bluuebunny/crossref_metadata_2025 huggingface dataset:
Elastic Modulus Adjustment Improved Convergement Schemes Using Variable Local Constraints
Authors: Adibi-Asl R. , Fanous Ihab F. Z. , Seshadri R.
Published: August 2008

Elastic modulus adjustment procedures (EMAP) have been employed to determine limit loads of pressure components. On the basis of linear elastic Finite Element Analysis (FEA) with non-hardening elastic properties, i.e., by specifying spatial variations in the elastic modulus, numerous set of statically admissible and kinematically admissible distributions can be generated, and both lower and upper bounds on limit loads can be obtained. Some methods such as the classical, r-node and mα methods provide limit loads on the basis of partly-converged distributions, whereas the accuracy of linear matching procedures rely on fully converged distributions. In this paper, a criterion for establishing the degree of convergence of EMAP is developed, and a simple procedure for achieving improved convergence is described. The procedure is applied to some practical pressure component configurations.
Asset Management Evaluation: A Pilot Case Study
Authors: Bauby C. E. , Charbonnier B. , Haik P. , Lacombe S. , Lonchampt J. , Remy E.
Published: August 2008

The long term management of a production asset raises several major issues among which rank the technical management of the plant, its economics and the fleet level perspective one has to adopt. Decision makers are therefore faced with the need to define long term policies (up to the end of asset operation) which take into account multiple criteria including safety (which is paramount) and performance. In this paper we first remind the reader of the EDF three-level methodology for asset management. As introduced in PVP 2003 and PVP 2004, this methodology addresses the component/technical level (how to safely operate daily and invest for the future), the plant level (how to translate technical decisions into plant-wide consequences including economic performance) and the fleet level (how to manage a large number of similar assets). We then focus on the software tool that implements this methodology in order to allow decision makers to define, evaluate and analyze long term plant operation and maintenance policies. Lastly we show how the methodology and the software tool were used on a pilot case study. The technical and economic results obtained at the plant level are described as well as the conclusions one can draw from them in order to help decision makers evaluate and analyze long term asset management strategies.
The Application of the Linear Matching Method to the Life Assessment Method R5: A Comparison
Authors: Chen Haofeng , Ponter A. R. S. , Ainsworth R. A.
Published: August 2008

The Life Assessment Method R5 is widely used for the assessment of power plant at high temperatures. The procedure involves the application of a sequence of rules that provide margins of safety against a range of possible failure modes, plastic collapse, ratchetting, fatigue failure, creep rupture and creep/fatigue interaction. Within R5 this is achieved by using limit load and shakedown methods, combined with Neuber’s rule and the evaluation of elastic follow-up factors. In recent years, the Linear Matching Method has been developed so that it is capable of providing optimal solutions for each of these criteria, thereby giving less conservative margins of safety, but adopting the same use of material data as R5. The paper describes a detailed comparison between the approach currently used in R5 and the result of the application of the Linear Matching Method, for the entire range of failure modes and for a simple example. The purpose of this comparison is to assess the circumstances where Linear Matching Methods may have a distinct advantage over current methods. The example consists of a square plate containing a circular hole. The plate is subjected to uniaxial loading and a radial varying temperature field. For all failure modes the Linear Matching Method gives less conservative results compared with standard R5 methods. The differences can be very significant, particularly for the prediction of strength limits; limit load, shakedown load and ratchet limit. Significantly lower and less conservation elastic follow-up factors were also obtained. The comparison demonstrates the advantages of the Linear Matching Method in this context.
Application of the Local Approach to Predict Load History Effects in Ferritic Steels
Authors: Mirzaee-Sisan Ali , Mahmoudi Amir-Hossein , Truman Christopher E. , Smith David J.
Published: August 2008

The effect of residual stresses on fracture behaviour of ferritic steels at low temperature has been investigated using a local approach based on the Beremin model [1]. The study aims to enhance the use of local approach in failure prediction when residual stress is present. A tensile residual stress field has been introduced in the laboratory specimens and their subsequent behaviour was investigated at low temperature. Local compression methods, including side-punching and in-plane loading, were employed to introduce residual stress fields. These methods are discussed and comprehensive range of experimental presented. The transferability of the Weibull parameters between the cracked specimens with different constraint, test temperatures and also from unstrained specimens to specimens with residual stresses are illustrated. The general scheme in failure prediction using the local approach is that the Weibull parameters in the Beremin type model calibrated to the as-received data should predict the failure following complex interaction of residual and applied stresses. The paper compares the predictions and the experimental results.
Characterization Improvement Through Signal Processing: Application to Radiography Inspection — 3D Reconstruction
Authors: Chatellier Laurence , Just Valery E. , Fournier Louis , Charbonnier Bruno , Robillard Lionel
Published: August 2008

During in-service inspections, experts are faced with the delicate task of establishing a complete diagnosis of defects from radiographs. Should a defect be detected, one must be able to demonstrate that the component still meets regulatory requirements. Thus, it is essential to be able to characterize precisely the defect, especially when the demonstration relies on mechanical calculus. However the characterization of the defect by only g or X-ray is sometimes very difficult, and the justification process can thus be jeopardized. In such cases, signal processing can be very helpful for the interpretation of the data and for the characterization (positioning and sizing) of the defect. This paper presents a 3-D reconstruction processing in hard conditions representatives of pipe inspections: the incidence angle is very reduced and thus the radiographs contain very little information along the vertical direction. The reconstruction process relies on the estimation of the attenuation. It is called inversion because it restores the attenuation from both data and prior information. The method has been tested on radiographs of a block with real defects and the performances were evaluated from a mock-up with several electro-drilled cylindrical defects. Even in the case of limited incidence, the method provides very useful 3D results. Moreover this process can be applied whatever the nature of the source. When a larger source is used in order to inspect thick components, signal processing allows to reduce the inevitable blur which leads to improved accuracy. In conclusion, signal processing and especially 3D reconstruction in the case of radiography can turn out to be a key step fur in-service inspection of major NPP components.
Application of the Master Curve Method to the Japanese RPV Materials
Authors: Funakoshi Hiroomi , Kanazawa Yasushi , Hirano Takashi , Kojima Naoki , Matsumoto Kouichi , Soneda Naoki , Miura Naoki
Published: August 2008

In application of the Master Curve (MC) method to the surveillance program of the BWR plants in Japan, we need to address two issues. The first one is to confirm the validity of the MC method to the steels typical to the Japanese reactor pressure vessels (RPVs), and the second is to find a procedure, which allows us to smoothly switch from the current Charpy-based surveillance program to the MC-based surveillance program. In this study, we present results to address the two different issues mentioned above. First, we study the application of the ASTM E 1921 procedure to determine the To values and the MCs for the unirradiated model steels for the Japanese BWR plants. We investigate the basic aspects of the MC method. Second, we investigate the applicability of instrumented Charpy impact data to estimate static fracture toughness Jc values.
Automated Parameter Determination of Advanced Constitutive Models
Authors: Rahman Syed M. , Hassan Tasnim , Ranjithan S. Ranji
Published: August 2008

Parameter determination of advanced cyclic plasticity models which are developed for simulation of cyclic stress-strain and ratcheting responses is complex. This is mainly because of the large number of model parameters which are interdependent and three or more experimental responses are used in parameter determination. Hence the manual trial and error approach becomes quite tedious and time consuming for determining a reasonable set of parameters. Moreover, manual parameter determination for an advanced plasticity model requires in-depth knowledge of the model and experience with its parameter determination. These are few of the primary reasons for advanced cyclic plasticity models not being widely used for analysis and design of fatigue critical structures. These problems could be overcome through developing an automated parameter optimization system using heuristic search technique (e.g. genetic algorithm). This paper discusses the development of such an automatic parameter determination scheme for improved Chaboche model developed by Bari and Hassan [4]. A new stepped GA optimization approach which is found to be more efficient over the conventional GA approach in terms of fitness quality and optimization time is presented.
Fluid Elastic Instability of Rotated Square Array Tube Bundle in Two-Phase Cross-Flow
Authors: Chung H. J. , Chu I.-C.
Published: August 2008

Fluid-elastic instability in air-water two-phase cross-flow has been experimentally investigated using two different arrays of straight tube bundles. Rotated triangular array tube bundle with p/d of 1.47 is for the counterpart test of existing work. Rotated square array tube bundle with p/d of 1.633 is for the investigation of fluid-elastic instability in KSNP steam generator. The present paper provides the experimental results of tube vibration response, damping ratio, hydrodynamic mass, fluid-elastic instability, and flow pattern in tube bundle. The damping ratio was very dependent on void fraction, as were in previous works. The tube vibration motion of the rotated square array was quite different from that of the rotated triangular array. Compared to the case of the rotated triangular array, tube-to-tube hydrodynamic coupling of the rotated square array is very weak, which might result in very high fluid-elastic instability constant. An ultrasonic through-transmission technique was successfully applied to identify the flow regimes in a tube bundle.
Evaluation of Strain and Stress Ranges for Proportional and Non-Proportional Fatigue
Authors: Itoh Takamoto , Sakane Masao , Hata Takahiro , Hirai Takanori
Published: August 2008

In non-proportional multiaxial fatigue, principal directions of strains and stresses change during a cycle. In such a case, strain and stress ranges and mean strain and stress cannot be easily determined. Developing an appropriate method of determining the ranges and mean values are needed for the design of pressure vessels. This paper proposes a method for calculating the strain and stress ranges and mean values in non-proportional multiaxial loading. Based on the method proposed, non-proportional strain and stress ranges are derived and the accuracy of life prediction method using the ranges is discussed for type 304 stainless steel fatigued under 14 non-proportional strain paths.
Effect of Rising and Falling K Profiles on SCC Growth Rates in High Temperature Water
Authors: Andresen Peter L. , Morra Martin M.
Published: August 2008

Effects of rising and falling stress intensity factor (K) profiles on the SCC growth rates of stainless steel and nickel alloys has been studied in high temperature water. Sophisticated test control software was used that changes loading (P) based on crack length (a) to achieve a specific K trajectory by controlling dK/da, not simply dP/dt. The majority of SCC problems develop adjacent to welds, which have a complex residual stress profile vs. wall thickness. This, coupled with the dependence of K on crack length, causes K to change as the crack grows, not per se with time (t). The effect of “K-dot” on crack tip strain rate and the associated crack growth rate is discussed, along with the repercussions to understanding and dispositioning SCC response.
Determination of Creep Constitutive Law of Solder Materials Using Indentation Creep Test
Authors: Sakai Shinsuke , Watanabe Yu , Izumi Satoshi , Iwasaki Atsushi , Ogawa Takeshi
Published: August 2008

Authors reported in the previous paper [1] that indentation creep test is effective to evaluate the creep power law from small specimen. We derived the formulation and applied the proposed method to turbine rotor materials. Eventually, it was shown that the method is applicable for the evaluation of creep constitutive law especially under high stress region. Under the low stress region, however, the applicability of the method was not confirmed since it requires much time for the confirmation. As most necessary property for creep deformation is that for low stress region, the extension of the proposed method for the low stress region is considered extremely important. For this purpose, it is not appropriate to use steel materials for the confirmation because it consumes too much time to conduct creep indentation test under low stress. For solder materials, however, it is rather easier to conduct the creep indentation test under low stress because the creep phenomena occur even at room temperature. In this paper, we report the results of creep indentation test for solder materials and examine the applicability of the proposed method. Besides finite element analysis is performed to evaluate stress conversion factor which enables us to evaluate Norton’s law only from creep indentation tests.
A Metallurgical Concept for Numerical Simulation of Arc Welding
Authors: Chau T. T.
Published: August 2008

Issued from two successive RD studies in 1991–1995 for shipbuilding in France, a simplified method of numerical simulation of arc welding has been developed and validated on samples in full scale executed inside the shipyard. The metallurgical concept of the methodology is based on two main characteristic diagrams of iron-carbon steel: metallurgical phase transformation diagram and thermal dilatation diagram. In this paper, the simplified methodology is described with on its basic assumptions.
Acoustic Vibration Behavior of Full Size Steam Generator and Tubular Heat Exhanger In-Line Tube Banks: A Brief Note
Authors: Eisinger Franktisek L. , Sullivan Robert E.
Published: August 2008

Based on recent laboratory experimental data by Feenstra et al. [1],[2] it has been determined that for larger test section widths, the maximum acoustic pressures generated during acoustic resonance were greater by more than a factor of four than those predicted by Blevins and Bressler [3]. We have evaluated a great number of resonant and non-resonant cases from inservice experience of full size steam generator and tubular heat exchanger tube banks in order to see the general vibratory behavior of the full size units. Fifteen vibrating and twenty-seven non-vibrating cases were evaluated and compared to the Feenstra et al. relationship. It is shown that on average the results from the full size units correlate well with the Feenstra et al. relationship. A gap exists between the vibratory and the non-vibratory cases. The non-vibratory cases produce acoustic pressures which are at or below the Blevins and Bressler relationship. From the results it can be concluded that the full size units, regardless of their size and also acoustic mode, produce high acoustic pressures at resonance, with the maximum acoustic pressure on average more than fifty to seventy five times higher than the input energy parameter defined by the product of Mach number and pressure drop through the tube bank. The results are tabulated and plotted for comparison.
Vibration Damping of Stabilized Steam-Generator Tubes
Authors: Janzen V. P. , Han Y. , Smith B. A. W. , Fluit S. M.
Published: August 2008

The integrity of steam-generator tubes is an important aspect of the long-term reliable operation of nuclear power plants. In situations where a tube is judged to be at risk, it must be either plugged, or removed, or reliably stabilized in some manner to avoid excessive motion of the tube due to flow-induced vibration. The present work describes measurements of the effect of an internal cable-type stabilizer on the structural damping of steam-generator tubes. The free-vibration response of unstabilized and stabilized tubes was analyzed to provide damping ratios from frequency-domain spectral responses, time-domain logarithmic decrement ratios and time-domain vibration decay-curves. The structural damping ratios typically increased from approximately 1.6% to approximately 4.3% with the addition of the stabilizer. This last value is somewhat less than recently published values for stabilized tubes from a different type of steam generator, suggesting that tube stabilization, while effective, has limitations that need to be conservatively assessed.
Fatigue Damage Evaluation for Thermal Striping Phenomena Using Analytical Spectrum Method
Authors: Okajima Satoshi , Sakai Shinsuke , Izumi Satoshi , Iwasaki Atsushi , Kasahara Naoto
Published: August 2008

It is well known that fatigue damage accumulates around T-junctions of piping system where two kinds of fluid with different temperatures are mixed. This phenomenon is called thermal striping, and simple method for evaluating the fatigue damage derived from this phenomenon is greatly important for both design and maintenance stages. However, the evaluation of the thermal stress derived from the thermal striping is rather difficult since the evaluation requires complicated analyses such as fluid mixing, heat transfer, and heat conduction. In addition, the closed-form solution to describe the stress-range distribution under random loading has not been established yet since the rainflow cycle-counting method is not suitable to the analytical treatment. Though numerical calculation may be available for the evaluation, it requires time-consuming work and is not practical for design stage. For this reason, an analytical method for evaluating the fatigue damage directly from stress power spectrum density (PSD) is desired. In this paper, the feasibility study to apply analytical methods to the evaluation of thermal-striping damage is examined. The analytical methods are applied to the fatigue damage evaluation from the stress PSD that was obtained by the thermal striping experiment. Finally, the applicability and problem of each method will be discussed. In order to apply the envelope PSD to design, the property of safety margin associated with the PSD is investigated too.
Application of a Classification Method to Obtain Primary Stresses Without Evaluation Sections to Perforated Structures
Authors: Nagashima Hideaki , Shibamoto Hiroshi , Inoue Kazuhiko , Kasahara Naoto , Sadahiro Daisuke
Published: August 2008

Conceptual design studies of Japanese commercialized Fast Breeder Reactor (abbreviated to FBR) are carried out. With the aim of reducing the construction cost, number and size of components are reduced. For example, Intermediate Heat Exchangers (abbreviated to IHXs) are integrated with primary pumps; number of main cooling loops is reduced from 3 to 2 and piping is shortened; the reactor vessel is simplified. Accompanying the reduction of cooling loops, Steam Generators (abbreviated to SGs) become larger, and semi-sphere perforated plates in 3-D structures are adopted for large sized steam generator to endure primary stress caused by pressurized steam. In “Design by Analysis” for these nuclear components, the stresses in these structures are classified into primary and secondary stresses. Conventional method of stress classification uses an evaluation section; however, it is difficult to define evaluation sections in 3-D structures. For this reason, an alternative evaluation method without evaluation sections, which can be easily applied to 3-D structures, is necessary. Primary stress is the so-called load controlled stress; this is decided as the stress in equilibrium with external loading and is independent of inelastic behavior of materials. Paying attention to above feature, there is an idea to obtain primary stress from a Re-distribution node (abbreviated to R-node), where stress is constant during stress re-distribution. One of concrete method is GLOSS (Generalized local Stress Strain) method proposed by R. Seshadri et al. [1–3]. To generate artificial stress re-distribution, this method needs to carry out two elastic analyses with different values of material constants. All stresses are different between two calculations except R-node and the stress at this point can be evaluated as the primary stress. This method adopts elastic analysis with special material constants that is determined from post-processing of stress distributions of the first elastic analysis due to the insufficient power of earlier computer. Recent progress of computer technologies reduces computational time and cost of inelastic calculation. Furthermore, it is not difficult for an elastic-plastic calculation to be carried out with classical constitutive equations provided by commercial codes. In this paper, an alternative method is proposed, which uses elastic-plastic analysis for artificial stress re-distribution instead of elastic analysis in GLOSS method. The proposed method is confirmed through the application to the example problem for a thick cylinder and the ability to be applied to actual structures is confirmed through the application to the example problem for a perforated plate.
Containment Sump Active Strainers
Authors: Mehta A. R. , Bilanin A. J. , Hamel J. , Kaufman A.
Published: August 2008

The containment sump, also known as emergency or recirculation sump, is part of the Emergency Core Cooling System (ECCS). Every nuclear power plant is required by regulations to have an ECCS to mitigate a design basis accident. The containment sump of a Pressurized Water Reactor (PWR) collects reactor coolant and chemically reactive spray solutions following a Loss of Coolant Accident (LOCA). The containment sump serves as the water source to support long-term recirculation. This water source, the related pump inlets and the piping between the source and inlets are all important safety components. Suppression pools in Boiling Water Reactors (BWRs) serve the same purpose as PWR containment sumps. Historically, a passive debris screen has been used to prevent debris from entering the ECCS suction lines surrounding the containment sump. Previous incidents demonstrated that the potential for excessive head loss across the containment sump screens exists because of the accumulation of debris on the containment sump. Because of this, the US Nuclear Regulatory Commission (NRC) has concluded that containment sump blockage is a potential concern for PWRs. US BWRs were required to conduct plant-specific evaluations of their suction strainer performance and, as required, modify their plant design. While all US PWRs are required to resolve this Generic Safety Issue (GSI-191), containment sump blockage continues to be a major concern for both BWRs and PWRs internationally. This paper describes the GE Active Strainer design, one of several strainers developed to resolve this generic safety issue. The Active Strainer presents an innovative and novel method of addressing containment sump blockage. This strainer employs a rotating, or “active”, plow and brush that sweep over a perforated surface. By keeping the perforated surface free of debris, fluid is allowed to pass through, providing sufficient coolant to the ECCS pumps to support long-term recirculation. Due to the unique method by which the Active Strainer filters coolant, a test program was developed to demonstrate its functionality and viability. Intrinsic differences between passive and active solutions make previous methods of testing obsolete for the GE Active Strainer. Moreover, the complex and varying geometries and conditions of actual plant containment sumps are difficult to replicate. Therefore, a methodology was developed to ensure prototypical test environment and strainer debris loads in a scaled test facility. This paper will discuss the GE Active Strainer design, the testing conducted and subsequent conclusions.
Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials
Authors: Shah V. N. , Shelton B. , Fabian R. , Tam S. W. , Liu Y. Y. , Shuler J.
Published: August 2008

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.
Non Destructive Testing of Unfired Pressure Vessels
Authors: Baylac Guy , Roberrts Ian , Zeelenberg Erik
Published: August 2008

This paper discusses the non destructive testing (NDT) of unfired pressure vessels made of ductile and tough steels, as contained in Part 5 of the European standard EN 13445:2002. The concept and use of testing groups along with “satisfactory experience” in welding are presented. Also the background and rationale for the determination of standards used for NDT methods, characterisation and acceptance criteria are discussed in detail. Benefits for the pressure equipment industry are emphasised.
Training in Quality Assurance for Radioactive Material Transportation Packaging
Authors: Fabian R. , Garrison L. , Shelton B. , Liaw J. , Shah V. , Tam S. W. , Liu Y. Y. , Shuler J.
Published: August 2008

The Department of Energy (DOE) has established guidelines for qualifications and training of the technical experts preparing and reviewing the safety analysis reports for packaging (SARP) and transportation of radioactive materials. One of the qualifications is working knowledge of, and familiarity with the quality assurance (QA) requirements in Subpart H of Title 10 of the Code of Federal Regulations Part 71. DOE is sponsoring a course on quality assurance for radioactive material transportation packaging. The objective of this paper is to describe the salient features of the course, the purpose of which is to provide QA training and practical experience that are required to develop and implement a QA plan or prepare the QA chapter of a SARP for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the packaging. The applicable QA requirements from DOE orders, federal regulations, and NRC regulatory guides will be highlighted, along with a graded approach to selected QA elements from Subpart H of 10 CFR Part 71. The paper will also briefly discuss ASME NQA-1 for Type B and fissile material packaging, current issues resulting from the different emphasis between a compliance-based QA program (in Subpart H, 10 CFR 71) for packaging and a performance-based QA program for DOE nuclear facilities (based on 10 CFR 830, “Nuclear Safety Management”), and the final rule changes in 10 CFR 71 that became effective on October 1, 2004.
Modeling Leakage in a Fuel Transfer Pipeline Using Computational Fluid Dynamics Techniques
Authors: Alexander Chris , Pordal Harbi
Published: August 2008

Stress Engineering Services, Inc. (SES) performed an assessment of leaks for an insurance company that occurred in a fuel transfer pipeline at a tank storage facility. Of specific interest were the duration and timing of the leaks, which occurred from a 30 foot section that entered an earthen containment berm. It was originally estimated that 28,900 gallons of gasoline and ethanol leaked periodically from two (2) pin holes in the pipeline during a two month period. Early analysis efforts were not able to estimate the conditions that were necessary to cause the leaking fluid to break through the surface of the earthen berm (a phenomenon known as daylighting). Consequently, SES performed a more rigorous investigation to determine what conditions were required to produce daylighting, the significance of which involved quantifying the estimates of leak duration and the petroleum volumes. This effort integrated assumptions and data from prior analyses to assess the effects of time-dependency using computational fluid dynamics (CFD) modeling techniques. The intent was to take the existing calculations and provide a more technically-defensible model to predict the timing and volume released using reasonable conditions. SES used soil permeability and actual pipe pressure data to simulate the pipeline leak and soil conditions. The results of the CFD analysis showed that it is possible for daylighting to occur within a two-month period. However, a specific combination of conditions associated with leak rates, leak duration, and soil permeability are required to generate daylighting in a relatively short period of time. The predominant observation is that there must be extended periods of continuous leaking involving leak rates of sufficient magnitude. The significance of this work is that it presents a proven analytical method for modeling leaks in pipelines and addressing the effects of specific variables on the amount of released products and the time required to achieve specific leak volumes.
J-Resistance Properties of Narrow Gap GTAW Welds for Ferritic Primary Loop Piping
Authors: Yoo One , Yoon Ki-Seok , Choi Tack-Sang , Lee Bong-Sang
Published: August 2008

J-Resistance properties of automatic narrow gap gas tungsten arc welds (NGGTAW) for ferritic primary loop piping were investigated. The piping diameter is 1066 mm and its thickness is 76 mm. To do this, tensile, micro-hardness and J-R tests were performed and the microstructures of the welds were examined. Tensile and J-R tests were conducted at 316°C and 177°C for temperature effect and at 1 mm/min and 1000 mm/min for strain rate effect. The results were compared with the weld cases which are fabricated by the conventional SMAW process. NGGTAW welds investigated in this paper have lower J-R curves than the SMAW welds do. Fracture surfaces of the NGGTAW welds have an unusual appearance with deep ditches and holes due to the inclusions and pores. It is believed that the high density of the inclusion and the pore which exists in the NGGTAW welds is responsible for this low toughness and the cracked surface. Therefore, for an improvement of the NGGTAW weld toughness, it is recommended that sulfur content should be limited to a considerably low level and the pore should be eliminated by modifying the welding procedure specification.
Inspection and Testing of Serially Produced Pressure Vessels
Authors: Roberts Ian , Baylac Guy , Zeelenberg Erik
Published: August 2008

The vast majority of applications of EN 13445-5:2002 will be for individually designed pressure vessels, but there will be a significant sector of serially produced pressure vessels, built in a continuous fabrication process to a single design. This paper presents Annex A of EN 13445-5 which addresses inspection and testing of these pressure vessels and gives the thinking and rationale behind these requirements.
An Assessment of Low Alloy Steel EAC Corrosion Fatigue Relationships for BWR Environment
Authors: Mehta Hardayal , Horn Ron
Published: August 2008

The fatigue crack growth rates for ferritic steels in water environments given in A-4300 of Appendix A, Section XI, ASME Code, were developed from data obtained prior to 1980. Subsequently, updated assessments by Eason, et al. and recent laboratory test results from Seifert and Ritter demonstrated that under certain conditions, ferritic steels exposed to oxygenated water environments may be susceptible to high fatigue crack growth rates that exceed the current disposition curves. In the light of ASME adopting Code Case N-643 for PWRs, there is a need for a similar Code Case for the BWR water environments (for both the normal water chemistry and hydrogen water chemistry/NobleChem) that takes into account these findings. This could mean modification of current EAC curves in the ASME Code. A joint program of EPRI and GE was developed to address this need for updated evaluations of the corrosion fatigue. The program’s first task has been to re-assess the role of rise time, environment, alloy, heat treatment and impurity levels on the established ASME codified disposition curves/methodologies. The data was then used as a basis to assess the impact of on modified cyclic curves on the disposition approaches that are currently used to evaluate postulated flaws in the BWR reactor pressure vessel or RPV head and the feed water nozzle regions. The presentation would include a discussion of the appropriate BWR plant transients and the GE process for performing evaluations. The role of the evaluations on the establishment of inspection intervals currently determined using NUREG-0619 and the latest BWROG Report would also be presented. Finally, the relationship between cyclic load and constant load behavior in these steels are discussed in the context of the mechanisms for environmentally assisted cracking.
The Effects of Fluid Media and Environmental Temperatures on the Set-Pressure of Safety Valves
Authors: Ezekoye L. Ike , Matty Terrence J. , Ferriera William Rego
Published: August 2008

The ASME OM Code provides guidance for set-pressure testing of safety valves. It requires that valves designed to operate on steam, including safety valves that are installed on water-filled loop seal, shall be tested with saturated steam. Alternative compressible fluids are permissible as the test media provided correlations can be developed to substantiate the test. The Code further requires that testing be performed in an environment that is consistent with the installed ambient temperature. This paper presents the results of set-pressure tests of safety valves with nitrogen and steam under controlled environmental temperatures. The results show the ASME Code concerns regarding the test media and ambient temperature considerations are valid.
An Overview of Ballast Water Treatment Methods
Authors: McCluskey D. K. , Holdo̸ A. E. , Calay R. K.
Published: August 2008

One of the greatest threats to the world’s oceans is the transfer and potential introduction of harmful organisms in environments to which they are not indigenous. Ballast water deposited by marine vessels has been identified as a key factor responsible for this transfer and introduction. The significance of transferral of species has led to diverse research and development into ballast water treatment methods. However, despite the number of ballast water management techniques in development, at present the exchange of ballast water in open oceans is the only management technique widely used. This paper provides an overview of Ballast Water Treatment techniques currently at a research or development stage. An assessment of the potential effectiveness of each method, its corrosion implications and economic cost are all presented together with preliminary recommendations and an outline of future work.
Fatigue Assessment of a Weld Joint Subjected to Thermal Stresses
Authors: Kalnins Arturs
Published: August 2008

Fatigue analysis is performed for a full-penetration weld joint, welded from both sides, connecting a cylindrical vessel to a flat head. It is subjected to a repeated transient cycle of ambient temperature inside the vessel. No pressure is cycled. Finite element heat transfer and elastic stress analyses are performed, and stress ranges are calculated. The allowable number of cycles for cyclic service is obtained by two structural stress methods. One is a method proposed for the ASME Boiler and Pressure Vessel code and the other is in the European Standard for unfired pressure vessels. It is found that for the same geometry and loading the former would permit 136,000 or 36,000 cycles while the latter would permit 8,000 cycles.
ENIQ European Framework on Risk Informed In-Service-Inspection
Authors: Gandossi Luca
Published: August 2008

The European Network for Inspection and Qualification (ENIQ) is a utility driven network working towards a harmonised European approach on reliable and effective in-service inspection (ISI). Its Steering Committee (SC) has one voting member for each EU member country with nuclear plants and for Switzerland. More specifically ENIQ works on qualification of ISI systems an on risk-informed in-service inspection (RI-ISI) within a European context. To reflect the latter, the SC set up a specific Task Group on Risk (TGR). The primary task for TGR was to produce a European framework document on RI-ISI, following the format of the well-established ENIQ document for qualification of NDT systems “European Methodology for Qualification, EUR 17299 EN”. The framework document for RI-ISI is a high-level document aimed at supporting utilities to implement RI-ISI and to interface with the nuclear regulators. Aim of this paper is to present the work of TGR so far and in particular the contents of the European framework document on RI-ISI, which has been recently approved for publication by ENIQ SC. It gives an overview of the philosophy behind a risk informed approach towards in service inspection and the principles of a risk informed in service inspection. It also defines how roles, responsibilities and interfaces between the different parties involved in a RI-ISI programme should be organised.
Failure of Resistance Thermometer Devices Due to Flow-Induced Vibrations
Authors: Husaini S. Mahmood , Qashu Riyad K. , Blevins Robert D.
Published: August 2008

Resistance Thermometer Devices (RTDs) are used to monitor the temperatures in sensitive locations of process piping, such as the hot and cold legs of the primary system of pressurized water reactor nuclear power plants. The RTDs are housed in thermowells that protrude in the cold and hot leg pipes. Eight of the twelve cold leg RTDs in the San Onofre Nuclear Plant have a direct safety function in the operation of the reactor. They provide the Core Protection Calculators with two temperature measurements from each of the cold legs. During the period 1997–2004, four RTD failures occurred in the cold legs of the San Onofre Nuclear Plant. Extensive investigations showed that all of the RTD failures were caused by cracking and fracturing of the platinum wires in the sensors. There was no damage to the nozzles or thermowells of the cold leg RTDs. Also, none of the RTDs in the hot leg were damaged. These failures occurred after the Cycle 9 refueling outage, when the Reactor Coolant System thermowells were replaced due to potential cracking of the Inconel 600 material. A root cause analysis was performed to identify the reasons for the failure of the RTDs in the reactor cold leg piping. The thermowells used for both the hot and cold leg RTDs are identical. The thermowells are 11.375” (289 mm) in length. Since the thickness of the cold leg pipe is 0.75” (19.1 mm) less than the thickness of the hot leg pipe, the cold leg thermowells protrude 0.75” (19.1 mm) more into the flow stream as compared to the hot leg thermowells. The protrusion lengths of the hot and cold leg thermowells in the pipe are 2.5” (63.5 mm) and 3.25” (82.6 mm), respectively. The increased protrusion length of the thermowells in the cold leg significantly lowers their natural frequency (as compared to the hot leg thermowells), which results in vortex-induced vibrations at lower flow velocities that cause failure. Also, the increased protrusion length results in increased amplitude of vortex-induced vibration of their tip that exceeds the capability of the installed RTDs. Thus, a recommendation was made to decrease the protrusion length into the cold leg pipe from 3.25” (82.6 mm) to 2.5” (63.5 mm). Since this recommendation was implemented more than eight months ago, there have been no RTD failures in the cold legs.
Numerical and Experimental Research of Pipelines Safety Including Aspect of Protection From Terroristic Threat or Warfare Conditions
Authors: Blyukher Boris , Neizgoda Tadeusz , Malachowski Jerzy , Szymczyk Wieslaw
Published: August 2008

The paper presents results of experimental and numerical tests over research and development of the numerical Finite Elements Method (FEM) based methods of pipeline strain/stress condition assessment. The research is inspired by certain circumstances accompanying domain of safe exploitation of pipeline systems used for gas and crude oil transportation [1]. Currently, we observe the increase in hazard originated not only by typical pipeline operations but the third party activities understood as terrorist threats and warfare activities, especially in regions that are of significant value for the stability of the world’s energy supplies. Methods used in the scope of the research may be applied also to such events as disturbances of normal pipeline operation by seismic or mining activities, as they need the analysis of shock waves influence on pipeline installations. Elements of these methods may be used in wider range of construction types like pipelines in power plants (especially nuclear) that are at leading locations on the list of any threats. The research is focused on the two topics: • establishing the method of fast numerical assessment of stress state in deformed pipes with the use of data possessed from inline inspection pipeline intelligent tool (PIG); • establishing methods of numerical simulations of influence of shock waves (caused by explosions, seismic as well as mining activities) on pipeline systems (buried, ground based, underwater). The research includes an investigation of geometric flaws which may be detected in oil or gas transporting pipeline by a caliper PIG during inline inspection. Such typical flaws are dents, folds and bucklings [2]. The research method is based on numerical simulation and analysis with the use of FEM (Finite Elements Method). But it must be verified experimentally. To achieve experimental verification of the numerical method, preliminary tests were conducted in a laboratory scale. Artificial dents, folds, and ovalizations were created with the use of material strength testing machine in the sample of weldless thin-walled pipes. The experiment was conducted under strict control, enabling measurements of deformations and strains in selected points located on the external surface of the pipe wall. Experimental research on deformations and strains, conducted for bended pipes, allows for verification of strain and stress distributions obtained from numerical calculations. This will allow improving methods of numerical simulations [3].
Economic Evaluation of Proactive Replacement of Upper Internals Guide Tube Support Pins (Split Pins)
Authors: Perdue Robert K. , Elder G. Gary , Gerzen Gregory
Published: August 2008

Certain nuclear power plants have “Rev B” reactor vessel upper internals guide tube support pins, commonly referred to as split pins, made from material with properties similar to Alloy 600 and known to be susceptible to primary water stress corrosion cracking (PWSCC). This paper describes a rigorous probabilistic methodology for evaluating the economics of a preemptive replacement of these split pins, and describes an application at four of Exelon Generation’s nuclear plants. The method uses Bayesian statistical reliability modeling to estimate a Weibull time-to-failure prediction model using limited historical failures, and a Westinghouse proactive aging management simulation tool called PAM to select a split pin replacement date that would maximize the net present value of cash flow to a plant. Also in this study is a sensitivity evaluation of the impact of zinc addition on split pin replacement timing. Plant decisions made based in part on results derived from applying this approach are noted.
Probabilistic Analysis of a Preemptive Replacement of an NPP Component With a Component of Superior Material
Authors: Perdue Robert K. , Woodcock Joel , Houssay Laurent
Published: August 2008

The Westinghouse proactive aging management tool, PAM, considers three major sets of variables when calculating the NPV or economic value of age replacement: (a) the projected time to failure, (b) the economic consequences of unplanned failure and (c) the cost of the replacement. All of these variables will typically be uncertain; particularly the time to part failure. A not uncommon complication in evaluating whether and when to replace a degrading component or part in a plant is that the replacement part is thought to have a longer expected life (be more resistant to degradation) but, to date, there is little field experience to substantiate that belief. This paper shows how two different approaches for statistical estimation of a Weibull failure distribution can be used in tandem to surmount this problem, and illustrates it within the context of the replacement of a nuclear power plant component tube bundle with a tube bundle expected to provide superior corrosion resistance.
Decision Advisor for Multi-Component Management of Alloy 600 Degradation Issues at the Exelon Braidwood and Byron Nuclear Power Plants
Authors: Elder G. Gary , Mycoff Clark W. , Gerzen Gregory , Perdue Robert K. , Ray Edward A. , Bamford Warren H.
Published: August 2008

This paper describes how the Westinghouse Decision Advisor Process was employed to develop and quantitatively evaluate the financial merits of alternative strategies to address Alloy 600 (& 82/182) degradation issues at the higher-susceptibility locations throughout four Exelon Generation Company nuclear power units. The objective was to help Exelon focus its resources over both locations and time. Recommendations were also provided for the actions to resolve these issues at these highly susceptible locations. The paper describes the actions taken by Exelon that were supported by this study.