceeds the maximum allowable working pressure (MAWP) value. Though these devices are responsible for numerous fatigue issues in process industries, there is a relatively low number of technical publications describing well-established facts about them, especially for heavy fluids. The present study deals with the investigation of the stability of a pressure relief valve when a pressure drop device is arranged upstream. The valve is a simple spring device, with a 1″1/2 inlet diameter and a set pressure equal to 3 MPa. The 12% to 66% range of relative opening for this valve exhibit an unstable static equilibrium of the plug, designed to achieve the so called “pop action”; as soon as the pressure set point is reached, a runaway process leads to the full opening. The statically stable regimes were observed in former studies with respect to the upstream pressure and to the plug position, with a test rig arrangement which ensured an almost constant pressure upstream. In the present study, high pressure drop devices are arranged upstream, in order to stabilize the hydraulic regimes. It is found that the upstream pressure drop devices significantly enlarge the range of steady state plug positions and upstream pressures. Pressure and plug position measurements are shown with a time resolution lower than 2 ms. Comparison with hydraulic regimes of the former studies indicate that the presence of an upstream pressure drop modifies the plug balance. It is proposed that the arrangement of pressure drop device upstream may significantly reduce the risk of valve instability in water pipes.
Implementation of the ASME Code and NQA-1 in U.S. Department of Energy Packaging Certification Program Training Courses
Authors: Gelder Lawrence F.
Published: May 2012

Under the authorization of the Department of Transportation, per 49 CFR Part 173.7(d), Type B and fissile radioactive materials packagings made by or under the direction of the U.S. Department of Energy (DOE) may be used for the transportation of Class 7 materials when evaluated, approved, and certified by DOE against packaging standards equivalent to those specified in 10 CFR Part 71. The DOE certificate is issued on the basis of a safety analysis report of the package design and application. The applicant must demonstrate to DOE the package meets the standards in the 10 CFR Part 71. Since the Type B and fissile radioactive materials packaging standards specified in 10 CFR Part 71 are performance based standards, guides and other tools are necessary to demonstrate how a package design meets the standards. Two essential tools used by packaging applicants and reviewers to quantify and demonstrate compliance with the safety standards/requirements of the CFR are the ASME Boiler and Pressure Vessel (B&PV) Code and ASME NQA-1. The DOE Packaging Certification Program develops and sponsors training courses for packaging applicants and reviewers. Many of these courses are required training by DOE for persons that manage or prepare safety analysis reports for package designs (i.e., applications) submitted to the DOE for certification. The ASME B&PV Code and NQA-1 are ubiquitous in the DOE core training courses. This paper provides an overview how the ASME B&PV Code and NQA-1 are implemented in DOE Packaging Certification Program training courses.
Structural Integrity Evaluation of a Corrugated Type Radiator for a Hermetically Sealed Oil Transformer
Authors: Yun Jin Chul , Lee Sung Won , Park Jung Yean , Lee Joon Yeob
Published: May 2012

An active part of a hermetically sealed oil transformer is a heat source increasing the temperature and the volume of insulation oil. The oil volume increase results in significant pressure variations to a tank of the transformer, and also can cause serious damage of the tank. For this reason, a corrugated type radiator is generally adopted to the transformer since it can provide prevention from the internal pressure rise by expansion of corrugated radiator fins. In this paper, design of a corrugated type radiator for the hermetically sealed oil transformer is suggested, and structural integrity of the radiator is evaluated by both numerical and experimental methods. The design of the corrugated type radiator has been proposed and variation of oil pressure inside the radiator is estimated for operating load case. The structural integrity of the proposed radiator is performed, and measured strain values are compared with the analyses results. In addition, a fatigue analysis of the radiator is conducted. From the results, it is concluded that the design of the radiator meets the requirements for structural integrity in the operating condition.
Semi-Active Control of Civil Structures With a Simultaneous Reduced-Order Modeling and a Tuning of the Control Law
Authors: Hiramoto Kazuhiko , Matsuoka Taichi , Sunakoda Katsuaki
Published: May 2012

Various semi-active control methods have been proposed for vibration control of civil structures. In contrast to active vibration control systems, all semi-active control systems are essentially asymptotically stable because of the stability of the structural systems themselves (with structural damping) and the energy dissipating nature of the semi-active control law. In this study, by utilizing the above property on the stability of semi-active control systems, a reduced-order structural model and a semi-active control law are simultaneously obtained so that the performance of the resulting semi-active control system becomes good. Based on the above fact any semi-active control laws derived from some models stabilize all real-existing structural systems that have structural damping. It means that the difference of dynamic behaviors between the real structural system and the reduced-order mathematical model in the sense of the open-loop response is no longer an important issue. In other words, we do not have to consider the closed-loop stability, which is one of the most important constraints in active control, in the process of the reduced-order structural modeling and the semi-active control design. We can only focus on the control performance of the closed-loop system with the real structure with the (model-based) semi-active control law in obtaining the reduced-order model. The semi-active control law in the present study is based on the one step ahead prediction of the structural response. The Genetic Algorithm (GA) is adopted to obtain the reduced-order model and the semi-active control law based on the reduced order model.
NuScale Approach to Passive Safety
Authors: Reyes Jose´ N.
Published: May 2012

The unprecedented seismic event and tsunami that caused the prolonged power outage at the Fukushima Daiichi nuclear plant have highlighted the importance of establishing stable long term cooling of the nuclear fuel and the containment following a complete station blackout. This paper presents an overview of the advanced passive safety systems implemented in the NuScale nuclear power plant. During normal operation, each NuScale containment is fully immersed in a water-filled stainless steel lined concrete pool that resides underground. The pool, housed in a Seismic Category I building, is large enough to provide 30 days of core and containment cooling without adding additional water. After 30 days, the decay heat generation is sufficiently small that natural convection heat transfer to air on the outside surface of the containment coupled with thermal radiation heat transfer is completely adequate to remove core decay heat for an indefinite period of time. These passive safety systems can perform their function without requiring an external supply of water, power, or generators. A complete assessment of the NuScale passive safety systems is being performed through a comprehensive test program that includes the NuScale integral system test facility at Oregon State University.
Plastic Instability Pressure of Conical Shells
Authors: Adibi-Asl R.
Published: May 2012

There several failure modes that are considered in the available codes and standards in field of pressure vessel and piping. One of these failure modes is plastic instability. This failure mode is defined as the pressure for which the components/structures approach dimensional instability (large deformation), i.e. unbounded displacement for a small increment in the applied load. In order to find this pressure both large deformation and strain hardening curve are considered. When the slope of numerically generated load-deformation curve approaching zero the corresponding applied load is considered as plastic instability load. Unlike cylindrical and spherical pressure vessels, available theoretical solution in range of plastic instability load for conical shells are very limited. Hence, it would be very useful to predict the behavior of these components with acceptable accuracy for design purposes. Analytical expressions are derived to determined plastic instability load of s conical shell subjected to internal pressure and when it is subjected to hydrostatic pressure (i.e., storage tanks). The geometrical changes can be estimated using the proposed solutions when considering the material strain hardening curve.
Optimal Shape Design Under Elastic-Plastic Behavior Based on Reference Volume Method
Authors: Adibi-Asl R.
Published: May 2012

The main objective of this paper is to determine the regions in a component or structure that directly participate in inelastic action (reference volume) using a new robust simplified method, namely the Elastic Modulus Adjustment Procedure (EMAP). The proposed method is based on iterative linear elastic finite element analysis that is implemented by modifying the local elastic modulus of the material at each subsequent iteration. The application of reference volume on optimum shape design is demonstrated through some practical examples including thick-walled cylinder, shank-head component and overlap joint weld. The results show that the reference volume concept can be used to optimize the shape of a body with respect to load carrying capacity and fatigue strength.
Reliability Assessment of Creep Rupture Life for Gr.91 Steel by an Interference Model
Authors: Kim Woo-Gon , Park Jae-Young , Yin Song-Nan , Kim Dae-Whan , Park Ji-Yeon , Kim Seon-Jin
Published: May 2012

This paper focuses on reliability assessment of creep rupture life under the service conditions at high temperatures and stresses for Gr. 91 steel which is considered as one of the prime structural materials for next-generation nuclear reactors. An interference model based on Z parameter, which is considered for the fluctuations of both service temperature and stress besides the scattering of rupture data, was analyzed and used to assess the reliability on creep rupture life of Gr. 91 steel. The scattering distribution of the creep rupture data of the Gr. 91 steel was investigated by using the Z parameter. It appeared that the Z parameter of creep rupture data for Gr.91 steel exhibited normal distribution. Using the normal distribution, a Monte-Carlo simulation (MCS) was carried out to generate a number of random variables for Zs and Zcr, and the reliability of the creep rupture life under the fluctuations of service temperature and stress conditions was estimated by using the interference model. It showed that the value of reliability decreased with increasing service time. Higher temperature caused the trend of faster deterioration. The value of reliability decreased rapidly at higher temperature fluctuation amplitude, and the reliability decreased as the scattering of the creep rupture data became serious.
Selective Laser Melting of Oxide Dispersion Strengthened Steels
Authors: Boegelein Thomas , Rao Ashwin , Jones Andrew R. , Tatlock Gordon J.
Published: May 2012

Oxide Dispersion Strengthened (ODS) alloys are a long established class of materials manufactured using powder metallurgy techniques. These alloys can offer exceptional high temperature strength and resistance to radiation damage, thus are envisioned to be used in a number of future nuclear and fossil energy power applications. However, due to the manufacturing steps involved, the overall cost to build components with these materials can be high. This paper presents work conducted to assess the feasibility of applying Selective Laser Melting (SLM) techniques to either coat or direct build on substrates with Fe-based Oxide Dispersion Strengthened (ODS) alloys. SLM is a rapid prototyping technique which can be used to manufacture near net-shape solid components from layered metallic powder beds. Two different geometries were of interest in this study — a simple button configuration with a nickel-base superalloy (IN939) substrate and a more complex hexagonal shaped wall with a mild steel substrate. Powders of PM2000 (a FeCrAl based ODS alloy) were deposited in both cases. Heat treatments were subsequently conducted on these structures to investigate effects of temperature on the bond characteristics and secondary recrystallisation. Electron microscopy examination revealed significant amounts of diffusion between the nickel and the ODS powders which enhances the bond strength. The studies have revealed the existence of a strong bond between the substrate and the interface even after prolonged exposure at elevated temperatures.
Small Bore Pipe Branch Connections Fatigue
Authors: Economou Jacques , Thebault Yannick , Costes Pierre-Emmanuel
Published: May 2012

Socket welds are used extensively for small bore pipe branch connections (up to a two-inch nominal diameter). For more than 15 years, numerous failures due to fatigue cracking occured in French NPPs (EDF operates 58 PWRs), mainly on safeguard systems as regards nuclear island safety-related systems. Such failures have significant impacts, on costs, due to repairs or In-Service-Inspections and sometimes, on plants availability. Destructive examination of cracked small bore pipe branch connections have led to a better understanding of failure mechanisms and have allowed to predict crack initiation locations depending on specific features such as main and small bore pipes thicknesses. Modifications implemented, based on empirical approaches, have turned out to be much or less efficient, depending on nuclear units. A three-stage methodology was devised in the early 2000’s and applied on the 8 most important safety-related systems. Based namely on vibration measurements, this methodology has allowed to identify sensitive small bore pipe branch connections, define a relevant In-Service Inspection program, determine vibrations origins and possible solutions. New modifications are currently underway on the 2 most affected systems. So far, they have shown to be effective. Solutions for the other systems should be decided at the end of 2011 and implemented on the whole French nuclear fleet.
How a Knowledge Data Base Assists the Process of Aging Management
Authors: Rothenho¨fer Horst , Scho¨ckle Friedrich , Ko¨nig Gu¨nter
Published: May 2012

The process of aging management in a nuclear power plant requires collaborative work of different departments. In this process information has to be gathered and evaluated systematically. At the beginning of the process it is important to classify all systems, structures and components (SSCs) into 3 categories of descending importance concerning safety where only the first two categories have to be considered for aging management. Basic information on the material and dimensions, operational parameters, quality status, analysis reports and relevant degradation mechanisms is stored in the data base for all SSCs of category 1 and 2. While a nuclear power plant is in service numerous events may occur which are related to aging and may have an influence on the quality of SSCs. Relevant events may be malfunction messages, maintenance results, operational experience, results of surveillance or events in external installations. It is important to evaluate these events concerning the aspect of aging and to decide which measures are required to prevent unexpected events of the same kind in the future. In some cases the new experience has to be assigned to comparable SSCs. These evaluations, the definition of measures and the tracking of their effectiveness have to be done by an aging management team that meets regularly every 3 months. The knowledge data base supports the application of the PDCA cycle and provides an overview and fast access to all quality documents related to each SSC including surveillance measures and results. It simplifies the generation of annual status reports which are valuable quality documents that reveal the effectiveness of aging management. Finally the procedure can be extended to SSCs which are important for the availability of the plant and thus take a step forward from aging management to plant life management.
Flowforce in Safety Relief Valve Under Incompressible, Compressible and Two-Phase Flow Conditions
Authors: Kourakos Vasilios , Chabane Sai¨d , Rambaud Patrick , Buchlin Jean-Marie
Published: May 2012

The use of relief valves is crucial for the safety of power plants. Indeed, these simple and robust in their design valves are the ultimate protection when all other systems are insufficient. This study is focused on valve opening characteristics which can be concluded through the determination of flowforces applied at the valve disk. A spring loaded safety relief valve (1 1/2″ G 3″) and transparent model of this valve are tested under static conditions. The spring is removed and the forces exerted at the valve disk for different inlet pressures and lift positions are measured. Compressible, incompressible and two-phase flow behaviors are investigated and compared. Inversed force of air and water is noticed above a certain value of valve lift. Numerical simulations are performed to verify experimental findings.
Emergency Relief of Internal Pressures From Aboveground Storage Tanks
Authors: Baker Mark , Morovich George L.
Published: May 2012

This paper covers the causes, frequency and risks of emergency internal pressure incidents that may be encountered by aboveground storage tanks and emergency pressure release alternatives. Methods for providing emergency relief of internal pressure from within aboveground storage tanks are desired as a safety precaution to prevent to shell uplift. Shell uplift may cause failure of the bottom to shell joint along with damage to piping, sumps and appurtenances; thereby, resulting in a release of a tank’s contents.
Large Ductile Tearing in Dissimilar Material Welds and Transferability Issues
Authors: Gilles Philippe , Brosse Alexandre , Mourgue Philippe
Published: May 2012

Dissimilar Metal Welds (DMW) are required in nuclear reactors to join low alloy steel components and stainless piping. The thermal and mechanical mismatch between the dissimilar material characteristics favors the stress concentration in the weld along the austenitic/ferritic interface. Assessing the ductile tearing resistance of DMWs is an issue even for crack initiation. But predicting large crack extensions is even more difficult for several reasons: crack path deviation, J-R curve determination, transferability from specimen to structures. This paper presents an approach to determine the resistance curve of a DMW on the basis of a ductile tearing simulation using a cavity growth model.
Thermal Hydraulic Studies on Helical Coil Steam Generator by CFD
Authors: Joh Sooyun
Published: May 2012

NuScale Power, Inc. is commercializing a 45 Megawatt electric light water nuclear reactor NuScale Power Module (NPM). Each NPM includes a containment vessel, a reactor vessel, a nuclear reactor core, an integral steam generator, and an integral pressurizer. The NuScale Power Module is cooled by natural circulation. The primary coolant in the Reactor Pressure Vessel is heated in the nuclear core, it rises through a central riser, it spills over and encounters the helical coil steam generator, it is cooled as steam is generated inside the steam generator, and it is again heated in the nuclear core. The Steam Generator also must be designed to provide adequate heat transfer, to allow adequate primary reactor coolant flow, and to provide adequate steam flow to produce the required power output. This paper presents the CFD results that describe the transport phenomena on the heat transfer and fluid flow dynamics in helical coil steam generator tubes. The ultimate goal of the CFD modeling is to predict the steam outlet conditions associated with the chosen helical coil tube geometries, solving the primary and secondary flow region together coupled with the helical coil tube. However, current studies are focused on the primary side with the heat flux boundary condition assigned on the outer surface of the helical coil steam generator. In this study, the ANSYS CFX v. 12.1 [1] was used to solve the three-dimensional mass, momentum and energy equations. The helical coil steam generator has complex geometry and modeling entire geometry requires the enormous memory that is beyond our hardware capability and is not practical. Therefore, geometry was limited to 1 degree of the wedge and 5% of the total length in the middle. Only external flow, single phase flow around the helical coils, is simulated using the standard k-ε model and shear stress transport model. From the results of the numerical simulation, the pressure drop and temperature profiles were determined. It is important to understand thermal hydraulic phenomena for the design and performance prediction of the reactor internal.
Testing and Analysis of Early Age Stress-Strain Development of Concrete Overlay for Reactor In-Situ Decommissioning
Authors: Guerrero Hector N. , Langton Christine A. , Restivo Michael L.
Published: May 2012

Non-uniform moisture distribution during early age drying and curing of protective concrete overlays for Savannah River Site reactors decommissioned in-situ were expected to result in differential shrinkage that lead to tensile stresses and possible cracking. The goal was to not exceed the concrete tensile strength of 2.8 MPa or compressive strength of 23.8 MPa. An experimental study was performed on the development of shrinkage and creep strains in concrete slab specimens embedded with strain gages and relative humidity sensors at controlled ambient temperature and relative humidity conditions. The experimental data was used to verify the validity of available physics-based methods and code calculations to predict early age shrinkage and creep strains, respectively. Data on the expansive effect of a shrinkage compensating admixture based on CaO was also obtained. Prediction of the surface stress in the test specimen suggests that without the shrinkage compensating admixture, the tensile strength would be slightly exceeded. However with the admixture, the combined effects of shrinkage, creep and expansion due to the admixture is predicted to result in all compressive stresses within the full scale slab, eliminating the possibility of cracking.
Life-Cycle Management of Pressurized Fixed Equipment
Authors: Osage David A. , Thornton David R. , Henry Philip A.
Published: May 2012

Many process plants will continue to operate pressurized equipment well beyond its intended design life. To ensure that the equipment operates safely and reliably requires adoption of an equipment Life-Cycle Management (LCM) process. During equipment design the LCM process requires identification of potential damage mechanisms, and a design that resists or mitigates the damage. For equipment that has been put into operation the LCM process continues with the use of prescriptive or Risk-Based inspection. An evaluation of the in-service inspection results reveals whether any anticipated (i.e., was considered in the initial design) or unanticipated damage has occurred. If the damage is anticipated and within the design limits, the equipment is returned to operation for a period of time that considers the equipment remaining life, with a new inspection at the end of the operational period. If unanticipated damage is discovered the LCM process requires identification of the damage mechanism and a subsequent Fitness-For-Service assessment to facilitate a decision to run as is, or to rerate, repair, or to replace the damaged components. To effectively implement the LCM approach, codes and standards must exist that address each aspect of the process. In addition, ensuring that similar analysis techniques are employed at the time of construction and when conducting in-service assessments requires coordination of the technology integration in these codes and standards. An overview of API, ASME, and other codes and standards is provided together with a discussion of the efforts to integrate technology to support the LCM process.
Statistical Assessment of Crack Initiation at Simulated Flaws due to Hydrided Region Overload in CANDU Zr-2.5%Nb Pressure Tube Material
Authors: Gutkin Leonid , Scarth Douglas A.
Published: May 2012

CANDU Zr-2.5%Nb pressure tubes are susceptible to formation of hydrided regions at the locations of stress concentration, such as in-service flaws. When the applied stress acting on a flaw with an existing hydrided region exceeds the stress at which the hydrided region has been formed, hydrided region overload may occur. Probabilistic methodology is being developed to evaluate in-service flaws in the pressure tubes for crack initiation due to hydrided region overload. Statistical assessment of relevant experimental data on the overload resistance of Zr-2.5%Nb has been performed as part of this development. The results of this assessment indicate that the critical nominal stress for crack initiation due to hydrided region overload increases with increasing the nominal applied stress during hydrided region formation, decreasing the stress concentration factor and increasing the threshold stress intensity factor for initiation of delayed hydride cracking. These findings are consistent with our fundamental understanding of hydrided region overload, as well as with the previous modeling work by E. Smith, as referenced in the paper. The overload resistance also appears to increase with the number of thermal cycles in the course of hydride formation. The results of this assessment have been used to develop a preliminary probabilistic model to predict the critical stress for crack initiation due to hydrided region overload under ratcheting hydride formation conditions, as well as a comprehensive experimental program to further investigate the overload behavior of CANDU pressure tube material.
Assessment of NDE Methods to Detect Lack of Fusion in HDPE Butt Fusion Joints
Authors: Crawford Susan L. , Doctor Steven R. , Cinson Anthony D. , Watts Michael W. , Moran Traci L. , Anderson Michael T.
Published: May 2012

Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, were 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 provided 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 assessments using ultrasonic and microwave nondestructive techniques and mechanical testing with the high-speed tensile impact test and the side-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. Some of the pipes were sectioned and the joints destructively evaluated with the high-speed tensile test and the side-bend test. The fusion parameters, nondestructive and destructive evaluation results have been correlated to validate the effectiveness of what each NDE technology detects and what each does not detect based on the limited testing that was conducted. No single NDE method detected all of the lack-of-fusion flaws; further a combination of NDE methods did not detect all of the flaws.