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ASME Conference Presenter Attendance Policy and Archival Proceedings

2014;():V001T00A001. doi:10.1115/PVP2014-NS1.
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This online compilation of papers from the ASME 2014 Pressure Vessels and Piping Conference (PVP2014) 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to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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: API 579/ASME Code Fitness-for-Service Activities

2014;():V001T01A001. doi:10.1115/PVP2014-28263.

Guidance for weld residual stress (WRS) in flaw assessments of in-service equipment is given in Annex E of the Fitness-for-Service (FFS) standard, API 579-1/ASME FFS-1. Recent progress in the simulation of weld residual stress, and in its verification and validation has provided the opportunity to both simplify and supplement Annex E. Specifically, a tiered approach is recommended with the baseline guidance simplified and harmonized with other existing guidance, while more complicated cases are recommended to be handled with an explicitly described simulation method. The simulation method is the subject of related papers, while this paper focuses on comparison and evaluation of existing guidance for the common cases of cylinder girth and seam welds. A complete draft of recommended guidance has been completed, and intended for publication in the 2014 revision of API 579-1/ASME FFS-1. A supporting basis document which details the underlying methodologies for recommendations has also been completed.

Commentary by Dr. Valentin Fuster
2014;():V001T01A002. doi:10.1115/PVP2014-28451.

The first edition of API 579 Recommended Practice for Fitness-For-Service was published in 2000, and subsequently recognized as the de facto international fitness-for-Service standard in the refining and petrochemical industry. The second edition of this document, API 579-1/ASME FFS-1 Fitness-For-Service, was published in 2007 as a joint standard of the American Petroleum Institute (API) and the American Society of Mechanical Engineers (ASME). The second edition included fitness-for-service assessment procedures applicable to other industries including fossil utility and pulp and paper. Work on the third edition of API 579-1/ASME FFS-1 has begun with many planned technical improvements to further address industry needs. These improvements include the edition of a new part on fatigue evaluation, updates to the assessment procedures for crack-like flaws and remaining life assessments for components operating at elevated temperatures, and a rewrite of residual stress solutions for use in the evaluation of crack-like flaws based on the latest state-of-the-art approaches. In addition, the third edition will be reorganized where by technical information currently placed in separate annexes that currently appear after all of the parts will be re-deployed as annexes to specific parts with a similar topic. This new organization will facilitate use and also simplify future updates to the document. An overview of proposed improvements to fitness-for-service technologies is provided along with a description of the new organization of API 579-1/ASME FFS-1.

Commentary by Dr. Valentin Fuster
2014;():V001T01A003. doi:10.1115/PVP2014-28839.

Recent progress in the simulation of weld residual stress (WRS), and in its verification and validation (V&V) has provided the opportunity to re-examine existing WRS guidance (Annex E of the Fitness-for-Service (FFS) standard, API 579-1/ASME FFS-1). Comprehensive review of existing guidance and corresponding basis documents, supplemented by validated weld residual stress analysis are utilized to assess the appropriateness of the existing Annex E, and to develop proposals for alternate approaches. The simulation techniques used to support the proposed guidance are described in detail here. The European NeT Program provides an important part of the V&V of all results, and is the focus of this paper. Results using the presented methods show excellent agreement with measurement, therefore providing a sound basis for supplemental analysis to extend available experimental results as necessary. The methods also provide a baseline from which to assess future modeling simplifications for special cases.

Commentary by Dr. Valentin Fuster
2014;():V001T01A004. doi:10.1115/PVP2014-28843.

Modeling of cyclic elastic-plastic material behavior (hardening) has been widely identified as a critical factor in the finite element (FE) simulation of weld residual stresses. The European Network on Neutron Techniques Standardization for Structural Integrity (NeT) Project has provided in recent years both standard test cases for simulation and measurement, as well as comprehensive material characterization. This has allowed the role of hardening in simulation predictions to be isolated and critically evaluated as never before possible. The material testing information is reviewed, and isotropic, nonlinear kinematic and combined hardening models are formulated and tested. Particular emphasis is placed on material model selection for general fitness-for-service assessments, as it relates to the guidance for weld residual stress (WRS) in flaw assessments of in-service equipment in Annex E of the FFS standard, API 579-1/ASME FFS-1.

Topics: Hardening
Commentary by Dr. Valentin Fuster

Codes and Standards: ASME Section III Activities for Codes and Standards

2014;():V001T01A005. doi:10.1115/PVP2014-28301.

To conduct an ASME III NG-3200 limit load strength assessment, it is required to determine the structure’s limit load under a particular loading configuration, and compare it against the applied loading represented as a static equivalent. Typically, the process is applied to static problems which have well-defined loading characteristics. When the limit load has been determined, often through the use of finite element (FE) based methods, the margin against plastic collapse is simple to calculate. For dynamically loaded structures, however, the process is more complicated since there are no ASME guidelines for expressing dynamic loads as their static equivalent. Thus, relating limit load analyses to dynamic events is not clear. This paper proposes an analysis technique which makes use of FE methods to apply the principles of limit load analysis to dynamically loaded structures. The primary benefit is that reserve factors against plastic collapse, in accordance with ASME III NG-3200 assessment criteria, can be calculated.

Commentary by Dr. Valentin Fuster
2014;():V001T01A006. doi:10.1115/PVP2014-28307.

In the ductile–to–brittle transition (DBT) temperature region, the specimen thickness B and fracture toughness Jc are correlated (JcB−1/2) [1, 2]; this is true even for standardized specimens. This has been described as the test specimen thickness (TST) effect based on the weakest link model [3]; however, it has been pointed out that Jc does not decrease indefinitely as B→∞, which contradicts this effect. This paper is an extension of our recent work [4] on the point that the TST effect on Jc. observed for the Shoreham reactor vessel steel (A533 Grade B) [5] was explained via the difference in crack tip constraint and demonstrating that the (4δt, σ22c) criterion can be applied to explain the effect.

Commentary by Dr. Valentin Fuster
2014;():V001T01A007. doi:10.1115/PVP2014-28352.

The current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C) because, at higher temperatures, it is not feasible to decouple plasticity and creep; which is the basis for the current simplified rules. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic analysis methods and which are expected to be applicable to very high temperatures. The proposed rules are based on the use of an elastic-perfectly plastic material model with a pseudo yield strength selected to ensure that the accumulated strain and creep-fatigue damage with meeting the currently specified limits in Subsection NH. For this phase of the verification process, the proposed rules have been compared using simplified example problems to the results obtained from application of the current Subjection NH rules for both simplified methods and full inelastic analysis. The Subsection NH 316 stainless steel properties data are used for these comparisons. Results of calculations for a testing program underway on Alloy 617 at 950C are given.

Commentary by Dr. Valentin Fuster
2014;():V001T01A008. doi:10.1115/PVP2014-28412.

Advanced materials are a critical element in the development of advanced sodium-cooled fast reactors. High temperature design methodology of advanced materials is an enabling reactor technology. Removal of unnecessary conservatism in design rules could lead to more flexibility in construction and operation of advanced sodium-cooled fast reactors. Developing mechanistic understanding and predictive models for long-term degradation phenomena such as creep-fatigue are essential to the extrapolation of accelerated laboratory data to reactor environments with high confidence, and to improve the American Society of Mechanical Engineers (ASME) code rules. This paper examines the cyclic softening and stress relaxation responses and associated plastic damage accumulation for Grade 91 ferritic-martensitic steel. Creep-fatigue experiments were conducted at 550°C in strain-controlled mode under various types of creep-fatigue loading conditions. Constitutive models were developed to describe the creep-fatigue interaction in G91.

Commentary by Dr. Valentin Fuster
2014;():V001T01A009. doi:10.1115/PVP2014-28444.

Alloy 617 has been selected as a reference material supporting the Very High Temperature Gas Cooled Reactor (VHTR). However, current simplified design methods in Subsection NH have been deemed inapplicable at very high temperatures because, at these conditions, it is not possible to decouple plasticity and creep which is the basis for the current methods. Also, the alternative use of inelastic analysis requires development and verification of material modeling at these very high temperatures. A test procedure has been developed and implemented to support verification of new simplified methods and material modeling of Alloy 617 at very high temperatures. The procedure is based on two bars tested in series using two coupled servo-controlled testing machines to achieve equal displacement and constant applied load, mimicking the behavior of a pressurized cylinder subjected to through wall thermal transients. The tests were conducted with a hold time at 950°C. The bars were heated and cooled out of phase to generated thermal induced loading superimposed on a constant mean stress. The results are presented for different mean stress levels, heating and cooling rates, and thermal histories.

Topics: Alloys
Commentary by Dr. Valentin Fuster
2014;():V001T01A010. doi:10.1115/PVP2014-28788.

An ASME Section III Task Group (TG) was formed in 2012 to develop alternate rules for the design assessment of Section III Class 1 nuclear components subject to fatigue service with environmental effects. Specifically, a flaw tolerance approach is being investigated based on similar methodology to that found in ASME Section XI Nonmandatory Appendix L. A key initial task of the TG (which reports to the Section III Working Group on Environmental Fatigue Evaluation Methods) was to develop and solve a detailed sample problem. The intent of the sample problem was to illustrate application of proposed rules, which will be documented as a Section III Code Case with a supporting technical basis document. Insights gained from round robin solution of the sample problem are presented and discussed in this paper. The objective of documenting the findings from the sample problem are to highlight the observed benefits and limitations of the proposed procedures, particularly how rules typically associated with in-service experience might be adapted into design methods.

The sample problem is based on a heavy-walled stainless steel nozzle that meets cumulative fatigue usage requirements in air (i.e., usage factor, U, without reactor water environment effects less than unity), but fails to meet usage factor requirements when environmental fatigue effects are applied.

The sample problem demonstrates that there is a class of problems dominated by severe thermal transients where fatigue initiation is predicted based on elastic methods including environmental effects, but fatigue crack propagation results are acceptable. Preliminary conclusions are drawn based on the results of the sample problem, and the next steps are also identified.

Commentary by Dr. Valentin Fuster

Codes and Standards: ASME Section XI Code Activities

2014;():V001T01A011. doi:10.1115/PVP2014-28049.

The stress intensity factor (SIF) solutions for circumferential through-wall cracks (TWCs) in cylinders are used for various fracture mechanics analyses. For example, it can be used to calculate the crack growth rate for stress corrosion cracking and to calculate the elastic J value which is needed to obtain the total J value for crack stability calculations. Thus, numerous SIF solutions have been published for circumferential TWCs in cylinders under axial tension and global bending. However, recently, it has been indicated that there is a need (e.g., for xLPR software code and ASME BPV Code Case N-513) to expand the solutions to wider ranges of crack lengths and cylinder geometries.

In this paper, solutions from Lacire et al., API 579-1/ASME FFS-1 and Zang (SINTAP) were compared against results from independent finite element (FE) analyses performed by the authors. From these comparisons, it was demonstrated that the Zang (SINTAP) solution provided the most accurate results. Hence, additional FE calculations were performed to expand the Zang (SINTAP) solution to cover Ri/t between 2 and 100 and crack length between 1% and 85% of the cylinder circumference. Furthermore, for practical applications, closed-formed solutions were developed for both axial tension and global bending loads. These new solutions were planned for use in the xLPR software code and ASME BPV Code.

Commentary by Dr. Valentin Fuster
2014;():V001T01A012. doi:10.1115/PVP2014-28222.

Analytical evaluation procedures for determining the acceptability of flaws detected during in-service inspection of nuclear power plant components are provided in Section XI of the ASME Boiler and Pressure Vessel Code. Linear elastic fracture mechanics based evaluation procedures in ASME Section XI require calculation of the stress intensity factor. In Article A-3000 of Appendix A of the 2013 Edition of ASME Section XI, the calculation of stress intensity factor for a surface crack is based on characterization of stress field with a cubic equation and use of stress intensity factor influence coefficients. The influence coefficients are only provided for a flat plate geometry.

The ASME Section XI Working Group on Flaw Evaluation is in the process of rewriting Article A-3000 of Appendix A. Major updates include the implementation of an alternate method for calculation of the stress intensity factor for a surface flaw that makes explicit use of the Universal Weight Function Method and does not require a polynomial fit to the actual stress distribution, and the inclusion of stress intensity factor influence coefficients for the cylinder geometry. Tabular data of influence coefficients for the cylinder geometry are available in API 579-1/ASME FFS-1 2007. Effort has been made to develop closed-form relations for the stress intensity factor influence coefficients for the cylinder geometry based on API data. With the inclusion of the explicit weight function approach and the closed-form relations for influence coefficients, the procedures of Appendix A for the calculation of stress intensity factors can be used more efficiently. The development of closed-form relations for stress intensity factor influence coefficients for axial ID surface flaws in cylinders is described in this paper.

Topics: Stress , Cylinders
Commentary by Dr. Valentin Fuster
2014;():V001T01A013. doi:10.1115/PVP2014-28355.

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been three revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the three revisions have been previously published.

There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fourth revision. These changes include addressing the current condition on the Code Case acceptance by the US Nuclear Regulatory Commission (NRC), clarification of the Code Case applicability limits and expansion of Code Case scope to additional piping components. New flaw evaluation procedures are given for through-wall flaws in elbows, bent pipe, reducers, expanders and branch tees. These procedures evaluate flaws in the piping components as if in straight pipe by adjusting hoop and axial stresses to account for the geometry differences. These changes and their technical bases are described in this paper.

Commentary by Dr. Valentin Fuster
2014;():V001T01A014. doi:10.1115/PVP2014-28397.

A previous PVP paper [1] identified suggested improvements to be made to ASME Code, Section XI, Nonmandatory Appendix G, “Fracture Toughness Criteria for Protection Against Failure” [2]. That paper also identified that the current version of Appendix G does not have any provisions for when the calculated operating stress (pseudo stress) exceeds the material yield strength. The treatment of stresses exceeding yield was included in earlier versions of Appendix G, but it was removed via Code Action ISI-94-40 in 1995. The specific reasons for removal of these provisions were not documented.

In some Appendix G postulated flaw evaluations for pressure-temperature (P-T) limits, the calculated total linear-elastic (or pseudo) stress (i.e., including the primary stress due to pressure loading and thermal stress) may exceed the material yield stress. The ASME Section XI Working Group on Operating Plant Criteria (WGOPC) decided that this provision needed to be more fully considered, with appropriate benchmarking and possible adjustments to Appendix G made consistent with the current state of knowledge in elastic-plastic fracture mechanics (EPFM) methods. This is appropriate since the state of knowledge in EPFM has significantly advanced since the time the technical basis for Appendix G was established, as documented in Welding Research Council (WRC) Bulletin No. WRC-175, which was published in 1972. Furthermore, EPFM provides an improved method for evaluating the effects of high stresses.

This paper describes the results of preliminary investigations of stresses exceeding the material yield stress in fracture toughness assessments associated with Appendix G. Also included in the technical evaluations presented are the temperature conditions for which upper shelf conditions are present and where EPFM methods are applicable.

Commentary by Dr. Valentin Fuster
2014;():V001T01A015. doi:10.1115/PVP2014-29111.

During routine inspection of an N5 feedwater nozzle dissimilar metal weld (DMW) of a BWR, an axial flaw was detected in the weld. The flaw was mitigated by applying a full structural weld overlay (WOL) repair consisting of Alloy 52M weld metal and using the gas tungsten arc welding process (GTAW). Because of the ductile nature and very high toughness of the Alloy 52M material, the limit load approach of ASME Code, Section XI, Appendix C was used for the sizing of the overlay. The use of limit load for the design of weld overlays for Alloy 52M material is supported by several studies documented in the industry [1].

In order to address other fracture failure modes such as failure by ductile tearing and brittle fracture, a failure assessment diagram (FAD) approach was used to evaluate the acceptability of the weld overlay design. FADs have been used for evaluating flaws in piping components, but not for acceptance of flaws in WOLs. The FAD evaluation was performed in accordance with the ASME Code, Section XI, Appendix H requirements which address brittle fracture, elastic-plastic fracture mechanics (ductile tearing), and limit load failure modes. The applicable stress combinations were used in combination with the materials JR resistance curve to determine flaw acceptability at the end of the evaluation period.

The evaluation considered the presence of both a circumferential flaw (360° around the circumference and 100% through the original pipe wall; the design basis for the full structural weld overlay) and an axial flaw. For both circumferential and axial flaws, several assessment points corresponding to the JR resistance curve were determined and plotted on the FAD curve for austenitic steels in ASME Section XI, Appendix H. The FAD curve in Appendix H was derived based on strength properties of stainless steel and is considered conservative in application to Alloy 52M since Alloy 52M has higher strength. All the assessment points were found to be below the FAD curve thus indicating the acceptability of the weld overlay with consideration to all the three possible fracture regimes.

Commentary by Dr. Valentin Fuster

Codes and Standards: Combination of Multi-Axial Loads in Design and Fitness-for-Service

2014;():V001T01A016. doi:10.1115/PVP2014-28141.

Pipe failure under internal pressure loading is understood as due to the onset of plastic instability. Law et al. [1] extended the Considere’s construction [2], which predicts the onset of plastic instability for tensile test specimen, so that it can predict the onset of plastic instability, and thus burst of pipes under internal pressure. They noted that “It is important to note that the reduced stress and strain where instability occurs is not a result of the biaxial stress state, but of the vessel geometry where increased stress comes from both increased inner diameter and reduced wall thickness” [1]. However, they intrinsically assumed that the cylinder plane that experiences instability is in plane strain state. Considering the fact that power plant pipes are often subjected to axial force from sources such as thermal expansion/contraction or fixing, the effect of deviation from the plane strain conditions should be considered.

For this purpose, the onset of plastic instability of the elastic-plastic internally pressurized 1) 2D plane strain pipe, 2) 3D pipe and 3) 3D pipe additionally subjected to axial force, were compared. The results showed that the onset of plastic instability could be monitored at the pressure when pRm−σθt > 0 for all the cases. Here, Rm, σθ, and t are the current mean radius, circumferential stress at mean radius and thickness at pressure p. However, the strain at this instability pressure showed non-negligible change due to presence of axial loading. On this meaning, the biaxiality affected the onset of plastic instability.

Topics: Stress , Pipes
Commentary by Dr. Valentin Fuster
2014;():V001T01A017. doi:10.1115/PVP2014-28288.

Piping components in power plants may experience combined bending and torsion moments during operation. There is a lack of guidance for pipe evaluation for pipes with local wall thinning flaws under the combined bending and torsion moments. ASME B&PV Code Section XI Working Group is currently developing fully plastic bending pipe evaluation procedures for pressurized piping components containing local wall thinning subjected to combined torsion and bending moments. Using elastic fully plastic finite element analyses, plastic collapse bending moments under torsions were obtained for 4 (114.3) to 24 (609.6) inch (mm) diameter pipes with various local wall thinning flaw sizes. The objective of this paper is to introduce an equivalent moment, which combines torsion and bending moments by a vector summation, and to establish the applicable range of wall thinning lengths, angles and depths, where the equivalent moments are equal to pure bending moments.

Topics: Torsion , Pipes , Collapse
Commentary by Dr. Valentin Fuster
2014;():V001T01A018. doi:10.1115/PVP2014-28791.

Recent studies on the effect of multi-axis loading on piping non-planar flaws indicate that the ASME Section XI, Appendix C pressure limit load approach for planar axial flaws can be used for evaluation of non-planar flaws provided that the effect of multi axial loading is to be considered. Finite element analysis results presented in this paper also indicate that bending limit load for planar circumferential flaws can be used for evaluation of non-planar flaws subjected multi axial loading. These studies used idealized uniformly thin rectangular flaws whose projected flaw geometry on pipe cross section and on pipe axial section are the same as the circumferential flaw geometry and the axial flaw geometry defined ASME B&PV, Section XI Appendix C. Sample problems with actual pipe wall thinning flaws due to flow accelerated corrosion and pitting in nuclear power plant are utilized for a comparison of the proposed methodology with various methodologies currently used by the industry for locally thinning pipe flaw evaluations.

Topics: Pipes
Commentary by Dr. Valentin Fuster

Codes and Standards: Environmental Fatigue Issues (Joint M&F)

2014;():V001T01A019. doi:10.1115/PVP2014-28066.

The first results of a detailed fatigue model for austenitic stainless steels in general and for the grades 1.4541 and 1.4550 are presented to describe the effect of the light water reactor (LWR) coolant environments on the fatigue life. The statistical evaluations are based on strain (and load) controlled test series from different institutions. The compiled fatigue data include not only results from America (Keller (1971), Conway (1975), Hale (1977), and Argonne National Laboratory (ANL)(1999–2005)), but also from Europe (Solin (2006), Le Duff (2008–2010), De Baglion (2011, 2012), Huin (2013),…) and Japan (Kanasaki (1997)). The fatigue life is defined as the number of cycles necessary for tensile stress to drop 25 percent from its peak value. Fatigue lives defined by other failure criteria are normalized to the load reduction of 25 percent, before the statistical analysis is performed. The fatigue data are expressed in terms of the Langer equation and the parameter “material variability and data scatter” is quantified.

Additionally, fatigue data in air of roughened specimens are compiled and discussed. A reduction factor of 2.5 on number of cycles is derived to cover the maximum allowed surface roughness.

Based on the derived best-fit curves, design-curves in air and, in a second step, environmentally assisted fatigue (EAF) curves for LWR environments, which consider temperature, strain rate, dissolved oxygen content, and hold-time effects, will be incorporated in the detailed fatigue model in the future.

Commentary by Dr. Valentin Fuster
2014;():V001T01A020. doi:10.1115/PVP2014-28098.

When defects are detected by in-service inspections of the nuclear power plants in Japan, allowable flaws are evaluated according to the Rules on Fitness-for-Service for Nuclear Power Plants of Japan Society of Mechanical Engineers (JSME maintenance rules). The fatigue crack growth analysis is an important part of the flaw acceptance evaluation. However FCGR curves for nickel base alloys in air are not provided in the current JSME maintenance rules.

This paper describes the evaluation of the fatigue tests in air of nickel base alloys and formulation of the FCGR curves for the JSME rules on Fitness-for-Service. From the test results, the temperature and stress ratio are not dominant factors. However, as based on the form of the FCGR curve of NUREG/CR-6721, the proposed curves are shown to be functions of these parameters. Threshold of stress intensity factor is introduced in the curves by taking into account crack closure.

Commentary by Dr. Valentin Fuster
2014;():V001T01A021. doi:10.1115/PVP2014-28190.

The influence of LWR coolant environment to the lifetime of materials has been discussed recent years. Nowadays the environmentally assisted fatigue (EAF) is under consideration in Codes and Standards like ASME by formula calculation and in the German KTA Rules (e.g. KTA No. 3201.2 and 3201.4) in addition by means of so called attention thresholds.

Basic calculation procedures in terms of quantifying the influence of LWR coolant environment by the so called Fen-factor were proposed in NUREG CR-6909 in 2007. Since this report the set of formulas have been adapted several times (e.g. in ANL-LWRS47-2011) and further changes in set of formulas are likely to occur. Subsequently a revision of the NUREG CR-6909 report is taking place incorporating recent changes.

Within the EPRI 2012 Technical Report “Guidelines for Addressing Environmental Effects in Fatigue Usage Calculations” some practical recommendations for the application of the EAF to real components are stated. In this report EPRI presented the calculation of sample problems in combination with sample transients carried out by the participants and verifying the applicability of the proposed approaches. However, the EPRI guideline provides a set of tools taking environmentally assisted fatigue into account. This guidance figures out to be quite helpful on one hand, but incorporates some challenges in terms of practical application on the other hand. Additionally, the EPRI Report gives no clear advice which specific combination of calculation methods to apply finally.

Within this publication the procedures proposed in the EPRI guideline will be applied to existing numerical approaches being published. Challenges when applying the methods defined in the EPRI guideline will be identified and solutions will be given. Additionally, recommendations will be stated based on the proposed procedures in terms of practical application especially in context of calculations for fatigue relevant primary circuit components.

Commentary by Dr. Valentin Fuster
2014;():V001T01A022. doi:10.1115/PVP2014-28207.

Environmentally assisted fatigue (EAF) is at the centre of numerous discussions and conferences on fatigue around the world. The methodology proposed in NUREG/CR-6909 [1] is being applied in the USA for new builds and for license renewal per the RG 1.207 [2], but there is still room for a better understanding of all the effects involved and for fine-tuning of the methodology. A roadmap by EPRI lists areas where testing is recommended [3].

In Europe, research is going on and the accumulated results show that NUREG/CR-6909 is very conservative [4]–[10]. Applicability of the proposed design curve and the Fen factors were questioned in previous papers and alternative approaches have been developed [11]–[14].

This paper summarizes different experimental, analytical and practical approaches taken by four main players in the European nuclear industry and compares results with the NUREG/CR-6909 methodology. This joint discussion demonstrates that, although the approaches have been different, a common conclusion is reached: the NUREG/CR-6909 methodology can be improved.

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2014;():V001T01A023. doi:10.1115/PVP2014-28300.

Environmentally assisted fatigue (EAF) evaluations for Class 1 components of an advanced PWR have been performed to comply with Regulatory Guide 1.207 and NUREG/CR-6909. The procedure consists of 3 steps. Step 1 is to use the maximum Fen values, Step 2 is to use the Modified Rate Approach and Step 3 is to use elastic-plastic Finite Element (FE) analysis with the Modified Rate Approach. This procedure can efficiently proceed with the evaluation. Also, the Class 1 components of an advanced PWR have satisfied with the code qualification requirement for EAF evaluation required by RG 1.207.

Commentary by Dr. Valentin Fuster
2014;():V001T01A024. doi:10.1115/PVP2014-28320.

There has been much focus in recent years on development of an empirical understanding of the effects of a range of variables on fatigue performance of nuclear plant materials in relevant operating environments — for example, the influences of surface finish, temperature, and cyclic waveform. To complement these empirical studies, a well-developed mechanistic understanding of Environmental Fatigue initiation behaviour is required to allow extrapolation of results beyond the variable ranges tested and to provide insight into the margins incorporated into fatigue analysis methods, including fatigue design curves.

Mechanistic understanding may be defined as a multi-scale description of physical and chemical effects determining fatigue behaviour through the stages of crack nucleation, small crack (Stage I) growth and the transition to long crack (Stage II) growth.

A review of literature related to mechanistic understanding of Environmental Fatigue initiation in nuclear plant materials in air and PWR primary coolant environments has been carried out. Detailed findings of that review are not presented here; however, high level conclusions are discussed. It was found that very little literature specific to these combinations of material and environment is available, although there exists a substantial body of work related to generic mechanistic understanding of fatigue and environmental effects in metallic materials.

Based on the conclusions of this literature review, a strategy for improvement of mechanistic understanding of Environmental Fatigue initiation has been developed by Rolls-Royce. This strategy is based around three strands of work: metallurgical investigation; numerical modelling; and laboratory testing.

Improvements in mechanistic understanding of Environmental Fatigue initiation offer significant opportunities for collaboration between interested parties. Indeed, attainment of a well-developed mechanistic understanding that may be used, in combination with empirical understanding, to inform advances in fatigue analysis methods is likely to be achievable only through collaborative effort.

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2014;():V001T01A025. doi:10.1115/PVP2014-28329.

In most industries and more particularly in the Nuclear one, the use of new materials or new manufacturing methods is really challenging because “safety” is still the key point. Among many characterizations required for a new steel grade or a new manufacturing process to be accepted and then introduced in a Nuclear design code, the fatigue properties must be determined with great care.

Nowadays, the consideration of the Pressurized Water Reactor (PWR) primary water environment effect on the Low Cycle Fatigue (LCF) behavior of Austenitic Stainless Steels is an important issue for both Nuclear Power Plants (NPP) lifetime extensions and new builds as described in NUREG/CR-6909 [1].

This study aims to present the LCF behavior in Air and in PWR water at 300°C of a type 304L steel manufactured by Powder Metallurgy coupled with Hot Isostatic Pressing process (PM/HIP) and to compare them with those observed on 304L nuclear grade products such as rolled plate or forged branch [2–5]. It appears that the LCF behavior in Air and in PWR water of this 304L HIP material is better or at least similar to the one observed on classical 304L steels.

Commentary by Dr. Valentin Fuster
2014;():V001T01A026. doi:10.1115/PVP2014-28408.

Environmentally Assisted Fatigue is receiving nowadays an increased level of attention for new builds and also for installed bases which are currently having their lives extended to 60 years in various countries. To formally integrate these effects, some international codes have already proposed code cases. More particularly, the ASME code has based itself on the NUREG/CR-6909 [1] to elaborate the Code Case N-792 [2] and suggests a modification of the fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which is to be multiplied by the usual fatigue usage factor.

In France, EDF and AREVA also aim at more explicitly integrating these effects in the RCC-M code. The initiative is technically supported by CEA and bases itself on international methodologies but also on results from French in-house testing campaigns [3] [4]. The approach is globally similar to the one in the ASME code: it will indeed consist in an update of the mean air and design fatigue curves as well as the calculation of an environmental penalty factor. Nevertheless, the methodologies differ in their detailed implementation, as was already hinted in previous papers discussing the French methodology [5] [6]. This paper is the sequel to the proposal already described in [7].

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2014;():V001T01A027. doi:10.1115/PVP2014-28465.

Our experimental research on fatigue performance of stainless steels and transferability of laboratory data to plant operational conditions focuses in niobium stabilized stainless steel (1.4550, X6CrNiNb1810mod) taken from a pipe manufactured as primary piping for a German NPP Good fatigue performance both in air and in PWR water was reported in previous PVP papers.

The NUREG/CR-6909 report proposes Fen factors based on stroke controlled experiments in hot water for non-stabilized steels. Since PVP2013-97500 we have new data in 200°C PWR water to compare with predictions by NUREG/CR-6909. Our strain controlled tests in 325°C and 200°C PWR water give longer lives resp. smaller Fen factors. For the slowest tested strain rate 4·10−6 in 325°C water the prediction according to NUREG/CR-6909 goes just below the current ASME design curve, but our results remains well above. Including also the relevant design temperature effect, our result Fen = 4 is well below the predicted Fen = 14,5. The gap is smaller for higher strain rates and low Fen values.

Simplified simulations of fatigue transients combined with normal operation indicated that relevant loading patterns as hold-time effects may result to notably longer lives than in standard laboratory tests. A concern was raised on transferability of data to thermal transients separated with months of normal operation. Cyclic strain (transients) followed by hot holds (normal operation) lead to time and temperature dependent hardening with reduction in cyclic plastic strain and fatigue usage, i.e. extension of life.

This paper reports new data, challenges met and our progress towards developing realistic design factors for effects both reducing and extending fatigue endurance in nuclear power plant operational conditions.

Commentary by Dr. Valentin Fuster
2014;():V001T01A028. doi:10.1115/PVP2014-28551.

During the past 30 years the main rules for fatigue analysis of pressure vessels were based on elastic approaches in order to evaluate cyclic strain amplitude and compare with an S-N fatigue curve for the corresponding material. After review of some rules in different Nuclear and Non-Nuclear Codes, like ASME Boiler & Pressure Vessel Code Section III, French RCC-M and RCC-MRx, European Standards EN 13445, the major conservatisms and uncertainties of different rules are discussed.

All these Codes propose simple rules to evaluate the strain amplitude based on elastic approaches and simplified correction factors (Ke and Kν), transient combination rules and damage cumulating procedure.

In the other hand, the material properties are based on standard fatigue tests done on the material associated to reduction factors to consider some particular effects like scatter, scale, surface roughness, mean stress or environmental effects to transfer them from small specimen to real structures.

Concerned components in this paper are mainly piping systems.

No existing Code covers all the aspects of fatigue with similar conservatisms that can affect the in-service inspection programs and the remaining life assessment of the corresponding components.

After the review of different rules, key factors that affect the results and predictions will be identified. Some proposals will be issued to progress in the near future.

Finally, a first set of recommendation on fatigue analysis will be presented to improve existing codes on harmonized way, associated to material properties needed, as fatigue curves associated to reduction factors.

Commentary by Dr. Valentin Fuster
2014;():V001T01A029. doi:10.1115/PVP2014-28573.

Published works on studying effects of specimen size on fatigue endurance limit and fatigue life were reviewed. Specimen size effect is apparent in many bending fatigue tests. However, axial tension and compression loading fatigue tests showed that the fatigue limit decreased slightly with increasing specimen size. Design margin for the size effect on fatigue limit under axial tension and compression loading condition that is the condition of the data used to construct Design Fatigue Curve (DFC) is not considered to be necessary. If it is considered conservatively, the value less than 1.1 is enough. Axial tension and compression loading fatigue tests using cylindrical specimens having different diameters showed that the fatigue life increased with increasing specimen diameter. This phenomenon is considered to occur due to the crack growth life which is not negligible in relatively large diameter specimens. Based on this, design margin factor on number of cycles is not necessary to be considered in DFC. These design margins based on the axial tension and compression loading fatigue test results are less than the values in current design codes. This indicates that the current design fatigue curves in the codes have large margins. These considerations will be verified after the large scale fatigue tests that are planned in Design Fatigue Curve (DFC) subcommittee in the Atomic Energy Research Committee in the Japan Welding Engineering Society.

Topics: Fatigue , Design
Commentary by Dr. Valentin Fuster
2014;():V001T01A030. doi:10.1115/PVP2014-28576.

Fatigue evaluation methods have been proposed based on environmental fatigue test results regarding parameters selected for simulating Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) conditions. The effects of strain wave form have been discussed by comparing experimental fatigue life with predicted fatigue life evaluated by modified rate approach (MRA) method. The applicability of the MRA method has been verified extensively by the environmental fatigue tests with strain rate changing conditions consisting of combined constant strain rates. However, different results have been obtained for a sine strain wave in simulated BWR and PWR conditions. More study for evaluating the applicability of MRA method was required by evaluating with continuous strain rate conditions such as a sine wave. For the purpose of verification, two approaches were applied. One is performing the environmental fatigue tests with the sine strain wave in simulated BWR condition. The other is to evaluate the low cycle thermal fatigue test performed in simulated BWR condition because the wave form of this test contains continuous strain rate changing condition. MRA method was indicated to be applicable to predict fatigue lives under these kinds of continuous strain rate changing conditions.

All of the studies including this study verifying the applicability of the MRA method were performed with small specimens having the well polished surfaces in the gage length. These results indicate that the evaluation by the MRA method includes the synergistic effect between the water environment and the transient. However, the synergistic effects with the surface roughness and the component size are not known. Design margin derived by the multiplication of the sub-factors of environment, surface roughness and component size may be conservative. The evaluation of the conservatism is considered to be beneficial.

Commentary by Dr. Valentin Fuster
2014;():V001T01A031. doi:10.1115/PVP2014-28601.

The published papers related to the effects of surface finish on fatigue strength are reviewed in order to formulate its factor in the design fatigue curve in air environment.

Firstly, some of regulations and literatures were examined to verify the surface finish effect on fatigue strength and formulation of that in design fatigue curve.

The fatigue strength of carbon and low alloy steels is decreased with an increase of its surface roughness and tensile strength but that of stainless steel is not decreased except for special conditions.

After screening the data of carbon and low alloy steels, a surface finish factor is formulated with these data which is a function of tensile strength, surface roughness and mean stress.

Commentary by Dr. Valentin Fuster
2014;():V001T01A032. doi:10.1115/PVP2014-28728.

The fatigue assessment of safety relevant components is of importance for ageing management with regard to safety and reliability. For cyclic stress evaluation, different country specific design codes and standards provide fatigue analysis procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria, the influence of different factors like e.g. environment, surface, temperature and data scatter must be taken into consideration in an appropriate way. In this context there is a need of consolidating and increasing the current knowledge.

In the framework of an ongoing three years German cooperative project performed by Materials Testing Institute MPA Stuttgart and AREVA GmbH (Erlangen) it is the aim to both improve the state of the art based on an experimental program on the factors mentioned above including hold-times at transient free static load and on the derivation of a practicable engineering fatigue assessment concept. Emanating from a review of the current state of the art the cooperative project is split up into three major parts:

1) Experimental investigations concerning the influence of loading parameters and environmentally assisted fatigue (EAF) effects (light water reactor environment) on the fatigue strength of ferritic steels including weldments.

2) Experimental investigations concerning the influence of long hold times and the EAF effects on the fatigue strength of austenitic and ferritic steels.

3) The results of the outlined experimental program and published results will constitute the input for the proposal of an engineering fatigue assessment concept. This concept includes the differentiation between numerous factors of influence as an essential feature. In this context the margins between mean data curves and design curves are to be discussed in detail considering the factors of influence in general and EAF in particular.

Based on a comprehensive consolidation of the state of the art and previous investigations in air and in light water reactor environment an experimental program is set up with the following key aspects:

- Strain controlled fatigue tests on welded (microstructure of the weldment excluding microscopic and macroscopic weld notch effects) and unwelded smooth laboratory specimens subjected to constant and variable strain amplitude loading in air and light water reactor environment.

- Strain controlled fatigue tests on notched specimens for the consideration of multi-axiality effects in air and light water reactor environment.

- Strain controlled fatigue tests on smooth round laboratory specimens in air and in light water reactor environment focusing on long (power plant relevant) hold time effects.

Commentary by Dr. Valentin Fuster
2014;():V001T01A033. doi:10.1115/PVP2014-28883.

For Class 1 components, the consideration of the environmental effects on fatigue has been suggested to be evaluated through two different methodologies: either NUREG/CR-6909 from March 2007 or ASME-Code Case N-761 from August 2010. The purpose of this technical paper is to compare these two methods. In addition, the equations from Revision 1 of the NUREG/CR-6909 will be evaluated.

For these comparisons, two stainless steel component fatigue test series with documented results are considered. These two fatigue test series are completely different from each other (applied cyclic displacements vs. insurge/outsurge types of transients). Therefore, they are producing an appropriate foundation for these comparisons.

In general, the severities of the two methods are compared, where the severity is defined as the actual number of cycles from the fatigue tests, including an evaluation of the scatter, divided by the number of design cycles from the two methods. Also, how stable the methods are is being evaluated through the calculation of the coefficient of variation for each method.

Commentary by Dr. Valentin Fuster
2014;():V001T01A034. doi:10.1115/PVP2014-29093.

In NUREG-1801 (GALL) Revision 0 and Revision 1, the US Nuclear Regulatory Commission (NRC) defined the locations evaluated in NUREG/CR-6260 as a minimum acceptable set for evaluation of environmentally assisted fatigue (EAF), in addressing license renewal for nuclear plant components. Within GALL Revision 2, the NRC revised the expectation, so that plants also investigate the possibility of other locations being more limiting. To address GALL Revision 2 and NUREG-1800 Revision 2, an EAF screening methodology was developed that considers all Safety Class 1 reactor coolant pressure boundary components in major equipment and piping systems that are susceptible to EAF, including those locations listed in NUREG/CR-6260. While the overall screening process steps are similar to those published by EPRI, elements of the detailed application of some steps were performed using alternative techniques. The screening process utilized the comprehensive database of plant component fatigue qualifications available in NSSS vendor documentation, and yielded a comprehensive list of lead indicator locations for EAF consideration. This paper describes the overall process and alternate methods in the context of a specific plant license renewal application.

Topics: Fatigue , Licensing
Commentary by Dr. Valentin Fuster

Codes and Standards: Fatigue Monitoring and Related Assessment Method

2014;():V001T01A035. doi:10.1115/PVP2014-28100.

Modern state-of-the-art fatigue monitoring approaches gain in importance not only as part of the ageing management of nuclear power plant components but also in the context of conventional power plants and renewables such as wind power plants. Consequently, lots of operators have to deal with demanding security requirements to ensure the safe operation of power plants and to cope with plant lifetime extension (PLEX) related issues.

AREVA disposes of a long tradition in the development of fatigue and structural health monitoring solutions. Nuclear and conventional power plant applications require the qualified assessment of measured thermo-mechanical loads. The methodology is transferable to mechanical loading conditions such as those of wind energy plants. The core challenge is the identification and qualified processing of realistic load-time histories. The related methodological requirements will be explained in detail.

In terms of the nuclear industry, the ageing management of power plant components is nowadays a main issue for all actors: states, regulatory agencies, operators, designers or suppliants. As regards fatigue assessment of nuclear components stringent safety standards imply the consideration of new parameters in the framework of the fatigue analysis process:

• new design fatigue curves, consideration of environmental fatigue (EAF) parameters and

• stratification effects.

In this general context AREVA developed the integral approach AREVA Fatigue Calculation (AFC) with new tools and methods in order to live up to operators’ expectations: Simplified Fatigue Estimation (SFE), Fast Fatigue Evaluation (FFE) and Detailed Fatigue Check (DFC). Based on real measured thermal loads and superposed mechanical loads the Fast Fatigue Evaluation (FFE) process allows a highly automated and reliable data processing to evaluate cumulative usage factors of mechanical components. Calculation and management of results are performed within the fatigue assessment software FAMOSi (FAatigue MOnitoring System integrated), thus impact of operating cycles on components in terms of stress and fatigue usage can be taken into account in order to plan optimized decisions relating to the plant operation or maintenance activities.

This paper mainly describes the fatigue and structural health methodologies developed within the AREVA Fatigue Concept (AFC).

Commentary by Dr. Valentin Fuster
2014;():V001T01A036. doi:10.1115/PVP2014-28178.

The assessment for adequacy in managing the effects of fatigue in the ASME Code Class-1 (pressure boundary) components is based on a calculated measure of the projected fatigue damage. This measure is the highest cumulative usage factor (CUF) in a given component under a specified set of cyclic loadings and their expected number of repetitions. The Code-based calculation of CUF and its adjustments for potential environmentally-assisted fatigue (EAF) damage accumulation utilize a multitude of inputs, and conservative assumptions and applied margins. To support the extended service life beyond the original design, or longer life of new designs, changes in inputs and/or conservative assumptions used in these deterministically calculated CUFs are often made to meet a deterministic performance criterion. This makes the impact of uncertainty in the inputs and/or changes in the conservative adjustments difficult to assess.

This paper presents a generic, engineering approach for estimation of the uncertainty distribution of CUF based on the expected statistical characteristics of input variables used in the calculation of EAF-based CUF. The approach does not involve Monte Carlo sampling. The proposed statistical approach analytically combines variances of the inputs leading to an acceptable estimation of the total variance of the CUF. The approach does not require specification of full probability distribution(s) for the input variables, nor is the dependence between variables a critical issue from the analytical point of view. Feasibility and limitations of the approach are discussed in relation to the NB-3200 and NB-3600 procedures of the ASME Code and the current Fen-based augmentation for environmental effects. This approach is further examined in the framework of stress–strength interference methodology to account for the uncertainty in the fatigue performance criterion, that can lead to a rational deterministic safety factor interpretation and its relation to a quantifiable measure of the probability of exceeding the fatigue performance criterion.

Topics: Fatigue , Uncertainty
Commentary by Dr. Valentin Fuster
2014;():V001T01A037. doi:10.1115/PVP2014-28191.

In recent years the Environmentally Assisted Fatigue (EAF) became an item, which has to be considered additionally in terms of ensuring a conservative determination of the actual component’s health status resp. the CUF. For practical application, the consideration of the so called Fen-factor leads to the reduction of the admissible cycles in fatigue calculations. Beyond that the influence of elevated temperatures has been identified as one parameter having a negative influence on the admissible cycles as well. For example the German KTA 3201.2 defines for austenitic steels separate fatigue curves for temperatures above 80°C and for temperatures below 80°C. In summary on the one hand parameters influencing component’s lifetime negatively have to be considered in terms of conservative calculations. On the other hand, there are other parameters which influence the component’s fatigue lifetime in a positive manner. As such positive effects are neglected so far, CUF allowing for EAF tend to become over conservative leading to oversized components. Therefore, positive effects should be considered as well in the framework of a comprehensive and detailed analysis making sure not to overdesign components.

When taking a closer look on the operational behavior of primary circuit components, fatigue loading is mainly defined by long steady-state periods with no significant changes in the loadings and by normally short outage periods with no thermal loading. For example fatigue of a PWR surge-line is mostly caused by short in-surge and out-surge events during start-up or shut-down of the plant. Normal operation transients mostly not cause fatigue relevant events in the surge-line. Fatigue of PWR spray-lines is primarily generated by very few spray-events during a one-year period of operation. Spray events are mainly caused by significant load ramps. Subsequently the fatigue status of primary circuit components is controlled by long periods with no fatigue relevant loading at operating temperature and few additional loading patterns in between. Experimental investigations have shown that hold time effects have a positive influence on fatigue lifetime of austenitic stainless steel materials.

Anyhow, no quantification of these effects has been published in recent years. Within this publication an engineering based approach will be developed to quantify the hold time effect based on literature and published data. On the basis of a practical example the influence of hold time effects will be quantified and a direct comparison to lifetime reducing effect of EAF and temperature will be drawn.

Commentary by Dr. Valentin Fuster
2014;():V001T01A038. doi:10.1115/PVP2014-28657.

In a scenario of an increasing use of renewable energy, conventional power plants will be more and more forced to compensate for the volatility of the natural resources. Even huge coal-fired units which have been designed for baseload operation will face an increased number of start-up/shutdown cycles and the requirement for faster load changes.

For the power plant operator that means a challenge as well as a chance: a challenge because the plant experiences higher alternating stresses which may reduce the lifetime. A chance because usually there are incentives for contributions to the grid stability which may give him additional profits. Coping with the challenges and making the best of the chances will require a detailed and quantitative assessment of the lifetime consumption in various modes of operation.

In this paper, an online software solution is presented that provides this kind of information right at the fingertips of the plant engineers. The innovative approach integrates the most recent European standards concerning the calculation of lifetime consumption from load cycling with state-of-the-art methods of predictive analytics and cutting-edge FEM technologies: the recent standards supported by FEM calculations allow an estimation of lifetime consumption which is unchained from unnecessary allowances. The predictive analytics easily correlate plant operation and lifetime consumption and allow for a reliable prediction of fatigue. This in turn gives the necessary information to make the best of the chances of load flexibilization while mitigating the risks of increased lifetime consumption. If the expected fatigue in a given period is exceeded due to the current mode of operation one can react with a more moderate mode of operation — or the other way round, if the expected fatigue is not reached.

Examples from German coal-fired power plants which have been put under economic pressure by the ongoing “Energiewende” (energy turnaround) are presented to demonstrate this approach.

Commentary by Dr. Valentin Fuster
2014;():V001T01A039. doi:10.1115/PVP2014-28716.

Modern state-of-the-art fatigue monitoring approaches gain in importance as part of the ageing management of conventional and nuclear power plant components (NPP) and particularly in the context of lifetime extension projects and new plants. A key feature of qualified fatigue monitoring is the measurement of realistic loads based on plant instrumentation and/or local temperature measurement. The major prerequisites for any subsequent fatigue assessment are the accurate component stress analysis and the identification of relevant cycles, using highly qualified and efficient cycle counting methods.

In the framework of the paper the pertinent cycle counting methods such as rain flow and peak-valley, are evaluated and allocated to the relevant design codes. The peculiarities of a nuclear power plant typical one year operational thermal load time history is discussed in terms of the resulting stress-time history. The peak-valley and the rain flow methods are applied to that typical power plant operational load sequence and compared to the reference methods. Both idealized model transients without knowledge of the exact sequence in time and the processing of real load time histories are compared and considered.

Design data and actual data for a given annual cycle of a Brazilian nuclear power plant (NPP) constitute the basis for a design code conforming (ASME Section III) fatigue assessment (calculation of partial usage factors for a reference piping component). Conventional and more advanced methodologies are subsequently applied for comparison purpose. The influence of the cycle counting method applied on the resulting partial fatigue usage factors are elaborated based on the given realistic plant data.

Commentary by Dr. Valentin Fuster

Codes and Standards: High Temperature Codes and Standards

2014;():V001T01A040. doi:10.1115/PVP2014-28022.

In order to extend the boiler lives at Advanced Gas-Cooled Reactor (AGR) nuclear power stations in the UK, new temperature measuring instrumentation to monitor reactor gas temperature has been proposed to install on the bore of an intact boiler tube to provide additional boiler operating data to support the station lifetime extension.

This paper details a creep-fatigue crack initiation assessment of the proposed installation of an instrument guide tube within the superheater header using the latest R5 high temperature assessment procedures based on detailed finite element thermal transient stress analysis values for a bounding start-up and shutdown cycle.

The fatigue damage at welds has been calculated based on both weld and parent material properties. The new approach for assessing weldments has been used in this paper. This new approach involves splitting the existing Fatigue Strength Reduction Factor (FSRF) into a Weldment Endurance Reduction (WER), which accounts for reduced fatigue endurance due to weld imperfections, and a Weldment Strain Enhacement Factor (WSEF), which accounts for material mismatch and local geometry.

The creep assessments of the weld material locations have been carried out on both parent and weld material properties including the welding residual stress.

The total creep-fatigue damage is then obtained as the sum of fatigue damage, Df, and creep damage, Dc.

Commentary by Dr. Valentin Fuster
2014;():V001T01A041. doi:10.1115/PVP2014-28349.

As the important lessons learned from Fukushima-nuclear power plant accident, preparation based on Probabilistic Risk Assessment (PRA) with adequate scenario is strongly recognized as essential countermeasures against severe accidents, which are possible in nuclear plants. IAEA requires design extension conditions (DEC) for considering severe accidents. From a view point of structural design, the strength evaluation approach for DEC is somewhat different from conventional one for design basis accident (DBA). There are additional failure modes by extreme loadings under DEC. Best estimation with possible scenarios is necessary for PRA and planning of accident management (AM).

This paper introduces study plan on failure modes and their mechanisms by extreme loadings under DEC.

First step is the list-up of possible failure modes which should be assumed for extreme loadings such as very high temperature, high pressure and great earthquakes. Next is clarification of failure mechanism and relevant limit strength.

One of examples is the failure modes of structural discontinuities under high pressure such as ductile fracture and local failure. Another example is ones of pipes under severe earthquake such as collapse and low cycle fatigue.

To clarify above questions, such different scale tests were planned and conducted as the fundamental tests with simulated materials, structural element tests and structural tests. Preliminary results of above tests and next plans are explained.

Commentary by Dr. Valentin Fuster
2014;():V001T01A042. doi:10.1115/PVP2014-28492.

This paper gives an application case of the RCC-MRx mechanical design code for nuclear components in the domain of significant creep. It could be seen as a guide for engineers who have to perform mechanical creep and creep-fatigue analyses with this code. The application case is a spherical shell with an internal radius of 1250 mm and a thickness of 50 mm, which is made in AISI 316L. The structure is assembled by manual arc welding of plates using 19Cr-12Ni-2Mo type rods. The shell is intended to operate at 550°C mean temperature under an internal pressure of 5 MPa due to argon gas (the atmospheric pressure is considered outside). During operating, the internal temperature is equal to 600°C and the external temperature is 500°C. The shell is periodically stopped for servicing: during these periods, the temperature is 50°C and uniform, and the internal pressure is equal to the atmospheric pressure. A cycle is defined by a 500 h dwell time at 550°C followed by a 24 h arrest at 50°C; transient temperatures, thermal shocks during starting and arrest are neglected. The application of the RCC-MRx is shown for different damages, such as:

- Excessive strain, plastic instability and rupture for primary loadings (negligible and significant creep);

- Ratcheting, fatigue, creep-fatigue for all type of loadings.

The effect of welds on expected life (creep, fatigue), and the comparison of calculated lives and allowed operating cycles for 316L and 316LN are described.

Topics: Creep
Commentary by Dr. Valentin Fuster
2014;():V001T01A043. doi:10.1115/PVP2014-28538.

In 2005, the American Petroleum Institute (API) initiated an effort to update existing yield, tensile and stress-rupture properties found in API Standard 530 Calculation of Heater Tube Thickness in Petroleum Refineries and add properties for alloys not yet covered. The design curves in API 530 until that time were based on data gathered 40 to 50 years earlier by the Materials Properties Council (MPC) and its predecessor, the ASTM-ASME Joint Committee on the Effects of Temperature on Materials. Later, MPC developed proprietary statistically sound algorithms to apply lot-centered regression for parametric analysis of large, unbalanced data sets of diverse heats tested under a variety of conditions. Subsequently, MPC built and maintained archives on the creep and stress-rupture data of alloys of interest to API. For some alloys the data sets contained over a thousand test results on over 100 heats. To assure that future designs will reflect the properties of materials produced using modern practices, API requested MPC to deliver design properties applicable to current materials. This paper presents the back ground, principles and results of the recent analyses performed by MPC that are now available for use by the API membership. The properties furnished in equation format are yield and ultimate tensile strengths for time-independent stresses and results of lot centered Larson-Miller Parameter analyses to obtain time-dependent average and minimum strengths. The properties and application examples of the equations are published as WRC Bulletin 541 Evaluation of Material Strength Data for Use in API Std 530.

Commentary by Dr. Valentin Fuster
2014;():V001T01A044. doi:10.1115/PVP2014-28797.

A recent high temperature steam header case study is extended here to include alternate methods of review, including elastic stress and isochronous strain analysis and accompanying limits. The previous creep analysis was formulated to be exactly consistent with the allowable stress basis, such that alternate design analysis methods and criteria could be rigorously compared. In the current work, the selection of the appropriate limits for elastic results is investigated, as discussed in previous literature, which is motivated by stress redistribution characteristics of the primary (and secondary) loading in a typical header. Next, use of isochronous stress-strain curves generated from the same consistent (Omega) creep model are used for analysis and compared to candidate strain limits. The analyses show that both elastic and isochronous analysis have potential for effective creep design in the context of current high temperature design modernization activities. Finally, multiaxial creep behavior and its effect on detailed creep, elastic and isochronous stress-strain analyses and corresponding limits is also introduced.

Topics: Temperature , Stress , Design
Commentary by Dr. Valentin Fuster
2014;():V001T01A045. doi:10.1115/PVP2014-28905.

In the presence of excessive plasticity, the fracture toughness depends on the size and geometry. For material under fully yielded conditions, the stresses near the crack tip are not unique, but depend on geometry. So Single-parameter; J-approach is limited to high-constraint crack geometry. J-Q theory has been proposed in order to decide crack geometry constraint. This approach assumes that the crack-tip fields have two degrees of freedom. In this paper, based on J-Q theory, crack-tip stress field of fully circumferential cracked pipe under combined load is investigated using FE analysis. Combined loads are tensile axial force and thermal gradient of radial direction. Q-stresses of a crack geometry and it’s loading state are used to determine constraint effect, and give a characteristic order for crack-tip constraint.

Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity Issues for Buried Pipe

2014;():V001T01A046. doi:10.1115/PVP2014-28249.

This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.

Commentary by Dr. Valentin Fuster
2014;():V001T01A047. doi:10.1115/PVP2014-28462.

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 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 improvements have been proposed for Code Case N-806. These include improvements to the analytical procedures for structural integrity evaluation under soil and surcharge loads. In addition, tables of soil properties and other parameters needed in the evaluation are proposed to be provided to improve ease of use. This paper presents the technical basis for the proposed revision to Code Case N-806.

Topics: Metals , Pipes
Commentary by Dr. Valentin Fuster
2014;():V001T01A048. doi:10.1115/PVP2014-28467.

Calculation of pressures in a soil body due to finite loads imposed on the soil surface is a necessary step in the design and analysis of buried commodities. This study compares two commonly-applied numerical methods used to develop the vertical soil pressure profiles applied to buried pipes.

The methods compared in this study differ in theory, basis, assumptions, complexity, and results, and therefore the comparison is meaningful. Provided is a comparison between design vertical forces on different sizes of buried pipes at a range of soil depth, determined using an integration of Boussinesq’s equation [1] and the method specified by AASHTO [2]. The Boussinesq equation is defined as a function of location in varying two dimensional soil planes and the integration is performed over the boundary of the pipe as well as the applied soil surface footprint. The soil surface loading considered in this study includes the AASHTO Design Truck and the AASHTO Design Tandem, positioned as required by the AASHTO LRFD code [2]. Recommendations for application of the results is provided based upon the resulting force magnitude calculated and ease of application of the methods. Consideration of the effects of redistribution of loading due to pavements or other rigid surfaces is outside the scope of this study.

Commentary by Dr. Valentin Fuster
2014;():V001T01A049. doi:10.1115/PVP2014-28766.

In the design of buried piping, resistance of the pipe-soil system to the applied vertical load is a significant design consideration. Application of a resultant vertical load perpendicular to the longitudinal pipe axis results in ovalization of the pipe, with resulting vertical pressures in the soil bedding and horizontal pressures in the soil backfill adjacent to the pipe, as well as internal forces and moments in the pipe wall. For pipes that are relatively rigid compared to the soil backfill, the developed internal forces in the pipe hoop and resulting internal moments in the wall tend to be higher than a more flexible pipe with a stiffer soil backfill. For many buried pipes, design to consider through-wall bending stress is not included and consideration of a buckling allowable is considered adequate. However, for safety-related pipes, investigation to include the through-wall bending stress is considered important.

Although estimation of the internal hoop stresses in a buried pipe is relatively straight forward, several different methodologies are presented in various industry documents for calculating the through-wall bending stress due to the resulting internal moments. Recommended approaches for calculating the through-wall bending stress vary depending on the pipe material, wall thickness, wall type, deflection, burial condition and other factors. A sensitivity study comparing various theories estimating the through-wall bending stress in a buried pipe subject to external loads with a vertical resultant force is performed. The range of variables is assessed to determine the applicability of the results to typical safety-related piping. This study presents the results of the sensitivity study and provides recommendations for the calculation and incorporation of through-wall bending stress in the design of buried safety related piping.

Commentary by Dr. Valentin Fuster
2014;():V001T01A050. doi:10.1115/PVP2014-28777.

This paper will provide an overview of a recent buried piping inspection project leveraging automated ultrasonic testing methods to evaluate internal raw water corrosion at a reliably-operating, nuclear power plant. Discussion of current piping inspection practices for resolving non-planar flaws in raw water piping will be provided, emphasizing inspection performance requirements and a pilot project using automated scanning to optimize inspection value while minimizing field time and costs.

Contents include past buried piping integrity inspection scope / method selection, details on techniques and critical success factors for the project relative to integrity needs, lessons learned from field implementation, further experience from subsequent related inspections using automated and manual techniques for comparison, and recommended actions for future integrity assessments and mitigating actions. Results provided clear justification for this new approach above traditional evaluations performed in the past and may represent a best industry practice for buried pipe assessment.

While targeting buried piping applications, this method can be applied to many systems / locations vulnerable to corrosion. Guidance on applications, limitations, and areas for improvement are also provided from a project / field perspective.

Commentary by Dr. Valentin Fuster
2014;():V001T01A051. doi:10.1115/PVP2014-28781.

This paper sets forth guidance on how to establish a justifiable internal corrosion rate following a first time inspection to predict re-inspection or replacement timing for raw water piping. A novel approach leverages actual plant-wide piping inspection data, leak history, repair history, and corrosion monitoring results together to inform integrity decisions based on experience at a reliably-operating, nuclear power plant. Data is applied on a risk-conscious basis to piping systems based upon failure consequence and uncertainty and differs from the typical approach of reporting location-specific, time-averaged rates. Excavation and in-plant inspection results can now inform commercially-friendly conservatism that reduces leak risks while also minimizing total inspection and maintenance costs.

While directly applicable to buried piping, this method can be applied to any corrosive system / location. Information is presented in a format that readers can readily follow to develop similar justifications for their own sites / systems. Guidance on applications, limitations, and areas for improvement are also provided.

Topics: Inspection , Corrosion , Pipes
Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity of Cast Stainless Steel Pipe

2014;():V001T01A052. doi:10.1115/PVP2014-28089.

Thermal aging of cast austenitic stainless steel (CASS) piping is a concern for long-term operation of nuclear power plants. Traditional conservative deterministic fracture mechanics analyses lead to tolerable crack sizes well below the sizes that are readily detectable in these large-grained materials. This is largely due to the conservative treatment of the scatter in material properties and the imposition of multipliers (structural factors) on the applied loads. In order to account for the scatter in the tensile and fracture toughness properties that enter into the analysis, a probabilistic approach is taken. Application of the probabilistic fracture mechanics (PFM) model to representative problems has led to questions regarding the dominant random variables and the influence of the tails of their distributions on computed failure probability. The purpose of this paper is to report the results of a study to identify the important random variables in the PFM model and to investigate the influence of the distribution type on the computed failure probability. Application of the PFM model to a representative piping problem to compute the depth of a part-through part-circumferential crack that will fail with a defined probability (10−6 for example) revealed that the fracture toughness was not a dominant variable and the distribution of the toughness did not strongly affect the results. In contrast to this, the flow strength (which enters into the calculation of the applied crack driving force — J) was important in that low flow strength was controlling the low probability failures in the Monte Carlo simulation. Hence, the low-end tail of the flow strength distribution was influential. Various types of distribution of flow strength consistent with the available data were considered. It was found that the distribution type has a marked, but not overwhelming, effect on the crack depth that would fail with a given probability. From this it is concluded that the PFM model is quite robust, in that it is not highly sensitive to uncertainties in the dominant input distributions.

Commentary by Dr. Valentin Fuster
2014;():V001T01A053. doi:10.1115/PVP2014-28233.

In this study, a method to predict fracture toughness of aged cast austenitic stainless steels (CASSs) using small punch (SP) test and finite element (FE) analysis is proposed. Grade CF8M is considered and thermally aged up to 5,000 hours at 400°C. SP tests and fracture toughness test using compact tension (C(T)) specimen are conducted with virgin (unaged) and aged CF8M. FE analyses performed in this study use ductile fracture simulation technique with ‘the multi-axial fracture strain model’. The multi-axial fracture strain model for each aged CF8M are determined from SP test data and FE analyses. Fracture toughness of aged CF8M are predicted by conducting fracture toughness test simulations using FE damage analyses. Predicted fracture toughness results are compared with C(T) data to validate the method suggested in this study. The predicted initiation toughness values are predicted well and fracture toughness values are slightly conservative compared to test data.

Commentary by Dr. Valentin Fuster
2014;():V001T01A054. doi:10.1115/PVP2014-28436.

Welds in cast austenitic steels (CASS) are very challenging to inspect using the current American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements. Supplement 9 of ASME Boiler and Pressure Vessel Code Section XI, Appendix VIII is still in the course of preparation, requiring inspectors to use ASME Code Section XI, Appendix III, which provides prescriptive ultrasonic testing (UT) requirements that are significantly less rigorous than UT techniques that have been demonstrated under Appendix VIII. The inability of licensees to demonstrate that the welds in CASS components meet ASME Code requirements has been an ongoing area of concern for the NRC staff. The lack of a reliable inspection method for welds in CASS materials has led to hundreds of relief requests over the past four decades. While no degradation mechanism has been found in CASS components to date, there is no guarantee that a new degradation mechanism affecting CASS welds will not emerge as nuclear power plants go beyond forty years of operation. Licenses need qualified procedures and personnel for the inspection of welds in CASS materials in order to put licensees into compliance with ASME Code, meet federal regulations, reduce the number of needed relief requests, and ensure the structural integrity of their welds.

Over the past decade there have been significant developments in nondestructive examination (NDE) technology. The use of encoded phased array techniques using low frequency ultrasound has been shown to be able to reliably find flaws greater than 30% through wall in CASS materials with a variety of microstructures. Additionally, an improved understanding of the fracture mechanics of CASS components is being developed that shows the flaw sizes that can be tolerated in CASS components. These advances in NDE techniques and fracture mechanics theory are converging on a path to allow for qualifications of procedures and personnel for the ultrasonic inspections of welds in CASS components.

Recent developments in ASME Code includes Code Case N-824, which provides guidance on the examination of CASS materials based on the advances in NDE technology and an improved understanding of the NDE techniques capable of finding flaws in CASS components as well as Code Case N-838 for flaw tolerance evaluations of CASS piping components. Finally, work on ASME Code Section XI Supplement 9 is progressing, with several important issues still to be addressed. The NRC staff sees a clear path forward and is working to ensure that qualified inspections of welds in CASS materials will be possible in the future.

Commentary by Dr. Valentin Fuster
2014;():V001T01A055. doi:10.1115/PVP2014-28745.

The Pacific Northwest National Laboratory (PNNL) has been involved with nondestructive examination of coarse-grained cast austenitic stainless steel (CASS) components for over 30 years. More recent work has focused on mapping the ultrasonic sound fields generated by low-frequency phased-array probes that are typically used for the evaluation of CASS materials for flaw detection and characterization. The casting process results in the formation of large-grained material microstructures that are nonhomogeneous and anisotropic. The propagation of ultrasonic energy for examination of these materials results in scattering, partitioning, and redirection of these sound fields. The work reported here provides an assessment of sound field formation in these materials and provides recommendations on ultrasonic inspection parameters for flaw detection in CASS components.

Confirmatory research conducted at PNNL consisted of acquiring sound field data from four CASS components containing columnar, equiaxed, and banded grain structures, and a fine-grained wrought stainless steel specimen used for benchmarking. Phased-array probes with center frequencies of 0.5, 0.8, and 1.0 MHz were used for sound field formation, with a pinducer being raster scanned over the end of the specimen face to capture the sound field energy. Data were collected at multiple refracted and skew angles, and imaging performed for analyses. A 6.4-mm (0.25-in.) thick slice of material was removed from the end of the CASS components and the beam mapping repeated. This slicing and mapping sequence was performed three times to produce multiple beam images through the specimens. Grain sizes were also measured at each mapped specimen face and compared to sound field characteristics. The acquired sound field images were characterized in terms of beam redirection from the theoretical position, beam scatter or coherence, and partitioning. A comparison of the fine-grained beam data to the CASS data is made and conclusions are presented.

Topics: Stainless steel
Commentary by Dr. Valentin Fuster

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

2014;():V001T01A056. doi:10.1115/PVP2014-28159.

There is a proximity rule to calculate residual fatigue life of components in nuclear power plants. It is easy to evaluate soundness of a structure member by using this rule. If a subsurface crack is located near free surface of the structure, this real subsurface crack is transformed to surface crack. The condition to transform subsurface crack to surface crack is defined by relationship between crack size and the distance from crack tip to free surface. However, many organizations proposed proximity rules which differ from each other. It is advisable to verify which rule is preferable in these rules by experiment, but it is difficult to introduce subsurface crack at an optional position. Therefore, numerical simulation is needed for this purpose. Especially, S-version FEM is very useful for as much as model of subsurface crack is independent of global structure, and crack growth is easily simulated.

Both subsurface and transformed surface crack growths are simulated by S-FEM. Subsurface crack grows toward free surface. When subsurface crack tip was touched to free surface, this crack was converted into surface crack by using stress intensity factor calculated at this time. In this way, crack growth behavior from subsurface to surface flaw is represented. By comparing the crack growth rate of surface to subsurface flaw with that of surface flaw transformed by each proximity rule, proximity rules can be verified by numerical simulation.

Authors had proposed the proximity rule at the ASME PVP 2013 conference [1]. However, new rule was proposed by numerical simulation only under cyclic tensile load. In addition, only two proximity rules were studied at the last conference. In this study, the number of proximity rules is increased, and this problem is simulated under other loading condition such as cyclic bending load. New proximity rule gives reasonable and conservative results in numerical simulation.

Commentary by Dr. Valentin Fuster
2014;():V001T01A057. doi:10.1115/PVP2014-28477.

In nuclear power plant, there is a proximity rule to evaluate multiple inner cracks [1]. Inner cracks are generated inside of the structure in different manners. There are many parameters which affects the growing processes of inner cracks. They are; locations, slant angles, aspect ratio of each inner crack and distances between adjacent inner cracks. When multiple inner cracks are detected, proximity rules are proposed. But due to the complexity of the problem, it is necessary to verify proposed proximity rules. But experimental study is very difficult due to existence of many parameters, and crack growth occurs inside of the structure. Numerical simulation is needed for this purpose.

This problem is simulated using S-version FEM [2]. Using S-FEM, inner crack is modeled independently from global structure, and crack growth is easily simulated. In maintenance code of nuclear power plant, initial defects are modeled as elliptical cracks in a normal plane to tension loading direction, and growth rate is estimated on this plane. But by using S-FEM, realistic defect shape is modeled, and crack growth by fatigue is simulated. Usually, such small defects are subjected to multi-axial loading, and crack growth behaviors are very complicated. Finally, detect shape becomes elliptical or circular crack in a plane normal to tension loading direction in the structure. Fatigue cycles for these growing processes are calculated, and conservativeness of this maintenance code is discussed. Parametric studies are conducted for this problem, and proximity rules are verified with numerical results.

Commentary by Dr. Valentin Fuster
2014;():V001T01A058. doi:10.1115/PVP2014-28513.

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

Commentary by Dr. Valentin Fuster
2014;():V001T01A059. doi:10.1115/PVP2014-28568.

A simplified method is proposed to evaluate limit load at incipient plastic collapse for a pipe made of ductile materials with multiple circumferential flaws. Fitness-for-service codes such as the ASME Boiler and Pressure Vessel Code Section XI specify limit load analysis for a pipe with a single flaw. Multiple flaws caused by stress corrosion cracking have actually frequently been detected in the same or adjacent weld lines. Therefore, some methods of evaluating limit load for a pipe with multiple flaws have been proposed. Bending stress at limit load i.e., collapse bending stress is evaluated regarding actual flaw depths, angles and locations of multiple flaws with these methods. Computational evaluation is required because of the enormous number of repetitive calculations. Thus, simple evaluation that can be applied to multiple flaws is required to incorporate into the code procedure. A simplified limit load analysis for multiple flaws in a pipe was developed in the present study based on a net-section approach. Multiple flaws were characterized as equivalent symmetric through-wall flaws on the basis of the sizes and locations of the original flaws. The collapse bending stress for the equivalent flaws was evaluated from simple equations on the basis of the net-section approach. The collapse bending stress for original flaws was determined from a lower value between the largest single flaw in the original flaws and equivalent flaws. The collapse bending stresses were evaluated based on previous and present studies. The results from a comparison demonstrated that the simplified limit load analysis could accurately be applied to a pipe with multiple flaws.

Topics: Stress , Pipes
Commentary by Dr. Valentin Fuster
2014;():V001T01A060. doi:10.1115/PVP2014-28892.

Ductile fracture behavior using GTN (Gurson-Tvergaard-Needleman) model in commercial FEA code was evaluated. The material properties of the GTN model for Type 304 SS were experimentally identified. Smooth-bar and notched-bar specimens were subjected to monotonic loading by tensile test, and load-displacement curves were measured. Then, the tensile test was simulated. Material properties of the GTN model were calculated from measured and simulated load-displacement curves with inverse analysis based on Bayes’ theorem. Simulated load-displacement curves using the GTN model of different curvature notched-specimens agreed well with the measured results. To verify the evaluation of ductile fracture behavior using the GTN model, flat-plate specimen with a single surface flaw and specimen with multiple through flaws were subjected to monotonic loading. Ductile fracture of the flat-plate specimen was simulated by FEA using the GTN model using the calculated material properties. The simulated load was less than the measured load at the same displacement. The analysis using the GTN model can estimate the load on the safe side and the GTN model can conservatively simulate ductile fracture behavior.

Commentary by Dr. Valentin Fuster

Codes and Standards: International Session for Fast Reactor Design and Construction

2014;():V001T01A061. doi:10.1115/PVP2014-28324.

The paper describes the design rules for ratcheting in the 2012 edition of the RCC-MRx Code issued in French and English versions by AFCEN (French Association for the rules governing the Design, Construction and Operating Supervision of the Equipment Items for Electro Nuclear Boilers).

For austenitic stainless steels, the RCC-MRx Code uses the efficiency diagram concept to evaluate an effective primary stress, Peff. Peff is defined as a virtual stress that applied alone would cause the same strain as the combination of the primary static stress and the secondary cyclic strain really applied. This concept is extended to significant creep domain and includes corrections to take into account structures cases presenting secondary membrane stresses (e.g. cylinders subjected to axial thermal gradients varying with time and space) or short duration overloads (as a level A seismic load, or an overload due to rapid drain-out caused by a sodium-water reaction).

An alternative 3Sm design rule is proposed for all materials in the case of non-significant creep damage.

Topics: Design , Damage
Commentary by Dr. Valentin Fuster
2014;():V001T01A062. doi:10.1115/PVP2014-28485.

The main objective of a standard or a code is to capitalize the technical feedback of constructions. It is the case of the RCC-MR code (now RCC-MRx) which has integrated the technical results of projects such as PHENIX, SUPERPHENIX, Jules Horowitz Reactor, ITER vacuum vessel. These rules are now planned to be used for new plants and it appears that they have to be updated to meet the needs of Gen IV projects like the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project or MYRRHA (Multi-purpose hybrid research reactor for high-tech applications) project.

This paper presents the ASTRID project approach to initiate in the earlier stage of the design an evaluation of the rules in order to improve as soon as possible the code. The result of this approach in terms of code modifications and improvement objectives are also presented.

Commentary by Dr. Valentin Fuster
2014;():V001T01A063. doi:10.1115/PVP2014-28689.

316FR and Mod.9Cr-1Mo steels are candidates as structural materials for Japanese Sodium Fast Reactor (JSFR). The design life and operation temperature of JSFR is 60 years and 550 °C, respectively. Time-dependent allowable stress is essential. The evaluation of allowable stresses to 500,000 h is a considerable item. Long term strength is evaluated from a viewpoint of microstructural evaluation related to fracture mechanism. In addition, degradation after long term operation at elevated temperature is important. Aging is considered as one of the degradation. The effect of aging on short term property is analyzed. Material strength standard is also necessary for very thick tube sheets of forgings and small diameter thin walled seamless pipes, which are made of Mod.9Cr-1Mo steel in steam generators. This paper summarized currently available data and information on the above items, and shows path forward to the development of material strength standard for 60 years design in JSFR.

Commentary by Dr. Valentin Fuster
2014;():V001T01A064. doi:10.1115/PVP2014-28853.

This paper overviews the current status of the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR), the demonstration fast reactor which is in the phase of conceptual study. Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept, a concept that materializes code rules that are most suitable to the reactor types they are applied to. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.

Commentary by Dr. Valentin Fuster
2014;():V001T01A065. doi:10.1115/PVP2014-28875.

The prevention of excessive deformation by thermal ratcheting is important in the design of high-temperature components of fast breeder reactors (FBR). This includes evaluation methods for a new type of thermal ratcheting caused by a traveling temperature distribution. Igari et al. [1] proposed a mechanism-based evaluation method to evaluate thermal ratcheting caused by temperature distributions traveling long and short distances.

In this paper, we simplify the existing method and propose a screening method to prevent thermal ratcheting strain in the design of practical components. The proposed method consists of two steps to prevent the continuous accumulation of ratcheting strain.

The first step is to determine whether all points through the wall thickness are in the plastic state. This is based on an equivalent stress, which comprises the primary stress, the thermal membrane stress, and the thermal bending stress. When the equivalent stress is less than the yield strength of the cylinder material, overall plastic deformation through the wall thickness does not occur. When the equivalent stress exceeds the yield strength in some regions of the cylinder, the ranges of these regions are measured for the second step. To prevent the acceleration of the plastic deformation due to creep, we define the upper limit of the equivalent stress based on the relaxation strength, Sr.

The second step is to determine whether the accumulation of the plastic strain saturates (i.e. if shakedown occurs). For this purpose, we define the screening criteria for the range of the plastic region. When the range of the plastic region is sufficiently small, residual stress is generated in the direction opposite to the plastic deformation direction. As a result of residual stress, further accumulation of the plastic deformation is suppressed, and finally shakedown occurs. If the range of the plastic region exceeds the defined criteria, a more detailed evaluation method (e.g. inelastic finite element analysis) may be used for the component design.

To validate the proposed method, we performed a set of elasto-plastic finite element method (FEM) analyses, with the assumption of elastic perfectly plastic material.

Commentary by Dr. Valentin Fuster

Codes and Standards: NDE Personnel Qualification (PQ)

2014;():V001T01A066. doi:10.1115/PVP2014-28885.

Implementation and sustained support of this central certification program for nondestructive examination and quality control personnel under the auspices of ASME requires coordination between ASME as the certification body, codes and standards, regulators, training organizations and employers. This paper discusses factors critical to ANDE implementation.

Commentary by Dr. Valentin Fuster
2014;():V001T01A067. doi:10.1115/PVP2014-28994.

The American Society for Mechanical Engineers (ASME) is developing and implementing a Nondestructive Examination (NDE) and Quality Control (QC) personnel qualification and certification program (ANDE). In keeping with ASME development of codes and standards the NDE program will be assembled by industry experts and administered centrally by ASME. A new ASME standard recognized by ANSI is being developed as the core of the program and will be based upon industry best practices. The total program is a significant paradigm change from the traditional employer based and other central qualification and certification programs, the new standard will include features such as the Systematic Approach to Training, Job Task Analysis and Qualification Cards to direct the development of training and experience. The program was designed to improve the skill level and consistency amongst NDE practitioners and will use the performance-based approach including psychometric practices in the development of both written and practical demonstration examinations. This paper will primarily focus on the development of the standard and its salient features.

Commentary by Dr. Valentin Fuster

Codes and Standards: Ratcheting and Fatigue Issues in Pressure Vessel and Piping Design

2014;():V001T01A068. doi:10.1115/PVP2014-28772.

This paper builds on PVP2013-98150 by Kalnins, Rudolph, and Willuweit [1], which documented two calibration processes for determining the parameters of the Chaboche nonlinear kinematic hardening (NLK) material model for stainless steel, and tested the material model using a pressurized cylindrical shell subjected to thermal cycling. The current paper examines (1) whether a Chaboche NLK model with only two terms (rather than four as in PVP-98150) is sufficiently accurate, (2) use of the ANSYS program for material model refinement and finite element analysis, and (3) analysis using temperature-dependent NLK model parameters, again using ANSYS.

Commentary by Dr. Valentin Fuster
2014;():V001T01A069. doi:10.1115/PVP2014-28820.

Design of components against incremental deformation or “ratcheting” under cyclic loading conditions is addressed in Article NB-3200 of Section III of the ASME Boiler and Pressure Vessel Code. The ratcheting rules, based on the Bree diagram, relate primary stress and secondary stress ranges that are calculated elastically and aim to approximate elastic-plastic material behaviors under cyclic loading. The Bree diagram was developed for cases with through-thickness thermal bending and constant primary membrane stress. It does not account for high thermal membrane stress that can occur near gross structural or thermal discontinuities. Cyclic thermal membrane stress combined with sustained stress can lead to ratcheting that is not accounted for in the current design rules.

This paper discusses the validation of proposed criteria for evaluating thermal stress ratcheting under high thermal membrane stress using an elastic analysis. These proposed criteria are confirmed by an analysis of a nozzle that is attached by a partial penetration weld to a vessel head and subjected to severe thermal cycling. Linearized stresses from an elastic analysis under pressure and thermal loadings typical for a nuclear power plant are compared to limits in NB-3200 for thermal stress ratcheting. Additionally, an elastic-perfectly plastic analysis is used to evaluate if the component will shakedown.

This analysis demonstrates that the proposed rules prevent ratcheting of a typical geometry with typical operating loads in a nuclear plant. The current thermal stress ratcheting rules evaluated on an elastic basis are enhanced to cover cases with high thermal membrane stress while not removing conservatism. Additionally, the evaluation of the simplified elastic-plastic rules for thermal stress ratcheting are simplified.

Commentary by Dr. Valentin Fuster
2014;():V001T01A070. doi:10.1115/PVP2014-28895.

To show the implications of using crack growth analysis to evaluate fatigue life, the case of a cylinder with an internal notch with a varying root radius is examined. The notch radius controls the stress gradient, where a steeper stress gradient is expected to result in slower crack growth and longer fatigue life. The notch root stress is made the same between specimens of different notch radius by scaling the applied load. As a result, the conventional fatigue analysis that calculates a fatigue usage factor from a fatigue curve based on stress at a point gives identical results for all specimens. A crack growth analysis, on the other hand, gives significantly different fatigue lives for the specimens because of the different stress gradients.

On this basis of allowable fatigue life, the traditional fatigue curve-based approach is compared with the crack growth-based flaw tolerance approach. The relative conservatism of the two approaches as a function of various parameters, including stress gradient, is discussed.

Commentary by Dr. Valentin Fuster
2014;():V001T01A071. doi:10.1115/PVP2014-28953.

The classical approaches in shakedown analysis are based on the assumption that the stresses are eventually within the elastic range of the material everywhere in a component (elastic shakedown). Therefore, these approaches are not very useful to predict the ratcheting limit (ratchet limit) of a cracked component/structure in which the magnitude of stress locally exceeds the elastic range at any load, although in reality the configuration will certainly permit plastic shakedown. The Non-Cyclic Method (NCM) has been proposed recently to determine both the elastic and the plastic ratchet boundary of a component or structure under cyclic loading by generalizing the static shakedown theorem (Melan’s theorem). The proposed method is based on decomposing the loading into mean (time invariant) and fully reversed components. When a cracked structure is subjected to cyclic loading, the crack and its vicinity behave differently (local) than the rest of the structure (global). The crack may propagate during the application of cyclic loading. This will affect both local and global behavior of the cracked structure.

This paper investigates global and local ratcheting of the cracked structures using the NCM and fracture mechanic parameters.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Chinese Codes and Standards

2014;():V001T01A072. doi:10.1115/PVP2014-28119.

In order to assess the safety of concentric reducers and eccentric reducers, the meridian stress and circumferential stress of the reducers subjected to in-plane bending moments are calculated based on the membrane theory. Solutions of limit bending moments are derived by using the approach with the equivalent stress and engineering coefficient. The solutions for both reducers are identical, as the formulas are composed of a base term multiplied by a bending moment coefficient. The limit bending moments are dominated by the small end of the reducers. The value of limit bending moment of the reducers is 14.6% more than that of a thin wall reducing elbow with same diameter and thickness when corresponding bending coefficient λ>3.0. But comparison with a straight pipe which has the same diameter and thickness as the small end of the reducers, the limit bending moment of the reducers is 11.1% lower and the constant in the formula for the reducers is 3.6 instead of 4.0 for straight pipe. It could be dangerous to determine the limit bending moment of concentric reducers and eccentric reducers directly from the formula of limit bending moment for a thin wall pipe.

Topics: Stress
Commentary by Dr. Valentin Fuster
2014;():V001T01A073. doi:10.1115/PVP2014-28186.

The influence of residual stress on stress corrosion behavior was studied by way of SSRT test on the base metal of X80 pipe with different residual tensile stress in the simulating soil environment aqueous solution; and the electrochemical test techniques were adopted to test the effect of the residual stress on electrochemical polarization behavior and AC impedance characteristics of the base metal. It is shown that the base metal specimens without introducing the residual stress of X80 UOE pipe and X80 SAWH pipe have different stress corrosion sensitivity in this solution. The base metal specimens with different residual stress of X80 pipe only occurs the anodic metal dissolution reaction on the material surface in this kind of solution without forming the passive film. The residual tensile stress leads to the increase of the electrochemical activity and the reduction of impedance of the base metal surface of X80 pipe, and it accelerates the nucleation and early propagation of the stress corrosion crack, which promotes the SCC sensitivity of the base metal.

Commentary by Dr. Valentin Fuster
2014;():V001T01A074. doi:10.1115/PVP2014-28243.

For many in-service pressure pipe welds with thinner thickness (less than 10mm), the conventional pulse echo testing technique is difficultly used for its less thickness and the radiographic testing is not conveniently performed because of the pipe position limitation. But, the creeping wave (CW) technology can play an important role in these pipe welds inspection because of its special characteristics. In the paper, the generation and attenuation feature of CW are detailed and the probe configuration and comparative block of CW testing are discussed. Finally, successful application of CW in pipe welds is demonstrated.

Commentary by Dr. Valentin Fuster
2014;():V001T01A075. doi:10.1115/PVP2014-28361.

The test of austenitic stainless steel specimens with strain control mode of pre-strain was carried out. The range of pre-strain is 4%, 5%, 6%, 7%, 8%, 9% and 10% on austenitic stainless steel specimens, then tensile testing of these samples was done and their mechanical properties after pre-strain were gotten. The results show that the pre-strain has little effect on tensile strength, and enhances the yield strength more obviously. According to the experimental data, we get a relational expression of S30408 between the value of yield strength and pre-strain. We can obtain several expressions about different kinds of austenitic stainless steel by this way. It is convenient for designers to get the yield strength of austenitic stainless steel after pre-strain by the value of pre-strain and the above expression.

Commentary by Dr. Valentin Fuster
2014;():V001T01A076. doi:10.1115/PVP2014-28490.

A series of seismic table tests about the large steel cylindrical liquid storage tank models with floating roof were taken in this study. Different direction seismic excitations were input to the experimental structure system under different working conditions to test and analyze the seismic response behavior. The effects of various factors, such as the liquid surface height, the floating roof, the different wave amplitudes and frequencies, as well as their combined effects to the seismic dynamic response were taken into account. Dynamic fluid pressure data was got by the tests, and a new method that rain flow counting method was used in this study, in order to consider mean pressure throughout the vibration process, the pressure amplitude and the effective amplitude of the cumulative number of cycles factors together. Through this method, the strength of the dynamic fluid pressure could be described more reasonably. In addition, the relationship between the test results and the tank uplift responses which were studied in our former work were discussed. A reliable basis could be provided for theoretical analysis, structural design and computer numerical simulation research through this investigation.

Commentary by Dr. Valentin Fuster
2014;():V001T01A077. doi:10.1115/PVP2014-28491.

Large tube type fuel heating furnaces are important and essential equipments, which have a wide range of applications in the petroleum and chemical industry, and failure modes of the furnace tubes in them are mostly caused by carburizing. It is important to improve effective detection technology for the measure of the tube carburized layer thickness, as well as the research on the comprehensive performance of furnace tubes considering the impacts of different conditions. The specifications of Φ70×6 mm HP40Nb cracking furnace tube was studied by experiment method to evaluate the effectiveness of an acoustic emission (AE) technique, which used to measure the carburized layer thickness. In the experiment, many factors were taking into account, such as different carburizing time, the response of acoustic emission attenuation to the organization changes, the magnetic field changes, and additional stress which caused by carburization are considered. The results show that the short periods strong carburizing has obviously impacts in the changes of the organization and magnetic field of the cracking furnace tube, however, it has little contribution to the acoustic emission signal attenuation.

Commentary by Dr. Valentin Fuster
2014;():V001T01A078. doi:10.1115/PVP2014-28497.

In order to study acoustic emission (AE) signals characteristics of pitting corrosion on carbon steel, Pitting corrosion process on carbon steel in 6% ferric chloride solution was monitored by AE technology. K-mean cluster algorithm was used to classify the monitored AE signals, in which the duration, counts, amplitude, absolute energy and peak frequency were analyzed as the AE signals characteristics, and different types AE sources were identified. The results showed that there were mainly three type AE sources during carbon steel pitting corrosion process in ferric chloride solution, and the different types AE sources could be classified by cluster analysis. The research results have some certain significance for AE monitoring of pitting corrosion on carbon steel.

Commentary by Dr. Valentin Fuster
2014;():V001T01A079. doi:10.1115/PVP2014-28521.

When pressure vessels are subjected to fire damage, they maybe deform partially and the mechanical properties of materials would be degraded. Fitness for service assessment can help to minimize reconstruction costs and allow safe resumption of unit operation as fast as possible. The safety of a packed column subjected to fire damage due to spontaneous combustion of FeS during a shutdown period is assessed according to API 579-1/ASME FFS-1. First, the highest temperature during the fire accident was estimated. The possible damage was examined by hardness test, dimensional checks, in-situ field metallography, ultrasonic test and penetration test. The Heat Exposure Zones were defined based on the results of visual inspection and test. The plates containing Widmanstätten or localized distortions were replaced. The column leaned with the vertical deviation of 85mm which could not satisfy the requirement of the national standard GB 50461-2008. The finite element method is adopted to analyze the influence of the vertical deviation with the consideration of wind loading. The analytical results show that the vertical deviation of 85mm has no significant effect on the safe operation of the column before the next inspection.

Commentary by Dr. Valentin Fuster
2014;():V001T01A080. doi:10.1115/PVP2014-28535.

Cr5Mo furnace tubes have served in a coking furnace nearly 15 years. For the purpose of risk reduction and to assure a safe long term stable operation of the furnace, these tubes have been investigated by measns of macrographic observation, chemical composition analysis, mechanical property analysis, hardness and metallurgical structure analysis, etc. The experiment results shows that after a long period serve at high temperature, the hardness of the tube has decreased in 15 years. However, the hardness of inner wall is higher than the others, from which we can infer that the inner wall of the tubes is carburized. Taking GB9948-2006 as a reference standard, the yield stress of the furnace tube in normal temperature has decreased by 16%, and the yield stress in high temperature is also failed to fit the standard. Metallographic analysis shows that the grain fineness is 6 grade and the nodulizing grade is 3.5 grade, pearlite spheroidization is getting severer and it will reduce the creep limit and creep rupture strength. Based on the result of high temperature endurance test, isotherm method and L-M extrapolation are adopted to predict the remaining life of furnace tubes, assessment shows that these furnace tubes are able to serve for another cycle under the operating temperature of 650°C and the stress of 18MPa.

Topics: Tubing , Furnaces , Damage
Commentary by Dr. Valentin Fuster
2014;():V001T01A081. doi:10.1115/PVP2014-28539.

Tube trailers with large capacity seamless steel cylinders have been used more and more widely in China to transport industrial gases. As an authorized tube trailer inspection organization by the government of China, China Special Equipment Inspection and Research Institute has carried out periodical inspection of tube trailers in China since 2004. A statistical analysis was done based on 4059 periodical inspection reports of tube trailers from 2004 to 2013, which include 3687 tube trailers and 31627 cylinders. Information of present status of tube trailers in China was obtained, such as annual amounts, usage regions of tube trailers, numbers of cylinders mounted on tube trailers, and specifications, filling mediums, typical flaws of the cylinders. The statistical results show that mounting threads of 16.16% of the cylinders were worn seriously, and repair or reprocess work had to be done on these threads. Reasons of the abrasion of the mounting threads were analyzed and discussed by taking into account the usage regions of the tube trailers, the detailed structures of the mounting threads, and the mounting location of cylinders on the tube trailers.

Topics: Inspection , China , Cylinders
Commentary by Dr. Valentin Fuster
2014;():V001T01A082. doi:10.1115/PVP2014-28569.

With the rapid development of Chinese economy, the pressure vessels are developing towards the direction of high parameters and have at least one of the following characteristics, one is the use of advanced technology and process, diversification and deterioration of raw stock, causing extremalization of service conditions, which is represented by higher pressure, higher or lower temperature, stronger medium corrosiveness; the other is promoting presence of pressure vessels with extreme dimensions in order to improve economic benefit, which is represented by increased diameter, wall thickness, length or height. Service environment and high parameterization of equipment dimensions cause expansion of design boundary, whereas expansion of design boundary will cause changes in failure modes and design criteria. Therefore, a group of Chinese scholars have conducted effective exploration on design, manufacturing and maintenance of high-parameter pressure vessels. This paper describes the expansion of design boundary and changes in failure modes brought by extramalization of service environment and high parameterization of large-sized equipment, reviews technical advancement in design, manufacture and maintenance of Chinese pressure vessels in recent decade, and forecasts recent opportunities and challenges for high-parameter pressure vessel.

Commentary by Dr. Valentin Fuster
2014;():V001T01A083. doi:10.1115/PVP2014-28667.

The skirt support structure of hydrogenation reactor works in the conditions of high temperature and high pressure. There is not only high mechanical stress but also high temperature difference stress in this zone. The service environment of the structure is in poor conditions. In this paper, temperature field analysis on the local structure of skirt hot box is carried out. The stress classification method and direct route method are used to calculate the skirt support structure of hydrogenation reactor. The calculation results are compared. Results of these two analysis and design methods are both meet with the requirements of relevant standards. The results obtained by two methods are discussed. Some meaningful conclusions are obtained.

Commentary by Dr. Valentin Fuster
2014;():V001T01A084. doi:10.1115/PVP2014-28740.

Pit defects due to corrosion may occur in the inner surfaces of metal liners of fiber wrapped composite cylinders during working process, which may affect their anti-fatigue performance. Finite element method was used to analyze the stress of hoop-wrapped composite cylinders with pit defects of different sizes based on element birth and death method to take the pre-stress resulted from autofrettage process into consideration. Numeral results show that stress concentration appears in the pit defects, but the portion of the liner far away from pit defects keeps the same stress level with the defect-free composite cylinder. Pit defects on liner reduce the fatigue performance of the composite cylinder. Goodman equation was utilized to calculate the equivalent alternating stress amplitude to evaluate the variation of the fatigue performance with the dimensional change of the pit defects in the inner surfaces of metal liners of hoop-wrapped composite cylinders.

Commentary by Dr. Valentin Fuster
2014;():V001T01A085. doi:10.1115/PVP2014-28878.

Liquid ammonia storage leakage accident occurred frequently in recent years. In this paper, the Infrared thermal imaging Technology had been presented by introducing the applications of liquid ammonia storage inspection. It was indicated that the infrared thermal imaging technology had irreplaceable advantages in equipment insulation condition assessment, funneled detection, heat level judgment, electrical equipment failure hidden. The experiments showed that infrared thermal imaging technology could provide scientific references for early diagnosis and preventive maintenance of devices. And it had widely application prospect in the ammonia refrigeration system safety inspections.

Topics: Inspection , Imaging
Commentary by Dr. Valentin Fuster
2014;():V001T01A086. doi:10.1115/PVP2014-29036.

Recently, the integrity management of large size atmospheric storage tank in China is still in the preliminary stage. The purpose and the concept of integrity management, the core technical system and the management system are discussed. Main system framework of integrity management about large size atmospheric storage tank is developed, combining with the characters including super-large inner volume, corrosive nature of storage media, dangerous of leakage explosion, strict request for non-interruption operating, and related national codes about integrity management. Two elements of the system, an integrity management process of large size atmospheric storage tanks which combines quality control, management of change, performance test, communication consulting, and an integrity management technology system of large size atmospheric storage tank which is composed of data collecting, risk evaluation, integrity evaluation, responses measures, are proposed. At the same time, the procedure of integrity management is applied to an oil depot in a petrochemical company. The risk evaluation and the integrity evaluation were made on 34 storage tanks. To reduce the risk, some measures are put forward to the 9 storage tanks in medium high risk. As a result, all the large size atmospheric storage tanks are controlled at a medium low level.

Topics: Storage tanks
Commentary by Dr. Valentin Fuster
2014;():V001T01A087. doi:10.1115/PVP2014-29045.

A life management approach on in-service device in refining unit was presented in this paper. By means of this method, corporate resources can reasonably be allocated. And a long-term operation of the equipment can be ensured. First, according to the type of the equipments, statistical analysis of equipments was carried out to obtain the life stages: early life stage, middle life stage, late life stage, extended life stage. Then, based on damage mechanism residual life of the equipments was calculated out. At the same time, the recommended maintenance cycle was given. If the operational life of the equipments cannot meet the expected life, the cause analysis should be done. The subsequent processing steps, such as maintenance, repair and renewal were proposed. Next, the equipments in the corporation were classified. For different class of the equipments, the different management methods were given. Finally, a re-evaluation of life was conducted.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in European Codes and Standards

2014;():V001T01A088. doi:10.1115/PVP2014-28092.

Failure of welded structures due to the presence of flaws is typically driven by a mixture of applied and residual stresses, yet in most cases only the former are known accurately. In as-welded structures, a typical assumption is that the magnitude of welding residual stress is bounded by the room temperature yield strength of the parent material. The UK flaw assessment procedure BS 7910:2013 also assumes that mechanical loading (either as a result of proof testing or during the initial loading of an as-welded structure) will bring about a relaxation in residual stress. Conversely, the UK structural assessment code for nuclear structures, R6, contains a warning on the ‘limited validation’ of the BS 7910 approaches for stress relaxation and suggests that they should be used ‘with caution’. The aim of this study was therefore to review the basis of the BS 7910 clauses on stress relaxation with a view to harmonising the BS 7910 and R6 rules for cases in which the original welding residual stress distribution is not known.

A companion paper describes the history of the residual stress relaxation clauses of BS 7910. A considerable programme of work was carried out in the late 1980s to justify and validate the clauses, using a range of experimental and numerical work. This included analysis of work carried out by the UK power industry and used in the validation of the R6 procedure. The full underlying details of the work have not hitherto been available in the public domain, although the principles were published in 1988. The approach proposed in BS 7910 combines ‘global’ relaxation of residual stress (Qm) under high mechanical load with ‘local’ enhancement of crack tip driving force through the adoption of a simplified primary/secondary stress interaction factor, ρ. This is different from the method adopted by R6, but appears to be equivalent to allowing negative values of ρ under conditions of high primary stress.

A re-analysis of the original TWI work, using the current version of BS 7910, has shown nothing to contradict the approach, which represents a workable engineering solution to the problem of how to analyse residual stress effects in as-welded structures rapidly and reasonably realistically when the as-welded stress distribution is unknown.

Commentary by Dr. Valentin Fuster
2014;():V001T01A089. doi:10.1115/PVP2014-28129.

The European Pressure Vessel Standard EN 13445 provides in its part 3 (Design) a simplified method for fatigue assessment (Clause 17) and a detailed method of fatigue assessment (Clause 18). Clause 18 “Detailed Assessment of Fatigue Life” is under revision within the framework of the European working group CEN/TC 54/WG 53 - Design methods. The latest amendments of Clause 18 are to be presented. All these amendments aim at a significant increase in user friendliness and clear guidelines for application. The following items are to be mentioned in particular in that context:

• Fatigue assessment of welded components based on structural stress and structural hot-spot stress approaches

• Detailed guidelines for determining relevant stresses and stress ranges

• Cycle counting proposals.

The fatigue assessment of welded components is part of paragraphs 18.6 “Stresses for fatigue assessment of welded components and regions” (determination of relevant stress ranges) and 18.10 “Fatigue strength of welded components” (relevant weld details and revised fatigue curves). The basic rules for unwelded components essentially remain unchanged. Stress analyses for clause 18 are usually based on detailed finite element analyses (FEA). As an essential amendment for the practical user the determination of structural stress ranges for the fatigue assessment of welds is further detailed in the new informative annex NA “Instructions for structural stress oriented finite elements analyses using brick and shell elements”. Here, different applicable methods for the determination of structural stresses are explained in connection with the requirements of the finite element models and analyses.

The cycle counting issue is comprehensively treated in the new informative annexes NB “Cycle Counting for a given Load History” and NC “Cycle Counting for Design Data Evaluation” including detailed proposals for implementation in an algorithmic programming framework.

Topics: Fatigue life
Commentary by Dr. Valentin Fuster
2014;():V001T01A090. doi:10.1115/PVP2014-28152.

Failure of welded structures due to the presence of flaws is typically driven by a mixture of applied and residual stresses, yet in most cases only the former are known accurately. In as-welded structures, a typical assumption is that the magnitude of welding residual stress is bounded by the room temperature yield strength of the parent material. The UK flaw assessment procedure BS 7910:2013 also assumes that mechanical loading (either as a result of proof testing or during the initial loading of an as-welded structure) will bring about a relaxation in residual stress. Conversely, the UK structural assessment code for nuclear structures, R6, contains a warning on the ‘limited validation’ of the BS 7910 approaches for stress relaxation and suggests that they should be used ‘with caution’. The aim of this study was therefore to review the basis of the BS 7910 clauses on stress relaxation with a view to harmonising the BS 7910 and R6 rules for cases in which the original welding residual stress distribution is not known.

The residual stress relaxation clauses of BS 7910:2013 date back to the 1991 edition of PD 6493 and have not changed substantially since then. A considerable programme of work was carried out by TWI at the time to justify and validate the clause, but the full underlying details of the work have not hitherto been available in the public domain, and are described in a separate companion paper. The approach proposed in BS 7910 combines ‘global’ relaxation of residual stress (Qm) under high mechanical load with ‘local’ enhancement of crack tip driving force through the adoption of a simplified primary/secondary stress interaction factor, ρ.

Commentary by Dr. Valentin Fuster
2014;():V001T01A091. doi:10.1115/PVP2014-28501.

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 and 2014, 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 new requirements 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
2014;():V001T01A092. doi:10.1115/PVP2014-28810.

2013 saw a number of significant issues of Standards and Guidelines associated with Bolted Joints and 2014 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 the recently published ASME PCC-1-2013. CEN have also been active in issuing new key documents: July 2013 saw the publication of the harmonized calculation standard. En1591-1. Late 2013 saw the training and competence standard EN1591-4 published. The paper will discuss 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

2014;():V001T01A093. doi:10.1115/PVP2014-28392.

The evaluation procedure for the reactor pressure vessel integrity of Japanese PWR plants against Pressurized Thermal Shock (PTS) events is prescribed in the Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” since 1991. The current procedure was developed based on the PTS verification test program, which was conducted as Japanese national project and the related studies in 1980’s.

Since much progress has been made on fracture mechanics, fracture toughness, in-service inspection techniques/results and so on, it is preferred to advance the current procedure for more credible evaluation by reflecting the latest knowledge.

This paper describes the outline of the studies to update the current procedure.

Commentary by Dr. Valentin Fuster
2014;():V001T01A094. doi:10.1115/PVP2014-28394.

When cracks are detected in piping in nuclear power plants during in-service inspections, the crack propagation is usually calculated using approximation formulas of stress intensity factor (SIF) provided in the ASME Code, the JSME Rules or the literature. However, when the crack is detected in complicated-shaped locations in components, finite element analysis (FEA) needs to be used to calculate the SIFs. Accordingly, a method of automatically conducting FEA for crack propagations in nuclear power plants is needed. Therefore, we, the Nuclear Regulation Authority (NRA, Japan) have developed an automatic 3D finite element crack propagation system (CRACK-FEM) for nuclear components. The developed CRACK-FEM uses three methods of SIF calculation: the Virtual Crack Extension Method (VCEM), the Virtual Crack Closure-Integral Method (VCCM) and the Domain Integral Method (DIM). Each method uses different meshes, so users can select a method which uses a suitable mesh for the problem. The software includes a geometry generator to create complicated weld models, and a mesh generator which can deal with interior boundaries formed between different materials. The functions and accuracy of the new software are demonstrated by solving several sample problems involving crack propagation.

The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Commentary by Dr. Valentin Fuster
2014;():V001T01A095. doi:10.1115/PVP2014-28602.

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement.

The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Commentary by Dr. Valentin Fuster
2014;():V001T01A096. doi:10.1115/PVP2014-28621.

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration.

In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Commentary by Dr. Valentin Fuster
2014;():V001T01A097. doi:10.1115/PVP2014-28645.

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

Commentary by Dr. Valentin Fuster
2014;():V001T01A098. doi:10.1115/PVP2014-28898.

The Master Curve approach for the fracture toughness evaluation is expected to be a powerful tool to ensure the reliability of long-term used RPV steels. In order to get sufficient number of data for the Master Curve approach coexistent with the present surveillance program for RPVs, the utilization of miniature specimens, which can be taken from broken halves of surveillance Charpy specimens, is important. CRIEPI developed the test technique for the miniature C(T) specimens (Mini-CT), whose dimensions are 4 × 10 × 10 mm, and verified the basic applicability of Master Curve approach by means of Mini-CT for the determination of fracture toughness of typical Japanese RPV steels. A round robin program is organized in order to assure the robustness of the testing procedure to the difference in testing machines or operators. The first and second round robin tests (PVP2012-78661 [1], PVP2013-97936 [2]) suggested that the reference temperature T0 evaluation technique by Mini-CT specimen potentially is fairly robust in regard to difference in testing machines and operators, and gives similar loading rate dependency to the larger C(T) specimens. As the final year of the round robin program, “blind tests” were carried out. Here, detailed material information such as the type of materials, estimated T0, existing fracture toughness data for the material, were not given with the specimen, and 6 organizations independently selected the test temperature based on Charpy full curve of the tested material. The selection of test temperature has the variation of −120 °C to −150 °C among the organizations. 8 to 20 specimens in a set were subjected to the Master Curve evaluation and all the 6 organizations successfully obtained valid T0. The scatter range in T0 was at most 16 °C, which was within the acceptable scatter range specified in ASTM E1921-10e1. The selection of test temperature seems to give limited effect as like as that in larger specimens.

Topics: Temperature
Commentary by Dr. Valentin Fuster
2014;():V001T01A099. doi:10.1115/PVP2014-29053.

This paper shows the technical basis of revision to the JSME Fitness-for-Service Code (the FFS code) for flaw evaluation methods of pipes which have a very shallow circumferential flaw. When a flaw in a pipe is very small, the allowable stress between the FFS code and the JSME Design and Construction Code (the design code) is mismatched. Fracture strength of a pipe with small flaw depends on the tensile strength accompanied by large deformation. Therefore, fracture mechanics is not applicable in such a case. This mismatch has been resolved for an axial crack assessment by improving definition of flow stress for shallow crack. In this study, the authors investigated this mismatch in the allowable stress in the flaw assessment for a pipe with a circumferential crack. Some past fracture test results of pipes showed that flawed pipes did not fracture at the flaw section when the circumferential flaw size was small and they failed by oval deformation or plastic buckling. Allowable stress for such behavior has been incorporated in some existing design codes as a restriction for plastic collapse. Through the reevaluation of the existing piping fracture test results, the applicability of fracture evaluation methods defined in the FFS code was examined for the case that flaw size was very small. As a result, the fracture evaluation method based on flow stress was found not to be applicable when flaw size was very small, and the failure criterion in this case depended on the collapse strength accompanying with ovalization. Revisions of the FFS code reflecting these examination results were proposed in this paper.

Commentary by Dr. Valentin Fuster

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

2014;():V001T01A100. doi:10.1115/PVP2014-28367.

Tsuruga unit-1, the first Japanese commercial BWR (BWR-II, 357MWe) started its operation in 1970, has a Mark-I type containment vessel and 3-loop reactor recirculation piping.

In the 33rd planned outage of 2011, most parts of the reactor recirculation piping were replaced to improve corrosion resistance and inspectability. This replacement work dealing with long pipe pieces had to be conducted in limiting condition and severe environment such as the narrow work space, only one small entrance hatch, and high dose rate. In addition, other maintenance works by four contractors were concurrently being carried out in the same area. Therefore, these work processes in the containment vessel were controlled carefully through close coordination with contractors and using advanced construction technique. As a result, it was achieved that efficient process which contributed to shortening work period and injury-free work in this challenging environment and condition.

Regarding radiation management, radiation dose rate was reduced substantially by applying chemical decontamination using ozone as chemical agent.

Experience obtained from this work is valuable in the current situation that many operators are trying to operate their aging nuclear plants continuously as long as economically viable.

Commentary by Dr. Valentin Fuster
2014;():V001T01A101. doi:10.1115/PVP2014-28470.

Since PWSCC has been observed in Alloy 600 used for butt welds between the low alloy steel RV-nozzles and stainless steel pipes in recent years, MHI has developed the inlay technology for outlet/inlet nozzles as a preventive maintenance and repair method. The inner surface, even if it has PWSCC, is exchanged from Alloy 600 to Alloy 690 which has excellent PWSCC resistance, by applying inlay.

This paper introduces the experiences of the MHI inlay System (RV-INLAY) and describes the merits as a counter measure against PWSCC. MHI has adopted the Cylindrical container for RV-INLAY. The Cylindrical container is installed into RV when the cavity is filled with water, then we can secure an atmospheric working space inside outlet/inlet nozzles. 3 or 4 manipulators are installed on the bottom of the container and carry out precise works like ambient temperature temper bead welding, UT, ECT, dimensional or visual inspection exchanging many kinds of tools (end-effectors) on the tip of the manipulators remotely. Heavy devices for machining, blast-decontamination or nozzle-plug handling are carried down into the container and set remotely using the special handling crane. Manipulators and other heavy devices move remotely to reduce the radiation exposure and simultaneously in 3 or 4 nozzles to shorten the working period. By using the Cylindrical container, we can directly access the inside outlet/inlet nozzles. This is different from the other preventive maintenance methods which need to access from outside of outlet/inlet nozzles. We can apply RV-INLAY from inside the nozzles without considering the difficulty in access due to the tight space and high radiation outside the nozzles. Furthermore when a RV is manufactured in a factory, we apply PT for the groove face before welding and the surface after welding. During RV-INLAY operation, we apply PT in the same manner because we can secure an atmospheric working space inside outlet/inlet nozzles. As a result, we can ensure the quality of welding same as manufacturing process in a shop. In addition, we don’t have to consider the crack propagation after RV-INLAY and we can perform ISI without preparing repair equipment against PWSCC. The Cylindrical container has a seal function same as the seal-plate used in usual outages, and it can separate the water inside the cavity from RCS. While the Cylindrical container is installed into RV, the RCS is drained and RCS piping works can be performed in parallel. In fact, we have shortened the outage duration by implementing RV-INLAY and RCS piping works in parallel. This is one reason why MHI has not selected under water welding but RV-INLAY.

Commentary by Dr. Valentin Fuster
2014;():V001T01A102. doi:10.1115/PVP2014-29060.

As a countermeasure against high residual stress, some residual stress improvement methods have been developed; Water Jet Peening (WJP) [1] [2] for components installed in water, Shot Peening by Ultrasonic-wave vibration (USP) for components installed in air, and outer surface irradiated Laser Stress Improvement Process (L-SIP)[3] for components that can only be accessed from the outer surface.

WJP is applied to Reactor Vessel (RV) outlet/inlet nozzle safe-end joints (Alloy600 weld metal), RV Bottom Mounted Instrument (BMI) inner surface and J-weld. Especially, it is difficult to apply the technology to BMI because BMI inner surface is a very narrow space (inner diameter; approximately 10–15mm) and BMI J-weld configuration is a complicated 3-dimensional form.

On the occasion of actual application, the verification tests have been carried out and checked that a stress improvement was effective as one of PWSCC mitigations. And the compressive stress induced by the WJP was verified to continue to exist under actual plant operation conditions.

Thus, in addition to replacing the material with Alloy 690, converting the residual stress to compressive one can prevent the occurrence of PWSCC.

Commentary by Dr. Valentin Fuster
2014;():V001T01A103. doi:10.1115/PVP2014-29062.

Seawater piping employed for cooling of emergency diesel generator and various components at Nuclear power plant, are internally lined for protection against seawater corrosion. (Fig. 1)

However, the lining materials such as rubber or polyethylene tend to incur peel-off or crack due to aged deterioration occurred by the operation, resulting in corrosion of the piping. (Fig. 2)

Internal piping integrity of seawater piping is usually performed by periodic visual inspection. But, for the 4B to 20B pipes it has been a challenge to detect the initial degradation because the inspector cannot get into the pipes.

Pinhole detection technology that enables detection of microscopic damages and cracks is available to use as means to detect its initial state of degradation, which seems effective as preventive maintenance of the lining.

Such being the case, we are developing Seawater Piping Inside Inspection Equipment applicable to the 4B to 20B sizes of pipes by evolving the conventional pinhole detection technology.

Commentary by Dr. Valentin Fuster

Codes and Standards: The James Farr Memorial Symposium on Structural Integrity of Pressure Components

2014;():V001T01A104. doi:10.1115/PVP2014-28004.

The rules on fitness for service for nuclear power plants of JSME are applied for flaw evaluation after detecting defects in the operating nuclear plants in Japan. The rules mainly focus on simple geometry such as straight pipes or vessels and do not provide the evaluation procedure for complex structures. The authors made a draft of flaw acceptance rule for J-groove weld of a bottom mounted instrumentation nozzle at application of the cap repair. The rule contains flaw modeling, fatigue and SCC crack growth calculation and flaw instability assessment. After detecting a defect on a J-groove weld, a flaw will be modeled in the whole of J-groove weld region because of fast SCC crack propagation in the weld region. Due to complex configuration of the evaluation location, FE analysis is needed for obtaining stress intensity factors (Ks) to calculate the crack growth and flaw instability. The proposed rule has a guidance for K calculation by FE analysis with the aim of decreasing dependence of individuals for calculation. The authors performed benchmark analyses to confirm the guidance applicability. The calculation results by three participants agreed within several percent.

Commentary by Dr. Valentin Fuster
2014;():V001T01A105. doi:10.1115/PVP2014-28032.

A previously-developed procedure applies to evaporator section of bottom-supported heat-recovery steam generators (HRSG): The procedure features methods to use on pressure internal parts designed according to Section I, as required by the Boiler & Pressure Vessel (B & PV) Code. It uses ASME Power Piping Code B31.1, which is for external piping, to evaluate additional concerns on pressure parts that have already met Section I. This mainly because other sections of B & PV code do not cover flexibility analysis: Procedure includes classical methods for the static analysis of the lower header and a computer-modeled flexibility analysis. They investigate the stability of the harp high-temperature components, which may be critical because of their longitudinal dimensions, thermal expansions and loading: Preliminary stress evaluation showed that Italian pressure formula is largely conservative.

Methods’ application to a worked example using both ASME and Italian pressure formulae shows now that the pressure contribution is the greatest. Also, the maximum stress found out on the header in sustained case appears consistent with that from numerical model, though bending contribution the greatest; Sustained case appears less critical than thermal. So, the design should be safe from risks of static instability due to longitudinal pressure stress: deformation occasionally observed in field on the lower headers may be correctly imputed to thermal expansion. Herein, a bottom-supported HRSG double-width unity of combined cycle power plant is considered as the worked example.

Commentary by Dr. Valentin Fuster
2014;():V001T01A106. doi:10.1115/PVP2014-28051.

API RP 581 gives detailed steps and data to help perform quantitative analyses for “Risk Based Inspection” of pressurized equipment. This includes probability of failure, consequence of failure and risk assessment. Unfortunately, the use of API RP 581 in its current form can sometimes lead to incorrect estimations of both probability and consequence of failure, leading to an incorrect risk assessment of the components/equipment being analyzed.

In this paper, we will highlight some of the issues that we ran across while performing an RBI analysis following the steps and methods prescribed in API RP 581, involving a level 1 consequence analysis, and suggest possible changes for future editions to remove/minimize the consequences of these issues.

Commentary by Dr. Valentin Fuster
2014;():V001T01A107. doi:10.1115/PVP2014-28067.

With some typical pressure vessel materials commonly used for fabrication of pressure retaining Code items and fracture mechanics assessment methods stipulated in Chinese standard GB/T19624, assessment for safety reserve factor for defect acceptance criteria stipulated in ASME Codes, some suggestion for revision to acceptance criteria has been pointed out.

Commentary by Dr. Valentin Fuster
2014;():V001T01A108. doi:10.1115/PVP2014-28081.

The ASME Codes and referenced standards provide industry and the public the necessary rules and guidance for the design, fabrication, inspection and pressure testing of pressure equipment. Codes and standards evolve as the underlying technologies, analytical capabilities, materials and joining methods or experiences of designers improve; sometimes competitive pressures may be a consideration. As an illustration, the design margin for unfired pressure vessels has decreased from 5:1 in the earliest ASME Code edition of the early 20th century to the present day margin of 3.5:1 in Section VIII Division 1. Design by analysis methods allow designers to use a 2.4:1 margin for Section VIII Division 2 pressure vessels.

Code prohibitions are meant to prevent unsafe use of materials, design methods or fabrication details. Codes also allow the use of designs that have proven themselves in service in so much as they are consistent with mandatory requirements and prohibitions of the Codes. The Codes advise users that not all aspects of construction activities are addressed and these should not be considered prohibited. Where prohibitions are specified, it may not be readily apparent why these prohibitions are specified. The use of “forged bar stock” is an example where use in pressure vessels and for certain components is prohibited by Codes and standards.

This paper examines the possible motive for applying this prohibition and whether there is continued technical merit in this prohibition, as presently defined. A potential reason for relaxing this prohibition is that current manufacturing quality and inspection methods may render a general prohibition overly conservative. A recommendation is made to better define the prohibition using a more measurable approach so that higher quality forged billets may be used for a wider range and size of pressure components.

Jurisdictions with a regulatory authority may find that the authority is rigorous and literal in applying Code provisions and prohibitions can be particularly difficult to accept when the underlying engineering principles are opaque. This puts designers and users in these jurisdictions at a technical and economic disadvantage.

This paper reviews the possible engineering considerations motivating these Code and standard prohibitions and proposes modifications to allow wider Code use of “high quality” forged billet material to reflect some user experiences.

Commentary by Dr. Valentin Fuster
2014;():V001T01A109. doi:10.1115/PVP2014-28093.

A lot of applications of elastic plastic FE analysis to flawed structural fracture behaviors of mode I have been investigated. On the other hand the analysis method has not been established for the case of the excessive cyclic torsion loading with mode II or III fracture. The authors tried simulating the fracture behavior of a cylinder-shaped specimen with a through-walled circumferential flaw subjected to excessive monotonic or cyclic loading by using elastic plastic FE analysis. Chaboche constitutive equation of the used FE code Abaqus was applied to estimate the elastic plastic cyclic behavior. As a result in the case of monotonic loading without crack extension, the relation of torque-rotation angle of the experiment was estimated well by the simulation. Also J-integral by the Abaqus’ function agreed with a simplified J-equation using the calculated torque-rotation angle relation. On the other hand under load controlled cyclic loading associated with ductile crack growth, the calculated torque-rotation angle relation did not agree with the experimental one because of high sensitivity of the used stress-strain curve. J-integral from Abaqus code did not increase regardless of the accumulated crack growth and plastic zone. Several simplified ΔJ calculations tried to explain the experimental ductile crack growth and it seemed that da/dNJ relation follows the Paris’ law. From these examinations an estimation procedure of the structures under excessive cyclic loading was proposed.

Commentary by Dr. Valentin Fuster
2014;():V001T01A110. doi:10.1115/PVP2014-28200.

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. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating the Structural Integrity of the RPVs in all operating modes, transients and accident conditions.

A large part of this demonstration consists of the Flaw Acceptability Assessment inspired by the ASME XI procedure but adapted to the nature and number of indications found in the Doel 3 and Tihange 2 RPVs. In particular, ASME XI IWB-3300 article requires combining closely spaced flaws in order to account for their mechanical interactions. However, it appeared early that the strict application of the current ASME XI proximity criteria for laminar flaws to the actual flaw indications found at Doel 3 led to unrealistic results and conclusions. Therefore, an alternative methodology to derive suitable characterization rules applicable to specific flaws observed at Doel 3 and Tihange 2 RPVs has been successfully developed, implemented and validated.

Commentary by Dr. Valentin Fuster
2014;():V001T01A111. doi:10.1115/PVP2014-28311.

In the early 2000s, ASME adopted Code Cases N-629 and N-631 [1–2], both of which permit the use of the Master Curve reference temperature (To) to define an reference temperature RTTo, as follows (in SI units, as are used throughout the paper):Display Formula

RTTo=To+19.4
The Code Cases state that “this reference temperature … may be used as an alternative to [the] indexing reference temperature RTNDTfor the KIcand KIatoughness curves, as applicable, in Appendix A and Appendix G [of Section XI of the ASME Code].” KIa is now only used in Appendix A. The functional form of the ASME KIc and KIa curves dictate that the temperature separation between them remains constant irrespective of the degree of neutron radiation embrittlement, as quantified by ΔRTNDT or ΔRTTo. However, data collected from the literature and new data reported by Hein et al. show that radiation embrittlement brings the KIc and KIa curves closer together as embrittlement increases. As a result, current Code guidance will not produce a bounding KIa curve in all situations when RTTo is used as an reference temperature. To reconcile this issue, this paper summarizes available data and, on that basis, concludes that use of the following reference temperature will ensure that the ASME KIa curve bounds currently available KIa data:Display Formula
RTKIa=RTTo-19.4+44.97×exp−0.00613×RTTo-19.4

Commentary by Dr. Valentin Fuster
2014;():V001T01A112. doi:10.1115/PVP2014-28347.

In recent years, Ultra supercritical power plant boiler in China is developing very rapidly, therefore, using the RBI risk evaluation method to carry out inspection of large power station boiler is very necessary. Operation practice shows that the membrane type water-wall is one of the main risk parts of large power plant boiler. Research has shown that typical failure area of membrane type water-wall is layer nearby the burner; Most of failure positions are in high heat flux region, the fins weld joint; Failure mechanism is comprehensive factors, such as, the welding deformation, weld defects and fins weld crack, foreign bodies blocked, blowing loss thinning, short-term over temperature overheating. On this basis, the application of RBI risk assessment method, analyzed the risk level and failure consequences of a power plant boiler membrane type water-wall, the risk matrix is obtained, proposed the corresponding test strategy, for the RBI technology application to evaluation of power plant boiler components provides a new method of failure risk.

Commentary by Dr. Valentin Fuster
2014;():V001T01A113. doi:10.1115/PVP2014-28540.

Section XI of the ASME Code provides models of the fracture toughness of ferritic steel. Recent efforts have been made to incorporate new information, such as the Code Cases that use the Master Curve, but the fracture toughness models in Section XI have, for the most part, remained unchanged since the KIc and KIa curves were first developed in Welding Research Council Bulletin 175 in 1972. Since 1972, considerable advancements to the state of knowledge, both theoretical and practical have occurred, particularly with regard to the amount of available data. For example, as part of the U.S. Nuclear Regulatory Commission’s (NRC’s) pressurized thermal shock (PTS) re-evaluation efforts the NRC and the industry jointly developed an integrated model that predicts the mean trends and scatter of the fracture toughness of ferritic steels throughout the temperature range from the lower shelf to the upper shelf. This collection of models was used by the NRC to establish the index temperature screening limits adopted in the Alternate PTS Rule documented in Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50.61a (10CFR50.61a). In this paper the predictions of the toughness models used by the ASME Code are compared with these newer models (that are based on considerably more data) to identify areas where the ASME Code could be improved. Such improvements include the following:

• On the lower shelf, the low-temperature asymptote of the KIc curve does not represent a lower bound to all available data.

• On the upper shelf, the de facto KIc limit of applicability of 220 MPa√m exceeds available data, especially after consideration of irradiation effects.

• The separation between the KIc and KIa curves depends on the amount of irradiation embrittlement, a functionality not captured by the ASME Section XI equations.

• The temperature above which upper shelf behavior can be expected depends on the amount of irradiation embrittlement, a functionality not captured in the ASME Section XI equations.

Commentary by Dr. Valentin Fuster
2014;():V001T01A114. doi:10.1115/PVP2014-28729.

Clauses UG-36 through UG-43 of ASME Section VIII Division 1 [1], describe the method of calculating the adequacy of compensation of openings in shells, using an area-replacement method. The method is based on determining and suitably replacing the missing metal area along any section, with metal available or provided, within the limits of reinforcement on the shell and nozzle.

Clause UG-36 (b) of ASME Section VIII Division 1 provides limits on the size of the opening for applicability of Clauses UG-36 through UG-43. If these limits are exceeded, supplemental rules of Clause 1-7 of Appendix 1 need to be complied with or alternatively the rules of Clause 1-10 of Appendix 1 may be applied.

The rules for large openings as stated in the Code are not dependent upon the absolute size of the nozzle and shell. For example, same calculations would be required to be carried out whether a nozzle of NPS 1 is attached to a shell of NPS 1.5 or a nozzle of NPS 16 is attached to a shell of NPS 24.

The work presented in this paper is an attempt to determine whether the additional calculations in Clause 1-7 need to be carried out for finished openings exceeding the limits of UG-36(b) irrespective of the absolute size of the nozzle and shell. This has been done by carrying out calculations for a wide range of nozzle-shell combinations and comparing the results so obtained with the results of a Finite Element Analysis.

Commentary by Dr. Valentin Fuster
2014;():V001T01A115. doi:10.1115/PVP2014-28913.

Inspections are widely used in the process industries to reduce risk related to failure on static mechanical equipment. Risk-based inspection is one of effective tools to optimize inspection and maintenance planning. Based on principle of the risk-based inspection methodology, the main failure modes and damage mechanisms of pressure vessels and pipelines in ethylene compression unit are identified. The risk assessment of pressure vessels and pipelines is carried out. All pressure vessels and pipelines in this unit are prioritized based on the level of risk. Risk mitigation measures and optimal inspection and maintenance strategy are proposed. The results of risk evaluation have highlighted a clear improvement in the quality of inspection and maintenance of the ethylene compression unit.

Commentary by Dr. Valentin Fuster
2014;():V001T01A116. doi:10.1115/PVP2014-29017.

The objective of this paper is to use finite element analysis in determining stresses in a Vortex Finder. Vortex Finder is tube projecting into central vortex of hydro cyclone or dense medium cyclone through which the classified fines or lighter specific gravity fraction of pulp leaves the system. Two Loading Conditions are considered. The stresses determined for the loading conditions are compared with allowable stresses for the material used for the Vortex Finder. Modifications are made to the Vortex Finder and in that also several design modifications are considered. Case 1 Original Design. Case 2 Modified model add eight Additional support connections and eight guests. Case 3 Same as Case 2 Double Elements to see effect on stresses. Case 4 Vortex Finder Modified Double Elements Air Velocity 100 ft. /sec. Case 5 Vortex Finder Modified Double Elements Change in Air Velocity 100 feet/sec. Add 6 support connections and guests removed. Case 6 Vortex Finder Modified Double Elements Air velocity 100 feet per sec. Add six more support connections and gussets Remove gussets move support points closer. For all the cases the stresses were compared with the allowable stresses. It was determined Case 6 with 24 total supports was the best. No Gussets. Support points moved closure. The maximum stress obtained by FEA Model meets the allowable stresses. This is robust design modification. This was implemented.

Commentary by Dr. Valentin Fuster
2014;():V001T01A117. doi:10.1115/PVP2014-29082.

Welding is a reliable and efficient joining process in which the coalescence of metals is achieved by fusion. Welding is carried out with a very complex thermal cycle which results in irreversible elastic-plastic deformation and residual stresses in and around fusion zone and heat affected zone (HAZ). A residual stress due to welding arises from the differential heating of the plates due to the weld heat source. Residual stresses may be an advantage or disadvantage in structural components depending on their nature and magnitude. The beneficial effect of these compressive stresses have been widely used in industry as these are believed to increase fatigue strength of the component and reduce stress corrosion cracking and brittle fracture. But due to the presence of residual stresses in and around the weld zone the strength and life of the component is also reduced. To understand the behavior of residual stresses, two 10 mm thick Fe410WC mild steel plates are butt welded using the Metal Active Gas (MAG) process. An experimental method (X-ray diffraction) and numerical analysis (finite element analysis) were then carried out to calculate the residual stress values in the welded plates. Three types of V-butt weld joint — two-pass, three-pass and four-pass were considered in this study. In multi-pass welding operation the residual stress pattern developed in the material changes with each weld pass. In X-ray diffraction method, the residual stresses were derived from the elastic strain measurements using a Young’s modulus value of 210 GPa and Poisson’s ratio of 0.3. Finite element method based, SolidWorks software was used to develop coupled thermal-mechanical three dimension finite element model. The finite element model was evaluated for the transient temperatures and residual stresses during welding. Also variations of the physical and mechanical properties of material with the temperature were taken into account. The numerical results for peak transverse residual stresses attained in the welded plates for two-pass, three-pass and four-pass welded joint were 67.7 N/mm2, 58.6 N/mm2, and 48.1 N/mm2 respectively. The peak temperature attained during welding process comes out to be 970°C for two-pass weld, 820.8°C for three-pass weld and 651.9°C for four-pass weld. It can be concluded that due to increase in the number of passes during welding process or deposition weld beads, the residual stresses and temperature distribution decrease. Also, the results obtained by finite element method agree well with those from experimental X-ray diffraction method.

Commentary by Dr. Valentin Fuster
2014;():V001T01A118. doi:10.1115/PVP2014-29083.

Welding is a reliable and efficient joining process in which the coalescence of metals is achieved by fusion. Welding is carried out with a very complex thermal cycle which results in irreversible elastic-plastic deformation and residual stresses in and around fusion zone and heat affected zone (HAZ). Due to the presence of residual stresses in and around the weld zone the strength and life of the component is largely reduced. An application of corner joint of square hollow section (IS 4923:1997 YSt 240) of 49.5mm × 49.5mm × 2.0mm in size is welded by the Metal Active Gas (MAG) process. To know the failure behavior of square hollow section corner welded joint various load test including bending test and shear test were performed on universal testing machine (UTM). Finite element based software SolidWorks Simulation was used to model the joint and for calculation of load stresses. Residual stress value was 512MPa and 516MPa, which is in good agreement with experimental, results (X-ray diffraction (XRD) technique). Welding parameters were optimized further using finite element model in order to reduce the residual stress. Failure behavior shows that welding has good strength both in bending and shearing test and small displacements at maximum applied load. By reduction in residual stress, displacement decreases five percentages in bending test and negligible change occurs in shear test.

Commentary by Dr. Valentin Fuster

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