Codes and Standards

2007;():3-15. doi:10.1115/PVP2007-26124.

Acceptance Standards in Section XI of the ASME Boiler and Pressure Vessel Code have an important role as the first step in the flaw evaluation procedure. When a flaw size is within the allowable flaw size in the Acceptance Standard, the flaw is acceptable and analytical evaluation is not required. Although ASME Section XI has Acceptance Standards for Class 1 piping in IWB-3500, there are no Acceptance Standards for Class 2 and 3 piping. Furthermore, the development of the current Acceptance Standards for Class 1 piping was based on flaw detectability by ultrasonic inspection and consideration of fracture mechanics. In this paper, the development of proposed new Acceptance Standards for Class 2 and 3 piping, as well as for Class 1 piping, is described. The development methodology is based on a fracture mechanics approach. For Class 1 piping with high fracture toughness, the allowable flaw sizes were determined by limit load solution. For Class 1 piping, the intent was to maintain overall consistency with the current Acceptance Standards. Proposed Acceptance Standards for Class 2 and 3 austenitic piping were also developed by the methodology used to develop the proposed new Acceptance Standards for Class 1 piping. Allowable flaw sizes for both surface flaws and subsurface flaws for preservice and inservice examinations were developed.

Commentary by Dr. Valentin Fuster
2007;():17-22. doi:10.1115/PVP2007-26189.

There is a rule in Acceptance Standards of Class 1 ferritic vessel in the ASME B&PV Code Section XI describing that allowable subsurface flaws near a component surface are transformed to allowable surface flaws. This is the flaw-to-surface proximity rule for the subsurface flaws. If the subsurface flaw is located closer to the component surface, the subsurface flaw not allowed close to the component surface but may be acceptable after transforming it to a surface flaw. This is a discontinuity of a subsurface flaw in the current Acceptance Standard for Class 1 ferritic vessels. Authors had developed a solution for the discontinuity and proposed the proximity rule at the last ASME PVP 2006 conference. After the discussion at the ASME Code Working Group on Flaw Evaluation, the proposed proximity rule was revised. This paper describes the revision of new flaw-to-surface proximity rule to minimizing the changes to the current allowable subsurface flaws of the Acceptance Standards.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2007;():23-29. doi:10.1115/PVP2007-26205.

ASME Code Section XI Nonmandatory Appendix C [1] formalized evaluation of flaws in piping for justification of continued service of piping components with an identified crack-like flaw. The revision of this appendix in 2004 was a significant improvement in the evaluation methodology for both flawed austenitic stainless steel and ferritic steel pipe depending upon the failure mode governed by limit load (fully plastic), elastic-plastic fracture mechanics, or linear elastic fracture mechanics. The appendix also provides a screening procedure to determine failure mechanism and a procedure for flaw modeling based on the estimated flaw size at the end of a specified evaluation period. The purpose of this paper is to propose an improvement to the limit load method applicable to screened-in carbon steel, wrought stainless steel base material, stainless steel weld material with nonflux weld, and cast products in which the ferrite content is less than twenty percent. In addition, changes in the formulation are proposed to extend the methodology to non-crack-like flaws. Both crack-like and non-crack-like circumferential flaws in the piping are analyzed to simplify formulation for flaw evaluation. The paper concludes that the proposed formulation improves efficiency of the application of Appendix C methodology for crack-like flaw and non-crack-like flaw evaluations.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2007;():31-37. doi:10.1115/PVP2007-26366.

Tabulation solution for ductile tearing failure of piping axial flaws was first implemented in the A90 Addendum to the 1989 Edition of the ASME B&PV, Section XI, Appendix H. The solution was based on J-Tearing analyses of pipe axial semi-elliptical part-through wall flaws in a generic ferritic material that has J1C > = 600 in-lbs/in2 . Analytical solution for pipe axial flaw has not been implemented in the Code. In this paper a general load multiflier Z-factor is developed for use in an analytical solution for pipe axial part-through wall flaw, which is compatible with the current Elastic-Plastic Fracture Mechanics (EPFM) tabular solution in the Section XI, Appendix C. The Z-curve is developed using the same J-Tearing analysis technique and the same generic material properties for ferritic materials that were used in the development of the Code tabulation solutions. The predicted failure stresses using the Z curve are fairly corelated and conservatively the actual failure stress in the available test specimens from the PIFRAC database.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2007;():39-48. doi:10.1115/PVP2007-26427.

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Commentary by Dr. Valentin Fuster
2007;():49-60. doi:10.1115/PVP2007-26696.

Cracking and occasional leaks have been reported in some Boiling Water Reactor (BWR) control rod drive (CRD) stub tubes. Roll expansion of the housing against the Reactor Pressure Vessel (RPV) bottom head penetration has been used successfully to provide a leak barrier. The recently approved ASME Code Case N-730 “Roll expansion of Class 1 Control Rod Drive (CRD) Bottom Head Penetrations in BWRs, Section XI, Division 1” provides the specific criteria for the application of roll expansion. The minimum roll band length in the Code Case was based on the requirement that the roll joint capability exceed the scram forces on the CRD. The roll joint capability was based on a simplified analytical model with assumed friction factors. The predictive model was then compared with the results of extensive testing on mockups. This paper describes the results of the testing that has been performed to determine the load capability of roll repairs for different roll band lengths, material combinations (stainless steel and Alloy 600), percent wall thinning, thermal cycling and surface condition. The mock-ups were rolled using procedures and rolling equipment similar to those used in actual plant application. The mock-ups were tested in a testing machine by applying a ‘push force’ on the housing. In addition to measuring the force using a load cell, strain gages were also used to measure the strains on the housings. LVDTs were used to monitor the displacement during the test. The results showed that the resistance of the rolled joint (i.e. the load capability) is proportional to the roll length. The load capability was not a strong function of wall thinning or thermal cycling. It was strongly affected by the surface condition (e.g. oxidation) and the housing material yield strength. The predictive model was consistent with the test results and confirmed that the roll expansion joint has substantial load capability. Thus, the roll joint is not only a leak barrier, but also a structural load-carrying joint that is sufficient to resist the upward scram loads on CRDs.

Commentary by Dr. Valentin Fuster
2007;():61-66. doi:10.1115/PVP2007-26733.

One of the ways that the ASME Section XI code incorporates elastic-plastic fracture mechanics (EPFM) in the Section XI Appendix C flaw evaluation procedures for circumferential cracks is through a parameter called Z-factor. This parameter allows the simpler limit-load (or net-section-collapse) solutions to be used with a multiplier from EPFM analyses. Traditionally the EPFM solution was determined by using the GE-EPRI J-estimation scheme to determine the maximum load by EPFM, and Z = limit load / EPFM solution. The Z-factor is a function of the material toughness as well as the pipe diameter. With the advent of primary water stress-corrosion cracks (PWSCC) in pressurized water reactor (PWR) dissimilar metal welds (DMW), there is a need to develop Z-factors for Alloy 82/182 nickel-based alloy welds that are susceptible to such cracks. Although there have been Z-factor solutions for cracks in stainless and ferritic pipe butt welds, the DMW are somewhat different in that there is a much lower yield strength material on one side of the weld (typically forged or wrought 304 stainless steel) and on the other side of the weld the low alloy steel has a much higher strength than even the weld metal. This paper shows how 3D finite element analyses were used for a particular pipe size to determine the sensitivity of the crack location in the Alloy 182 weldment (crack in the center of weld, or closer to the stainless or low alloy steel sides), and how an appropriate stress-strain curve was determined for use in the J-estimation schemes. A Z-factor as a function of the pipe diameter was then calculated using the LBB.ENG2 J estimation scheme using the appropriate stress-strain curves from the finite element analysis. The LBB.ENG2 analysis was used rather than the GE-EPRI estimation scheme since it has been found that the LBB.ENG2 analysis is more accurate when compared with full-scale pipe tests. From past work, the GE-EPRI method was found to be the most conservative of the J-estimation schemes in predicting the maximum loads for circumferential flaws when compared to full-scale circumferentially cracked-pipe tests. The proposed Z-factor relationship should be restricted to normal operating temperatures (above 200C) with low H2 concentrations, where the Alloy 182 weld metal exhibits high toughness.

Commentary by Dr. Valentin Fuster
2007;():69-86. doi:10.1115/PVP2007-26798.

The commercial Light Water Reactors operating within the United States have been in service from about 20 to 35 years. These plants include buried Service Water piping systems primarily made from low carbon steel. This piping has been subject to aging over the years, resulting in degradation and corrosion that will require replacement of the piping. Due to the advantageous cost and durability of High Density Polyethylene (HDPE) piping (as demonstrated in other commercial industries), ASME code inclusion of this piping is logical. Duke Power industry has expressed interest in replacing a portion of their steel buried Service Water Piping in Nuclear Power Stations with HDPE pipe. To assist in this effort EPRI has funded and supported the work summarized in this paper to develop design criteria for HPDE Pipe and has teamed with EPRI to develop appropriate ASME Code requirements. Other nuclear utilities will follow once HDPE piping is included in the ASME Code. This paper includes proposed allowable limits of all modes of failure and provides design criteria for HDPE pipe made from PE 3408 resin. It also provides the technical basis for the proposed criteria. This paper deals primarily with the actual design of the piping. The methods included comply with ASME Power Piping Code, B31.1-2004 and Section III of the ASME Boiler and Pressure Vessel Code. Extensive use was made of industrial research, data and experience over 40 years of use of high-density polyethylene piping. Allowable stresses are based on data published in these sources for Design and Service Levels A-D.

Commentary by Dr. Valentin Fuster
2007;():87-96. doi:10.1115/PVP2007-26799.

The commercial Light Water Reactors operating within the United States have been in service from about 20 to 35 years. These plants include buried Service Water piping systems primarily made from low carbon steel. This piping at several plants has been subject to aging over the years, resulting in degradation and corrosion that may require replacement of the piping. Due to the advantageous cost and durability of High Density Polyethylene (HDPE) piping (as demonstrated in other commercial industries), the nuclear power industry has expressed interest in replacing steel buried Service Water Piping in Nuclear Power Stations with HDPE Pipe. To assist in this effort EPRI has funded and supported the work summarized in this paper to develop design criteria for HPDE Pipe. The paper provides design criteria for High Density Polyethylene (HDPE) pipe made from PE 3408 resin. It also provides the technical basis for the proposed criteria. This paper deals primarily with the design of the piping in relation to its interface with the soil in which it is buried. The criteria primarily is derived from current analysis methodology for steel and concrete buried pipe while incorporating changes required to account for the properties and behavior of HDPE pipe. The proposed analysis methodology described herein has evolved into a proposed ASME Boiler and Pressure Vessel Code, Section III, Division I, Design Code Case for consideration by the Section III, Subcommittee on Nuclear Power.

Commentary by Dr. Valentin Fuster
2007;():97-108. doi:10.1115/PVP2007-26800.

The commercial Light Water Reactors operating within the United States have been in service from about 20 to 35 years. These plants include buried Service Water piping systems primarily made from low carbon steel. This piping has been subject to aging over the years, resulting in degradation and corrosion that will require replacement of the piping. Due to the advantageous cost and durability of High Density Polyethylene (HDPE) piping (as demonstrated in other commercial industries), the industry has expressed interest in replacing steel buried Service Water Piping in Nuclear Power Stations with HDPE piping. To assist in this effort EPRI has funded and supported the work summarized in this paper to develop design criteria for HPDE Pipe. This paper provides an example problem demonstrating the application of recently developed design criteria for HDPE piping. The technical bases of these criteria are presented in separate papers and are not repeated in this discussion.

Commentary by Dr. Valentin Fuster
2007;():111-122. doi:10.1115/PVP2007-26100.

Low cycle fatigue life of structural materials diminishes remarkably as functions of various parameters in high temperature water simulating LWR coolant. Such reduction was estimated by the fatigue life reduction factor (Fen ) and the equations to calculate Fen were developed and have undergone revision over the past ten years. The authors have endeavored to establish the method assessing fatigue damage at LWR power plants for the past 13 years in the Japanese EFT (Environmental Fatigue Tests) project under the financial support from the JNES (Japan Nuclear Safety Organization). The project terminated at the end of March in 2007. Final proposals of Fen equations were established for carbon, low-alloy, and austenitic stainless steels and nickel base alloys based on all the data obtained in the project. As the results, a small change in saturated strain rate for carbon and low-alloy steels in highly dissolved oxygen water and newly revised equations including slight change in saturated strain rate for stainless steels in BWR water as well as those for nickel base alloys were proposed. The difference between revised and previous model is essentially not large.

Commentary by Dr. Valentin Fuster
2007;():123-131. doi:10.1115/PVP2007-26101.

The fatigue life reduces remarkably with reduction in strain rate in simulated light water reactor (LWR) water but the effects of strain wave form on this reduction are still not clear. This paper provides fatigue life data obtained from stepwise strain rate change tests, sine wave tests and strain holding tests. The effects of varying strain rate on fatigue life reduction can be estimated very well by the modified rate approach (MRA) method in the case of the step wise strain rate changing as shown in authors’ previous papers [1, 2, 3, 4, 5]. In the case of sine wave, however, the fatigue life reduction is much less compared to that predicted by the MRA method. The mechanism of such difference is not clear and the quantitative assessment of the fatigue life reduction caused by irregular strain wave form in actual transient seems impossible. The current MRA method gives always conservative assessment for sine wave straining and thus it is judged that this method need not be revised any more. The fatigue life reduction caused by strain holding at the peak of straining cycle in simulated BWR water had been reported in the previous paper [6]. In actual thermal transients, however, strain is not usually held at the peak of straining cycle but at the point somewhat reduced from the peak after the stabilization of temperature. In considering this phenomenon, additional fatigue tests in which the strain was held at the point somewhat reduced from the peak were carried out. In such conditions, the fatigue life reduction caused by strain holding disappeared. The similar fatigue tests with peak strain holding were also carried out in simulated PWR water and no fatigue life reduction can be observed. Considering the effects of strain holding on fatigue, the model for evaluating fatigue life reduction in LWR water was revised.

Commentary by Dr. Valentin Fuster
2007;():133-141. doi:10.1115/PVP2007-26143.

Recently published Draft Regulatory Guide DG-1144 by the NRC provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light water reactor (LWR) environment. The analytical expressions and further details are provided in NUREG/CR-6909. In this paper, the environmental fatigue rules are applied to a BWR feedwater line. The piping material is carbon steel (SA333, Gr. 6) and the feedwater nozzle material is low alloy steel (SA508 Class 2). The transients used in the evaluation are based on the thermal cycle diagram of the piping. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. In both cases the cumulative fatigue usage factors are shown to be less than 1.0.

Commentary by Dr. Valentin Fuster
2007;():143-149. doi:10.1115/PVP2007-26185.

Reference fatigue crack growth rate (FCGR) curves for ferrite and austenitic stainless steels in light water reactors environments are prescribed in JSME S NA1-2004 in Japan. The reference FCGR curves in the environment in pressurized water reactors (PWR) are determined as functions of the stress intensity factor range, temperature, load rising time and stress ratio. However, similar reference FCGR curve for nickel-based alloys for PWR environment are not prescribed. In order to propose reference curve in PWR environment, fatigue tests of nickel-based alloys in a simulated PWR primary water environment were conducted. The results of the study show that FCGR in a PWR primary water environment increases with decreasing cyclic loading frequency f, increasing stress ratio R, and increasing temperature Tc .

Commentary by Dr. Valentin Fuster
2007;():151-159. doi:10.1115/PVP2007-26186.

Japanese reference fatigue crack growth rate (FCGR) curves for ferrite and austenitic stainless steels in light water reactor environments are prescribed in JSME S NA1-2004. However, similar reference FCGR curve for nickel-based alloys for pressurized water reactors (PWR) are not prescribed. In order to propose reference FCGR curve for nickel-based alloys, under high stress ratio and low rising time, the effect of the welding method, the effect of specimen orientation and low stress intensity range fatigue crack propagation tests of nickel-based alloys 600, 132 and 82 weld metals were conducted as part of the Environmental Fatigue Test (EFT) projects of Japan Nuclear Energy Safety Organization (JNES). The results show that the effect of heat, welding methods, specimen orientations and environmental water conditions on the FCGR was not significant for Alloys 600, 132 and 82. The FCGR increased with increase of stress ratio, and cyclic loading frequency. According to the procedure for determining the reference FCGR curve of austenitic stainless steels in PWR environment of nickel-based alloys is proposed based on the reference data and the results of this study. The reference FCGR curve for nickel-based alloys in PWR environment are determined as a function of stress intensity factor range, temperature, load rising time and stress ratio.

Commentary by Dr. Valentin Fuster
2007;():161-172. doi:10.1115/PVP2007-26210.

As a part of the 2006 ASME Code support being provided by the Materials Reliability Program (MRP) Fatigue Issue Task Group (ITG), and later the Technical Support Committee (TSC), it is desired to develop a solution that establishes the most severe transient for design purposes when environmental fatigue rules are considered. This problem does not have an obvious answer, since the environmental fatigue multiplier (e.g., Fen ) expressions depend on the strain rate during a transient. The strain rate in a thermal transient is dependent on the ramp time of the temperature change. Classically, fatigue analysis has been performed by conservatively considering that temperature changes are instantaneous (e.g., ramp time = zero). This results in maximizing the stress response. But, all other things being equal, Fen effects are minimum at instantaneous changes. Previous work performed by the MRP to support 2005 ASME Code activities has investigated how Fen × Usage varies with ramp time and has concluded that Fen × Usage maximizes at a small but definitely non-zero ramp time [1]. The implications of non-zero transient ramp times are that difficulties arise in both specifying ahead of time and qualifying the component for appropriate ramp times and, at the same time, not creating a situation whereby plant operations are required to proceed at a specified pace to remain design compliant. Therefore, it is desirable to have the qualifying fatigue analyses cover all conceivable ramp times such that the operator neither has to: (a) be limited to a minimum pace, nor (b) confirm through observations that the pace is at least as fast as assumed in the design. This paper summarizes qualifying fatigue analyses that have been performed using piping methodologies to define bounding ramp times for a variety of piping geometry and material configurations. The intent of these analyses is to provide the component designer with a set of parametric tools that can be used to easily design components without the need for iterative fatigue analyses to determine the bounding conditions when Fen rules are considered. In addition, the tool developed to perform the parametric analyses is available for future use by the designer should more specific analyses be required.

Topics: Fatigue , Pipes
Commentary by Dr. Valentin Fuster
2007;():173-190. doi:10.1115/PVP2007-26211.

Draft Regulatory Guide DG-1144 “Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors”, July 2006 [1], and Associated Basis Draft Document NUREG/CR-6909 (ANL-06/08), “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials”, July 2006 [6] provided methods for addressing environmentally assisted fatigue (EAF) in all new nuclear plant designs. In these documents, a new model was proposed that more accurately accounts for actual plant conditions. The new model includes an EAF correction factor, Fen , which is different from Fen methods previously and currently being considered for adoption into the ASME Code. The Fen methods proposed in DG-1144 are also different than the Fen methods utilized by license renewal applicants, as required by the Generic Aging Lessons Learned (GALL) report [2], as documented in NUREG/CR-5704 [4] (for stainless steel) and NUREG/CR-6583 [3] (for carbon and low alloy steels).

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2007;():191-199. doi:10.1115/PVP2007-26247.

In order to introduce environmental effects into fatigue evaluation of design and construction codes, the environmental fatigue evaluation method should not only be established, but the conservativeness of the codes, such as safety factors of design fatigue curve and simplified elastic-plastic analysis method (Ke factor), etc. should also be reviewed. Then plant designers should optimize total system of fatigue evaluation according to the objective of the codes by properly selecting design transient conditions and stress analysis methods used in fatigue evaluations as necessary. In addition, investigation of measures for reducing fatigue should be performed to mitigate possible fatigue initiators and alternative evaluation methods in case that the evaluation result should exceed the criteria specified in the design and construction codes. This paper discusses the present status in the review of these items for the Japanese PWR plants and future prospects to tackle on the application of environmental fatigue evaluation in design stage of plant construction.

Topics: Fatigue , Optimization
Commentary by Dr. Valentin Fuster
2007;():201-212. doi:10.1115/PVP2007-26356.

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

Commentary by Dr. Valentin Fuster
2007;():213-218. doi:10.1115/PVP2007-26379.

Fatigue tests were conducted on the notched specimens of Ni-base-alloy NCF600 in high temperature water. Notch root strain was analyzed by finite element method (FEM) to calculate the notch root strain rate and the environmental fatigue correction factor Fen . Relationship between notch root fictitious stress amplitude and corrected fatigue life in water Fen Nwater are compared with the fatigue data of smooth specimen in air (i.e. best-fit curve) and it was found that the corrected fatigue life data in water shows a little shorter but almost agree with best-fit curve in air for the specimen with Kt = 1.4A. For the specimen with Kt = 3, corrected fatigue life in water is longer than that in air and the difference between both lives becomes larger with decreasing stress amplitude. The longer fatigue life in sharp notched specimen than that in dull notched specimen at the same notch root stain amplitude is thought to be dominated by the difference in the crack propagation life since the stress distributions on the cross section are decisively different. It is concluded that the fatigue life of notched specimen in high temperature water is adequately predicted using the modified rate approach method when the notch root strain is appropriately estimated e.g. by FEM analysis even though it gives excessively conservative prediction for sharp notch at low stress amplitude. The aspect of environmental fatigue of notched specimens are summarized on carbon steel and stainless steels with Ni-base alloy studied in Japan Nuclear Energy Safety Organization (JNES) Environmental Fatigue Test (EFT) project.

Topics: Fatigue , Alloys
Commentary by Dr. Valentin Fuster
2007;():219-229. doi:10.1115/PVP2007-26422.

The environmental fatigue life of carbon steel is influenced by BWR conditions such as water temperature, dissolved oxygen concentration, water flow rate and so on. These parameters change inconstantly during operation of BWR plants and strain rate changes in the structural components due to the temperature change. In general, fatigue life evaluation equations have been formulated based on the fatigue data obtained under constant conditions. To apply these equations to evaluate the fatigue life of actual components, a study is necessary to confirm the applicability of the proposed equations to the changing conditions. In this study, fatigue tests were performed under changing conditions of strain amplitude, strain rate, temperature, dissolved oxygen concentration and water flow rate. It was confirmed that the proposed fatigue life equation could predict the fatigue life under changing conditions.

Commentary by Dr. Valentin Fuster
2007;():231-242. doi:10.1115/PVP2007-26423.

Fatigue tests in simulated LWR environment of carbon and stainless steels were performed under high water flow rates between 7 to 10 m/s. For carbon steel, high flow rate of water clearly mitigated the environmental effect on the fatigue life at the high sulfur concentration of 0.016% which caused high environmental effect on a fatigue life. On the contrary, high flow rate of water slightly enhanced the environmental effect at the low sulfur concentration at or less than 0.008% which caused very low environmental effect. These results suggested that the environmental fatigue life under various flow rate conditions should be determined by the combination between the mitigating effect caused by flushing of the severe local environment and the enhancing effect caused by increase in corrosion potential. Low alloy steel showed the similar behavior as carbon steel. For stainless steel, flow rate had little effect on the fatigue life of type 316NG stainless steel. It suggested that there was no role of water flushing. For type 304 and 304L stainless steel, fatigue life has a tendency to decrease with increase in water flow rate. Fatigue lives of type 304 stainless steel under high flow rate of 7 to 10 m/s were shorter than those predicted by proposed fatigue life prediction equation by the Japanese EFT committee. This effect should be considered in an evaluation of environmental fatigue. No water flow effect was found in cast stainless steel.

Commentary by Dr. Valentin Fuster
2007;():243-251. doi:10.1115/PVP2007-26517.

Fatigue crack propagation behaviors in PWR environment can be evaluated by using crack growth rate (CGR) curves which are given in JSME code on Fitness-for-Service for nuclear power plant. The CGR curves, however, are only defined in crack growth region and crack growth thresholds are not considered. Since there is a case that stresses in low ΔK region is applied to the components in case of fatigue, it is needed to investigate near-threshold fatigue CGR to establish fatigue assessment. In this study, CGR tests for stainless steels were carried out, and CGR in the region and ΔKth were obtained in simulated PWR primary water. It was found that CGR was accelerated in high temperature water compared to that in air and ΔKth existed even in water environment. ΔKth was influenced by temperature, stress ratio and frequency, independently of materials. Oxide-induced-crack-closure has an important role in high temperature water. ΔKth was formulated and ΔKth evaluation method, whose accuracy were ±25% between experimental data and evaluation value, was proposed.

Commentary by Dr. Valentin Fuster
2007;():253-268. doi:10.1115/PVP2007-26659.

Crack growth rate (CGR) data have been obtained in boiling water reactor environments on several grades of austenitic stainless steels, including weld heat-affected-zone and cast materials, that were irradiated up to 2.0 × 1021 n/cm2 (E > 1 MeV) (≈3 dpa). Crack growth tests were conducted on 1/4-T compact tension specimens subjected to either a sawtooth waveform with load ratios up to 0.7 and rise times up to 1000 s, or a constant load with or without periodic partial unloading. The results indicate significant enhancement of crack growth rates in the irradiated steels. The results are compared with data obtained from other studies. The existing CGR data are also reviewed to evaluate the effects of material composition, irradiation, and water chemistry on the CGRs in austenitic SSs. The significance of specimen size criteria is discussed.

Commentary by Dr. Valentin Fuster
2007;():269-286. doi:10.1115/PVP2007-26709.

Structural integrity assessment of reactor components requires consideration of crack growth. A key input to this is the development of reference stress corrosion crack (SCC) growth rate curves for use in the structural evaluation. The ASME Section XI Task Group on SCC Reference Curve is looking into available SCC data for stainless steel and nickel based alloys and associated weldment in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. The test data show significant data scatter in crack growth rates (CGR). The conservative approach is to develop reference curves that bound all available data so that upper bound crack growth predictions. While this approach may be conservative, it may lead to excessive estimates of crack growth and result in unrealistic (and often meaningless) structural margin predictions. Selection of the appropriate SCC reference curves requires realistic interpretation of test data so that the predictions are consistent with field behavior and provide reasonable, but conservative assessment. This paper describes crack growth assessment for stainless steel piping and Alloy 600 safe end components with Alloy 182/82 welds in BWR environment. The results from the crack growth analysis for piping can be used to determine whether a proposed reference curve provides reasonable results. The objective is to use the piping and safe end crack growth predictions to develop optimal SCC Reference Curves for use in ASME Code evaluations.

Commentary by Dr. Valentin Fuster
2007;():287-302. doi:10.1115/PVP2007-26755.

The rate of growth of flaws in reactor circuit components by fatigue is usually determined using the reference crack growth curves in Section XI of the ASME Boiler and Pressure Vessel Code. These curves describe the rate of crack propagation per cycle (da/dN) as a function of the applied stress intensity factor range (ΔK). No reference curves for water-wetted defects in austenitic stainless steels are currently available. This paper describes the results of testing of austenitic stainless steel and weld metal in simulated PWR primary coolant over a range of temperatures and mechanical loading conditions. Previous data presented by the authors on wrought stainless steel demonstrated that crack growth rates can be significantly enhanced by the PWR primary environment at temperatures between 150°C and 300°C. The current study extends these data to weld metal and also investigates the impact of other loading waveforms (e.g. trapezoidal loading) on the degree of environmental enhancement. The environmental enhancement increases significantly with reducing loading frequency and decreases with decreasing water temperature. The environmental influence on fatigue is shown to be independent of load ratio over the range R = 0.1 to R = 0.8. The level of enhancement is frequently smaller at very high R ratio (≥0.85) with the enhanced rates of fatigue frequently being unsustained at these high load ratios. There is a strong correlation between the rise time and the level of enhancement of crack growth rate over inert crack growth rates at all temperatures tested. Weld metal has been shown to exhibit similar behavior to wrought material over the whole temperature range studied although the apparent rates of enhancement relative to average inert crack growth rates are lower than found for wrought material. For complex loading waveforms (e.g. trapezoidal loading with hold periods at maximum or minimum load) it is possible predict the level of enhancement on the basis of the test data generated using simpler saw tooth loading regimes.

Commentary by Dr. Valentin Fuster
2007;():305-311. doi:10.1115/PVP2007-26214.

It is still difficult to select the proper constants of crack growth rate corresponding to the loading condition for R-ratio has significant effect on crack growth rate. An unique crack growth rate equation, taking equivalent stress intensity factor range (ESIFR) as the driving force, has been examined for steels and as welded joints. An empirical equation of β for structural steels has been presented. The crack growth rate data expressed by ESIFR instead of SIFR (stress intensity factor range) condensing the crack growth data under different R-ratios to the curve corresponding to R = 0. The most commonly tested crack growth rate constants under R = 0∼0.1 are sufficient in fatigue crack growth calculation under different loading condition. Furthermore, the crack growth rate constants for fatigue life calculation under different R-ratios can be obtained directly from the crack growth rate data tested under any given constant R-ratio test conditions according to the relationship between ESIFR and SIFR. This deduction can greatly reduced the requirement of quantity and cost of fatigue tests for determining the constants of crack growth rate. The gape between the curves corresponding to R<0.5 and R≥0.5 respectively in BS7910 can be filled by present model. For fatigue life prediction of as welded joints, the crack growth rate constants corresponding to R = 0 and R = 0.5 can be used.

Commentary by Dr. Valentin Fuster
2007;():313-331. doi:10.1115/PVP2007-26364.

Thermal fatigue cracking has been observed for thick perforated spacer rings used as part of a thermal fatigue test loop operating at Bechtel Bettis, Inc. The perforated rings are used for instrumentation access to the fluid flow at the test specimen inlet and outlet, and are subject to alternating hot and cold forced flow, low oxygenated water every three minutes so that rapid changes in water temperature impart a thermal shock event to the inner wall of the rings. Thermal and structural three dimensional elastic and elastic-plastic finite element analyses (FEA) were conducted for the ring and the results used to predict fatigue crack initiation using strain-based fatigue-life algorithms. Predicted cycles-to-crack initiation agreed well with the observed cracking when alternating shear strain intensity analogous to the Tresca stress was used. This analysis qualifies the use of FEA for thermal fatigue assessments of complicated three-dimensional components.

Topics: Fatigue testing
Commentary by Dr. Valentin Fuster
2007;():333-341. doi:10.1115/PVP2007-26440.

Ratcheting tests are conducted on stainless steel 304 under uniaxial, torsional, and combined axial-torsional loading. The ratcheting strain is predicted based on the constitutive theory that incorporates a modified Ohno-Wang kinematic hardening rule and Tanaka’s isotropic hardening model. The results show that the main features of the stress-strain response can be simulated with the constitutive model. The experimental and predicted ratcheting strains for nonproportional paths are found in decent correlation. Ratcheting strain depends highly on the loading path and load level, and less on cyclic hardening or softening of the material. The torsional ratcheting strain under mean shear stress with (or without) fully reversed axial strain cycling is found close to the axial ratcheting strain under equivalent mean stress with (or without) torsional strain cycling.

Topics: Stainless steel
Commentary by Dr. Valentin Fuster
2007;():343-350. doi:10.1115/PVP2007-26474.

The concept of material selection map proposed by Ashby [1] was modified to statistical maps for reliability design of high temperature components. The proposed procedure for making maps contained following steps; (i)define a performance index such as cost, (ii)define constraint conditions such as material strength, (iii)mapping the performance index under the constraint conditions and statistical distribution of material properties. The statistical distribution of material properties were expressed as the function of normalized parameters to obtain unified regressions for wide variation of material heats. The material properties used were tensile, creep and low cycle fatigue related ones for several ferritic and austenitic heat resistant steels mostly referred to NIMS (National Institute for Materials Science) database. This proposed mapping procedure was applied to the design of pipes and flanges under creep and thermomechanical low cycle fatigue conditions. The maps could show the statistically upper and lower bounds of allowable geometrical dimensions. The maps showed the effectiveness for optimizing the dimensions from the design aspects of creep and fatigue reliability.

Commentary by Dr. Valentin Fuster
2007;():351-361. doi:10.1115/PVP2007-26550.

Current fatigue analyses of metallic structures undergoing variable amplitude loading, including pressure vessels, are mostly based on linear cumulative damage concepts, as proposed by Palmgren and Miner. This type of analysis neglects any sequential effects of the loading history. Several studies have shown that linear cumulative damage theories can produce inconsistent fatigue life predictions. In this paper, both fatigue damage accumulation and cyclic elastoplastic behaviors of the P355NL1 steel are characterized, using block loading fatigue tests. The loading is composed by blocks of constant strain-controlled amplitudes, applied according to two and multiple alternate blocks sequences. Also, loading composed by blocks of variable strain-controlled amplitudes are investigated. The block loading illustrates that fatigue damage evolves nonlinearly with the number of load cycles, as a function of the block strain amplitudes. These observations suggest a nonlinear damage accumulation rule with load sequential effects for the P355NL1 steel. However, the damage accumulation nonlinearity and load sequential effects are more evident for the two block loading rather than for multiple alternate block sequences, which suggests that the linear Palmgren-Miner’s rule tend to produce better results for more irregular loading histories. Some phenomenological interpretations for the observed trends are discussed under a fracture mechanics framework.

Topics: Fatigue , Steel
Commentary by Dr. Valentin Fuster
2007;():363-377. doi:10.1115/PVP2007-26622.

Experimental results for the fatigue testing of several welded flat head geometries are reported. These tests are similar to those previously reported by Hinnant (2006) [1] and focus on the fatigue behavior of full penetration welds with cover fillet welds. Fatigue calculations according to several fatigue design methods are compared against the experimental results, as are the mean fatigue curves of several of the design methods. Of particular interest for these new tests is the effect of plate thickness, testing environment, and geometric effects. Nominal plate thickness values ranging from 0.0625" (1.59 mm) to 0.1875" (4.76 mm) have been tested and correlated. Four additional fatigue tests were conducted using air to determine if previous testing in room temperature tap water resulted in decreased fatigue life.

Commentary by Dr. Valentin Fuster
2007;():379-384. doi:10.1115/PVP2007-26691.

The paper distinguishes between FSRFs that are used for two different purposes. One is to serve as a guideline for an initial estimate of the fatigue strength of a welded joint. That is the purpose of the FSRFs that are given in the ASME B&PV Code and various accompanying documents. If that estimate renders the fatigue strength inadequate, an FSRF can be sought that is limited to the joint under consideration. The paper shows how such FSRFs can be determined from fatigue test data. In order to make it possible to read the allowable cycles from the same design fatigue curve as that used for the FSRFs of the guidelines, a Langer curve [defined by equation (2) in the paper] is used to curve fit the data. The appropriate FSRF is obtained by minimizing the standard deviation between this curve and the data. The procedure is illustrated for girth butt-welded pipes. The illustration shows that for the data used in the analysis, a constant FSRF is applicable to less than one million cycles but not to the high-cycle regime.

Commentary by Dr. Valentin Fuster
2007;():385-391. doi:10.1115/PVP2007-26705.

This paper studies the cyclic ratcheting for two materials under multiaxial stress state. The two materials are SUS304 austenitic stainless steel and A1070 pure aluminum. The former material is known as a material that gives strong additional hardening and the latter material shows little additional hardening under nonproportional cyclic loading. The ratcheting behavior under 12 stress-strain waveforms was extensively studied using hollow cylinder specimen. Ratcheting strain depended on the material and stress-strain waveform. Anisotropic ratcheting was found in A1070 but isotropic ratcheting was observed in SUS304 steel.

Commentary by Dr. Valentin Fuster
2007;():395-401. doi:10.1115/PVP2007-26564.

The RCC-MR creep-fatigue rules were developed and written in the framework of studies for the first SFR (Sodium Fast Reactors). These reactors were characterized by low primary loads and moderately high temperatures. The rule thus has to be improved with the aim of decreasing its conservatisms in case of higher temperatures and/or higher pressures (for GEN IV Gas Cooled Reactors). Studies were realized to improve the rule on the following points: - the position of the temperature dwell time in the cycle : the current rule always considers that the dwell time is located at one of the extremes of the cycle, what can be very conservative in some cases, - the symmetrisation effect of the stabilized cycle, - the case where the primary loads vary during the cycle, - the primary and secondary stresses combination during the temperature dwell time for the evaluation of the stress relaxation. These works are based on viscoelastoplastic calculations of stabilized cycles and the new proposals are applied on different tests. The consequences on creep-fatigue damage evaluation can be very significant.

Topics: Creep , Fatigue
Commentary by Dr. Valentin Fuster
2007;():403-412. doi:10.1115/PVP2007-26569.

The R5 procedures have been developed within the UK power generation industry to assess the integrity of nuclear and conventional plant operating at high temperatures. Within R5, there are specific procedures for assessing creep-fatigue crack initiation in initially defect-free components (Volume 2/3) and for assessing components containing defects (Volume 4/5). This paper first describes in outline the current R5 Volume 2/3 and Volume 4/5 procedures. Attention is then focused on recent and proposed future developments in these procedures.

Commentary by Dr. Valentin Fuster
2007;():415-419. doi:10.1115/PVP2007-26011.

Fully plastic failure stress for a single circumferential flaw on a pipe is evaluated by the limit load criteria in accordance with Appendix E-8 in the JSME S NA-1-2004 and Appendix C in the ASME Code Section XI. However, multiple flaws such as stress corrosion cracking are frequently detected in the same circumferential cross section in a pipe. If the distance between adjacent flaws is short, the two flaws are combined as a single flaw in compliance with combination rules. If the two flaws separated by a large distance, it is not required to combine two flaws. However, there is no evaluation method for two flawed pipe in the JSME and ASME Codes. Failure stress for pipes with two circumferential flaws based on net-stress approach had been proposed by one of the authors. The present paper is concerned with the comparison of experimental data and the proposed theoretical method for pipes with circumferentially multiple flaws.

Topics: Stress , Pipes , Failure
Commentary by Dr. Valentin Fuster
2007;():421-427. doi:10.1115/PVP2007-26055.

In recent years, Japanese nuclear power plants meet the problems due to SCC (Stress Corrosion Cracking) that may seriously affect the integrity of the important components such as the pressure boundary or the reactor internals. Repair or replacement of the components with SCC has been taken in the past. However, establishment of a reliable evaluation method for the SCC crack propagation is expected, especially for surface cracks, under such complicated stress fields as residual stress. The authors have developed a software system called “SCAN” to evaluate the stress intensity factor, K, and to simulate fatigue crack propagation for surface cracks for arbitrarily distributed surface stress. In this paper we have extended its function to estimate SCC crack propagation based upon the propagation law described by the JSME (Japan Society of Mechanical Engineers) standard. And we call it “SCAN SCC Version”. In order to investigate the reliability of the developed system a simplified simulation model has been analyzed and the results are compared with those obtained by the algorithm of API and JSME standards. It has been found that it works well, that is, we can estimate reliable results easily by the proposed system. It is very useful because it can be applied to the problems of more complicated stress fields. Further, we have developed “SCAN circumferential crack version” which is effective to estimate the residual life of SCC crack propagation, for internal fully circumferentially cracked pipes. In order to take account of the effect of pipe compliance, we have proposed a simplified model to estimate K-values for the displacement controlled problem. An example problem shows how important it is to take account of the effect of pipe compliance when we think of the SCC crack propagation.

Commentary by Dr. Valentin Fuster
2007;():429-434. doi:10.1115/PVP2007-26221.

The influence of interaction between two surface cracks on the limit load (LL) was examined by finitie element analysis. The cicumferential surface cracks were assumed to be on the straight pipe that was subjected to uniform tensile load. Change in the LL due to relative spacing of cracks, crack and pipe geometries were evaluated. The LL was equivalent to coalesced crack when the cracks were on the same plane or their offset and horizonatal distance were the same, although the LL decreased as the offset distance increased in the other cases. The magnitude of the interaction was large compared with that of the stress intensity factor and J-integral value under the same relative spacing. Based on the analysis results, assessment criteria for evaluating the influence of the interaction was discussed.

Commentary by Dr. Valentin Fuster
2007;():435-449. doi:10.1115/PVP2007-26688.

Boiling water reactor (BWR) components made of austenitic stainless steel are exposed to the high temperature water environment. Under the combination of susceptible material condition (sensitization due to welding or cold work), weld residual stresses and the oxygenated high temperature water environment, reactor internals such as the core shroud and internal core spray piping experience stress corrosion cracking. The cracking is typically in the weld heat affected zone on both sides of the weld and is parallel to the weld direction. The determination of the structural capability of the component with parallel offset cracks is the subject of this paper. The Section XI, ASME code position on parallel planar flaws states that there is no interaction (i.e. no change in load capability due to multiple parallel cracks) for cracks that are separated by 12.5 mm (1/2 inch) or more. While this is reasonable under linear elastic fracture mechanics conditions, it is not conservative for ductile failure under limit load conditions. Alternatively, assuming that the parallel cracks are in a single plane for the purpose of determining the load capability is over-conservative and underestimates the structural capability. This paper is an extension of earlier work on the interaction of parallel flaws and considers the combines load capability as a function the crack separation. The interaction rules developed here were based on analysis and validated by comparison with extensive test data from different sources. For cracks in two parallel planes, the proposed rules for ductile materials (where limit load is the governing failure mechanism) allow them to be considered as separate cracks if the distance between the two planes is greater than 3t (where t is the shell thickness) and the limit load for cracking in each plane is determined separately. For planes separated by less than t, they are combined and assumed to be in one plane and the limit load is calculated for the combined crack. Linear interpolation is used when the separation distance, d is such that t≤d ≤3t.

Commentary by Dr. Valentin Fuster
2007;():453-456. doi:10.1115/PVP2007-26188.

Radiographic Test (RT) has been widely used in industries to detect inner defects of welded structures or any other significant components. Especially in the nuclear industry, film radiography is the dominant and standardized procedure in performing radiographic testing. Lately emphasis is on digital radiography. One of the most serious concerns for digitization is the lack of image resolution standardizing device like resolution gauge which would determine imaging parameters such as modular transfer function (MTF). This paper proposes line pair type image quality indicator (IQI) corresponding to the current IQIs for both hole and wire type. The advantage of this IQI is to enable easier calibration of testing conditions and quantification of digital RT image quality with required MTF that should be clearly defined in the examination procedures. Furthermore, to acquire “resolution-ensured” digital image of existing RT films, we developed line pair type standardization film. Prototypes of line pair type IQI and line pair type standardization film are currently in the validation study and trial implementing process. These results are also reported in this paper.

Commentary by Dr. Valentin Fuster
2007;():457-471. doi:10.1115/PVP2007-26375.

Many utilities select critical welds in their main steam (MS) and hot reheat (HRH) piping systems by considering some combination of design-based stresses, terminal point locations, and fitting weldments. The conventional methodology results in frequent inspections of many low risk areas and the neglect of some high risk areas. This paper discusses the use of a risk-based inspection (RBI) strategy to select the most critical inspection locations, determine appropriate reexamination intervals, and recommend the most important corrective actions for the piping systems. The high energy piping life consumption (HEPLC) strategy applies cost effective RBI principles to enhance inspection programs for MS and HRH piping systems. Using a top-down methodology, this strategy is customized to each piping system, considering applicable effects, such as expected damage mechanisms, previous inspection history, operating history, measured weldment wall thicknesses, observed support anomalies, and actual piping thermal displacements. This information can be used to provide more realistic estimates of actual time-dependent multiaxial stresses. Finally, the life consumption estimates are based on realistic weldment performance factors. Risk is defined as the product of probability and consequence. The HEPLC strategy considers a more quantitative probability assessment methodology as compared to most RBI approaches. Piping stress and life consumption evaluations, considering existing field conditions and inspection results, are enhanced to reduce the uncertainty in the quantitative probability of failure value for each particular location and to determine a more accurate estimate for future inspection intervals. Based on the results of many HEPLC projects, the author has determined that most of the risk (regarding failure of the pressure boundary) in MS and HRH piping systems is associated with a few high priority areas that should be examined at appropriate intervals. The author has performed many studies using RBI principles for MS and HRH piping systems over the past 15 years. This life management strategy for MS and HRH critical welds is a rational approach to determine critical weldment locations for examinations and to determine appropriate reexamination intervals as a risk-based evaluation technique. Both consequence of failure (COF) and likelihood of failure (LOF) are considered in this methodology. This paper also provides a few examples of the application of this methodology to MS and HRH piping systems.

Commentary by Dr. Valentin Fuster
2007;():473-480. doi:10.1115/PVP2007-26661.

The development and implementation of risk-informed inservice inspection (ISI) approaches, providing an alternative to ASME Section XI Code requirements for the selection of examination locations in nuclear power plant piping systems, has been recently identified by industry leaders as the most successful voluntary application of risk technology in the United States. This technology improves the effectiveness of examination of piping components, i.e. concentrates inspection resources and enhances inspection strategies on high-safety-significant locations, and reduces inspection requirements on others while maintaining or enhancing overall plant safety (in terms of core damage and large, early release frequency). Risk-informed ISI has been successfully implemented in more than 90% of U.S. reactors and in nuclear power plant ISI programs in at least eight other countries. Beginning almost 20 years ago, Dr. Spencer H. Bush played an instrumental role in the development of this technology as a Steering Committee member of the ASME Research Project on Risk-Based Inspection Guidelines. He later became a member of the ASME Section XI Working Group on Implementation of Risk-Based Examination participating in the development and review of ASME Code Cases allowing for trial use of this new technology. Dr. Bush, having a long leadership role with ASME Section XI, played an instrumental role in the development of an overall structure and process for integrating the technologies inherent to a risk-informed ISI program, including piping failure data and non-destructive examination reliability results. He also played a key role in garnering ASME and regulatory acceptance of this alternative approach. The authors, along with many other colleagues, had the honor and privilege of working closely with Dr. Bush on this initiative over the last two decades, and via this paper, the authors would like to highlight some of Dr. Bush’s key contributions to this successful development in his memory.

Commentary by Dr. Valentin Fuster
2007;():481-484. doi:10.1115/PVP2007-26770.

Long-range guided-wave technology is a recently developed inspection method for surveying large areas of structures and is widely used as a screening tool for corrosion defects in piping. Because of its ability to examine a structure over a long distance, this technology with permanently installed probes is ideal for on-line long-term structural health monitoring of pressure vessels or piping for improved safety, operation, and maintenance. The magnetostrictive sensor (MsS) technology is a guided-wave technology developed at authors’ institution. It uses a probe that consists of thin ferromagnetic strip and coil for guided-wave generation and detection. The MsS probe is inexpensive and can be used on components up to 300°C. Applications of the MsS for long-term monitoring of pressure vessels and piping are presented, including the configuration of MsS probes, types of guided-wave mode used, defect types detectable with guided waves, and example data.

Topics: Sensors , Waves , Pipes , Vessels
Commentary by Dr. Valentin Fuster
2007;():485-489. doi:10.1115/PVP2007-26776.

Piping and pressure vessels are on occasion subjected to overload that may affect their ability to perform the role for which they were designed. A combination of velocity change and frequency analysis for near surface longitudinal (LCR ) waves is used to evaluate load damage past the yield point in 4140 steel. Prior work has shown that loading past the yield point affects the velocity in the bar. Further, frequency analysis has been shown to predict the fracture toughness in AISI 4137 steel used in offshore drilling. The present project joins these two approaches and uses velocity measurement and frequency analysis at different locations in the steel bar. Some locations were subjected to post-yield loading and others were not. Discussion on applying the technique to detecting hydrogen attack is included.

Topics: Steel , Stress
Commentary by Dr. Valentin Fuster
2007;():491-497. doi:10.1115/PVP2007-26788.

In this paper, lock-in thermography techniques for quantitative nondestructive evaluations developed by the present authors are reviewed. Self-reference lock-in thermography was developed for remote nondestructive testing of fatigue cracks. This technique is based on the measurement of thermoelastic temperature change due to stress change. Cracks can be identified from significant temperature change observed at crack tips due to the stress singularity. For accurate measurement of the thermoelastic temperature change under random loading, a self-reference lock-in data processing technique was developed, in which a reference signal was constructed by using the temperature data simultaneously taken at a remote area. Thermoelastic temperature change in a region of interest was correlated with that at the area for reference signal construction. It enabled us to measure the relative stress distribution under random loading without using any external loading signal. The self-reference lock-in thermography was applied for fatigue crack identification in welded steel plate specimens and actual steel structures. It was found that significant temperature change was observed at the crack tip in the self-reference lock-in thermal image, demonstrating the feasibility of the proposed technique. Lock-in thermography technique was also applied to quantitative nondestructive evaluation of material loss defects. Transient temperature data under pulse or step heating were measured by infrared thermography. Temperature data were processed by the lock-in analysis scheme based on the Fourier series expansion, in which Fourier coefficients synchronizing with sine and cosine waves were correlated with defect parameters. Experimental investigations were conducted using steel samples with artificial material loss defects. It was found that the defect parameters can be quantitatively determined from the Fourier coefficients, demonstrating the feasibility of the proposed technique.

Commentary by Dr. Valentin Fuster
2007;():499-502. doi:10.1115/PVP2007-26802.

The ASME construction code books specify materials and fabrication procedures that are acceptable for pressure technology applications. However, with few exceptions, the materials properties provided in the ASME code books provide no statistics or other information pertaining to material variability. Such information is central to the prediction and prevention of failure events. Many sources of materials data exist that provide variability information but such sources do not necessarily represent a consensus of experts with respect to the reported trends that are represented. Such a need has been identified by the ASME Standards Technology, LLC and initial steps have been taken to address these needs: however, these steps are limited to project-specific applications only, such as the joint DOE-ASME project on materials for Generation IV nuclear reactors. In contrast to light-water reactor technology, the experience base for the Generation IV nuclear reactors is somewhat lacking and heavy reliance must be placed on model development and predictive capability. The database for model development is being assembled and includes existing code alloys such as alloy 800H and 9Cr-1Mo-V steel. Ownership and use rights are potential barriers that must be addressed.

Topics: Pressure , Databases
Commentary by Dr. Valentin Fuster
2007;():503-505. doi:10.1115/PVP2007-26842.

Spencer Bush was an energetic and creative synthesizer of technology into effective practices and thereby made unique and lasting contributions for more effective Codes and Standards. Dr. Bush’s participation in ASME, PVRC and international activities are discussed.

Commentary by Dr. Valentin Fuster
2007;():509-515. doi:10.1115/PVP2007-26023.

As a part of an ongoing activity to develop ASME Code rules for the hydrogen infrastructure, the ASME Boiler and Pressure Vessel Code Committee approved new fracture control rules for Section VIII, Division 3 vessels in 2006. These rules have been incorporated into new Article KD-10 in Division 3. The new rules require determining fatigue crack growth rate and fracture resistance properties of materials in high pressure hydrogen gas. Test methods have been specified to measure these fracture properties, which are required to be used in establishing the vessel fatigue life. An example has been given to demonstrate the application of these new rules.

Commentary by Dr. Valentin Fuster
2007;():517-525. doi:10.1115/PVP2007-26756.

The American Standards Association (ASA) B31.3-1959 Petroleum Refinery Piping Code [1] grew out of an ASA document that addressed all manner of fluid conveying piping systems. ASA B31.3 was created long before widespread engineering use of computer “mainframes” or even before the inception of piping stress analysis software. From its inception until recent times, the B31.3 Process Piping Code [2] (hereafter referred to as the “Code”) has remained ambiguous in several areas. This paper describes some of these subtle concepts that are included in the Code 2006 Edition for Appendix S Example S3. This paper discusses: • the effect of moment reversal in determining the largest Displacement Stress Range, • the impact of the average axial stress caused by displacement strains on the Example S3 piping system and the augmenting of the Code Eq. (17) thereto, • a brief comparison of Example S3 results to that of the operating stress range evaluated in accordance with the 2006 Code Appendix P Alternative Requirements.

Topics: Pipes , Computers
Commentary by Dr. Valentin Fuster
2007;():527-531. doi:10.1115/PVP2007-26759.

Code Case N-659 Revision 0 was approved in 2002 to allow ultrasonic examination (UT) an alternative to radiography (RT) for nuclear power plant components and transport containers under Section III of the ASME Code. The Nuclear Regulatory Commission has not approved N-659 and its subsequent revisions (currently N-659-2) for general use, but they have been used on a case-by-case basis mainly where logistic problems or component configuration have prevented the use of radiography. Like the parallel Code Case 2235 for non-nuclear applications under Section I and Section VIII, Code Case N-659 requires automated, computerized ultrasonic systems and capability demonstration on a flawed sample as a prerequisite for using UT in lieu of RT. Automated ultrasonic examination can be significantly more expensive than radiography, so a cost-benefit evaluation is a key factor in the decision to use the Code Case. In addition, the flaw sample set has recently become an issue and a topic of negotiation with the NRC for application of the Case. A flaw sample set for a recent radioactive material transport cask fabrication project was successfully negotiated with the NRC. The Code Case N-659 approach has been used effectively to overcome barriers to Code required radiography. Examples are examination of welds in an assembled heat exchanger and in a radioactive material transport cask assembly where internal shielding prevented radiography of the weld. Future development of Code Case N-659 will address sample set considerations and application-specific Code Cases, such as for storage and transport containers, will be developed where NRC concerns have been fully addressed and regulatory approval can be obtained on a generic basis.

Commentary by Dr. Valentin Fuster
2007;():535-540. doi:10.1115/PVP2007-26126.

Since 1989 ISPESL has emanated dispositions for pressure equipment designed to operate in the creep range according to time dependent mechanical properties. During the years, on the basis of the results of in-field examinations, a series of new provisions has been improved and sharpened the original procedure. During the last two years ISPESL has gone to revise and update the emanated regulations. Now a new entire legislation covers all the subjects and it is set as an advanced means of investigation for equipment working in the creep range: ISPESL technical procedure N. 48/2003. The elaborating of the legislation in matter has benefited from the previous experience, the indications of users and the European and international standards. The procedure is widely applied in Italian plants.

Topics: Pressure , Creep
Commentary by Dr. Valentin Fuster
2007;():541-546. doi:10.1115/PVP2007-26421.

This paper describes the Creep Amendments which will be implemented in EN 13445, the European Standard for Unfired Pressure Vessels, in 2007. It will address four topics: 1) Specifications for Materials and Weldments. 2) Specifications for Design By Formula and Damage Accumulation rules. 3) Specifications for Design By Analysis — Direct Route (Creep Rupture, Excessive Creep Strain, Creep and cyclic Fatigue Interaction). 4) Specifications for Inspection and Non-Destructive Testing (NDT).

Topics: Creep
Commentary by Dr. Valentin Fuster
2007;():549-556. doi:10.1115/PVP2007-26059.

This paper describes effects of a probability of detection (POD) of a stress corrosion crack on the reliability of the piping in a nuclear power plant (NPP). Various POD curves were proposed using the results of Japanese study on the detection for the stress corrosion crack which is frequently observed in austenitic stainless steel piping of boiling water reactor (BWR). Based on the proposed POD curves, the reliability of a flawed pipe was analyzed using probabilistic fracture mechanics (PFM) code. The results suggest that the detectable crack depth and oversight probability (e.g. human error) is important on the reliability of piping with the stress corrosion crack. The reliability of piping depends on the detectable crack depth rather than the oversight probability when the detectable crack depth is larger than 3mm. Meanwhile, it depends on the oversight probability, when the detectable crack depth is 3mm or less.

Topics: Reliability , Pipes
Commentary by Dr. Valentin Fuster
2007;():557-563. doi:10.1115/PVP2007-26169.

Allowable flaws for Class 1 vessels and piping are defined in the Acceptance Standards of JSME (The Japan Society of Mechanical Engineers) Code on Fitness-for-Service (FFS). When detected flaws are subsurface flaws located near component surface, the subsurface flaws are transformed into surface flaws. This is a proximity rule of subsurface flaws. This paper describes the inconsistencies of the subsurface flaws in the Acceptance Standards of vessels and piping. Authors have developed method to improve the inconsistencies and compared it with the proximity rules of the ASME Code Section XI.

Commentary by Dr. Valentin Fuster
2007;():565-568. doi:10.1115/PVP2007-26187.

In 2002, Japanese Industrial standard JIS Z 2340 [1] has issued to provide the instruction of new methodology to calibrate the observing conditions of surface testing such as liquid penetrant and magnetic particle testing applying visual calibration gauges. The visual calibration gauges are the transparent plates, on which the line pairs printed to be used to confirm the resolution of visual testing observing view. The concept of line pair has been used to evaluate the resolutions of optical instruments, and the line pair value is a numerical value that shows how many line pairs can be distinguished as separate lines within 1 millimeter. In this paper, the background of the development and the general outlines of visual calibration gauges are introduced at first. And then applications of the gauges and detail process to calibrate the surface testing observing view conditions are also described.

Topics: Gages , Testing , Calibration
Commentary by Dr. Valentin Fuster
2007;():569-574. doi:10.1115/PVP2007-26256.

In response to the pipe wall thinning damage experienced in power plants in 2004, the Japan Society of Mechanical Engineers (JSME) has started activities to develop technical standards on the pipe wall thinning management. The first edition of the JSME rules on pipe wall thinning management for thermal power generation facilities (JSME S TB1-2006 [1]) was issued in March 2006, and its latest edition will be issued in 2007, which describes the technical requirements to meet the JSME performance-based rules for pipe wall thinning management (JSME S CA-1 2005 [2]). Based on 24,774 inspection data obtained at the thermal power plants in Japan, the latest JSME rules will show the specific attention to the need for inspection of piping systems that are susceptible to the wall thinning damage. The JSME rules describe the selection of thickness measurement locations such as downstream of piping configurations that produce turbulence, downstream of orifices, downstream of control valves, and they describe the periodic inspections including the first inspection to be scheduled taking the wall thinning rate data at the equivalent locations into consideration. The JSME rules stipulate some available inspection methods such as ultrasonic scanning, radiographic profile, eddy current and potential drop technique. This paper presents outline of the JSME rules including basic philosophy, technical requirements on the inspection and testing practices and the relation with the regulations in Japan.

Commentary by Dr. Valentin Fuster
2007;():575-587. doi:10.1115/PVP2007-26410.

The newly-developed p-M diagram provides a means for readily evaluating the collapse load of pressure equipment with external flaws simultaneously subjected to internal pressure, p and external bending moment, M due to earthquake, etc. In this paper, numerous experiments and FEAs for a cylinder with an external flaw were conducted under (1) pure internal pressure, (2) pure external bending moment, and (3) subjected simultaneously to both internal pressure and external bending moment, in order to determine the plastic initiation load and plastic collapse load by applying the twice-elastic slope (TES) as recommended by ASME. It has been clarified that the collapse (TES) loads are much the same as those calculated under the proposed p-M line based on the measured yield stress. The p-M line adopted in the Ibaraki FFS rule based on the specified minimum yield stress with a safety factor of 1.5 indicates that the safety margin for the plastic initiation loads at LTA is about 1.0–3.0, about 1.5–4.0 for the TES loads at LTA and 2.5–6.5 for the plastic instability (break) loads.

Commentary by Dr. Valentin Fuster
2007;():589-597. doi:10.1115/PVP2007-26437.

A simplified assessment procedure using the p-M diagram, which can evaluate the plastic collapse load for pressure equipment such as vessels, piping and storage tanks with an internal surface flaw simultaneously subjected to internal pressure, p, and external bending moment, M, due to earthquake, etc., is derived by taking into account the influence of internal pressure acting on the flaw surface. For an internal surface flaw subjected to pressure, the already-proposed p-M diagram for an external flaw can be applied if the parameters for an internal surface flaw proposed in this paper are used. And the plastic collapse loads derived from the p-M diagram method are being verified by comparison with experimental results. It has been clarified that the parameters for internal surface flaws are also the same as those for external surface flaws where the ratio of thickness to outer radius of a vessel is significantly smaller than unity and internal pressure is small.

Topics: Collapse , Vessels
Commentary by Dr. Valentin Fuster
2007;():601-608. doi:10.1115/PVP2007-26048.

The objective of this paper is to develop the vessel design pre-qualification for a typical annular tank. The pressures and pressure combinations being considered are first combined maximum working internal pressure plus static pressure second combined maximum accidental internal pressure plus static pressure and third combined internal working pressure plus static pressure plus seismic. The FEA modeling method is used to determine if the Annular Tank’s deflection remain within the criticality deflection tolerance and ASME Code stresses are not exceeded. If the tank’s deflection and/or stresses are exceeded, tank original design is modified and new analysis is performed. This process is repeated until the Annular Tank’s design is within the acceptable range of ASME Code stresses and deflection.

Commentary by Dr. Valentin Fuster
2007;():609-617. doi:10.1115/PVP2007-26071.

This paper presents a critical review of the newly published ASME BTH-1-2005, which is intended to be a companion to ASME B30.20, Below the Hook Lifting Devices, a safety standard. The very limited structural design criteria contained in the latter standard was previously addressed in the literature by the current author and was compared against the various national and international regulations, codes, and standards in regard to the presumed factors of safety inherent in the designs of vessel lifting lugs. Based upon the criteria previously outlined and addressed, the current American National Standard ASME BTH-1-2005 is critically reviewed and the commentary that is now incorporated in such is analyzed in an effort to determine the adequacy of the updates in meeting and exceeding the current regulations in both the United States and Canada. The statutory and provincial regulations in both the United States and the province of Alberta, Canada are also reviewed and discussed with respect to the too often utilized phrase “factor of safety” (FOS). The implied implications derived from the chosen FOS are also outlined. Exemplar lugs on vessels are defined and the finite element analyses and closed form Hertzian contact problem solutions are presented and interpreted in accordance with the new ASME BTH-1-2005 structural design criteria. These results are again highlighted against the very limited design information contained within ASME B30.20. A review of the author’s prior recommendations made to revise the ASME B30.20 Below the Hook Lifting Devices safety standard are presented and discussed in light of the examples and technical justification presented in the following paragraphs. Contact stresses that are well known to exist between a lifting pin and clevis type geometry are also discussed in light of the new structural design criteria contained within ASME BTH-1-2005. Additional recommendations are provided for the design and analysis of vessel lifting lugs in consideration of current regulations.

Topics: Design , Vessels
Commentary by Dr. Valentin Fuster
2007;():619-625. doi:10.1115/PVP2007-26209.

For Leak-Before-Break (LBB) analysis of nuclear piping, a circumferential through-wall crack (TWC) with the crack front parallel to the cylinder radius is typically postulated, i.e., an idealized TWC. Such assumption simplifies the LBB analysis significantly. However, in reality, an internal surface crack grows through the wall thickness and penetrates through the wall thickness at the deepest point. Hence, a TWC with different crack lengths at inner and outer surfaces is formed. Such a TWC is referred to as a “slanted TWC” in the present study. Leak rates as well as SCC and fatigue crack growth rates of slanted TWC are expected to be quite different from those of postulated idealized TWC. In this context, characterization of the actual TWC shape during crack growth due to fatigue or stress corrosion cracking is essential for accurate LBB analysis. Based on detailed 3-dimensional (3-D) elastic finite element (FE) analyses, the present paper provides stress intensity factors (SIFs) for plates and cylinders with slanted TWCs. As for loading conditions, axial tension was considered for the plates, whereas axial tension and global bending were considered for the cylinders. In order to cover the practical range of crack sizes, the geometric variables affecting the SIF were systematically varied. Based on FE analysis results, SIFs along the crack front, including the inner and outer surface points, were provided. The SIFs of slanted TWC can be used to evaluate the fatigue crack growth of a TWC and to perform detailed LBB analysis considering a more realistic crack shape.

Commentary by Dr. Valentin Fuster
2007;():627-633. doi:10.1115/PVP2007-26234.

This paper presents the third of a series of solutions to the buckling of imperfect cylindrical shells subjected to an axial compressive load. In particular, the initial problem reviewed is the case of a homogeneous cylindrical shell of variable thickness that is of an axisymmetric nature. The equilibrium equations as first introduced by Donnell over seventy years ago are discussed and reviewed in establishing a basis for embarking upon a solution that utilizes finite difference methods to solve the resulting equilibrium and compatibility equations. The ultimate objective of these calculations is to achieve a quantitative assessment of the critical buckling load considering the small axisymmetric deviations from the nominal cylindrical shell wall thickness. Clearly in practice, large diameter, thin wall shells of revolution that form stacks are never fabricated with constant diameters and thicknesses over the entire length of the assembly. The method and results described herein are in stark contrast to the “knockdown factor” approach currently utilized in ASME Code Case 2286-1. The results obtained by finite difference method agree well with those published by Elishakoff and Williams for the prediction of buckling load.

Topics: Pipes , Buckling , Thickness
Commentary by Dr. Valentin Fuster
2007;():635-642. doi:10.1115/PVP2007-26433.

The minimum hydrostatic test pressure for class 2 and 3 components has been reduced from 1.5 to 1.25 times the Design Pressure in ASME B&PV Sec. III, Division 1 since 1999 addenda. If these requirements are applied to the system hydrostatic test as they are, the minimum hydrostatic test pressure of components and system becomes identical. Therefore it may happen that the test pressure imposed on components installed at low locations in the system exceeds the maximum permissible pressure due to the static head during the system hydrostatic test. PWHT temperature requirements for P-No.4 materials in various Construction Codes, such as ASME B31.1, Sec. I and Sec. VIII, except Sec. III have been unified and the minimum PWHT temperature became 649°C (120°F) since 2004 edition. When considering the mechanical properties of the weld, the minimum PWHT temperature of 593°C or that of P No. 4 materials in Sec. III is too low to reduce hardness and to increase toughness. When PWHT is performed on dissimilar material joints (e.g., between P-No.1 or P-No. 3 and P-No. 4) at 649°C (1200°F) in accordance with Sec. VIII etc., it is possible that the strength of the lower P-No. materials is decreased below the design strength because the PWHT temperature will exceed the tempering temperature of the lower P-No. materials. In this study, the cases of system hydrostatic test in UC-3, 4 units and Steam Generator Nozzle to feedwater pipe joint in Korea Standard Nuclear Power Plants (e.g., UC-3, 4 units, YK-3, 4 units and YK-5, 6 units) were reviewed and analyzed. And then problems of two cases were presented. It is suggested that the minimum system hydrostatic test pressure in Sec. III NC, ND should be decreased by the reduction rate of test pressure for components and the minimum PWHT temperature for P No. 4 materials in Sec. III should be 630°C (1166°F).

Commentary by Dr. Valentin Fuster
2007;():645-651. doi:10.1115/PVP2007-26516.

Intergranular stress corrosion cracking (IGSCC) has been still one of concerns at the weld zone in boiling water reactors (BWRs). Therefore, Weld Overlay (WOL) process has been developed and applied to repair of BWR pipe joints. To understand residual stress and crack growth behavior is important to evaluate the reliability of pipe joints with WOL. In this paper, the residual stresses were calculated by using thermal elasto-plastic analysis by finite element method (FEM). The analytical model was assumed the primary loop recirculation (PLR) pipe joint with WOL followed by the Japanese guideline. The tensile hoop residual stress was changed to compressive stress on the inner surface of the pipe. On the other hand, tensile axial residual stress was occurred on the inner surface of the pipe by butt-welding and some WOL cases increased the tensile axial stress. In addition, the stress intensity factor for fully circumferential cracks was evaluated using calculated residual stress distributions. As a result, the effect of application of WOL on the crack growth behavior is insignificant in PLR pipe joint.

Commentary by Dr. Valentin Fuster
2007;():653-659. doi:10.1115/PVP2007-26636.

Weld overlays have been used to provide repair and mitigation to stress corrosion cracking (SCC) susceptible butt welds in nuclear power plant piping. Among the several advantages associated with weld overlays are the beneficial compressive residual stresses that are developed in the inner portion of the component after application of the overlay. These compressive stresses can provide significant mitigation against SCC in these welds. To determine the residual stresses resulting from the weld overlay process in analytical modeling, a weld repair during original fabrication of the butt weld is typically assumed before application of the weld overlay. If the fabrication records are available, the details of the weld repair can be simulated in the analysis. However, in most cases, the weld records are not easily accessible and in instances where they are available, the quality and completeness of the information are questionable. As such, various conservative assumptions are made on the extent of the weld repair to be simulated in the analytical modeling. In this paper, the residual stress results of an axisymmetric finite element simulation of a bimetallic weld subjected to an inside surface weld repair followed by a weld overlay repair are presented. Three through-wall weld repair sizes (25%, 50% and 75% of the wall thickness without the overlay) assumed to be full 360° around the circumference were considered in the study. The results indicate that for all three weld repair cases, the inside of the configuration is very tensile after the weld repair indicating that regardless of the size of the weld repair, SCC is a possibility. The post weld repair stress distribution of the 50% and the 75% repair cases are similar indicating that an assumed 50% repair is fairly representative of repairs that can be assumed for analysis purposes. The application of the overlay resulted in favorable compressive stresses on the inside portion of the configuration for all the three weld repair cases indicating that regardless of the size of the initial weld repair, the application of the weld overlay provides mitigation against SCC.

Commentary by Dr. Valentin Fuster

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