ASME Conference Presenter Attendance Policy and Archival Proceedings

2015;():V01BT00A001. doi:10.1115/PVP2015-NS1B.

This online compilation of papers from the ASME 2015 Pressure Vessels and Piping Conference (PVP2015) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Codes and Standards: High Temperature Codes and Standards

2015;():V01BT01A001. doi:10.1115/PVP2015-45270.

The 2015 edition of the RCC-MRx Code will be issued, by the end of the year, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), and to components operating at high temperature and to the Vacuum Vessel of ITER (coming from the RCC-MR 2007).

A significant work has been performed since 2012, first edition of the RCC-MRx, in order to improve the rules, to facilitate its use and understandability, and also to have a better fit with the feedbacks of the main projects (ASTRID, JHR, ITER).

This paper presents the technical evolutions in the 2015 edition and also the other initiatives led upstream (prenormative works) and downstream (publication of the rules background - criteria) of the publication of rules.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A002. doi:10.1115/PVP2015-45381.

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress.

Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes.

The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant.

The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints.

The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A003. doi:10.1115/PVP2015-45553.

As the important lessons learned from Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure were strongly recognized against severe accidents (SA) and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premised under design extension conditions (DEC) such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, the best estimation for these failure modes is required for preparing countermeasures and management. Therefore, this study focused on identification failure modes under design extension conditions. To realize best estimation, it is prerequisite to clarify failure modes with ultimate structural strength under extreme loadings such as very high temperature, pressure and great earthquakes.

The authors attempt to clarify unclear failure mechanisms by extreme loadings under DEC using numerical simulations. In this paper, relations between failure modes and extreme loadings were investigated by the numerical simulation using the cylindrical model which is a typical structure of nuclear reactor structures (for example, Formed Head, Nozzle, Instrument Tube, Guide Tube, Support Skirt, etc.). Moreover, it was shown that failure modes change with an effect of structural discontinuities. Local failure dominates than ductile fracture at locally constraint portions where stress triaxiality becomes high.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A004. doi:10.1115/PVP2015-45988.

Lower headers of bottom-supported heat-recovery steam generators (HRSG) may be critical because of their longitudinal dimensions, thermal expansions and external loading (the harp’s weight): Present work considers the creep analysis of the high-temperature-section (superheater /reheater) headers: they may be critical because of the long continued service (175000 hours or twenty years), larger dimensions and the external loads, including a negligible steam-drum weight fraction. The aim of the work is to compare life results from the Italian creep code with those predicted by the American standard API 579-1. This work also checks the compatibility of results coming from the two polynomial models in both Italian and API 579-1 procedures. Classical methods, applied using both ASME and Italian pressure formulae, show that, as for the evaporator-section header, the pressure contribution to longitudinal stress may be greater than bending alone; considering now the increased header’s weight, the stress ratio is also comparable to the evaporator’s. Consistency of results from numerical-model stress analysis (elastic) is good, confirming the pressure contribution is greatest. For the Level-1 assessment (B31.1 stresses), the Italian procedure and the API 579-1 return consistent creep life results, though the API 579-1 results appear more conservative than the Italian-procedure’s. Level-1 assessment, acted through an elastic finite element analysis (FEA), uses Larson-Miller parameter (LMP)-approach method with minimum stress-to-rupture data: the Italian procedure and API 579-1 return consistent creep life results when evaluated on the tubehole branch side, Italian-procedure’s appearing little more conservative than the API 579-1’s. For the Level-2 assessment (FEA stresses), again the Italian procedure and the API 579-1 return consistent creep life results with the Italian-procedure ones again a little more conservative than the API 579-1’s for both sides of the intersection. Level-3 assessment (incorporating creep, plasticity and relaxation) shows (short) creep lives similar to Italian-procedure’s.

Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity Issues for Buried Pipe

2015;():V01BT01A005. doi:10.1115/PVP2015-45276.

The consideration of a geometrical nonlinearity is a common practice for the thin-walled structures. The relevance here are two well-known cases treated in ASME codes. First one is accounting for reduction of the pipe bends flexibility due to the inner pressure. The second one is the retarded increasing (and subsequent saturation) of additional local bending stress with increasing of inner pressure in a pipe with initial cross section form distortion. In both cases the rerounding effect and decreasing of local flexibilities take place. The crack can be treated as the concentrated flexibility and it is quite natural to expect that the stress intensity factor should grow nonlinearly with applied load.

Two cases of SIF calculation for 1-D long axial surface crack in a pipe loaded by inner pressure are considered here: a) cross section has an ideal circular form: b) the form has a small distortion and crack is located in the place of maximal additional bending stresses.

The theoretical analysis is based on: a) the well known crack compliance method [1] and b) analytical linearized solution obtained for deformation of the curved beam in case of action of fixed circumferential stress due to pressure written in the form convenient for transfer matrix method application.

It was shown that for moderately deep crack (crack depth to the wall thickness ratio is 0.5 and bigger) and typical dimensions of pipes used for oil and gas transportation (radius to thickness ratio is 25–40) and loading which can reach up to 200 to 300 MPa, the effect investigated can be quite noticeable and can lead to 5–15 percent reduction of calculated SIF as compared with linear calculation.

The analytical results are supported by nonlinear FEM calculation.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A006. doi:10.1115/PVP2015-45646.

Innovative Method for Installing Pipe Depth Smart Test Stations at Exelon Nuclear Generation Stations to Quantify Corrosion Rates, Evaluate Cathodic Protection Effectiveness, Quantify Corrosion Potentials and Perform Subsurface Soils Analysis to Support Aging Management Plans and Reasonable Assurance Commitments on In-Service Buried and Underground Pipe and Tanks

For several decades ANSI and NACE direct assessment (DA) standards for monitoring cathodic protection and coating quality (SP0502) have been relied upon by pipeline operators and regulators to accurately monitor the impact of corrosion on transmission pipelines. These same standards have been proven far less useful when deployed in nuclear power plant subsurface environments.

In order to meet regulatory and industry commitments the challenge has been to develop a cost effective method to accurately monitor nuclear plants’ buried pipe. The in-scope pipe is comprised of a dense complex of multiple systems in close proximity at depths sometimes exceeding 30-ft. Most nuclear plants have been in service for over 30 years and rely on original buried pipe systems to perform as designed. These systems are manufactured using various metals, which are bonded together to reduce electrical step and touch potential hazards. Each of these factors reduce the confidence factor of DA, but combined they make the procedure even less effective.

Smart Monitoring Test Stations can be installed without excavation at depths of up to forty feet directly into the subject pipeline local environment. This provides access to critical condition and assessment information that was previously unavailable. Smart test stations have been designed with the following components: two Electrical Resistance Probes (ER), two CP coupons and a stationary reference electrode.

This paper will describe the planning and obstacles encountered during the installation, operation and interpretation of data from recent experiences related to the installation of test stations at Exelon Nuclear Generating Stations. We will describe the process from planning to successful completion including:

• Smart Monitoring Test Station features and function

• Value of precise location insertion of probes

• Digitizing plant buried and underground pipe and tanks

• Smart monitoring test station site location selection

• Push probe method to capture soil samples

• Analyzing and interpreting soil samples

• Evaluating potential impacts from groundwater

• Push probe method of installing smart test stations at or near pipe depth

• Monitoring data from test stations

• Application of information to support AMP and NEI commitments

Commentary by Dr. Valentin Fuster
2015;():V01BT01A007. doi:10.1115/PVP2015-45920.

ASME Section XI Code Case N-806, for evaluation of metal loss in Class 2 and 3 metallic piping buried in a back-filled trench, was first published in 2012. This Code Case has been prepared by the ASME Section XI Task Group on Evaluation Procedures for Degraded Buried Pipe. The Code Case addresses the nuclear industry need for evaluation procedures and acceptance criteria for the disposition of metal loss that is discovered during the inspection of metallic piping buried in a back-filled trench. A number of additional improvements have been proposed for Code Case N-806. These include expanded guidance for the determination and validation of a corrosion rate and other clarifications to improve ease of use. This paper presents an update of details of the proposed revisions to Code Case N-806 and their technical basis.

Topics: Metals , Pipes
Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity of Cast Stainless Steel Pipe

2015;():V01BT01A008. doi:10.1115/PVP2015-45191.

ASME Code, Section XI Code Case N-838 provides analyses methods that may be used when performing flaw tolerance evaluation of Class 1 and 2 cast austenitic stainless steel (CASS) piping components with ferrite content exceeding 20%. Tolerable flaw sizes are provided in the Code Case that cannot be exceeded during the evaluation period from an initial postulated flaw. The tolerable flaw sizes were determined based on probabilistic fracture mechanics (PFM) techniques employing toughness and flow strength correlations developed at Argonne National Laboratory (ANL) that estimated the crack growth resistance (J-R) curve using material composition with the room temperature Charpy energy as an intermediate variable. Since the publication of the technical basis for this Code Case, additional toughness data have been identified, some of which are significantly lower than the toughness for CASS CF-8M materials considered in the ANL model.

This paper reports the results of evaluations and sensitivity studies performed to examine the additional toughness and tensile data and determine the effects on the Code Case tolerable flaw size tables. The current study directly employs an expanded set of data of actual measured fully saturated high-temperature crack growth resistance curve, thereby bypassing the material composition-to-Charpy-to-toughness correlations in the ANL model. In this study, both deterministic fracture mechanics and PFM analyses were performed using the assembled data of toughness and tensile properties. The results obtained in the current PFM and deterministic analysis which use the actual toughness and tensile data are found to always provide larger tolerable flaw sizes than those obtained from the previous analyses which used ANL material correlations, thereby demonstrating the conservative nature of the technical basis for the Code Case.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2015;():V01BT01A009. doi:10.1115/PVP2015-45325.

JSME rules for fitness for service have flaw acceptance rules for cast austenitic stainless steel (CASS) pipes. They allow applying two-parameter and elastic-plastic fracture mechanics methods using Z-factor. However they do not clearly describe whether limit load method is applicable for the case of no or low thermal aging condition. The authors performed tensile fracture tests using flat plate specimens with a surface flaw and confirmed that limit load method is applicable in the conditions of no thermal aging and even fully saturated thermal aging with high ferrite number. Also the plate with a shallow flaw ruptured at the critical stress defined by nominal stress at rupture-flaw depth curve in the code case which was determined by the similar flat plate tests of stainless steel or nickel alloy specimens. These results will be reflected to the revision of the code.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A010. doi:10.1115/PVP2015-45434.

In this study, fracture toughness of cast austenitic stainless steel aged at 300–450°C for up to 15000 h were investigated using fracture toughness test, Charpy impact test and indentation hardness test. Test material was statically casted grade CF-3M stainless steel. As a result of the tests, it was found that the fracture toughness and the Charpy absorbed energy tended to decrease with the increase of aging time. However, the behavior of thermal embrittlement varied at each aging temperature. In particular, the fracture toughness of the specimens aged at 300°C was almost the same as that of the unaged specimens. At elevated temperature, the differences of the fracture toughness between unaged specimens and aged specimens were smaller than that of tested at room temperature.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A011. doi:10.1115/PVP2015-45478.

This report summarizes the results of a scoping fracture toughness tests at high and low temperature for thermally aged cast austenitic stainless steels (CASSs) in a pressurized water reactor (PWR) environment. CF8M (ferrite content = 10.1%, 18.9%) and CF8 (ferrite content = 10.5%) were thermally aged up to 5,000 hours at 465°C. Tensile tests, Charpy impact tests and fracture toughness tests were conducted in air at 325°C and 50°C. Fracture toughness tests were also performed in simulated PWR primary water. Although the effect of 325°C and 50°C in simulated PWR primary water and dissolved hydrogen on the fracture toughness (JIc and J-Δa relationship) were slightly observed, fracture toughness was greater than that predicted by the thermally aged fracture toughness prediction method (Hyperbolic-Time-Temperature-Toughness (H3T) model).

Commentary by Dr. Valentin Fuster
2015;():V01BT01A012. doi:10.1115/PVP2015-45790.

Thermal embrittlement of cast austenitic stainless steels (CASS) occurs at reactor operating temperatures during the reactor design lifetime of 40 years leading to a reduction in their toughness and an increase in strength. Additionally most US nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. As the fracture toughness reduces due to thermal embrittlement, some aged CASS materials have the potential to have exceedingly low toughness. CASS can also show high toughness variability due to the variability of its microstructure. Recently an ASME Section XI Code Case N-838 has been proposed to evaluate the flaw tolerance based on probabilistic fracture mechanics (PFM). An assessment of mechanical-property degradation is an input to perform the flaw evaluation procedure in CASS components. There are at least four different models for predicting the change in J-R curves in CASS due to thermal aging. One model is proprietary and the other three are the Argonne/NUREG-CR/4513R1, the French/EDF and a Japanese model.

In this work, two of the thermal aging models were reviewed, reproduced and validated against their example cases for each individual model. Both models were then utilized to assess the fully aged conditions for cases that covers a large spectrum of CASS J R curves with high COV (coefficient of variance). Finally, J-R curves distributions using both Argonne and French models were established by examining the actual chemical compositions of CASS materials found in some US PWR plants. The J-R curves distributions include 21 pipes/fittings in primary pipe loop as well as data from an EPRI report. The calculated toughness variability in a single LBB plant is compared using the Argonne and French models. Additionally the relationship of the “C” and “m” parameters used in the power-law J-R curve equations (J = C×Δam) was explored to determine the proper way to statistically vary the J-R curve in probabilistic analyses.

Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity of Reactor Pressure Vessels and Internals for Codes

2015;():V01BT01A013. doi:10.1115/PVP2015-45065.

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5].

To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6].

Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A014. doi:10.1115/PVP2015-45307.

This paper evaluates current guidance concerning conditions under which the analyst is advised to transition from a linear-elastic fracture mechanics (LEFM) based analysis to an elastic-plastic fracture mechanics (EPFM) based analysis of pressure vessel steels. Current guidance concerning the upper-temperature (T>c) for LEFM-based analysis can be found in ASME Section XI Code Case N-749. Also, while not explicitly stated, an upper-limit on the KIc value that may be used in LEFM-based evaluations is sometimes taken to be 220 MPa√m (a value herein referred to as KLIM). Evaluations of Tc and KLIM were performed using a recently compiled collection of toughness models that are being considered for incorporation into a revision to ASME Section XI Code Case N-830; those models provide a complete definition of all toughness metrics needed to characterize ferritic steel behavior from lower shelf to upper shelf. Based on these evaluations, new definitions of Tc and KLIM are proposed that are fully consistent with the proposed revisions to Code Case N-830 and, thereby, with the underlying fracture toughness data. Formulas that quantify the following values over the ranges of RTTo and RTNDT characteristic of ferritic RPV steels are proposed:

• For Tc, two values, Tc(LOWER) and Tc(UPPER), are defined that bound the temperature range over which the fracture behavior of ferritic RPV steels transitions from brittle to ductile. Below Tc(LOWER), LEFM analysis is acceptable while above Tc(UPPER) EPFM analysis is recommended. Between Tc(LOWER) and Tc(UPPER), the analyst is encouraged to consider EPFM analysis because within this temperature range the competition of the fracture mode combined with the details of a particular analysis suggest that the decision concerning the type of analysis is best made on a case-by-case basis.

• For KLIM, two values, KLIM(LOWER) and KLIM(UPPER), are defined that bound the range of applied-K over which ductile tearing will begin to occur. At applied-K values below KLIM(LOWER), ductile tearing is highly unlikely, so the use of the KIc curve is appropriate. At applied-K values above KLIM(UPPER), considerable ductile tearing is expected, so the use of the KIc curve is not appropriate. At applied-K values in between KLIM(LOWER) and KLIM(UPPER), some ductile tearing can be expected, so it is recommended to give consideration to the possible effects of ductile tearing as they may impact the situation being analyzed.

These definitions of Tc and KLIM better communicate important information concerning the underlying material and structural behavior to the analyst than do current definitions.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A015. doi:10.1115/PVP2015-45382.

Optimization calculation method determining wall thickness for large oil storage tank made of high strength steel is investigated in this paper. Taking three oil storage tanks with different volumes of 10×104 m3, 15×104 m3 and 20×104 m3 for examples, the wall thickness calculation methods of API 650, GB 50341, JIS B 8501 and BS EN 14015 have been analyzed and compared. Results show that as the volume of oil storage tank increases, some wall thickness calculation results of the standards have been larger than the allowable value, leading to the unreasonable distribution of the wall circumferential stress. The wall thickness calculation result applying the method of API 650 is more reasonable than other standards. While for the tanks made of high strength steel, like 12MnNiVR (GB 50341), the yield ratio of the steel has reached 0.803, which is larger than the upper limit value of API 650. In order to make up the deficiency, an optimization method based on API 650 is presented, which considers the effects of yield strength, tensile strength and yield ratio on the determination of allowable stress. Taking the 20×104 m3 oil storage tank and selecting a proper welded joint efficiency, the wall thickness is calculated by the presented optimization method. The wall thickness calculation result is more reasonable and the circumferential stress distribution is more homogeneous when the safety factor of tensile strength is taken to be 2.4. Results show that the optimization method is applicable to the thickness calculation of oil storage tanks made of high strength steel.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A016. doi:10.1115/PVP2015-45850.

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case N-830. This revision incorporates a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all the fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J-R) from the lower shelf through the upper shelf. A paper presented at the 2014 ASME Pressure Vessel and Piping Conference described most of these models; a companion paper at this conference describes the J-R model. This paper also supports the WGFE effort by performing an assessment of the appropriateness of Wallin’s Master Curve model to represent toughness data on the lower shelf, and by comparing the Master Curve with the current Code KIc curve on the lower shelf.

The work presented in this paper supports the following conclusions:

1. The Master Curve provides a reasonable representation of cleavage fracture toughness (KJc) data at lower shelf temperatures. A statistical evaluation of a large database demonstrates that the Master Curve works well to temperatures approximately 140 °C below To or, equivalently, approximately 160 °C below RTTo.

2. The percentile of cleavage fracture toughness data falling below a KIc curve indexed to RTTo varies considerably with temperature. At lower shelf temperatures as much as half of the data lie below the KIc curve, while at temperatures close to RTTo this percentage falls to approximately ≈ 1.5%. The current guidance of Nonmandatory Appendix A to Section XI to use structural factors of √10 or √2 is one means of addressing this inconsistency.

3. The inconsistent degree to which the KIc curve, with or without structural factors, bounds fracture toughness data cannot be fixed within the current Code framework for two reasons: the KIc curve does not reflect the actual temperature dependence shown by the fracture toughness of ferritic RPV steels, and the ratio of a mean or median toughness curve to a fixed percentile bound is not a constant value. It is for these reasons that in the next revision of Code Case N-830 the ASME WGFE is moving away from use of the KIc curve coupled with structural factors and, instead, is adopting models of fracture toughness that represent both the temperature trends and the scatter in the data with high accuracy.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A017. doi:10.1115/PVP2015-45915.

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.

Commentary by Dr. Valentin Fuster

Codes and Standards: Interaction and Flaw Modeling for Multiple Flaws (Joint With M&F)

2015;():V01BT01A018. doi:10.1115/PVP2015-45210.

Growth behaviors of multiple inner cracks, which are located near to structure surface, are simulated using S-FEM. Coalescence of cracks occurs before penetration to surface in some condition. In other condition, penetration occurs before coalescence. By changing relative locations of two inner cracks, these phenomena are observed, and criteria for coalescence and penetration are discussed.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A019. doi:10.1115/PVP2015-45451.

During an in-service inspection, if multiple cracks have been found in a nuclear component, the crack interaction effect due to adjacent cracks should be taken into account to characterize the detected multiple cracks into equivalent single combined crack or independent single crack. However, there must be many factors to be considered to quantify crack interaction effect, many experimental and numerical works should be made to propose robust guidelines on crack interaction effect depending on material characteristics of interest.

Although many works have been made during the past few years to evaluate crack interaction effect of steam generator tubes with multiple cracks, the robust guidelines are still lacking.

In this study, systematic 3-dimensional (3D) elastic-plastic finite element (FE) analyses are performed for steam generator tubes with multiple through-wall cracks. As for geometries of multiple through-wall cracks, four different cases are considered; axial collinear cracks, axial parallel cracks, circumferential collinear cracks, and circumferential parallel cracks. The geometric variables affecting the Pc (coalescence pressure), i.e. crack length and distance between multiple cracks, are systematically varied in the present study.

Based on the coalescence pressure evaluation model proposed by authors in the previous study and the present FE results, the Pc of steam generator tubes with multiple cracks are investigated.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A020. doi:10.1115/PVP2015-45792.

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the RPVs.

In the frame of the Structural Integrity demonstration of these RPVs according to ASME XI principles, ASME XI IWB-3300 article requires combining closely spaced flaws in order to account for their mechanical interactions. However, it appeared early that the characterization rules were adapted neither to quasi-laminar flaws nor to such densities of flaws.

Therefore, an alternative methodology to derive characterization rules for quasi-laminar flaws has been developed, implemented and validated.

This work, based on 2D eXtended Finite Element Method (X-FEM) calculations and presented during ASME PVP 2014, has led to the proposed ASME Code Case N-848 “Alternative characterization rules for quasi-laminar flaws – Section XI, Division I”.

This 2D approach, even though better suited to quasi-laminar flaws, results however in very conservative proximity rules.

Therefore, it appeared that more realistic — although still conservative — proximity rules based on 3D X-FEM calculations could be developed.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A021. doi:10.1115/PVP2015-45901.

If a subsurface flaw is located near a component surface, the subsurface flaw is transformed to a surface flaw in accordance with a subsurface-to-surface flaw proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the specific criteria of the re-characterizations are different among the FFS codes. Recently, the authors have proposed a new subsurface-to-surface flaw proximity rule based on the experiments data and the interaction of stress intensity factors. In this study, extended Finite Element fatigue crack growth calculations were carried out for thick wall component like vessels with subsurface flaws, using the proposed subsurface-to-surface flaw proximity rule and the proximity rule provided in the current ASME Code Section XI.

Different, flaw aspect ratios and ligament distances from subsurface flaws to inner surface of vessel were taken into account. As the results, the current proximity rule and proposed one provide relatively similar fatigue lives, whatever the aspect ratios of the initial subsurface flaws. However, when the thickness of the component decreases this similarity between both proximity rules appears not to be valid anymore.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A022. doi:10.1115/PVP2015-45946.

A subsurface flaw located near a component surface is transformed to a surface flaw in accordance with a flaw-to-surface proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the criteria of the re-characterizations are different among the FFS codes. In addition, the proximity factors in the rules are defined by constant values, irrespective of flaw aspect ratios. This paper describes the stress intensity factor interaction between the subsurface flaw and component free surface, and proposes a proximity factor from the point of view of fatigue crack growth rates.

Topics: Fatigue cracks
Commentary by Dr. Valentin Fuster

Codes and Standards: Master Curve Method for Fracture Toughness

2015;():V01BT01A023. doi:10.1115/PVP2015-45154.

Evaluation of integrity and lifetime of reactor pressure vessels is usually based on fracture mechanics approach using empirical correlation between transition temperatures from impact tests and static fracture toughness test results in the form of “design curve”. Moreover, material degradation during operation is also usually monitored by impact surveillance specimen testing under the assumption that shifts in temperature dependencies if impact toughness and static fracture toughness are the same.

To verify this assumption, study of the correlation between these two shifts has been performed on WWER steels — 15Kh2MFA (Cr-Mo-V) and 15Kh2NMFA (Ni-Cr-Mo-V) types. Several sources of results have been used : (a) reconstitution of tested remains of Charpy V-notch impact test specimens from irradiated programs was performed to obtain pre-cracked Charpy size specimens for three point bending type fracture toughness testing, (b) comparison of tests results from surveillance programs irradiated by similar fluences, (c) experimental irradiation programs with accelerated irradiation in research reactor.

Additionally, some results from the recent study of irradiation embrittlement of high nickel weld are included — its behavior shows to some extraordinary tendency.

Thanks to the use of reconstitution, both series of specimen were irradiated under the same conditions — temperature and neutron fluence and comparison is reliable. Results show that transition temperatures from fracture toughness testing are larger than those from Charpy impact tests. Similar results have been obtained also for other two groups of results.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A024. doi:10.1115/PVP2015-45253.

At the 2014 ASME Pressure Vessel and Piping Conference, these authors and others presented a paper that drew together a number of models describing the fracture toughness of ferritic reactor pressure vessel (RPV) steels. That paper summarized models of both the temperature dependence and scatter in a number of fracture toughness metrics (i.e., KJc, KIa, JIc, and J0.1). That paper also provided equations that quantify the interrelationships between these toughness metrics, and how these interrelationships are affected by hardening. Significantly, all of these models and interrelationships are linked via a single parameter: the Master Curve index temperature, To, which can be measured as described in ASTM Standard Test Method E1921. Work is currently underway within the ASME Section XI Working Group on Flaw Evaluation (WGFE) to develop a revision to Code Case N-830 that incorporates all of these models, and provides information on how to apply them in a flaw evaluation. As part of that work, an effort was initiated to augment these models by the addition of a model that can be used to predict the temperature variation of, and the scatter in, J-R curve behavior. A J-R curve model is also expected to support on-going WGFE efforts to in development of acceptance criteria for flaws in ferritic components operating in the upper shelf temperature range.

The work presented in this paper provides a model of the J-R behavior of ferritic RPV steels. When combined with other fracture toughness models to be published in Code Case N-830-1, this model allows prediction of the mean J-R curve, confidence bounds on the mean, and the temperature dependence of J-R all based only on input of To. The J-R model described herein has equivalent or better accuracy to other models described in the literature, and generally has fewer fitting parameters than those other models. Because the full J-R curve is predicted, this model is also useful for prediction of J0.1.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A025. doi:10.1115/PVP2015-45279.

Reactor pressure vessel (RPV) multilayer welding seams show an inhomogeneous structure. It raises concerns that the evaluation of non-uniform material might not be amenable to the statistical analysis methods on which the Master Curve approach is based. In particular with regard to weld metals, it can be expected that the cleavage fracture toughness is strongly influenced by the orientation of the Charpy size SE(B) specimen. The T-L oriented SE(B) specimen (axis axial and crack propagation in circumferential direction) comprises of various welding beads along the crack front whereas in a L-S specimen (axis axial and crack propagation through the thickness) the crack tip is located in one welding bead with an approximately uniform structure.

The paper summarises fracture toughness results measured on welding seams of decommissioned and non-commissioned RPVs of WWER type nuclear reactors and the non-commissioned Biblis-C RPV. Specimens of T-L and T-S orientation were tested. The results show, that in general the cleavage fracture toughness values, KJc-1T, follow the Master Curve description. However, the number of KJc-1T data outside the 2% and 98% tolerance bounds is larger than predicted by the underlying model, which indicates non-uniform material.

There is a large variation in the evaluated through thickness T0 values of the investigated multilayer beltline welding seams. Within the sampling range of the surveillance specimens, T0 values vary with a span of 30 to about 60 K depending on the applied welding technology. The fracture toughness strongly depends on the intrinsic weld bead structure. Hence, the position of the fatigue crack tip of the pre-cracked SE(B) specimen at the multilayer welding seam is crucial and defines the cleavage fracture toughness. Modified Master Curve based evaluation procedures like the MC based approach of the SINTAP procedure were applied to get fracture toughness values which are representative for the most brittle fraction the test series.

Despite of the pronounced non-homogeneity of the micro-structure along the crack front of T-L specimens, crack initiation sites are randomly distributed along the crack front. This means that one of the basic assumptions in ASTM E1921, i.e. the uniform distribution of initiation sites, is fulfilled also for the T-L specimens from the multilayer weld metal.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A026. doi:10.1115/PVP2015-45502.

Reactor pressure vessels of PWR/BWR/WWER type reactors are covered by austenitic cladding made by welding on their inner wall. Austenitic materials usually have no transition temperature behavior as they have fcc crystallographic structure. But, austenitic cladding made by welding contain usually up to 8 % of delta-ferrite that results in some transition behavior of fracture properties. This transition can be observed in temperature region below room temperature.

Surprisingly, this transition behavior in static fracture toughness of both cladding layers can be well described by “Master curve” approach.

Results from testing austenitic cladding for WWER type reactors will be shown and discussed, ether in unirradiated as well as irradiated conditions — only small changes in fracture toughness properties in this transition region are observed as a result of irradiation.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A027. doi:10.1115/PVP2015-45554.

Although Ultra High Strength Steels (UHSS) with nominal strengths up to 1500 MPa have been available on the market for many years, the use of these steels in the civil engineering industry is still rather uncommon. One critical point limiting the use of UHSS steels lies in their rather poorly documented fracture properties in relation to more conventional steels covered by the codes. The major concept governing the assessment of steels is the Master Curve (MC) methodology. It provides a description for the fracture toughness scatter, size effect and temperature dependence in the ductile to brittle transition region. It enables a complete characterization of brittle fracture toughness of a material based on only a few small size specimens. The method combines a theoretical description of the scatter, a statistical size effect and an empirically found temperature dependence of fracture toughness. The fracture toughness in the brittle fracture regime is thus described with only one parameter, the transition temperature T0. At this temperature the mean fracture toughness for a 25.4 mm thick specimen is 100 MPa√m. The Master Curve method as defined in ASTM E1921-13a is applicable to ferritic structural steels with yield strength between 275 MPa and 825 MPa. Very few studies have been made with respect to the applicability of the Master Curve to Ultra High Strength Steels with yield strengths in the excess of 900 MPa. This is the topic of this work. Focusing on novel directly quenched high performance steels, the applicability of the Master Curve methodology with special emphasis on the temperature dependence will be investigated. Possible improvements to the Master Curve will be proposed for further consideration.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A028. doi:10.1115/PVP2015-45706.

This contribution deals with determination of the reference temperature of JRQ steel using miniaturized specimens. The dimensions of used miniaturized specimens were 3 × 4 × 27 mm (thickness × width × length). This specimen type offers the utilization of limited amount of test material or broken halves of precracked Charpy and larger specimens. The test material comes from the broken halves of 0.5T SEB specimens previously tested for purposes of the reference temperature determination in Ciemat, Madrid. The fracture toughness tests of specimens were performed in the transition region of the steel according to the recommendations of standard ASTM E1921 and according Wallin’s recommended temperature range for miniaturized specimens. The determined reference temperature of the Master Curve was very similar to the determined ones from three-point-bend specimen of sizes 0.2T, 0.4T and 0.5T. The obtained results confirm a necessity of conduct of tests at low temperatures and testing sufficient number of specimens in order to generate enough valid data for determination of the reference temperature.

Topics: Temperature , Steel
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in ASME Codes and Standards

2015;():V01BT01A029. doi:10.1115/PVP2015-45023.

External pressure charts of ASME Boiler & Pressure Vessel Code do not account for reduction of buckling strength due to creep under long-term loads at elevated temperatures. This restricts the design of ASME boiler and pressure vessel components in the creep range. Although the design factors and equations for creep buckling are available in various other literatures, the external pressure charts in the creep range are not available at present time. External pressure charts are developed for most commonly used elevated temperature materials (Carbon & 1.25Cr-0.5Mo and 2.25Cr-1Mo low alloy steels) in the refinery and petrochemical industry up to a temperature of 550 °C. API 579-1/ASME FFS-1 isochronous stress-strain curves from Omega model, which does not include the effects of plasticity and primary creep, are used. Thus adjustment factors are proposed based on available isochronous stress-strain curves which considers the plasticity and primary creep effects. The developed external pressure chart for 2.25Cr-1Mo steel at 538 °C is validated with ASME B&PV Code Case 2676. The loads and load combinations to be considered in the design and examples illustrating the application of the developed external pressure charts in the creep range are presented.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A030. doi:10.1115/PVP2015-45347.

Stresses in Class 1 branch connections that consist of large bore run pipe with a reinforced branch nozzle should rarely be limited by the run–branch interface stresses. The end of the “branch nozzle – branch pipe” interface is the location on the branch nozzle one would expect to see the limiting stress. Therefore, it is important that reasonable Design Rules are maintained in the ASME Code Section III for the stress analyses of the Class 1 Piping branch connections to avoid over-predicting the “run pipe - branch nozzle” interface stresses. This will allow the analysts to concentrate on load reductions needed in a logical manner.

In Class 1 Piping Design, the calculation of the branch total stress due to the moments is the result of the sum of the stresses from the run moment and of the stresses from the branch moment with these branch moment stresses being calculated using either the branch pipe cross-section or the branch nozzle cross-section. This in itself is already severe, when compared with other Piping Design Rules for branch connections. In addition, starting with the ASME Code year 2002, the branch-side moment stress is based exclusively on the branch pipe cross-section, which leads to a higher moment stress, and this higher moment stress is still absolutely added to the run-side moment stress. As indicated in that ASME Code year 2002 and beyond, this addition is independent of the length of the branch nozzle reinforcement. This leads to total moment stresses that are the sums of moment stresses that do not occur at all at the same location.

The purpose of this technical paper is to compare a) the stresses calculated with the earlier more correct Class 1 Piping methodology from 2001 and before 2001; b) the stresses calculated with the more recent and more severe Class 1 Piping methodology; and c) the stresses from finite-element analyses.

Conclusions are provided on what should be done for the future Class 1 Piping Design methodology of branch connections.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A031. doi:10.1115/PVP2015-45696.

In conducting a Class 1 piping analysis per the simplified rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Article NB-3600, a fatigue analysis is required per paragraph NB-3653 for both Service Level A and Service Level B. The fatigue analysis provides two options. The options are dependent on Equation 10 of subparagraph NB-3653.1. If this equation is met for a given load set pair under consideration, then the analysis proceeds directly to subparagraphs NB-3653.2 through NB-3653.5. If however, Equation 10 is exceeded, the Code allows the use of a simplified Elastic Plastic Analysis as delineated in subparagraph NB-3653.6. The first requirement of NB-3653.6 is that both Equation 12 and Equation 13 must be met. The changes in the seismic design in the last 25+ years have not been appropriately reflected in the subparagraph NB-3653.6(b) Equation 13. Also, the Code provides no clear guidance on seismic anchor motions in paragraph NB-3650. In 2012 ASME Code Committees undertook an action to address these issues. This paper provides the background and basis for Code changes that are anticipated will be implemented in the near future in paragraph NB-3653.6 of the ASME Boiler and Pressure Vessel Code, Section III, Division 1 that will address both of these issues. This implementation will make the Elastic Plastic Fatigue rules of NB-3653.6 consistent with the design by analysis approach of NB-3228.5.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A032. doi:10.1115/PVP2015-45697.

When the piping seismic design rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsections NB, NC, and ND were updated in 1994 a strain based criterion was incorporated in Article NB-3200. Due to objections to this criterion by the USNRC they were subsequently removed and replaced by guidance which stated that for Article NB-3200 for Service Level C and D reversing dynamic loads use the simplified design rules of Article NB-3600. The use of simplified Article NB-3600 rules in the NB-3200 design by analysis Article is inconsistent with the philosophy of NB-3200 and precludes the analyst from using detailed analysis methods for piping for Service Level C and D reversing dynamic loads. In 2012 an effort was undertaken to develop a more detailed design by analysis criteria for reversing dynamic loads. This paper provides the background and basis for changes that are anticipated to be implemented in the near future in the ASME Boiler and Pressure Vessel Code, Section III, Division 1 for the design by analysis rules of Article NB-3200 for reversing dynamic loads in piping systems.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Chinese Codes and Standards

2015;():V01BT01A033. doi:10.1115/PVP2015-45060.

Due to hazardous chemicals with pressure inside and service in different areas, the transportable pressure vessels (TPV) are of high risks. Therefore, compulsory supervision should be carried out for the sake of security of TPVs all over the world. In China, TPV contains 5 categories, which are railway tank car, road tanker, tank container, tube trailer and tube container. In this paper, Chinese code and standard system of TPVs were introduced in detail by 5 levels, which are law, regulation, divisional regulation, safety technical regulation and standard. The scope and main content of the documents were presented briefly. The differences of the Chinese code and standard system were compared with that of EU and the United States. The causes of the differences were analyzed and some advices were brought forward. At last, basing on the current situation of the design, manufacture and inspection technology, the trends on development of Chinese codes and standards, tank material, design load, safety factor and allowable stress etc. of TPV were presented.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A034. doi:10.1115/PVP2015-45118.

This paper mainly introduced the Chinese codes and standards of glass-lined pressure vessels. The manufacture and major maintenance of pressure vessels shall be subject to supervision and inspection by the authorized inspection institutions in accordance with the safety technical codes. Design, materials, weld, nondestructive testing, glass-lined process, hydrostatic test and leak test were investigated. In addition, test specimens for calculating wall thickness, Welding Procedure Qualification and Glass-lined Procedure Qualification were analyzed. TSG R7004-2013 Regulation on Supervision Inspection for Pressure Vessel specifies Glass-lined Procedure Qualification shall be carried out before glass-lined. Glass-lined procedure qualification is used to control temperature range, heating rate, holding time, and cycle times. The glass-lined process is absolutely critical to the longevity of the glass-lined vessel.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A035. doi:10.1115/PVP2015-45152.

Large capacity cylinders installed in the frames of tube trailers subject to torques caused by their own curvatures and out of roundness under the action of gravity. The torques may induce additional stresses to the ends of the cylinders connecting to the frames. If the torques are big enough, the cylinders may rotate significantly, which can deform or even break the gathering and transferring pipes connecting to the cylinders. An explicit calculation method was presented to predict the maximum torques introduced by the degrees of curvatures and out of roundness of the cylinders under gravity, and the torques calculated agree with the tested results of 5 steel cylinders with a water capacity of 2250 L, which verifies the usability of the calculation method.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A036. doi:10.1115/PVP2015-45203.

The fracture characteristics of several types of tensile specimens of 9%Ni steel at low temperature was studied, and the Notch (Crack) Sensitivity and Notch (Crack) Relative Strength decreased (increased) Ratio of notch (crack) specimen were studied, while the crack preparation method of the crack specimen was discussed. The results indicate that the Notch (Crack) Sensitivity and Notch (Crack) Relative decreased (increased) Ratio can be used to evaluate the low temperature low stress brittle fracture resistance of the material. 9%Ni steel has good brittle fracture resistance at low temperature. The crack preparation principle of the crack tensile specimen was put forward.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A037. doi:10.1115/PVP2015-45228.

As the localized temperature drop induced by the Joule-Thomson cooling effect in a leak causes a reduction in the fracture toughness at the crack front, leak-before-break approach that does not take this effect into account may be unconservative. In this paper, argon was selected as the experimental gas used in the experiment due to its incombustibility and similar Joule-Thomson coefficient to that of methane. An experimental pressure vessel with a design pressure of 250 bar was designed and fabricated. Liquid nitrogen cracking method was employed to fabricate a realistic through-thickness crack in a test plate. Under the condition of ambient temperature of 30 °C and maximum internal pressure of 91 bar, the temperature of argon at the exit of the crack and the measured lowest temperature of metal near the crack are −9.2 °C and 7.9 °C, respectively.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A038. doi:10.1115/PVP2015-45330.

The degeneration of mechanical properties is one of the main concerns in assessment of fire damaged pressure vessels. This study investigates the influence of fire exposure on mechanical properties of Q345R steel which is widely used for pressure vessels in China. Heat treatment with different temperatures and holding times was conducted to simulate various heat exposure conditions in fire event. Hardness testing, metallographic analysis and tensile tests were carried out to investigate the effects of fire exposure temperature and duration. The experimental results indicate that the inflection temperature for mechanical property degeneration of Q345R steel is 700 °C. The decline of hardness, yield and tensile strengths due to spheroidization become more obvious with increasing heat exposure duration. A linear correlation is indicated by fitting the tensile strength and hardness. For the assessment of fire damaged component, the mechanical properties of Q345R steel at room temperature can be determined combining on-situ field metallographic examination and hardness testing with.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A039. doi:10.1115/PVP2015-45352.

In recent years, Chinese petrochemical plants have been continuously developing towards the direction of extreme conditions such as high temperature, high pressure, aggressive medium, large scale, etc. Once the seal structure of the pressure equipments for petrochemical plants leaks, it will not only cause material loss and nonscheduled shutdown, but also may cause disastrous accidents such as fire, poisoning, environment pollution, personal casualties, etc. This paper focuses on the flange seal structure of pressure equipment for petrochemical plants, makes comparison analysis of the standards/codes of different countries, points out the deficiency of Chinese current standards/codes and gives the advice that controlling leak rate of volatile organic compounds (VOCs) to 2×10−5mg/s • mm as the target to conduct relevant research on design, manufacture, installation and maintenance of flange seal that satisfies leakage rate control requirement and improvement of design standard.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A040. doi:10.1115/PVP2015-45383.

Double-deck floating roof has gained more and more popularity in large oil storage tanks. Wind load is one of the main factors that cause failure of the floating roofs. Because of complexity of structure and difficulty in theoretical calculation, wind load on double-deck floating roof has not been considered into the general design standards. An engineering calculation method based on finite element method for the strength and stability of double-deck floating roof under wind load is established. The relation between pressure distribution on floating roof and liquid height is studied when the double-deck floating roof locates at different levels. Results show that pressure difference on the roof reaches maximum when the liquid height is 100%. The finite element model of double-deck floating roof is established based on large deformation theory, the overall displacement, deformation and stress are obtained by applying pressure distribution on the roof. The first anti-sinking criterion and axial compression stability criterion of long bar are selected to verify the strength and stability of floating roofs under wind load. Results show that the engineering calculation method presented can be effectively used to analyze the strength and stability of double-deck floating roof under wind load.

Topics: Stability , Stress , Roofs , Wind
Commentary by Dr. Valentin Fuster
2015;():V01BT01A041. doi:10.1115/PVP2015-45829.

The models of Ti clad steel tube sheet and carbon steel tube sheet are established in this paper. By the finite element analysis software ANSYS Workbench, the disciplinarian of stress distribution is investigated under thermal and mechanical loadings of tube sheet. Through the optimization, the effects of base layer, clad layer thickness and tube wall thickness on clad steel tube sheet are obtained. Results show: the overall stress distribution of clad steel tube sheet is more complex than that of carbon steel tube sheet, and the connection surface of base and clad layer presents the phenomenon of stress concentration for clad steel tube sheet. The increase of clad layer or tube wall thickness has an unfavorable effect on the quality of clad steel tube sheet. Through reasonably decreasing thickness of clad layer or tube wall, stress concentration can be improved and performance of clad steel tube sheet can be enhanced. The results provide some reference for the design of clad steel tube sheet.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A042. doi:10.1115/PVP2015-45870.

Currently the reliability of rotating equipments in typical petrochemical plants, such as turbo machineries, centrifugal compressors, reciprocating compressors, fans, pumps, etc. plays an important role in a long-term operation of the plants. The important tasks for management technology research of rotating equipments are to build a new type of maintenance management platform and take an advantage of the reliability analysis technique for fault diagnosis and maintenance strategy optimization. In this paper, firstly, the development of maintenance management technology in China and overseas was introduced. The problems of traditional reliability analysis techniques were pointed out. Secondly, the key technologies of the new type of maintenance management system were introduced. Furthermore, the structure and work process of the new type of maintenance management system were proposed. Finally, the key contents and functions of this system were given.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A043. doi:10.1115/PVP2015-45882.

Using the Discrete Phase Model (DPM) model in fluent software, a simulation study for elbow erosion situation was performed for different inlet particle concentrations, inlet velocities and curvature radiuses based on the similarity theory. The distribution of the serious erosion location and the fluid field of elbow can be determined. The results of numerical simulations show that the maximum speed is in the region of the small and big bend surfaces of the elbow. The small bend surface has a minimum pressure and the big bend surface has a maximum pressure. The most serious erosion region is mainly located in the 52°–60° region of the big bend surface near the outlet. The position and shape of the serious erosion almost does not change with the inlet concentration and velocity. Finally, based on the dimensional analysis the simulation analysis was carried out. The empirical formula on the inlet concentration, inlet velocity and bend radius ratio are obtained. The error of the formula is acceptable by verification. The research results can provide an effective help for selection of the elbow, determination of dangerous operational conditions and life prediction of the elbow.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A044. doi:10.1115/PVP2015-45890.

In this paper, a failure analysis was performed on a high-pressure steam pipeline, which was burst after service in a chemical plant for 12 years. The burst field was investigated and the fragments were observed. Microstructures of the pipeline were analyzed by optical microscope (OM) and scanning electron microscope (SEM). The mechanical properties tests, including tensile testing and Charpy V notch impact testing, were performed on specimens removed from the high-pressure steam pipeline. The impact testing results showed that toughness of the material was poor and little resistance against cracking extension. According to further analysis, strain aging was the dominant mechanism of the failure. Some inspection approaches were recommended to prevent similar failures.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A045. doi:10.1115/PVP2015-45892.

A crack was observed on an outlet elbow of the pre-converter in coal gasification unit during operation. This paper details the investigation into the failure and highlights the most probable cause of failure based on available documents and experimental analysis. Visual examination, chemical components analysis, energy spectrum analysis, fracture analysis, metallurgical analysis, mechanical properties test and residual stress measurement were performed. The experimental results show that the primary crack initiated from inside and propagated to outer surface of the elbow. The content of titanium element was lower than the requirement in GB/T 14976-2002. Corrosion products were rich in O and S elements. Amounts of secondary cracks and strain induced martensite were observed. Furthermore, the residual stress on the inner surface near the crack tip was extremely high. According to the experiment results and the analysis of operating condition and history, the failure mechanism of the elbow is stress corrosion cracking. Sensitization of the stainless steel due to low Ti content and the faulty heat treatment contributed to the intergranular stress corrosion cracking.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in European Codes and Standards

2015;():V01BT01A046. doi:10.1115/PVP2015-45018.

Transition welds have been used in some Advanced Gas-Cooled Reactors to attach a 2¼Cr1Mo spool piece to a hot reheat header, which is fabricated from 316L material. These welds were made using Inconel-82 filler and were post-weld heat-treated (PWHT) at around 705°C. In this paper, a creep-fatigue crack growth assessment has been carried out on the 2¼Cr1Mo ferritic side of the weld using the R5 procedure.

As the welds have been PWHT rather than solution heat treated, therefore there are residual welding stresses present. On the 2¼Cr1Mo ferritic side of the weld, a 60MPa through-wall bending stress has been applied in the axial direction and a 60MPa membrane stress in the hoop direction. These are consistent with the recommendations of R5 Volume 7.

Due to the differing thermal expansion coefficients of Inconel and ferritic materials, thermal mismatch stresses induced from changes in temperatures have been considered. It has been assumed that at the PWHT temperature of 705°C there are no thermal mismatch stresses. Any reduction from this temperature generates thermal mismatch stresses which are accounted for by using secondary Stress Intensification Factor values in the calculation.

On the 2¼Cr1Mo ferritic side of the weld, a 5mm deep by 50mm long circumferential defect (semi-elliptical surface-breaking surface defect) postulated to exist on both internal and external surfaces has been used. Limiting defect sizes have been calculated following the R6 procedure.

Four cases have been investigated. After presenting the investigation results, conclusions are drawn.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A047. doi:10.1115/PVP2015-45033.

The paper describes the general approach followed by AFCEN, the French Society for Codified Rules for Design, Construction and In-Service Inspection of Nuclear Island Components, from the technical and organizational points of view. The RCC-M code, reissued in 2012 and modified with addenda in 2013 2014, and 2015 is presented. The main new topics of activity of the RCC-M Subcommittee are considered: conformity with regulation(s), use of International Standards, equivalence with other codes and harmonization, and latest publication for the quality management system.

The paper highlights how industrial experience is currently being integrated into the RCC-M code, and how the code is evolving to take into account the enlargement of the AFCEN Membership, new AFCEN organization rules, and the international environment, and best practices. The processes for dealing with requests for modifications and interpretations are described.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A048. doi:10.1115/PVP2015-45095.

The underlying purpose of this paper is to evaluate whether the CEN CWA 15627 “Small Punch Test Method for Metallic Materials” first published in 2006 has indeed succeeded in providing a stimulus for a wider implementation of the small punch test technique in industrial applications throughout Europe and indeed worldwide. A wealth of research progress has been apparent, as strongly evidenced in three dedicated SSTT (Small Specimen Testing Techniques) conferences held in Europe over the last five years, but also in the wider literature. In particular it is important to mention the recent publication of a Japanese standard and the announcement of parallel progress in China. The present paper concentrates on progress within Europe from the launch of the Code to the present day. In particular attention is focused on the need for industrial acceptance of the test methodology and methods for evaluating the results. Some scepticism still seems to prevail within sectors of the conventional power generation industry, an industry which can potentially benefit most from successful remanent lifetime extension tools of which small punch testing can be considered as a prime candidate. In spite of this, it is demonstrated that a major proportion of the Small Punch testing research of the last decade has been carried out on power plant steels. Meanwhile it is shown that there is evidence that the original remit of the methodology in assessing the integrity of irradiated nuclear plant remains active, new interest is developing for aerospace and next generation nuclear applications enhancing further the credibility of the Code.

Topics: Testing
Commentary by Dr. Valentin Fuster
2015;():V01BT01A049. doi:10.1115/PVP2015-45817.

Tubesheets are usually designed according to different national or international design codes. The great majority of these standards is based on Gardner’s theory, elaborated more than 60 years ago. On the other hand these pressure components are critical in heat exchanger design, because they are subject to characteristic service damages which require a correct dimensioning and appropriate inspections during service.

This paper is aimed at comparing the different design methods, analyzing the theoretical background behind the rules.

Main focus is made to the alternative design method contained in Annex J of the European Standard EN13445-3.

With reference to the typical configuration of a fixed tubesheet exchanger with flanged connections, the results of the different design approaches are compared in order to find out the optimal configuration.

Benchmark examples are carried out using a commercial computer code with reference to heat exchangers with the tubesheets welded to the shell and bolted to the channel.

The results show the advantage of using Annex J which allows smaller thicknesses of the tubesheet in respect of the conventional approach used by TEMA, ASME Section VIII and EN 13445-3 clause 13.

Topics: Design
Commentary by Dr. Valentin Fuster
2015;():V01BT01A050. doi:10.1115/PVP2015-45874.

A number of standards and guidelines associated with integrity of bolted joints were updated or introduced in 2013 and 2015 will see the release of “The Energy Institute Guidelines for the Management of Integrity of Bolted Joints in pressurized systems.” The first guidelines were introduced in 2002 in response to a drive to reduce Hydrocarbon leakage in the UK Offshore Industry; the third edition sees the industry once again targeting Hydrocarbon Leak Reduction. This paper will give an overview of the Guidelines — and discuss the key changes on the previous versions and alignment with ASME PCC-1-2013 and ASME PCC-2-2011. Ongoing work is in progress on European Standards: July 2013 saw the publication of the harmonized calculation standard. En1591-1. Late 2013 saw the training and competence standard EN1591-4 published and this standard is now being implemented. A guide and sample calculations to EN1591-1 are being worked on under EN1591-6. The paper will discuss other developments in Europe with the aim of complementing and harmonizing with the work of ASME.

Topics: Bolted joints
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Japanese Codes and Standards

2015;():V01BT01A051. doi:10.1115/PVP2015-45262.

According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure and the current code design allowable stresses are very conservative. Since the stress assessment based on the elastic analysis does not reflect actual response of piping systems including plastic region, rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load are expected to be developed for piping seismic design applications.

With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a research activity has been planned. Through the activity, the authors intend to establish two kinds of guidelines; 1) a guideline of a standard analysis procedure to evaluate elastic-plastic behavior of piping systems under extreme seismic loads with rational and conservative margins, and 2) a guideline that provide criteria for the seismic safety assessment of piping systems by the standard analysis to evaluate elastic-plastic behavior established by the above guideline. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test.

In this paper, the outline of the research activity and the preliminary results of benchmark analyses for a pipe element test are described.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A052. doi:10.1115/PVP2015-45265.

Allowable stress for code is determined based on material tensile strength SU using design margin, where the design margin is normally taken to be 5.0, 4.0 or 3.5. ASME B&PV Codes Sections I, III and VIII Division 1 had improved the design margin from 4.0 to 3.5 in 1998. Japanese industrial codes such as JIS (Japanese Industrial Standards), HPI (High Pressure Institute) codes and Japanese government rules has two design margins of 4.0 and/or 3.5. One of the Japanese government rules is discussing to introduce the design margin of 3.5 for thermal power boilers. The allowable stress based on 3.5 increases to 14% larger that based on 4.0. The objective of the paper is to demonstrate that minimum flaws by NDE (non destructive examination) technologies can be well identified, and permissible flaws by NDE are harmless to cause fractures, even when the design stresses increase by 14%.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A053. doi:10.1115/PVP2015-45275.

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014.

In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A054. doi:10.1115/PVP2015-45411.

When the structural integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) events is assessed, an underclad crack is postulated at the inner surface of RPV and the stress intensity factor (SIF) corresponding to the driving force of non-ductile crack propagation, is evaluated for this crack. On the inner surface of RPV, cladding of the stainless steel is overlay-welded as a means for corrosion protection. Because the cladding is a ductile material, it is important to evaluate the SIF for postulated underclad crack considering the plasticity of cladding. A SIF evaluation method, which takes the effect of plasticity into account using a plastic correction method, has been established in France. In this study, we examined the SIF evaluation method established in France for underclad cracks during PTS transients. The elastic and elastic-plastic analyses based on the finite element method considering PTS events and inner pressure were performed using three-dimensional models including an underclad semi-elliptical crack with different geometry. We discussed the conservativeness of plastic correction method based on the analysis results.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A055. doi:10.1115/PVP2015-45497.

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A056. doi:10.1115/PVP2015-45499.

The use of miniature C(T) specimens makes it possible for the direct determination of the reference temperature of reactor pressure vessel steels, because they can be taken from the broken halves of the Charpy specimens used for surveillance program to monitor neutron irradiation embrittlement. Fracture toughness tests using C(T) specimens usually need the measurement of load-line displacement, however, it is difficult to mount a clip gauge inside the miniature specimen due to the limitation of the specimen size. A pair of knife edges is machined at the front face of the miniature C(T) specimen to mount a clip gauge with razor blade tips, and the front-face displacement is translated to the load-line displacement. When front-face displacement measurements are made, the load-line displacement can be inferred by multiplying the measured values by the constant 0.73. This conversion factor has been simply derived from the assumption of the linear deformation around a supposed point of rotation. In this study, the conversion factor was directly evaluated by using a three-dimensional elastic-plastic finite element analysis for the miniature C(T) specimens, and the adequacy of the conversion factor was investigated.

Topics: Stress
Commentary by Dr. Valentin Fuster
2015;():V01BT01A057. doi:10.1115/PVP2015-45721.

This paper shows a benchmark result regarding elastic-plastic seismic response analysis and fatigue evaluation for of piping model excitation test. National Research Institute for Earth Science and Disaster Prevention (NIED) has a lot of experimental results of shaking table test of piping. To propose advance seismic evaluation method for piping under severe seismic event, a task in JSME Code committee performs benchmark activity for elastic plastic seismic response analysis and evaluation for piping by using NIED’s experimental results [1].

Authors are taking part in the task and have been performed a benchmark of seismic response analysis and seismic evaluation by comparing with the experimental results. For fatigue evaluation, the strain ranges and the cycles of each range obtained from strain time history were evaluated. The response acceleration and displacement was evaluated for plastic collapse.

Topics: Fatigue , Pipes
Commentary by Dr. Valentin Fuster
2015;():V01BT01A058. doi:10.1115/PVP2015-45853.

Carbon steel STS410 (JIS Standard), which is widely used for high pressure piping components, exhibits cyclic hardening under repeated loading. Extreme seismic loading can cause repetitive large strains, eventually leading to the failure of components. For failure assessment of such components, inelastic analyses using cyclic plasticity constitutive models are needed. In this paper, a multilayer kinematic hardening model for cyclic plasticity, equipped with a set of standard stress-strain characteristics, is developed for STS410 under isothermal condition of various temperatures. This model can express not only the nonlinearity of stress-strain relations, but cyclic hardening of a material by introducing a generic stress-strain relation composed of a combination of monotonic and steady state cyclic stress-strain curves. Finite element large deformation elastic-plastic analyses with this model are conducted for a cyclic in-plane bending test of an elbow. The proposed constitutive model predicted well characteristic features of global deformation and local strain behaviors of the elbow.

Commentary by Dr. Valentin Fuster

Codes and Standards: Repair, Replacement and Mitigation for Fitness-for-Service Rules

2015;():V01BT01A059. doi:10.1115/PVP2015-45483.

Operational small leakage is occasionally observed in a nuclear power plant, and the leak forces an operator to decide whether to shut down the plant or not. Even if the leakage is just a little, it might draw the considerable attention in the society, so that the operator sometimes gets into the situation to judge more severely than technical judgment. Furthermore, at the time of plant restart and the system leak test just after maintenance, even the operator doesn’t accept any leakage considering the long management for the leakage up to the next outage.

On the other hand, once the operator shut down the plant, it sometimes takes long time to restart again because of the difficulty to obtain new pipes and valves in short time. The temporary repair techniques referred to the JSME code might be able to be applied to maintain the plant operation, however some difficulties exist in a practical process.

One of the authors has faced with many cases in which the operational small leakage had to be dealt at Tsuruga nuclear power station. This paper shows some cases of them and discusses lessons which are related to the codes and standards.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A060. doi:10.1115/PVP2015-45903.

Chapter of Repair, Replacement Activities (RRA) in Fitness-For-Service (FFS) Code of the Japanese Society of Mechanical Engineers (JSME) includes rules of article RB-3000 for temporary repair techniques in the use of covering leakage during operation of a Nuclear Power Plant (NPP). Temporary repair techniques are RB-3010 Plating, RB-3020 Adhesion and RB-3030 Infill. In this paper, rules of these temporary repair techniques are summarized as well as the meaning of ‘temporary’, intention and benefit of them.

On the other hand, these rules of temporary repair techniques were provided by Thermal and Nuclear Power Engineering Society (TENPES) at first in 1986, about thirty years ago. “Plant Operation and Maintenance Standards (POMS) Plan” developed by Japan Power Engineering and Inspection Corporation (JAPEIC) in 1996 incorporated those rules of temporary repair techniques, then JSME FFS Code has taken over them from POMS. Because the design rules of these temporary repair techniques have had much margin since origin, it may result in excessive design, or time-consuming procurement of parts. Especially, since these temporary repair techniques are often applied to the leakage around valve gland and flange, simpler and more practical modified rules could provide more benefit for effective repair activities. In this paper, an orientation of possible revision is described on temporary repair techniques rules in JSME FFS Code.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A061. doi:10.1115/PVP2015-45924.

Intelligent ECT (Eddy Current Testing) probe (I-Probe) has a capability to inspect tubes more productively and more accurately than conventional probes for Steam Generators (SG) as shown in Fig. 1. And I-Probe provides effective “One-Pass” solution by combination with MHI (Mitsubishi Heavy Industries, Ltd.) Intelligent Data Analysis System (MIDAS) [1]. It has been applied to full length tube inspection in Japan since 2003 and there is much field experience to date.

On the other hand, a phenomenon in SG secondary side is known that scale deposits clog holes of quatrefoil type tube support plate (TSP). Water level oscillation of secondary side caused by excessive blockage of scale deposits may decrease operation efficiency and it can be a cause of unintended troubles. Therefore, periodical monitoring of blockage condition and applying countermeasures like secondary side cleaning are very important. A new method has been developed to evaluate blockage condition of every TSP around all tubes automatically using I-Probe data. Also, a new software has been developed to visualize the evaluation results as color maps.

This new method may contribute to increase the reliability of SG maintenance like cleaning, because the new method makes it possible to monitor a whole SG condition more precisely.

Topics: Maintenance
Commentary by Dr. Valentin Fuster

Codes and Standards: Risk-Informed Methods for Structural Integrity Assessment (Joint With M&F)

2015;():V01BT01A062. doi:10.1115/PVP2015-45746.

Codes and standards are often filled with deterministic equations and relationships that are presented without explicit quantification of the uncertainty inherent in their application or derivation. That is not to say that such equations and relationships are not conservative, but to the contrary, that whatever conservatism is dutifully included sometimes goes unquantified. The gamut of Failure Assessment Diagrams (FAD) used for assessing the risk of crack-like flaws provides an excellent example of useful criteria that may benefit from uncertainty quantification. The Level 2 FAD from API 579-1/ASME FFS-1 is used as an exemplar in the probabilistic extension of existing codes and standards. As a first step toward modeling the true probability of failure for assessment points nominally below the FAD envelope, assumptions are made regarding the variance of operating conditions, geometry, and flaw sizes, and these assumptions are utilized to estimate the probability that a nominal assessment point will in fact correspond to conditions exceeding the FAD envelope. This estimation procedure is used to generate contour lines of constant probability within the existing FAD envelope, providing an easily interpretable visualization of the results that may be coupled with risk matrices or other risk-based inspection methodologies.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A063. doi:10.1115/PVP2015-45997.

CANDU(1) Zr-2.5%Nb pressure tubes are susceptible to formation of hydrided regions at the locations of stress concentration, such as in-service flaws. Hydrided region overloads occur when the applied stress acting on a flaw with an existing hydrided region exceeds the stress at which the hydrided region has been formed. The overload events may potentially result in crack initiation and its subsequent growth by the mechanism of delayed hydride cracking. Therefore, evaluating the in-service flaws in the pressure tubes for crack initiation due to hydrided region overloads is required by the Canadian Nuclear Standards, and methodology is being developed to perform such evaluations. As part of this development, the resistance of pressure tube material to crack initiation due to hydrided region overloads was modeled statistically. In the proposed modeling framework, the overload resistance is expressed as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading. This approach fundamentally relies on the concept of a dual process zone introduced by E. Smith, as discussed in the paper. Both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. The overload crack initiation coefficient also varies with the extent of stress reduction prior to hydride formation and with the number of non-ratcheting hydride formation thermal cycles. The developed model is suitable for use as a predictive model in probabilistic assessments of CANDU reactor core, and has been proposed for implementation into the scheduled revision (2015) of the Canadian Nuclear Standard CSA N285.8.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A064. doi:10.1115/PVP2015-46010.

A fracture mechanics based calculation procedure is provided in the CSA Standard N285.8 to define pressure-temperature limits for fracture protection of CANDU Zr-Nb pressure tubes. The calculated pressure-temperature limits are used to construct a plant-specific operating envelope for pressure tubes under reactor heat-up and cool-down conditions. The current calculation procedure to define pressure-temperature limits for fracture protection is deterministic, and makes use of conservative inputs, including an axial through-wall crack of 20 mm in length, lower-bound fracture toughness, and a safety factor of 1.3 on the calculated critical internal pressure. The deterministic procedure is straightforward to use, and has a long history of successful applications of protecting pressure tubes from rupture. However, the deterministic procedure will potentially impose challenging operational constraints on pressure tubes at late life conditions, due to the predicted low fracture toughness of pressure tubes with high levels of hydrogen equivalent concentration.

As an alternative, the CSA Standard N285.8 also permits probabilistic evaluation of fracture protection, which implies the acceptability of using risk-informed pressure-temperature limits for pressure tubes under reactor heat-up and cool-down conditions. The feasibility of developing a risk-informed procedure to define pressure-temperature limits for fracture protection of pressure tubes under heat-up and cool-down conditions is described in this paper. The intent is to use the risk-informed methodology to develop alternate pressure-temperature limits that allow more operational flexibility and still satisfy safety goals. The proposed risk-informed approach is consistent with risk-informed approaches that have been used in the U.S. nuclear industry.

Commentary by Dr. Valentin Fuster

Codes and Standards: Structural Integrity of Pressure Components

2015;():V01BT01A065. doi:10.1115/PVP2015-45002.

The performance of regular inspections of pipe supports and hangers at refineries, power plants and other industrial sites is often overlooked or ignored. The API, ASME and other Codes and Standards have long included recommended or required practices to perform regular monitoring of pipe supports and restraints, especially as it affects the condition of the piping. This is true not only of existing plants but also of new plants. The condition of pipe supports and restraints is an external barometer of hidden problems with the piping and attached equipment. Recognizing pipe support and restraint distress can help prioritize pipe inspections and equipment maintenance. This paper reviews and identifies applicable sections of the various Codes, Standards and Recommended Practices to provide the reader with a source for such information.

Topics: Inspection , Pipes
Commentary by Dr. Valentin Fuster
2015;():V01BT01A066. doi:10.1115/PVP2015-45564.

Clause 4.5 of ASME Section VIII Div. 2 [1] provides rules for compensation of openings in cylindrical shells having fitted nozzles.

There appears to be no definition of “nozzle” in either ASME Section VIII Div. 2 or ASME Section VIII Div. 1 [2].

The rules provided in Clause 4.5.5 of ASME Section VIII Div. 2 are based on pressure area method which ensures that the reactive force provided by the vessel material is greater than or equal to the load from the pressure.

The key element in applying this method is to determine the dimensions of the reinforcing zone, i.e, the length of the shell, height of the nozzle and reinforcing pad dimensions (if reinforcing pad is provided), that resist the applied pressure.

There appears to be no restriction on the physical dimensions of the nozzle or shell, so long as the required area AT is obtained and the stresses are within allowable limits.

It is therefore possible that all of the required area AT is obtained from the nozzle or from the shell. While both these alternatives would be acceptable, the actual stresses at the shell/nozzle junction may vary considerably.

The work reported in this paper was undertaken with a view to determining if there needs to be any restriction on the proportion of area contributed by shell or nozzle to ensure that actual stresses were within allowable limits.

Topics: Dimensions , Stress , Nozzles , Pipes
Commentary by Dr. Valentin Fuster
2015;():V01BT01A067. doi:10.1115/PVP2015-46006.

The purpose of this study is to evaluate the design verification of the welded type 45° lateral tee for the steam pipe in power plants. For it, first, the stress analysis was carried out under design condition in accordance with ASME Sec. VIII Div. 2 in order to evaluate the possible occurrence of plastic collapse and local failure. And next, the creep-fatigue damage analysis was performed under the normal operating condition in accordance with ASME Sec. III Subsection NH considering the service temperature of 566°C. From the results, it was found that the welded type 45° lateral tee satisfies the design criteria corresponding to the plastic collapse and the local failure. However, it has a probability of creep rupture during the design life due to the high stress localized in the crotch region. Therefore, a welded type 90° lateral tee was also evaluated with the same analysis procedures to consider the influence of the geometry at the crotch region. Based on the results, the welded type 90° lateral tee satisfies the design criteria of the plastic collapse, local failure and the creep-fatigue strength. This result indicated that an optimal shape design of the crotch region shall be required in order to secure the creep strength of the welded type 45° lateral tee having high service temperature.

Commentary by Dr. Valentin Fuster
2015;():V01BT01A068. doi:10.1115/PVP2015-46012.

Rivets are widely used as a means of fastening in airframe construction industry. Among the other types of fasteners riveted joints are preferred in such applications due to their permanence after installation and their economical advantages. In a riveted joint, it is known that residual stresses are present as a result of the installation process. Furthermore, during the flight of an aircraft, the fuselage comes across pressurization and depressurization cycle. During one flight pressurization-depressurization cycle is completed and such cycles are repeated throughout the service life of the aircraft. As a result, the panels and the rivets are subjected to fatigue type loading. The integrity of the joint must be maintained against this combination of service loads and the residual stresses. The present study is aimed to develop and analyze three-dimensional finite element model of riveted lap, and then the numerical analysis (SolidWorks Simulation) are carried out to calculate the residual stress values and fatigue values in the riveted lap joint under the effect of varying temperature. The result shows that the fatigue life varies inversely proportion to residual stresses whereas damage varies directly proportion to residual stresses. The maximum residual stress obtained is 292 MPa at temperature of 150°C and the minimum residual stress obtained is 15 MPa at temperature of −50°C. Maximum damage is 60% at 150°C and minimum is 8% at −50°C. Maximum life is 234346 cycles at −50°C and minimum life is 33111 cycles at 150°C.

Commentary by Dr. Valentin Fuster

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