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IN THIS VOLUME


2002;():3-10. doi:10.1115/PVP2002-1215.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: evaluating variations in plastic strain increments and evaluating variations in the width of elastic core. To verify the validity of these criteria, calculations were performed for several typical models in C-TDF [2]. This paper shows the results of a simple cylinder model. Cyclic plastic analyses were performed applying sustained internal pressure and alternating linear temperature distribution through the wall. Analyses were performed with various load ranges to evaluate the precise ratchet limit and its behavior across the limit. Both pressure and thermal stress were given parameters. In the analyses, Elastic-Perfectly-Plastic (EPP) material was used and also strain hardening material for comparison. The ratchet limit in the Code [3] is based on Miller’s theoretical analysis [4] for a cylinder assuming a uni-axial stress state, whereas real vessels are in multi-axial stress state. By our calculations, we also examined the ratchet limit in real vessels. The results show that for the cylinder in a multi-axial stress state, the ratchet limit rises 1.2 times the ratchet limit by the Code. The evaluation results show that variations in equivalent plastic strain increments can be used for ratchet criterion and ratcheting can be assessed by confirming the presence of elastic core in the second cycle.

Commentary by Dr. Valentin Fuster
2002;():11-16. doi:10.1115/PVP2002-1216.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1],[2] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: (1) Evaluating variations in plastic strain increments, and (2) Evaluating the width of the area in which Mises equivalent stress exceeds 3Sm . To verify of these criteria, we selected notched cylindrical vessel models as prime elements. To evaluate the effect of the local peak stress distribution on these criteria, cylindrical vessels with a semicircular notch on the outer surface were selected for this analysis. We used two notch configurations for our analysis, and the stress concentration factor for the notches was set to 1.5 and 2.0. We conducted elastic-plastic analysis to evaluate the ratchet limit. Sustained pressure and alternating enforced longitudinal displacements which causes secondary stress were used as parameters for the elastic-plastic analysis. We found that when no ratchet was observed, the equivalent plastic strain increments decreased and the area in which Mises equivalent stress exceeds 3Sm are below the certain range.

Commentary by Dr. Valentin Fuster
2002;():17-22. doi:10.1115/PVP2002-1217.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: evaluating variations in plastic strain increments and evaluating variations in the elastic core region. To verify the validity of these criteria, calculations were performed for several typical models in C-TDF [2]. This paper shows calculations and evaluation results of a Flat Head Vessel for shakedown. To study shakedown criteria for gross structural discontinuity, a flat head vessel is surveyed. The flat head vessel consists of a stiff flat head and a shell and is subjected internal pressure and thermal cycle. The elastic shakedown area and the plastic area are compared and plastic strain increments are surveyed. A shakedown evaluation method based on distribution of elastic-plastic strain range is proposed.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2002;():23-29. doi:10.1115/PVP2002-1218.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: evaluating variations in plastic strain increments and evaluating variations in the elastic core region. To verify the validity of these criteria, calculations were performed for several typical models in C-TDF [2]. This paper shows calculations and evaluation results of 2-dimensional and 3-dimentinal nozzles for shakedown and Ke-factors, as defined by equation 2. Two models are used. One is a 2-dimensional (2D) axi-symmetric model of a typical nozzle. The other is a 3-dimantional (3D) model of the nozzle of which shell radius is half of 2-dimensional model. The primary and secondary stress, shakedown analyses using elastic-plastic FEA and Ke-factors which are directly calculated from elastic-plastic FE analyses are surveyed. The results show that the alternative criteria are applicable for those models. The analysis results of the 2D model show good relation to those of the 3D model.

Topics: Nozzles
Commentary by Dr. Valentin Fuster
2002;():31-37. doi:10.1115/PVP2002-1219.

This paper discusses the background of a new approach to check the admissibility of pressure vessels via the Direct Route (DR) to Design by Analysis (DBA), and the various design checks required, in general and in detail. Emphasis is on the various tools and procedures used in applications, and on typical applications.

Topics: Design
Commentary by Dr. Valentin Fuster
2002;():39-46. doi:10.1115/PVP2002-1220.

The rules in codes such as the ASME Boiler and Pressure Vessel Code Section III Division 1 and Section VIII Division 2, provide the concept of stress categorization to prevent inelastic failure modes based on the elastic analyses. The categorization of the stresses obtained by the FEM analysis, however, is not always clear and the Three Dimensional FEM Stress Evaluation in JPVRC (TDF committee) has been developed alternative criteria to dispense with the stress categorization. As for the evaluation of the primary plus secondary stress, criteria based on the concept of the Cyclic Yield Area (CYA) have been developed. In this paper, the recent results obtained in the committee are summarized to evaluate the validity and the usability of the criteria.

Commentary by Dr. Valentin Fuster
2002;():47-55. doi:10.1115/PVP2002-1221.

ASME B&PV Code directives for shakedown and ratcheting evaluation are reviewed. The objective is to assess their effectiveness when executed with plastic finite element analysis (FEA) and to propose procedures when they are inadequate. At first, they are applied to cases involving cyclic primary loading only, for which shakedown is evaluated. It is found that FEA is not suitable to determine whether plastic shakedown is achieved for a given loading. An alternate approach based on physical grounds leads to the conclusion that if only primary loading is cycled, and if the loading does not exceed the design limit load, plastic shakedown is always achieved. Then the directives are applied to cases in which cyclic thermal or other displacement-controlled loading is superimposed on primary loading with a mean, which may lead to ratcheting. Three ratcheting measures are applied to a thermal transient example. It is found that FEA is not suitable for establishing the absence of ratcheting of a vessel for a given loading by requiring zero growth of its dimensions. The ratcheting check is modified by specifying an acceptable limit on the increments of the ratcheting measure that are predicted by plastic FEA within a specified number of cycles, which makes it practical for design purposes. A decision is required on an acceptable growth of the diameter of a vessel that would not endanger the serviceability of the vessel during its life.

Commentary by Dr. Valentin Fuster
2002;():57-64. doi:10.1115/PVP2002-1222.

This paper discusses the evaluation criteria for alternating loads utilizing partial inelastic analyses and free from the stress classification. As finite element analysis becomes popular, it has been noticed by designers that in some cases the conventional stress classification does not work well. The stress classification itself had been engineered as a practical tool to evaluate the integrity of a structure by elastic analyses, which actually could have inelastic behavior. For example, primary stress limits were determined reflecting the stress level at collapse. Therefore, the problem concerning the stress classification can be solved recalling how it had been engineered. In other words, the key to solve the problem is the inelastic evaluation method corresponding to each stress category. From this point of the view, the application of the inelastic analyses becomes widely studied. Consequently, as for primary loads, it has been proven that the collapse load evaluation by Limit or Plastic Analysis is effective and practical for design analyses. On the other hand, as for the alternating loads, it is not sufficiently discussed how the alternative criteria should be without stress classification. In this paper, the following are discussed based on the calculation results in the Committee on Three Dimensional Finite Element Stress Evaluation in JPVRC. 1. Prerequisite of the elastic-plastic analysis for shakedown evaluation, and the evaluation criteria based on plastic strain increment and its distribution. 2. The advantage to use simplified elastic-plastic analysis method than to perform fully elastic-plastic analyses, and the calculation procedure for Ke factors to be used with. The associated code rules are proposed.

Commentary by Dr. Valentin Fuster
2002;():65-72. doi:10.1115/PVP2002-1223.

The paper discusses the load carrying capacity of toroidal shells with closed circular cross-section and loaded by static external pressure. Details about the manufacturing, pre-experiment measurements and testing of three, nominally different, steel toroids are provided. Two of them were manufactured from mild steel by spinning two halves and then welding them around the inner and outer equatorial perimeters. The third one has been assembled by welding four 90 deg, stainless elbows. The outer diameter of these models was about 300 mm and the wall thickness varied from 2.0 mm to 3.0 mm. The hoop radius-to-thickness ratio, r/t, varied from about 15 to 30. The experimental collapse pressures were in the range from 4 MPa to 8 MPa. Comparisons with numerical results are also provided.

Topics: Collapse
Commentary by Dr. Valentin Fuster
2002;():73-82. doi:10.1115/PVP2002-1224.

This paper proposes a simple two-surface model for cyclic incremental plasticity based on combined Mroz and Ziegler kinematic hardening rules under nonproportional loading. The model has only seven material constants and a nonproportional factor which describes the degree of additional hardening. Cyclic loading experiments with fourteen strain paths were conducted using Type 304 stainless steel. The simulation has shown that the model was precise enough to calculate the stable cyclic stress-strain relationship under nonproportional loadings.

Commentary by Dr. Valentin Fuster
2002;():83-88. doi:10.1115/PVP2002-1225.

Creep damage of high energy piping (HEP) systems in fossil fuel power plants results from operation at creep range temperatures and stresses over many years. Thermal expansion stresses are typically below the yield stress and gradually relax over time. Consequently, the operating stresses in a piping system are typically below the yield stress and become load controlled. Conventional designs of HEP systems use the American Society of Mechanical Engineers B31.1 Power Piping Code. The Code is a general guideline for piping system design. Utilities typically determine examination sites by performing Code piping stress analyses and selecting locations that include the highest sustained longitudinal stress, highest thermal expansion stress, and terminal points. However, the Code does not address weldment properties, redistribution of thermal stresses and time-dependent life consumption due to material creep degradation. As an alternative, a high energy piping life consumption (HEPLC) methodology was used to predict maximum material damage locations. The methodology was used to prioritize expected creep damage locations, considering applicable affects such as weldment properties, field piping displacements, time-dependent operating stresses, and multiaxial piping stresses. This approach was applied to the main steam piping system at Cholla Unit 2. The locations of highest expected creep damage would not have been selected by a conventional site selection approach. Significant creep damage was found at the locations of maximum expected creep damage using the HEPLC methodology.

Topics: Creep , Welded joints , Pipes
Commentary by Dr. Valentin Fuster
2002;():91-98. doi:10.1115/PVP2002-1226.

Fatigue crack growth rate tests were performed on a 304 stainless steel compact tension (CT) specimen in water with 40–60 cc/kg H2 . Data in the literature for CT tests show minor environmental effects in hydrogenated water, but higher effects in oxygenated water. However, the PWR data presented by Bernard, et al (1979) were taken at low stress ratios (R = 0.05) and high stress intensity levels (ΔK = 16–41 MPa√m). The purpose of these tests is to explore the crack growth rate characteristics of 304 SS in hydrogenated water at higher R values (0.7 and 0.83) and lower ΔK values (11.0 and 7.7 MPa√m) Each set of R, ΔK conditions were tested at frequencies of 0.1, 0.01 and 0.001 Hz. The results show a pronounced effect on crack growth rates when compared to available literature data on air rates.

Commentary by Dr. Valentin Fuster
2002;():99-107. doi:10.1115/PVP2002-1227.

The fatigue life of steel in elevated temperature water is strongly affected by the composition of the environmental water, temperature and strain rate. The effects of these parameters on fatigue life reduction have been investigated experimentally in these years. One problem to be discussed is the fact that the previous experiments which leaded main conclusions on the environmental effects were generally executed by keeping a set of experimental parameters constant. In the transient condition in an actual plant, however, such parameters as temperature and strain rate are not constant. In order to evaluate fatigue damage in an actual plant on the basis of experimental results under constant temperature and constant strain rate conditions, the modified rate approach method was developed. The method can be applicable to changing temperature condition and strain rate condition separately. In the present study, an additional model is proposed with considering that both temperature and strain rate change simultaneously in an actual plant. The applicability of this method is discussed and verified experimentally. The fatigue lives predicted by this method are scattered within the factor of 2 or 3 bands against test results even when several parameters changed synchronously.

Commentary by Dr. Valentin Fuster
2002;():109-117. doi:10.1115/PVP2002-1228.

The fatigue life of austenitic stainless steel has recently been shown to undergo remarkable reduction with decrease in strain rate and increase in temperature in water. Either of these parameters as a factor of this reduction has been examined quantitatively and methods for predicting the fatigue life reduction factor Fen in any given set of conditions have been proposed. All these methods are based primarily on fatigue data in simulated PWR water owing to the few data available in simulated BWR water. Recent Japanese fatigue data in simulated BWR water clearly indicated the effects of the environment on fatigue degradation to be milder than under actual PWR conditions. A new method for determining Fen in BWR water was developed in the present study and a revised Fen in PWR water is also proposed based on new data. These new models differ from those previously used primarily with regard to the manner in which strain amplitude is considered to affect Fen in the environment.

Commentary by Dr. Valentin Fuster
2002;():119-132. doi:10.1115/PVP2002-1229.

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components and specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, and flow rate, on the fatigue lives of these steels are summarized. Statistical models are presented for estimating the fatigue ε–N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue design curves.

Commentary by Dr. Valentin Fuster
2002;():133-142. doi:10.1115/PVP2002-1230.

This paper examines the mechanism of fatigue crack initiation in austenitic stainless steels (SSs) in light water reactor (LWR) coolant environments. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, level of dissolved oxygen in water, and material heat treatment on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The influence of reactor coolant environments on the formation and growth of fatigue cracks in polished smooth SS specimens is discussed. Crack length as a function of fatigue cycles was determined in air and LWR environments. The results indicate that decreased fatigue lives of these steels are caused primarily by the effects of the environment on the growth of cracks <200 μm and, to a lesser extent, on enhanced growth rates of longer cracks. A detailed metallographic examination of fatigue test specimens was performed to characterize the fracture morphology. Exploratory fatigue tests were conducted to enhance our understanding of the effects of surface micropits or minor differences in the surface oxide on fatigue crack initiation.

Commentary by Dr. Valentin Fuster
2002;():143-150. doi:10.1115/PVP2002-1231.

The flow rate of water flowing over a steel surface is considered to be one of the most important factors influencing the fatigue life of the steel, because the water flow produces differences in the local environment. The effect of the water flow rate on the fatigue life of carbon, low alloy, and austenitic stainless steels was therefore investigated experimentally. Fatigue testing of low (S = 0.008 wt%) and high (S = 0.016 wt%) sulfur content carbon steels and a low alloy steel was performed at 289°C for various dissolved oxygen concentrations (DO) of less than 0.01 and 0.05, 0.2, and 1 ppm, and at various water flow rates. Three different strain rates of 0.4, 0.01, and 0.001%/s were used in the fatigue tests. For high sulfur carbon steel (S = 0.016 wt%), the effect of a high water flow rate on mitigating fatigue life reduction was more clearly observed at a lower strain rate, irrespective of the DO. This effect of high water flow rate was most notable at a DO of 0.2 ppm, which was the DO level that produced a significant sulfur effect. This indicates that the mechanism responsible for the mitigation of fatigue life reduction is the flushing effect of the water, which eliminates the locally corrosive environment. For high sulfur carbon steel (S = 0.016 wt%), no benefit of a high water flow rate was found at a DO of 0.01 ppm. This was because the environmental effect is insignificant at this low DO level. For low sulfur carbon steel (S = 0.008 wt%) and low alloy steel (S = 0.008 wt%), a high water flow rate had little effect on mitigating fatigue life reduction even at a DO of 0.2 ppm. This indicates that the sulfur is much less influential in low sulfur steel than in high sulfur steel. Fatigue testing of Type 316 nuclear grade stainless steel (316NG) and Type 316 stainless steel (SUS316) was performed at 289°C and 320°C for DO levels of less than 0.01 and 0.05, and 0.2. For austenitic stainless steel, no mitigating effect at a high water flow rate was found. It should be noted rather that there is a possibility that a high water flow rate decreases the fatigue life because a tendency to a slight decrease in fatigue life with an increasing flow rate was observed.

Commentary by Dr. Valentin Fuster
2002;():151-164. doi:10.1115/PVP2002-1232.

Fatigue crack propagation (FCP) rates for 304 stainless steel (304SS) were determined in 24°C and 288°C air and 288°C water using double-edged notch (DEN) specimens of 304 stainless steel (304 SS). Tests performed at matched loading conditions in air and water at 288°C with 20–60 cc H2 /kg H2 O provided a direct comparison of the relative crack growth rates in air and water over a wide range of crack growth rates. The DEN crack extension ranged from short cracks (0.03–0.25 mm) to long cracks up to 4.06 mm beyond the notch, which are consistent with conventional deep crack tests. Crack growth rates of 304 SS in water were about 12 times the air rate. This 12X environmental enhancement persisted to crack extensions up to 4.06 mm, far outside the range associated with short crack effects. The large environmental degradation for 304 SS crack growth is consistent with the strong reduction of fatigue life in high hydrogen water. Further, very similar environmental effects were reported in fatigue crack growth tests in hydrogen water chemistry (HWC). Most literature data in high hydrogen water show only a mild environmental effect for 304 SS, of order 2.5 times air or less, but the tests were predominantly performed at high cyclic stress intensity or equivalently, high air rates. The environmental effect in low oxygen environments at low stress intensity depends strongly on both the stress ratio, R, and the load rise time, Tr , as recently reported for austenitic stainless steel in BWR water. Fractography was performed for both tests in air and water. At 288°C in water, the fracture surfaces were crisply faceted with a crystallographic appearance, and showed striations under high magnification. The cleavage-like facets on the fracture surfaces suggest that hydrogen embrittlement is the primary cause of accelerated cracking.

Commentary by Dr. Valentin Fuster
2002;():165-170. doi:10.1115/PVP2002-1233.

A test apparatus was developed to study the interaction between corrosion fatigue (CF) and stress corrosion cracking (SCC) in high-temperature water simulated boiling water reactor environment. Tests were conducted using 1/2T-CT samples of both low alloy and sensitized stainless steels under 3 different types of loading at 0.2–8 ppm in dissolved oxygen concentrations at 563 K in water. Type 1 was a normal cyclic loading test of constant amplitude, Type 2 a monotonic constant loading rate test, and Type 3 a combination of Type 1 + Type 2 loading modes. In the low alloy steel, no striking interaction was observed between CF and SCC, whereas in the case of Type 3 loading condition crack growth rates of the sensitized stainless steel were as much as 3 times higher than those for Type 1 + Type 2. The mechanism of the CF and SCC interaction is discussed.

Commentary by Dr. Valentin Fuster
2002;():171-190. doi:10.1115/PVP2002-1234.

Fatigue design rules for welds in the ASME Boiler and Pressure Vessel Code are based on the use of Fatigue Strength Reduction Factors (FSRF) against a Code-specified fatigue design curve generated from smooth base metal specimens without the presence of welds. Similarly, Stress Intensification Factors (SIF) that are used in the ASME B31 Piping Codes are based on component S-N curves with a reference fatigue strength based on straight pipe girth welds conducted by Markl et al in 1950s. Typically, the determination of either the FSRF or SIF requires extensive fatigue testing to take into account the stress concentration effects associated with various types of component geometry, weld configuration, and loading conditions. As the fatigue behavior of welded joints is being better understood, it has been generally accepted that the difference in fatigue lives from one type of weld to another is dominated by the difference in stress concentration. However, general finite element procedures are currently not available for effective determination of such stress concentration effects. This is mainly due to the fact the stress solutions at a notch (e.g., at weld toe) are strongly influenced by mesh size at and near a weld, resulted from notch stress singularity. In this paper, a mesh-insensitive structural stress method is used to re-evaluate the S-N test data ranging from Markl et al in 1950s, to those by Heald and Kiss on nuclear piping in 1970, to the most recent piping weld S-N data by Scavuzzo et al in 1998. The major findings are as follows: (a) The mesh-insensitive structural stress method provides a simple and effective mean for characterizing stress concentrations at vessel and pipe welds. (b) The structural stress based parameter provides an effective measure of stress intensity at welds, which can be related to fatigue lives. (c) Once the mesh-insensitive structural stress are used, the S-N data processed thus far can be reasonably consolidated into one narrow band. Therefore, single master S-N curve for vessel and piping welds can now be established, regardless of piping weld types or geometries (straight pipe girth welds, different types of flange welds, elbow welds, Mitre bends, etc.), and can be used to general a master fatigue design curve.

Commentary by Dr. Valentin Fuster
2002;():191-198. doi:10.1115/PVP2002-1235.

A procedure to calculate crack initiation probabilities by creep-fatigue damage is explained in this paper with a calculation example. Material properties of 316FR are determined as probabilistic distributions from test data. As the result yield stress, fatigue property, cyclic stress-strain relation and creep property are input into a creep-fatigue evaluation as the probabilistic distributions. The crack initiation probability is calculated with the condition for the Japanese commercialized sodium cooled fast breeder reactor. As the result, the allowable thermal cycles in the present design becomes 5 × 10−9 cumulative crack initiation probability.

Commentary by Dr. Valentin Fuster
2002;():199-208. doi:10.1115/PVP2002-1236.

A database of fatigue strength in air and high temperature water environments for the materials used in LWR structural components has been developed. It includes the fatigue data acquired in the EFT project entrusted by METI, Japan, as well as public data obtained from domestic and overseas literature. One of the features of the database is that it includes about 2,500 images such as scanning electron micrographs of the fractured surface of the specimen after fatigue test. Another feature is that the program can search data, draw graphs such as S-N curves and hysteresis loops automatically, and make calculations of the hysteresis energy equivalent to the area of hysteresis loop, for example. The structure of the database, the contents of the data compiled, and the functions of the program are presented with actual examples, and the characteristics of the database are discussed.

Commentary by Dr. Valentin Fuster
2002;():209-217. doi:10.1115/PVP2002-1237.

To test the flow accelerated corrosion in nuclear power plant environment with ease, High Temperature Rotating Cylinder Electrode (HTRCE) was developed. The main design concept of HTRCE is to assure stable operation of working electrode up to highest possible rotation speed in a severe environment, to insulate electrode housing except working electrode surface against external fluid, and to extract corrosion parameter from the rotating cylinder to outside of the autoclave safely. From the results of corrosion experiment at high temperature water, HTRCE has been proved as an effective device to evaluate the velocity sensitivity in high temperature water environment.

Commentary by Dr. Valentin Fuster
2002;():221-227. doi:10.1115/PVP2002-1238.

There are numerous instances in which in-service flaws due to various kinds of damage and deterioration are found in equipment as a result of in-service inspections. The proper evaluation of such flaws is extremely important. Fitness-for-Service (FFS) codes, such as ASME B&PV Code Sec. XI and JSME S NA1 for nuclear power generation facilities and BS 7910 and API-RP579 for general industrial facilities, are available. In light of such circumstances, the High Pressure Institute of Japan (HPI) has prescribed its code “Assessment procedure for crack-like flaws in pressure equipment” for conducting quantitative safety evaluations of flaws detected in common industrial pressure components such as pressure vessels, piping, storage tanks, and so on designed and fabricated in accordance with Japanese codes and regulations such as JIS B8265 and High Pressure Gas Safety Law. The FFS code consists of Level 1 assessment (whereby assessment can be conducted without extensive knowledge of fracture mechanics) and Level 2 assessment (which enables more detailed fracture mechanics analyses and is currently being studied). The allowable flaw size is specified in accordance with the plate thickness. The required impact absorbed energies based on material strength, whether or not PWHT has been done and the orientation of the flaw in relation to the weld seam, are also specified. An approximated equation of stress intensity factor for an embedded flaw near the surface has been derived. The re-characterization procedure for assessing an embedded flaw has been clarified. The flaw can be judged to be acceptable if its size is less than that of an allowable flaw and the equipment is to be used at temperatures exceeding the temperature (MAT) at which the material absorbed energy meets the required impact absorbed energy.

Commentary by Dr. Valentin Fuster
2002;():229-240. doi:10.1115/PVP2002-1239.

In year 2000, ASME Code (Section VIII – Div. 1), CODAP (French Code) and UPV (European Code for Unfired Pressure Vessels) have adopted the same rules for the design of U-tube tubesheet heat exchangers. Three different rules are proposed, based on different technical basis, to cover: • Tubesheet gasketed with shell and channel. • Tubesheet integral with shell and channel. • Tubesheet integral with shell and gasketed with channel or the reverse. At the initiative of the author, a more refined technical approach has been developed, to cover all tubesheet configurations. The paper explains the rationale for this new design rule which is being incorporated in ASME, CODAP and UPV in 2002. This is substantiated with comparisons to TEMA Standards and a benchmark of numerical comparisons.

Commentary by Dr. Valentin Fuster
2002;():241-264. doi:10.1115/PVP2002-1240.

This is part 2 of a two part paper that outlines the 101 essential elements that need to be in place, and functioning well, to effectively and efficiently , preserve and protect the reliability and integrity of pressure equipment (vessels, exchangers, furnaces, boilers, piping, tanks, relief systems) in the refining and petrochemical industry. Part 1 of this paper was published in the proceedings of the 2000 ASME PVP Conference. Each of the two parts outline half of the 101 essential elements of pressure equipment integrity management (PEIM). This paper is not just about minimum compliance with rules, regulations or standards; rather it is about what needs to be accomplished to build and maintain a program of operational excellence in pressure equipment integrity that will permit owner-users to make maximum use of their physical assets to generate income. Compliance is not the key to success in pressure equipment integrity management (PEIM ); operational excellence is. Each of the 101 work processes outlined in this two part paper, is explained concisely to the extent necessary, so that owner-users will know what needs to be done to maintain and improve their PEIM program. This paper does not prescribe how each of these 101 key elements is to be accomplished, as that description would result in a book rather than a paper. This paper simply outlines all the fundamentals that are necessary to avoid losses, avoid safety incidents, and maintain reliability of pressure equipment. It pulls together a complete overview of the entire spectrum of programs, procedures, and preventative measures needed to achieve first quartile performance in maintaining pressure equipment integrity (PEI ).

Commentary by Dr. Valentin Fuster
2002;():265-272. doi:10.1115/PVP2002-1241.

This paper introduces a newly started study that aims to expand the basic concept of the system based code for structural integrity, which has been proposed by Asada et al [1,2], for application to fast breeder reactors (FBRs) [3]. The System Based Code for FBR (FBR System Based Code) offers more reasonable structural integrity assessment methods for relatively important components of FBR than current codes and standards do, to enable profound improvement both in reliability and lifetime cost-effectiveness. It will be realized by margin optimization through an integrated evaluation of all the technical fields that affect structural integrity at the stage of design. Those fields are the prerequisites of design, material, design analysis, fabrication and installation, preservice inspection, operation, inservice inspection, and repair and replacement. For margin optimization, three promising methods, failure probability assessment, application of quality assurance index, and the introduction of “systematized design factors” were proposed. The FBR System Based Code will consist of a control code and partial codes. The former takes care of margin optimization while the latter offers various options of technologies and engineering tools among which a designer can chose the most appropriate one according to various requirements to be fulfilled. Partial codes will be developed for each technical field that is dealt with in the code. Technologies that need to be developed for the development of the FBR system Based code were also clarified.

Commentary by Dr. Valentin Fuster
2002;():275-281. doi:10.1115/PVP2002-1242.

The objective of this paper is to design and analyze a horizontal tank on saddle supports. The horizontal vessel is to store various chemicals used in today’s industry. The over all dimensions of the horizontal vessel are determined from the capacity of the stored chemicals. These dimensions are first determined. The design function is performed using the ASME Code Sec VIII Div 1. The horizontal tank design is broken up into (a) shell design, (b) two elliptical heads and (c) two saddle supports. The designed dimensions are used to recalculate the stresses for the horizontal vessel. The dimensioned horizontal vessel with saddle supports and the saddle support structure is modeled using STAAD III finite element software. The stresses from the finite element software are compared with the stresses obtained from calculated stresses by ASME Code Sec VIII Div 1 and L. P. Zick’s analysis printed in 1951. The difference in the stress value is explained. This paper’s main objective is to compare the code design to the finite element analysis. The design is found to be safe for the specific configuration considered.

Commentary by Dr. Valentin Fuster
2002;():283-291. doi:10.1115/PVP2002-1243.

The compression ring at a tank’s roof-shell junction has been designed using the API 620 [1] rules. The geometrical configuration of the ring is non-conventional compared to the typical shapes recommended in API 620. In order to verify the structural integrity of the roof-shell junction, a plastic FEA was conducted to establish the maximum allowable working pressure (MAWP) using a limit load analysis procedure. The FEA results indicated that the tank’s roof-shell junction’s MAWP is 18.7 psi which exceeds the calculated MAWP of 15 psi using the API 620 rules. This paper presents the comparisons of the MAWPs calculated using the API 620 rules and the plastic FEA.

Commentary by Dr. Valentin Fuster
2002;():293-299. doi:10.1115/PVP2002-1244.

Recent experience has shown that many utilities have installed new feedwater flow measurement instrumentation, which is designed with new, proven technology that more accurately monitors feedwater flow and therefore, allows for improved thermal power level calculations. As a result of this new approach, many utilities have been able to extend limited power upgrades to the tune of 1.4%, and yet with no noticeable additional environmental impact. These electric power increases are generally attributable to the large design margins included in the original nuclear power plant designs and in addition, to the technological advances that have been made to the nuclear industry. Power increases are now carried out by several nuclear utilities in the United States with the NRC’s concurrence. Current nuclear power plants are providing cheap, reliable and affordable electricity to help meet current energy growing demands. As a result of this, power uprate increases are encouraged. Current economic conditions strongly favor power uprates and plant life extensions. In the early 1990s, a limited power uprate program was initiated, but due to complex design reviews, this was not steadfastedly recommended during this time period. Nuclear Steam Supply Systems (NSSS) vendors performed a key role in this nuclear power uprate program. This paper discusses structural and piping qualification review required to achieve these power uprate programs, currently being performed by nuclear utilities in the United States.

Commentary by Dr. Valentin Fuster
2002;():303-305. doi:10.1115/PVP2002-1245.

A recent revision to A-4300 of Appendix A to Section XI of the ASME Boiler and Pressure Vessel Code specifies a limiting threshold value of the stress intensity factor range below which crack propagation under cyclic loading can be neglected. The threshold value decreases with increasing R ratio, where R is defined as the ratio of the minimum to the maximum stress intensity factor. This paper reviews the crack closure phenomenon that is the underlying reason for this effect, summarizes the low-alloy pressure vessel steel test data upon which the threshold limit is based, and proposes an empirical equation for the threshold stress intensity factor range that is a reasonably conservative limit. The threshold equation incorporated into Appendix A will also replace the more restrictive threshold formula currently in Code Case N-643.

Commentary by Dr. Valentin Fuster
2002;():307-312. doi:10.1115/PVP2002-1246.

If the flaws detected during in-service inspection are multiple discrete flaws that are in close proximity to one another, the flaws are evaluated as to whether they are combined or not, in accordance with combination rules in the ASME Code. The combination rules require that multiple flaws shall be treated as a single combined flaw if the distance between the adjacent flaws is equal to or less than the dimension of the flaw depth. After the coalescence of the multiple flaws, the flaw length becomes larger and then the stress intensity factor of the combined flaw would be expected to be significantly larger. Stress intensity factors for two surface flaws and a combined flaw under membrane stress and bending stress were analyzed using influence function method. From the calculation results of the stress intensity factors for two flaws and the combined flaw, it is shown that less conservative combination rules are appropriate, as compared to the existing combination rules in the ASME Code.

Topics: Stress
Commentary by Dr. Valentin Fuster
2002;():313-318. doi:10.1115/PVP2002-1247.

This paper will review the development of the visual examination requirements of Section XI of the ASME Boiler and Pressure Vessel Code. The original visual requirements were ‘one visual exam fits all’ – to detect physical damage, physical displacement, and evidence of leakage. Resolution requirements were those of Section V of the Code. These requirements evolved over the next 20 years to become several specific types of requirements, each with specific resolution, illumination, and proximity restraints. Review indicated that these separate visual rules are now converging. This paper provides recommendations for revisions to Section XI that consolidate and simplify these requirements.

Commentary by Dr. Valentin Fuster
2002;():321. doi:10.1115/PVP2002-1248.
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The ASME does not require design calculations for Section X, Class I vessels. Design qualification is by destructive testing of a prototype vessel. The candidate vessel undergoes repetitive pressure testing up to the design pressure for as many as 100,000 cycles, depending on the type of FRP laminate in the vessel. If the vessel passes the fatigue test, it is then pressurized to six times the design pressure. If it also passes this test, vessels identical to the prototype may be built and receive the code mark. The prototype does not receive a code stamp. Rigorous quality assurance requirements insure that the production vessels are the same as the prototype. The ASME does require design calculations for Class II vessels. It also has stringent quality assurance and inspection requirements. Every Section X Class II vessel must also pass an acoustic emission examination before it receives a code stamp, thereby providing experimental verification of the structural integrity of the vessel. Class I and Class II vessels have different size and pressure scopes. Class I has no geometric limitations. The maximum design pressure for Class I is limited to 150 psig for bag-molded, centrifugally cast, and contact molded vessels, 1500 psig for filament-wound vessels, and 3000 for psig filament-wound vessels with polar boss openings. Class II vessels may be designed by rules or by stress analysis. For vessels designed by rules the diameter must lie between 6 in. and 96 in., and the maximum design pressure is 75 psig. Vessels that are designed by stress analysis must have diameters between 6 in. and 144 in. Their maximum design pressure varies with diameter: from 6 in. to 36 in. the maximum design pressure is 200 psig., and from 36 in. to 144 in. the design pressure is given by P = 7200/D where P is in psi. and D is in inches. RTP-1 has no size restrictions, but design pressure is limited to 15 psig. In both classes of Section X and in RTP-1, external design pressure must be less than 15 psig. Design calculations for FRP differ from calculations for metal vessels because FRP behaves differently from steel and other vessel metals. The material behaviors are compared and the resulting calculation differences are discussed.

Commentary by Dr. Valentin Fuster
2002;():323. doi:10.1115/PVP2002-1249.
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Over the past several years, many industries have grown to recognize that Fiber Reinforced Plastic (FRP) pressure vessels must be built to established industrial safety standards to help ensure consistently safe products. End Users and Engineers familiar with Section VIII of the ASME Code typically turn to Section X as the standard recognized to govern the fabrication of fiber-reinforced vessels. However there tends to be confusion concerning Section X and how design integrity is maintained. There is a belief held by some that a composite pressure vessel designed in accordance with the Section X, Class I meets the essence of the Code. The feeling is that complete compliance is an unnecessary expense and third party certification is of minimal value. Section X is very specific in pointing out the fundamental error in this thinking. Section X recognizes that, unlike metal construction, the fabricator of a fiberglass vessel is responsible for the creation of a new and very temperamental material every time a part is fabricated. With this chance of inconsistency, even a fundamentally sound design can be executed poorly and with disastrous results. The purpose of this paper will be to describe the design and procedure qualification process used for Class I pressure vessels and how the integrity of the design in maintained throughout the fabrication of ASME Code Stamped pressure vessels.

Commentary by Dr. Valentin Fuster
2002;():325. doi:10.1115/PVP2002-1250.
FREE TO VIEW

The ASME accredits RL Industries to fabricate both Section X, Class II and RTP-1 tanks and vessels, one of only two manufacturers in the world to hold this distinction. The design, quality assurance, inspection and testing requirements for the two stamps are similar, but not identical. The same manufacturing processes can be used to produce equipment to either standard, but the record keeping required by the ASME is somewhat different. Similarly, the design methods are not identical. This presentation describes the similarities and differences. Both standards are much newer than the other Sections of the Boiler and Pressure Vessel Code, and RL Industries pioneered in their introduction. The manufacturing, quality assurance, testing and inspection of FRP to the standards will be described. Vessels with a Section X or RTP-1 stamp are as safe and reliable as equipment built to other Sections of the Code.

Commentary by Dr. Valentin Fuster
2002;():329. doi:10.1115/PVP2002-1251.
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The nuclear industry has been aggressively pursuing risk-informed regulation for the past seven years. In this pursuit, the Nuclear Regulatory Commission (NRC) has established a basic framework for reviewing and granting risk-informed regulations. Focused regulatory guides have been specifically prepared in the areas of Plant Technical Specifications, In-Service Testing, In Service Inspections and Graded Quality Assurance. While these regulations have not supplanted traditional deterministic regulations, they have been used to support focused changes to the plant design basis. As the industry continues to mature the use of PSA information in day to day plant operations will grow. Today, risk insights are required by federal regulation for all plant maintenance activities. In addition, the regulator reserves the right to include consideration of risk in applications where the outcome could have a significant impact on risk. Despite, the major strides made in development and use of risk information, the industry and the regulator were operating without an agreed to PSA standard. Over the past three years the ASME has formed a committee of stakeholders, both commercial and regulatory, to develop a workable standard for the development and utilization of PSA data in the nuclear industry. The ASME PSA standard has recently been issued. The current standard has been developed to support licensing applications and is focused on the development and use of the Level 1PSA and the calculation of the Large Early Release Frequency (LERF). The ASME standard is unlike most standards in that it is tiers, and includes guidance for using results when specific items in the standard are not in complete compliance with specific standard elements. The tiers included in the standard are reflective of the level of detail in the PSA elements. The ability to use PSAs with many elements acceptable only at the lowest tier will be more limited than for more sophisticated PSAs and therefore, applications may be limited in scope and would likely involved strong deterministic support as well. As PSA tiers increase the reliance of the decision on the PSA may increase. The acceptability of the PSA elements is established via peer review process. It is the intent of this panel discussion to explore the implications of the recently released ASME PSA standard, and other focused standards under development on the nuclear industry and the role of the ASME standard in the associated regulatory process. The panelists will explore expectations of the industry, needs of the regulator and challenges of the PSA peer review process.

Commentary by Dr. Valentin Fuster
2002;():331. doi:10.1115/PVP2002-1252.
FREE TO VIEW

After several years of intense labor by many industry people, ASME is about to issue its newly approved PRA standard. This standard is for probabilistic risk assessment (PRA) for nuclear power plant applications. It is not a standard on how to build a PRA model; although, that could be inferred from the standard’s technical requirements. This Standard sets forth requirements for PRAs used to support risk-informed decisions related to design, licensing, procurement, construction, operation, and maintenance. It also prescribes a method for applying these requirements depending the degree to which risk information is needed and credited.

Commentary by Dr. Valentin Fuster
2002;():335-344. doi:10.1115/PVP2002-1253.

In this paper both dynamic shaking and cyclic static component test data are used to develop a modified ASME Code moment capacity equation for seismic loading on piping systems. The aim of this modified equation is to conservatively achieve a seismic capacity margin corresponding to a 1% probability of failure of less than about two.

Commentary by Dr. Valentin Fuster
2002;():345-352. doi:10.1115/PVP2002-1254.

The ASME Code for seismic design of piping system was revised in 1994 based on evaluation of pipe component test results submitted by the Piping and Fitting Dynamic Reliability Program (PFDRP). PFDRP indicated piping component failure to result in most cases from fatigue with ratchet. Excessive progressive deformation was noted by this program to ultimately occur in test #37, #39 and #40 piping models. This mode of failure was considered due to superposition of bending moment arising from vertical load produced by weight and horizontal seismic inertia force. To clarify the conditions leading to such failure, elastic-plastic dynamic analysis using the general-purpose FEM Code was carried out on the test#37 piping model of PFDRP. The analytical model and method were verified as effective means of study by comparison of analytical results with those obtained experimentally. The parameter determinations were made under condition of variation in dead weight stress and excitation and its dominant frequency. 1994 ASME seismic stress limits were shown effective for preventing excessive progressive deformation in tests #37, #39 and #40.

Commentary by Dr. Valentin Fuster
2002;():353-362. doi:10.1115/PVP2002-1255.

This paper presents an analytical study of frequency effects on seismic margins of piping components. The study is based on response data obtained as part of a joint Electric Power Research Institute (EPRI) and NRC Piping and Fitting Dynamic Reliability (PFDR) Program. The majority of the PFDR component tests were performed using a narrow-banded earthquake excitation input that was tuned to have a frequency slightly lower than the fundamental frequency of the test components. However, the natural frequency of a piping system in an actual plant may vary over a wide range. Therefore the seismic margins at off-resonance conditions are of importance. Two seismic margin definitions are examined. The primary objective of this study was to extrapolate the PFDR test margins to other frequency regions and to investigate the effects of various parameter changes on the margins.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2002;():363-367. doi:10.1115/PVP2002-1256.

The new rules for seismic piping design in Section III that were developed and included in the requirements in 1994 Addenda of the ASME Boiler and Pressure Vessel Code (B&PV Code) generated considerable discussion within the industry and from the United States Nuclear Regulatory Commission, (USNRC). The USNRC initiated a review of the results of the previous EPRI/NRC experimental program and the Japanese industry started its own experimental program. To accommodate and address developments resulting from these efforts, the ASME, B&PV Code established a Special Working Group (SWG) to continue the review and study of the questions and information generated. This paper reports on the efforts of this SWG which resulted in refinements of the revised rules. These refinements have been accepted for inclusion in Section III of the ASME, B&PV Code.

Commentary by Dr. Valentin Fuster

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