bluuebunny/crossref_metadata_2025 huggingface dataset:
Integrity Management of Aged Equipment in Sour Oil and Gas Service: Kuwait Oil Company Approach
Authors: Al-Mithin Abdul Wahab , Al-Ahmad Abdul Wahab , Sardesai Vinayak , Kumar G. Santhosh
Published: January 2011

The professionals across the Oil & Gas industry have strived to combat the adverse effects of Hydrogen Sulphide and maintain the integrity of vital facilities. The detrimental effects of Hydrogen Sulphide on process equipment has remained as one of the prime subject of research in Oil & Gas Industry, fostering the advancement of inspection philosophies and necessitating adoption of appropriate mitigating measures. The challenge of Integrity Management increases with the ageing of equipment and increased influx of sour media. This paper discusses: • The change in operating environment — from non sour to sour. • The actions taken to protect the equipment in the changed operating environment. • Consequent changes in the inspection strategies and techniques to detect material degradation. • Material protection methodologies deployed to manage the age old facilities for safe operation. This paper gives an insight into the various methodologies adopted in KOC to address asset/plant life extension issues of an operating company to tackle the challenges caused by the changed composition of fluids handled. The paper illustrates adoption of appropriate inspection strategies, application of effective internal coating systems, periodic online monitoring of the pressure vessels to improve confidence level with respect to integrity of the vessels without necessitating major changes in the facilities.
Potential for Stress Corrosion Cracking of A537 Carbon Steel Nuclear Waste Tanks Containing Highly Caustic Solutions
Authors: Lam Poh-Sang , Stripling Craig S. , Fisher Donald L. , Elder James B.
Published: January 2011

The evaporator recycle streams of nuclear waste tanks may contain waste in a chemistry and temperature regime that exceeds the current corrosion control program, which imposes temperature limits to mitigate caustic stress corrosion cracking (CSCC). A review of the recent service history found that two of these A537 carbon steel tanks were operated in highly concentrated hydroxide solution at high temperature. Visual inspections, experimental testing, and a review of the tank service history have shown that CSCC has occurred in uncooled/un-stress relieved tanks of similar construction. Therefore, it appears that the efficacy of stress relief of welding residual stress is the primary corrosion-limiting mechanism. The objective of this experimental program is to test A537 carbon steel small scale welded U-bend specimens and large welded plates (30.48 × 30.38 × 2.54 cm) in a caustic solution with upper bound chemistry (12 M hydroxide and 1 M each of nitrate, nitrite, and aluminate) and temperature (125 °C). These conditions simulate worst-case situations in these nuclear waste tanks. Both as-welded and stress-relieved specimens have been tested. No evidence of stress corrosion cracking was found in the U-bend specimens after 21 days of testing. The large plate test was completed after 12 weeks of immersion in a similar solution at 125 °C except that the aluminate concentration was reduced to 0.3 M. Visual inspection of the plate revealed that stress corrosion cracking had not initiated from the machined crack tips in the weld or in the heat affected zone. NDE ultrasonic testing also confirmed subsurface cracking did not occur. Based on these results, it can be concluded that the environmental condition of these tests was unable to develop stress corrosion cracking within the test periods for the small welded U-bends and for the large plates, which were welded with an identical procedure as used in the construction of the actual nuclear waste tanks in the 1960s. The absence of evidence of stress corrosion cracking and general corrosion in the laboratory-scaled specimens indicate that this type of nuclear waste tank is not susceptible to highly caustic solutions up to 12 M hydroxide at 125 °C when sufficient nitrite inhibitor is present.
The Reliability Investigation on the Assessment Criteria for Local Metal Loss Based on API 579-1/ASME FFS-1
Authors: Kaida Takuyo , Mogami Yuichi , Izumi Satoshi , Sakai Shinsuke
Published: January 2011

In order to rationalize the safety margin of pressurized components subject to local metal loss, we applied a probabilistic reliability method. The modeling of the local metal loss is based on API 579-1/ASME FFS-1 and we investigated the safety margin for the assessment criterion quantitatively. We also proposed an assessment curve based on the reliability index for local metal loss. The assessment criterion of API 579-1/ASME FFS-1 can be expressed on the probabilistic basis. The proposed assessment curve based on the probabilistic reliability method can be used to the local metal loss based on the quantified safety margin and it is expected to expand the allowable conditions for the local metal loss compared with those by API 579-1/ASME FFS-1.
Applicability of J Estimation Procedures for Overmatched SE(T) Fracture Specimens
Authors: Paredes Lenin M. , Carvalho Henrique S. S. , Ruggieri Claudio
Published: January 2011

This work focuses on the evaluation procedure to determine the elastic-plastic fracture toughness J-integral based upon the η-method for welded SE(T) fracture specimens. The primary objective of this investigation is to address the significance of the η-factor on accurate and robust estimates of J clamped single edge notch tension (SE(T)) specimens using load-displacement records. Very detailed non-linear finite element analyses for plane-strain models provide the evolution of load with increased load-line displacement to define the applied load as a separable function dependent upon crack geometry, strength mismatch level and specimen deformation. The analyses reveal that η-factors for clamped SE(T) fracture specimens based on load-line displacement (LLD) records and plastic area are relatively insensitive to weld strength mismatch.
Stress Concentration Factors for Crossbores in Thick Walled Cylinders and Square Blocks
Authors: Watanabe Katsumi , Terada Susumu
Published: January 2011

In this study, finite element analyses were performed to investigate the stress concentration factors around cross-bores in thick walled cylinders and square blocks. As for the results, the stress concentration factors are presented relating the wide range of ratios of the diameters of main and sub bores (DH/DI) to some specific values of the wall ratios (Y). The effects of boundary conditions were also investigated. As these obtained stress concentration factors seems to show some differences from these in Appendix J of ASME Section VIII Div.3 [1], the revisions of these stress concentration factors in Div.3 are proposed.
Acoustic Power and Acoustic Pressure in Piping Systems Handling High Velocity Steam and Gases Through a Pressure Reducing Device
Authors: Eisinger Frantisek L. , Sullivan Robert
Published: January 2011

Piping systems handling high-pressure and high-velocity steam and various process and hydrocarbon gases through a pressure reducing device can produce severe vibration of the piping system and noise and pulsation in the surroundings. Utilizing the data published by Carucci, V.A., and Mueller, R.T., 1982, “Acoustically Induced Piping Vibration in High Capacity Pressure Reducing Systems”, ASME Paper No. 82-WA/PVP-8, we develop a relationship for acoustic power and acoustic pressure as a function of the product of the Mach number M and pressure drop Δp (MΔp) through the system. Thirty six cases were evaluated to formulate this relationship.
Shakedown Behavior of a Two-Layer Beam Considering Ductile Damage Under Thermo-Mechanical Loadings
Authors: Zheng Xiao-Tao , Xuan Fu-Zhen
Published: January 2011

In this paper, the nonlinear kinematic hardening material based on Armstrong-Fredrick model coupled with ductile damage was used to estimate the shakedown behavior of a two-layer beam under constant bending moment and cyclic temperature loads. The shakedown limit loads were determined by the accumulation of plastic straining after enough cycles under various loads combination and presented by the Bree-type diagram. The results indicated that the twice yield stress limit of shakedown behavior was considerable conservative compared to the shakedown domain obtained by nonlinear kinematic hardening model without damage. However, shakedown limit loads considering ductile damage decreased remarkably compared with the undamaged material modal, In this case, the twice yield stress limit of shakedown behavior may be not safe for engineering application, especially when the thermal load is prominent.
Finite Element Analysis of the Mechanical Behavior of Multi-Bolted Joints Subjected to Shear Loads
Authors: Fukuoka Toshimichi , Nomura Masataka , Kamihira Takahiro
Published: January 2011

When subjected to external forces, bolted joints behave in a complex manner especially in the case of the joints being clamped with multiple bolts. Friction type joints are widely used for the joints subjected to shear loads. Bearing type joints, which support the shear load on the surface of bolt body, are used less frequently, since its mechanical behavior is too complicated to accurately estimate the load capacity. In this study, mechanical behavior of multi-bolted joints subjected to shear loads is comprehensively analyzed by three-dimensional FEM. Load distribution patterns of the bearing type joint are compared to those of the friction type joint. It has been found that the shear load applied to a bearing type joint distributes with a concave shape along the load direction as clamping bolts are increased, and a fair amount of the shear load is also supported by friction. Additionally, a simple method that calculates the shear load distribution is proposed using elementary theory of solid mechanics, which can estimate the shear load distribution with sufficient accuracy for the case of small friction coefficient.
Study on Collapse Characteristics and CSRF Method for Cylinder-to-Cylinder Intersections by Plastic Analysis
Authors: Kataoka Shunji , Miyakawa Asako , Sato Takuya
Published: January 2011

It is known that the collapse strength of complex three-dimensional structures is hard to be evaluated accurately with elastic analysis, and more accurate results require the use of inelastic analysis. A typical example is cylinder-to-cylinder intersections. These are basic structures of process plant, and can be seen in the branch piping and nozzles of pressure vessels. In the previous work, the authors proposed CSRF (: Collapse strength reduction factor) method which is an alternative to the SIF (: Stress Intensification Factor) method [1]. The CSRF in the previous work was based on the limit load analysis and was defined as the ratio of the simple cylinder limit pressure to the cylinder-to-cylinder limit pressure, which was represented as CSRFL in the paper. The SIF is based on the elastic analysis and is defined as the ratio of the local primary membrane stress (PL) to the general primary membrane stress (Pm). Comparing the SIF with CSRFL, it was revealed that the proposed concept of CSRF method provides more reasonable design than SIF method. The CSRF in the previous work (CSRFL) was, as mentioned above, based on the limit load analysis, and elastic-perfectly plastic constitutive material model was applied. Therefore effect of the material constitutive material model as strain hardening behavior after yielding on the CSRF has not been investigated. In this paper, the concept of CSRF method was extended to general plastic materials in order to confirm the robustness of this method. The elastic plastic analyses were conducted on the cylinder-to-cylinder intersection model considering material and geometric nonlinearity. The values of CSRFPP and CSRFLP were newly proposed. CSRFPP is purely the strength ratio between an intersection and a simple cylinder. CLRFLP is defined as the strength ratio to the simple cylinder based on limit load analysis. The results showed that the collapse pressure of intersections considering material characteristics is larger than that of limit load analysis. As a result, CSRFPP and CSRFLP are smaller than CSRFL in almost every case. Therefore, independently of material characteristics, CSRFPP and CSRFLP methods give more accurate design than CSRFL method.
Evaluation of Torque vs. Closure Bolt Preload for a Typical Containment Vessel Under Service Conditions
Authors: Smith Allen C.
Published: January 2011

Radioactive material package containment vessels typically employ bolted closures of various configurations. Closure bolts must retain the lid of a package and must maintain required seal loads, while subjected to internal pressure, impact loads and vibration. The need for insuring that the specified preload is achieved in closure bolts for radioactive materials packagings has been a continual subject of concern for both designers and regulatory reviewers. The extensive literature on threaded fasteners provides sound guidance on design and torque specification for closure bolts. The literature also shows the uncertainty associated with use of torque to establish preload is typically between 10 and 35%. These studies have been performed under controlled, laboratory conditions. The ability to insure required preload in normal service is, consequently, an important question. The study described here investigated the relationship between indicated torque and resulting bolt load for a typical radioactive materials package closure using methods available under normal service conditions.
Regulatory Issues Associated With Shipment of Small Quantities of Radioactive Material
Authors: Smith Allen C. , Abramczyk Glenn A. , Nathan Stephen J.
Published: January 2011

Following decertification of the ubiquitous and simple Department of Transpsortaion (DOT) 6M specification package, radioactive materials package Shippers have been faced with the need to use Certified Type B packagings. Many Department of Energy (DOE), commercial and academic programs have a need to ship small masses of radioactive material, where the identity of the material or radionuclides is know but the individual activity of some may not be known. For quantities which are small enough to be fissile exempt and have adequate shielding to ensure low radiation levels, these materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the American National Standards Institute (ANSI) N14.5 definition for leak-tight (≤ 1×10−7 ref cm3 air/sec), the 9975, 9977, and 9978 are capable of transporting contents requiring the highest standard of containment. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of such radioactive material are discussed and the logical basis for certification for such contents is described.
Technical Basis for a Flaw Tolerance Option for Code Case N-770-1 for Large Diameter Cold Leg Piping to Main Coolant Pump Welds, With Obstructions
Authors: Bamford Warren , Ganta Reddy , Hall Gordon , Kelley Matthew
Published: January 2011

An extensive series of evaluations have been performed on the Alloy 82/182 dissimilar metal butt welds located at the safe end regions of the CE designed reactor coolant pump suction and discharge nozzles. These nozzles present inspection coverage challenges, which hinder the likelihood of obtaining the required inspection coverage of MRP-139, and the successor document, ASME Code Case N-770. Furthermore, the geometry of the region also contributes to the difficulty of performing standard mitigation techniques. However, these nozzle regions operate at cold leg temperatures, nominally 550°F, and have a very high resistance to the potential for PWSCC, and a low predicted crack growth rate, if such a flaw were to exist in the region. This leads to the suggestion that the required inspection regimen may be too strong for these regions, and the study described herein was structured to investigate that possibility and develop a technical basis for proposing changes to inspection requirements consistent with the flaw tolerance of the region. Specifically, changes to Code Case N-770 are proposed herein, to take advantage of the flaw tolerance of the region. These proposed changes are described in this paper, and the technical basis for them is described in the remainder of the paper. The technical basis rests on three complementary findings: 1. The probability of a flaw existing or initiating in this region is very low; 2. There is a significant margin between the size flaw which would leak at a detectable rate, and the size flaw which would cause the pipe to fail. This provides a significant level of defense in depth for the region; and 3. The flaw tolerance of the region, for both axial and circumferential flaws, is very high, as measured by the size flaw which could grow to the Section XI allowable flaw size for either flaw type.
Technological Advances for Improved Performance and Operation of Fiber-Reinforced Polymer Piping
Authors: Mertiny Pierre , Bashar Mohammad , Parashar Avinash , Juss Kulvinder
Published: January 2011

Fiber-reinforced polymer (FRP) piping has been recognized for excellent corrosion resistance and high specific properties such as its strength-to-weight ratio. Despite the positive characteristics, FRP piping has limited, albeit growing utilization in industrial service. This is in part due to initial cost when compared to conventional metallic pipe. Reduced life cycle expenditures in conjunction with operational advantages may foster an increased implementation of FRP piping. This may be achieved through installation procedures, longevity and operational capabilities that are superior to those related to metallic piping. The present article reviews recent technological advances relating to these attributes, namely improved joining methods; enhanced wear, corrosion and damage resistance; and embedded monitoring systems for wear and other parameters.
Effect of Overload on the Threshold Stress Intensity Factor Range of SUS316 as a Function of Crack Size
Authors: Ando Kotoji , Takahashi Koji , Hashikura Yasuaki , Houjyou Keiji
Published: January 2011

The effects of overload on the threshold stress intensity factor range (ΔKth) in SUS316 were studied. Tensile overload was applied to a CT(Compact tensiton) specimen of SUS316. Then, fatigue tests were carried out to determine the resultant threshold stress intensity factor range (ΔKth). The mechanism of the improvement of ΔKth by the tensile overloading was analyzed using the Dugdale model. It was clarified that the value of ΔKth increases as increasing the overloading. The compressive residual stress intensity factor (Kr) is the key factor controlling the value of ΔKth.
Effect of Shot Peening on Fatigue Limit of Surface Flawed Sample
Authors: Ando Kotoji , Takahashi Fumio , Tange Akira , Ono Yoshiki
Published: January 2011

The effects of shot peening on the fatigue strength of specimens having a semicircular notch of varied surface length, 2a (aspect ratio=1.0), are investigated. In the case of un-peened specimens, the fatigue limit of specimens having a notch of a=0.05mm was equal to that of the un-notched specimens. However, the fatigue limit of a=0.3mm was 46% smaller than that of the un-notched specimens. On the contrary, in the case of peened specimens, the fatigue limit of a=0.2mm was equal to that of the un-notched specimens and furthermore, that of a=0.3mm was only 5% smaller than that of the un-notched specimens. Multiple non-propagating cracks were observed in a peened specimen after fatigue testing. The stress intensity factor of the maximum non-propagating crack size was corresponding to that of a=0.2mm notch. In other words, shot peening increases fatigue limit and decreases the likelihood that a surface flaw will result in failure. The experimental data agree well with the calculation results. This study would be industrially useful for the decision of allowable defect size.
Effect of Crack Extension in a Compact Tension C(T) Specimen Containing a Residual Stress on the Stress Intensity Factors
Authors: Kamel Simon , Davies Catrin , Lee Hyeong , Nikbin Kamran
Published: January 2011

Compact tension C(T) specimens with a blunt notch are mechanically pre-compressed to introduce a tensile residual stress ahead of the notch. The specimens are then thermally soaked which results in crack extension due to the presence of the tensile residual stress. Neutron diffraction measurements are performed on the pre-compressed specimens and after thermal soaking, to determine the redistribution of the residual stress as the crack extends. Stress intensity factors (SIF’s) are estimated using an upper-bound linear distribution to model the redistribution of residual stress as the crack extends. These are compared with the SIF’s using a membrane stress of yield magnitude which ignores redistribution of residual stress with crack extension.
Measurement of the Strain Distribution in Notch Tip Hydrides With High Energy X-Ray Diffraction and the Impact on Overload Modeling
Authors: Kerr Matthew , Tracy Stephanie , Daymond Mark R. , Holt Richard A. , Almer Jonathon D.
Published: January 2011

The formation of notch-tip hydrides in CANDU® Zr-2.5Nb pressure tubes can significantly reduce their resistance to fracture, particularly during overload conditions. This paper outlines recent high energy X-ray diffraction measurements of notch tip strain fields in Zr-2.5Nb specimens, during both hydride growth and overload. The use of this data to validate continuum Finite Element (FE) and possible inclusion in ‘Process Zone’ models of hydride fracture are also discussed.
Experimental Design and Data Collection for Dynamic Fragmentation Experiments
Authors: Weiderhold Joseph , Lambert David E. , Hopson Michael
Published: January 2011

Experiments have been conducted to investigate the fracture and fragmentation characteristics of a liquid phased sintered (LPS) tungsten and high strength steel alloys. Metal cylinders, each of which was 20.32 cm tall and 5.08 cm inner/5.88 cm outer diameter, were explosively driven to failure. Two complimentary types of experiments were conducted in this series to determine input parameters for a related continuum mechanics based modeling effort. Open air experiments utilized ultra-high speed framing photography and a photonic Doppler velocimetry system (PDV). The information from these experiments provided a case wall velocity, relative time of breakup and strain-rate during the stress loading timeframe. Complimentary experiments were conducted in a water tank to perform a soft recovery of the fragments. The fragments were subsequently cleaned, massed, and characterized according to their mass and failure strain distributions. Various methods of analyzing the data (Mott & Weibull distributions) are discussed along with the calibration of the continuum damage model parameters. Results of the failure strain analysis, fragment distribution, and damage model are then supplied for use in subsequent modeling and application designs. Further details of the modeling and simulation approach are outlined in a complimentary set of two papers presented by Lambert [1] and Hopson [2].
A Risk-Uncertainty Formula Accounting for Uncertainties of Failure Probability and Consequence in a Nuclear Powerplant
Authors: Fong Jeffrey T. , Gosselin Stephen R. , Marcal Pedro V. , Filliben James J. , Heckert N. Alan , Chapman Robert E.
Published: January 2011

This paper is a continuation of a recent ASME Conference paper entitled “Design of a Python-Based Plug-in for Benchmarking Probabilistic Fracture Mechanics Computer Codes with Failure Event Data” (PVP2009-77974). In that paper, which was co-authored by Fong, deWit, Marcal, Filliben, Heckert, and Gosselin, we designed a probability-uncertainty plug-in to automate the estimation of leakage probability with uncertainty bounds due to variability in a large number of factors. The estimation algorithm was based on a two-level full or fractional factorial design of experiments such that the total number of simulations will be small as compared to a Monte-Carlo method. This feature is attractive if the simulations were based on finite element analysis with a large number of nodes and elements. In this paper, we go one step further to derive a risk-uncertainty formula by computing separately the probability-uncertainty and the consequence-uncertainty of a given failure event, and then using the classical theory of error propagation to compute the risk-uncertainty within the domain of validity of that theory. The estimation of the consequence-uncertainty is accomplished by using a public-domain software package entitled “Cost-Effectiveness Tool for Capital Asset Protection, version 4.0, 2008” (http://www.bfrl.nist.gov/oae/ or NIST Report NISTIR-7524), and is more fully described in a companion paper entitled “An Economics-based Intelligence (EI) Tool for Pressure Vessels & Piping (PVP) Failure Consequence Estimation,” (PVP2010-25226, Session MF-23.4 of this conference). A numerical example of an application of the risk-uncertainty formula using a 16-year historical database of probability and consequence of main steam and hot reheat piping systems is presented. Implication of this risk-uncertainty estimation tool to the design of a risk-informed in-service inspection program is discussed.
A Transferability Model for Cleavage Fracture Toughness in Strength Mismatched Welds Using the Weibull Stress Approach
Authors: Ruggieri Claudio
Published: January 2011

This work describes the development of a toughness scaling methodology incorporating the effects of weld strength mismatch on crack-tip driving forces. The approach adopts a nondimensional Weibull stress, σw, as a the near-tip driving force to correlate cleavage fracture across cracked weld configurations with different mismatch conditions even though the loading parameter (measured by the J-integral) may vary widely due to mismatch and constraint variations. Application of the procedure to predict the failure strain for an overmatch girth weld made of an API X80 pipeline steel demonstrates the effectiveness of the micromechanics approach. Overall, the results lend strong support to use a Weibull stress based procedure in defect assessments of structural welds.
Initial Development of a High-Speed Thermocouple Data Acquisition System
Authors: Nye Jennifer , Reich Alton , Shaw James
Published: January 2011

The Arnold Engineering Development Center (AEDC) testing complex includes more than 50 wind tunnels, test cells, arc heaters, and other specialized test facilities. Of these, 27 units have capabilities that are unmatched in the United States, and 14 are unmatched in the world. These unique facilities pose equally unique testing challenges, including several related to test preparation. This paper discusses the initial development of a high-speed thermocouple data acquisition system that is based on smart sensor design principals. The initial effort focused on developing an overall system architecture concept and prototyping the thermocouple interface.
Calculation of Working Pressure for Cylindrical Vessel Under External Pressure
Authors: Brar Gurinder Singh , Hari Yogeshwar , Williams Dennis K.
Published: January 2011

Initial geometric imperfections have a significant effect on the load carrying capacity of asymmetrical cylindrical pressure vessels. This research paper presents a comparison of a reliability technique that employs a Fourier series representation of random asymmetric imperfections in a defined cylindrical pressure vessel subjected to external pressure. Evaluations as prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 rules are also presented and discussed in light of the proposed reliability technique presented herein. The ultimate goal of the reliability type technique is to statistically predict the buckling load associated with the cylindrical pressure vessel within a defined confidence interval. The example cylindrical shell considered in this study is a fractionating tower for which calculations have been performed in accordance with the ASME B&PV Code. The maximum allowable external working pressure of this tower for the shell thickness of 0.3125 in. is calculated to be 15.1 psi when utilizing the prescribed ASME B&PV Code, Section VIII, Division 1 methods contained within example L-3.1. The Monte Carlo method as developed by the current authors and published in the literature is then used to calculate the maximum allowable external working pressure. Fifty simulated shells of geometry similar to the example tower are generated by the Monte Carlo method to calculate the nondeterministic buckling load. The representation of initial geometric imperfections in the cylindrical pressure vessel requires the determination of appropriate Fourier coefficients. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion that appears in the form of a double Fourier series. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries and the associated asymmetric imperfections as a function of position within a given cylindrical shell. The method and results described herein are in stark contrast to the dated “knockdown factor” approach currently utilized in ASME B&PV Code.
ASME Code Qualification of Pipe Bends With Localized Wall Thinning
Authors: Abdelsalam Usama , Vijay Dk
Published: January 2011

This paper addresses the application of the design by analysis rules of the ASME B&PV Code SEC III Division 1 NB-3200 criteria on degraded Class 1 piping. A tight radius pipe bend with a local thin area (LTA) located on the inner surface and super imposed on general thinning region is considered using a detailed FEA model implementing idealized smooth axial and circumferential thickness profiles. A location dependent thinning rate function is developed (based on the smoothed profiles and the assumed original thickness distribution) to predict the wall thickness distribution at the end of an arbitrary evaluation period. Internal pressure and dead weight loads are statically applied. Linear elastic analysis is performed and the results are checked against the ASME Code criteria for the primary stress intensity. It is demonstrated that a “local” wall thickness considerably below the pressure based thickness for the corresponding straight pipe segment meets the requirements of the ASME Code SEC III for the primary stresses. The effect of the extent of the local thin area in the axial direction is explored. This paper also compares the allowable pressure obtained from elastic analysis of NB-3221 and limit analysis of NB-3228.1.
Finite Element Analysis of Residual Stresses in Butt Welding of Stainless Steel Plates by GTAW
Authors: Brar Gurinder Singh , Kumar Rakesh
Published: January 2011

Welding is one of the most commonly used permanent joining processes in the piping and pressure vessel industry. During welding a very complex thermal cycle is applied to the weldment, which in turn causes irreversible elastic-plastic deformation and consequently gives rise to the residual stresses in and around fusion zone and heat affected zone (HAZ). Presence of residual stresses may be beneficial or harmful for the structural components depending on the nature and magnitude of stresses. The beneficial effect of compressive stresses have been widely used in industry as these are believed to increase fatigue strength of the component and reduce stress corrosion cracking and brittle fracture. In large steel fabrication industries such as shipbuilding, marine structures, aero-space industry, high speed train guide ways and pressure vessels and piping in chemical and petrochemical industry the problem of residual stresses and overall distortion has been and continue to be a major issue. It is well established fact that material response of structural components is substantially affected by the residual stresses when subjected to thermal and structural loads. Due to these residual stresses produced in and around the weld zone the strength and life of the component is reduced. As AISI 304 stainless steel has excellent properties like better corrosion resistance, high ductility, excellent drawing, forming and spinning properties, so it is almost used in all types of application like chemical equipment, flatware utensils, coal hopper, kitchen sinks, marine equipment etc. But because of the problems of residual stresses during the time of welding it is very essential to understand the behavior and nature of AISI 304 stainless steel material. So in order to overcome all these problems a 3-dimensional finite element model is developed in a commercially available FEA code by drafting an approximate geometry of the butt welded joint and then the finite element analysis is performed, so that one can understand the complete nature of residual stresses in butt welding of AISI 304 stainless steel plate. In this paper, butt welding simulations were performed on two AISI 304 stainless steel plates by gas tungsten arc welding (GTAW). Analysis of butt welded joint by commercially available finite element analysis code showed that butt weld produced by GTAW resulted in 782.84 MPa of residual stress in plates. In addition, the residual stress is plotted against axial distance to have a clear picture of the magnitude of residual stress in and around weld area.
Break Exclusion in French EPR: A Step by Step Analysis
Authors: Faidy Claude
Published: January 2011

After a lot of analysis in USA, USNRC has accepted a leak before break procedure developed in NUREG 1061 [4]. An updated version is under preparation in particular for operating plant to include field experience, but also for future plants. In France, EDF and AREVA have developed a new “break exclusion” methodology in 2 steps: - the safety consideration and defense in depth principle, - a modernized procedure to justify applications of break exclusion on a given system. Finally, large piping double-ended guillotine break are not considered in EPR-FA3 as a design transient. The Large Break Loss of Coolant Accident (LB-LOCA) is not considered as a design transient, only safety analysis with realistic data and methods are performed. For EPR-FA3, the larger break considered is the double end guillotine break of the surge line (16″ instead of 28 or 32″). The paper will present the details of these 2 step procedure and will discuss possible extension to other piping systems.
Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics
Authors: Chen Bo-Yi , Huang Chin-Cheng , Chou Hsoung-Wei , Liu Ru-Feng , Lin Hsien-Chou
Published: January 2011

The Chinshan boiling water reactor (BWR) units 1 and 2, owned by Taiwan Power Company (TPC), started commercial operations in 1978 and 1979, respectively. The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast-neutron fluence exposure. This effect should be considered in the life extension and license renewal application. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture Analysis of Vessels – Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, chemistry components, neutron fluence and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08, is found to have the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging analysis results for the life extension and the license renewal applications.
Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event
Authors: Chou Hsoung-Wei , Huang Chin-Cheng , Chen Bo-Yi , Liu Ru-Feng , Lin Hsien-Chou
Published: January 2011

With the development of probabilistic fracture mechanics (PFM) methods in recent years, the risk-informed approach has gradually been used to evaluate the structural integrity and reliability of the reactor pressure vessels (RPV) in many countries. For boiling water reactor (BWR) pressure vessels, it has been demonstrated that it is not necessary to perform the inservice inspections of beltline circumferential welds to maintain the required safety margins because their probability of failure is orders of magnitude less than that of beltline vertical welds, thus may well reduce the associated substantial cost and person-rem exposure. In Taiwan, however, the inservice inspections of shell welds still have to be performed every ten years per ASME Boiler and Pressure Vessel Code, Section XI inspection requirements for a BWR type Chinshan nuclear power station. In this work, a very conservative PFM model of FAVOR code consistent with that USNRC used for regulation is built with the plant specific parameters concerning the beltline shell welds of RPVs of Chinshan nuclear power station. Meanwhile, a hypothetical transient of low temperature over-pressure (LTOP) event which challenges the BWR RPV integrity most severely is also assumed as the loading condition for conducting the PFM analyses. Further, the effects of performance of inservice inspection are also studied to determine the benefit of the costly inspection effort. The computed low probability of failure indicates that the analyzed RPVs can provide sufficient reliability even without performing any inservice inspection on the circumferential welds. It also indicates that performing the inservice inspections can not promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on RPV degradation and also be helpful for the regulation of BWR plants in Taiwan.
Buckling Sensitivity Analysis and Optimal Design of Two-Layered Shells Under Circumferential Strain Load
Authors: Papadakis George
Published: January 2011

Shell buckling under circumferential strain loading appears often in many biomedical problems, such as collapse of asthmatic airways, esophagus etc. Currently there is no analytic solution from which the critical buckling load can be calculated from the geometrical and material properties of the shell layers. The theoretical analysis leads to the numerical solution of a generalized eigenvalue problem. The purpose of the present paper is to present a method with which the sensitivity of the buckling load (minimum positive eigenvalue) to uncertainties of the geometrical and material parameters can be calculated. It is shown that the sensitivity can be evaluated with the aid of the left (adjoint) eigenvectors. The proposed method is validated against separate global computations using the finite difference method. Finally, we employ the information on sensitivity to optimize the thickness of the outer shell layer in order to maximize the buckling load.
Seismic Safety Margin Assessment for Electrical Components Utilizing Full-Scale Test
Authors: Iijima Toru , Nakagawa Masaki , Shibuya Akira , Ebisawa Katsumi , Kameda Hiroyuki
Published: January 2011

In Japan, the Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities (2006) requires that residual risk for earthquakes beyond design base be considered. Moreover, in the Niigata-ken Chuetsu-oki earthquake (NCOE:2007), the earthquake motion exceeded the seismic design condition of the Kashiwazaki-Kariwa nuclear power plants (K-K NPPs). In response to these issues, there is a growing demand for quantitative clarification of seismic safety margin. The Japan Nuclear Energy Safety Organization (JNES) started a study on the seismic safety margin. JNES defined the term “seismic safety margin” in this study. The seismic safety margin is based on the probability distribution of seismic response and seismic capacity of equipment. Regarding the seismic capacity, JNES has carried out seismic capacity tests on various types of equipment whose malfunction would significantly affect core damage frequency in terms of seismic probabilistic safety assessment (seismic PSA). Our seismic safety margin is effective to understand the quantitative margin related to the failure of equipment. This paper presents examples of seismic safety margin evaluation. Specifically, JNES has applied the evaluation concept of the seismic safety margin to a metal-clad switchgear and a reactor protection rack utilizing full scale model test results.
Seismic Safety Margin Assessment for Thin-Wall Cylindrical Tanks Utilizing Ultimate Strength Test
Authors: Iijima Toru , Nakagawa Masaki , Shibuya Akira , Ebisawa Katsumi , Kameda Hiroyuki
Published: January 2011

In Japan, the Seismic Design Guideline for Nuclear Power Plants revised in 2006 requires that residual risk for earthquakes beyond design base be considered. Moreover, in the Niigata-ken Chuetsu-oki earthquake (NCOE:2007), the earthquake motion exceeded the seismic design condition of the Kashiwazaki-Kariwa nuclear power plants (KK NPPs). In response to these issues, there is a growing demand for quantitative clarification of seismic safety margin. The Japan Nuclear Energy Safety Organization (JNES) started a study on the seismic safety margin. JNES defined the term “seismic safety margin” in this study. The seismic safety margin is based on the probability distribution of seismic response and seismic capacity of equipment. Regarding the seismic capacity, JNES has carried out seismic capacity tests on various types of equipment whose malfunction would significantly affect core damage frequency in terms of seismic probabilistic safety assessment (seismic PSA). Our seismic safety margin is effective to understand the quantitative margin related to the failure of equipment. JNES applied the concept of seismic safety margin to thin wall cylindrical tanks utilizing ultimate strength test results, and compared it with the design margin. This paper reports examples of the seismic margin evaluation of cylindrical tank.
Improving the Scaling of Two-Phase Damping by Local Observations in a Tube Bundle
Authors: Pascal-Ribot Sylviane , Blanchet Yves
Published: January 2011

Experimental data are reported to investigate the dissipation mechanisms that govern two-phase damping and fluid-elastic instability of a single flexible tube in a rigid array. The working fluid is an air-water mixture and the void fraction and interfacial velocity are measured using a bi-optical probe (BOP) positioned upstream of the flexible tube. The present work aims at revisiting the problem of fluid-elastic instability by developping various scaling models of two-phase fluid damping before the onset of instability. For most of the experiments, the measured damping factor was seen to increase with increasing bubble chord length, with decreasing superficial liquid velocity, and with decreasing amplitude of vibration. The Connor’s approach has been generalized to the two-phase flows provided that the reduced velocity is calculated with gas velocity and with mixture density deduced from the local void fraction measured inside the bundle. The collapse of the fluid-elastic data is more satisfactory than when using the Homogeneous Equilibrium Model (HEM). Void fraction, gas velocity, relative velocity, liquid superficial velocity, bubble chord length, vibratory frequency are shown to be relevant parameters to reduce the two-phase damping data. The use of these parameters in non-dimensional numbers such as Capillary number, Reynolds number, pressure ratio, mass ratio leads to helpful observations as well as several promising approaches to the reduction of two-phase damping.
Fracture Toughness of Pressure Boundary Steels With Higher Yield Strength
Authors: Gupta Karan K. , Hoffmann Christopher L. , Hamilton Andrew M. , DeLose Frank
Published: January 2011

ASME Section III Appendix G provides a fracture mechanics methodology for evaluating the fracture resistance of pressure vessel materials based on comparing the applied stress intensity from service induced loadings on the assumed flaw to a material fracture toughness (KIR) curve. The applicable fracture toughness curve described in Appendix G is defined as a lower bound curve for static, dynamic, and crack arrest fracture toughness tests for a database that includes a number of pressure vessel materials having specified minimum yield strengths of 50 ksi or less. For materials which have specified minimum room temperature yield strength greater than 50.0 ksi but not exceeding 90.0 ksi, the same fracture toughness curve may be used provided fracture mechanics data points are obtained on at least three heats of the material on a sufficient number of specimens to cover the temperature range of interest, including the weld metal and heat affected zone, and provided that the data points are equal to or above that of the fracture toughness curve of Appendix G of the ASME Code (Fig. G-2210-1). At present, the pressure boundary components of steam generators and pressurizers typically use SA-508 Grade 3 Class 2 forgings and SA-533 Type B Class 2 plate material with minimum yield strength of 65 ksi and 70 ksi respectively. The fracture toughness for these materials is not readily available. This technical paper is compilation of fracture toughness of forging and plate material, weld metal, and heat affected zone of such higher strength forgings and plate materials. The paper includes toughness data of weld metal and heat affected zone resulting from high heat input process and includes the toughness test data with long term postweld heat treatment. All dynamic fracture toughness values for the ASME forgings with a minimum yield strength of 65 ksi and plate material with a minimum yield strength of 70 ksi, heat affected zone, and weld metals exceed the ASME specified minimum KIR given in Article G-2000 of the ASME Boiler and Pressure Vessel Code. This data concludes that the ASME specified minimum reference KIR curve can be used conservatively for the forging and plate materials with minimum yield strength of 65 ksi and 70 ksi that are currently specified for construction of steam generators and other Section III, Class 1 components.