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ASME Conference Presenter Attendance Policy and Archival Proceedings

2016;():V01BT00A001. doi:10.1115/PVP2016-NS1B.
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This online compilation of papers from the ASME 2016 Pressure Vessels and Piping Conference (PVP2016) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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: Integrity Issues for Buried Pipe

2016;():V01BT01A001. doi:10.1115/PVP2016-63923.

Nuclear power plants and many other industries are required to periodically inspect their buried piping to determine its fitness-for-continued service (FFS). The FFS process requires that both the general corrosion rate and the rate of maximum penetration for localized corrosion (e.g., pitting) be estimated so that the remaining lifetime and/or time until the next inspection can be determined. Revision 1 to ASME Code Case N-806, “Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a Back-Filled Trench” [1] provides 4 options for estimating the corrosion rates:

a. Wall thickness measurements from the current examination and from one or more previous examinations of the same metal loss region.

b. Repeat measurements at two or more times from another location that has a predicted metal loss rate greater than or equal to the rate of the metal loss region under evaluation.

c. Repeat measurements using corrosion coupons, linear polarization probes, or electrical resistance probes

d. Generic historical data

Each of these methods has its uses and limitations, and it is generally preferable to consider results from 2 or more of the methods.

This paper examines historical data gathered by the National Bureau of Standards (NBS, renamed in 1988 as the National Institute of Standards & Technology - NIST) at ∼ 70 locations around the US in the 1930s – 1950s. Maximum penetration and weight loss (general corrosion) data from each site were placed in one of four soil texture groups for both carbon steel and cast iron. A regression analysis was performed to determine the median rates and 80% and 95% probabilistic values. It was found that results within each soil texture group were relatively similar and that the corrosion rates in the first 3 years after burial tended to be much higher than rates in years 5–18. The coefficients of determination were determined to quantify differences within each soil texture group.

It is proposed that the steady state rates provided herein are an option to be used as the Historical Rates for FFS evaluations as described in [1].

Commentary by Dr. Valentin Fuster
2016;():V01BT01A002. doi:10.1115/PVP2016-63973.

Non-planar flaw such as local wall thinning flaw is a major piping degradation in nuclear power plants. Hundreds of piping components are inspected and evaluated for pipe wall loss due to flow accelerated corrosion and microbiological corrosion during a typical scheduled refueling outage. The evaluation is typically based on the original code rules for design and construction, and so often that uniformly thin pipe cross section is conservatively assumed. Code Case N-597-2 of ASME B&PV, Section XI Code provides a simplified methodology for local pipe wall thinning evaluation to meet the construction Code requirements for pressure and moment loading. However, it is desirable to develop a methodology for evaluating non-planar flaws that consistent with the Section XI flaw evaluation methodology for operating plants. From the results of recent studies and experimental data, it is reasonable to suggest that the Section XI, Appendix C net section collapse load approach can be used for non-planar flaws in carbon steel piping with an appropriate load multiplier factor. Local strain at non-planar flaws in carbon steel piping may reach a strain instability prior to net section collapse. As load increase, necking starting at onset strain instability leads to crack initiation, coalescence and fracture. Thus, by limiting local strain to material onset strain instability, a load multiplier factor can be developed for evaluating non-planar flaws in carbon steel piping using limit load methodology. In this paper, onset strain instability, which is material strain at the ultimate stress from available tensile test data, is correlated with the material minimum specified elongation for developing a load factor of non-planar flaws in various carbon steel piping subjected to multiaxial loading.

Topics: Carbon steel , Stress , Pipes
Commentary by Dr. Valentin Fuster
2016;():V01BT01A003. doi:10.1115/PVP2016-64025.

Because there is no closed-form solution to predict ovalization of buried pipe with an uneven pattern of wall loss that is subjected to soil and surface loads, a conservative screening approach has been to assume a uniform wall loss around the pipe circumference, with a wall thickness equal to the thinnest, most corroded spot. To improve on this grossly conservative approach, this paper numerically studies the effects of uneven corrosion patterns around the circumference on the ovality of the pipe, using finite element simulation of the corroded pipe and surrounding soil. A second topic addressed in this paper is the effect of trench fill material on the ovality of the pipe, in the case where the trench fill has different soil moduli, caused by variations in compaction or materials, including controlled low-strength material (CLSM).

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

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

2016;():V01BT01A004. doi:10.1115/PVP2016-63486.

When conducting structural integrity assessments for reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, the stress intensity factor (SIF) is evaluated for a surface crack which is postulated near the inner surface of RPVs. It is known that cladding made of the stainless steel is a ductile material which is overlay-welded on the inner surface of RPVs for corrosion protection. Therefore, the plasticity of cladding should be considered in the SIF evaluation for a postulated underclad crack. In our previous study, we performed three-dimensional (3D) elastic and elastic-plastic finite element analyses (FEAs) for underclad cracks during PTS transients and discussed the conservatism of a plasticity correction method prescribed in the French code. In this study, additional FEAs were performed to further investigate the plasticity correction on SIF evaluation for underclad cracks. Based on the 3D FEA results, a new plasticity correction method was proposed for Japanese RPVs subjected to PTS events. In addition, the applicability of the new method was verified by studying the effects of the RPV geometry, cladding thickness and loading conditions. Finally, it is concluded that the newly proposed plasticity correction method can provide a more rational evaluation with a margin to some extent on SIFs of underclad cracks in Japanese three-loop RPVs.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A005. doi:10.1115/PVP2016-63821.

The applicability of a stress intensity factor (SIF) solution for a flat plate with a surface flaw (flat plate model) to nozzle crotch corner flaw is investigated by using a 3-D finite element method (FEM). Flaws due to stress corrosion cracking (SCC), fatigue, etc. are frequently postulated or detected at discontinuity regions. To evaluate the structural integrity of flawed components, it is important to calculate fracture mechanics parameters rationally and appropriately. SIF is one of the key fracture mechanics parameters for linear elastic fracture mechanics evaluation. Therefore, almost all fitness-for-service codes, e.g. ASEM BPVC Section XI, JSME FFS Code and R6 provide SIF solutions. However, they essentially provide SIF solutions only for flat plates or cylinders with flaws. Since nozzle crotch corner suffers from stress concentration and there are a lot of nozzle geometry parameters, there is no direct SIF solution for arbitrary nozzle geometries. To estimate SIFs for nozzle crotch corner flaws rationally and appropriately, the applicability of the SIF solution for the flat plate model needs to be verified. SIFs were calculated by using 3-D FEM for a flawed nozzle, and the results were compared with those calculated by equations and coefficients in the code solution. Our comparison showed that the flat plate model is applicable to nozzle crotch corner flaws rationally and conservatively.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A006. doi:10.1115/PVP2016-63823.

In the nuclear industry, demands on the structural integrity reliability of metal components are always increasing. The quantification of allowable defects in pressure vessels should therefore draw on advanced structural integrity assessment procedures. In the UK, R6 [1] is the main procedures used for defect tolerance assessment (DTA). In this paper, the overall evaluation procedure of DTA using R6 applied to the Main Steam (MS) nozzle crotch corner of the Advanced Boiling Water Reactor (ABWR) is presented. At the nozzle crotch corner region, high stresses, including through-wall bending stresses from the local structural discontinuity, were present. These bending stresses have been categorised as secondary. R6 conservatively implies such bending stresses may need to be categorised as primary, to allow for the possibility of elastic follow-up. To support application as a secondary stress, an elastic-plastic finite element analysis has been performed to evaluate the J-integral for the nozzle crotch corner. The resulting values of J, when compared to the stress intensity factor and collapse solutions used for the assessment, showed that treating the bending stress as secondary maintained sufficient margin, indicating conservatism. Finally, the DTA results of the nozzle crotch corner are presented to determine the defect tolerance criteria. This includes calculating the limiting defect size at the start of plant life when considering the end of life critical defect size and through life Fatigue Crack Growth (FCG).

Commentary by Dr. Valentin Fuster
2016;():V01BT01A007. doi:10.1115/PVP2016-63971.

Baffle Former Bolt (BFB) is a fastening part of Reactor Vessel Internals (RVI) of PWR. BFB is made of type 347 or 316CW (cold work) stainless steel and it is known to have the risk of cracking caused by Irradiation Assisted Stress Corrosion Cracking (IASCC) under high neutron flux and tensile stress. To evaluate the time to crack of BFB, BFB’s time-dependent stress change caused by irradiation creep (relaxation) and by the swelling deformation of a baffle structure should be obtained. The authors have developed the finite element (FE) analysis method to calculate time-dependent stress of BFB considering the irradiation effects. The method combines two kinds of models; “global model” to calculate the deformation of whole baffle structure and “local model” to calculate the peak stress at the stress concentrated area under the bolt head. Incorporating the above calculation method, a new BFB inspection and evaluation guideline has been established in Japan. The concept of the guideline is also outlined in the paper.

Commentary by Dr. Valentin Fuster

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

2016;():V01BT01A008. doi:10.1115/PVP2016-63429.

Multiple flaws are often observed in engineering structures and components. Interactions between multiple flaws can increase the stress intensity factors. Although fitness-for-service codes provide combination rules to account for these effects, they do not unify the criteria for flaw combinations. To establish a reasonable combination rule, we must identify the parameters that dominate the interaction. This study investigates the interactions of stress intensity factors in two coplanar subsurface flaws in a plate. The plate was subjected to a remote tension acting normal to the flaw surface. For varying shapes and distances of elliptical subsurface flaws, we solved the intensity factors by finite element analyses. Flaw distance and area were identified as important parameters for characterizing the interaction factor between two subsurface flaws.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A009. doi:10.1115/PVP2016-63483.

When multiple flaws are detected in structural components, remaining lives of the components are estimated by fatigue flaw growth calculations using combination rules in fitness-for-service codes. ASME, BS7910 and FITNET Codes provide different combination rules. Fatigue flaw growth for adjacent surface flaws in a pipe subjected to cyclic tensile stress were obtained by numerical calculations using these different combination rules. In addition, fatigue lives taking into account interaction effect between the two flaws were conducted by extended finite element method (X-FEM). As the calculation results, it is found that the fatigue lives calculated by the X-FEM are close to those by the ASME Code. Finally, it is worth noticing that the combination rule provided by the ASME Code is appropriate for fatigue flaw growth calculations.

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2016;():V01BT01A010. doi:10.1115/PVP2016-63768.

If a subsurface flaw is located near a component surface, the subsurface flaw is transformed to a surface flaw in accordance with a subsurface-to-surface flaw proximity rule.

The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes in different countries. However, the specific criteria of the recharacterizations are different among the FFS codes.

The authors have proposed a new subsurface-to-surface flaw proximity rule based on experimental data and equivalent fatigue crack growth rates.

Recently, the authors have highlighted through numerous fatigue crack growth calculations that, on one hand, the proximity rule provided in the current ASME Boiler and Pressure Vessel Code Section XI (ASME Code Section XI) can provide non conservative fatigue lives for thin wall components like pipes and, on the other hand, for thick wall components like vessels, the current proximity rule and the proposed one provide relatively similar fatigue lives.

It appears therefore that the flaw-to-surface factor should be updated according to the thickness of the component or according to the type of component i.e. pipe or vessel.

In this study, fatigue crack growth calculations were carried out on additional flaw configurations in thick wall pipes and thin wall vessels in order define the best limit for the thickness-dependence of the fatigue lives.

Finally, a new subsurface to surface proximity rule depending on the thickness of the component is proposed.

Topics: Pipes , Vessels
Commentary by Dr. Valentin Fuster

Codes and Standards: Master Curve Fracture Toughness and Other Small Specimen Mechanical Properties

2016;():V01BT01A011. doi:10.1115/PVP2016-63078.

Cleavage fracture initiates usually at single locations in front of the fatigue crack in some position along the crack front. If the crack driving force along the crack front is uniform, one should expect the initiation sites to be randomly located along the crack front. Finite element analyses have, however, shown that the crack driving force varies along the crack front. Thus, the location of the cleavage initiation sites should reflect this variation in crack driving force.

Fracture toughness specimens differ both in geometry and size. Also, the specimens may be side grooved or plane sided. All this can be expected to affect the local crack driving force along the crack front. The local crack driving force for cleavage fracture initiation can be divided into two components. The local KJ value describes the local effective stress intensity, whereas Q or Tstress describes the local constraint. To make things even more complicated, the local constraint is also affected by the local effective stress intensity. All of these are also affected by any ductile tearing occurring prior to cleavage initiation.

The testing standards contain specific limitations on specimen sizes and their measuring capacity in order to ensure that the crack driving force in different specimens is sufficiently similar to make the results from different specimen types and sizes comparable.

Classically, the fracture toughness test specimens have been comparatively large. Recently more and more work has been devoted to diminish the size of the specimens, to save material. One very promising specimen type is the miniature C(T) specimen with a 4 mm thickness and total height of 10 mm. Based on a recent international round-robin, the miniature C(T) specimen appears to provide compatible Master Curve T0 values as large specimens, but further validation regarding the similitude of the cleavage initiation is required, since the Master Curve is based on the assumption that specimen size does not affect this similitude.

In this work, the location of cleavage initiation sites along the crack front are examined for different size and type of fracture toughness specimens, focusing on the miniature C(T) specimen. The location distributions are evaluated in terms of load level, specimen type, size and possible side grooving. It is shown that, as long as the standard requirements are fulfilled, the initiation location distributions for the miniature C(T) specimens are similar to larger conventional specimens. Side grooving is shown to have a minor effect on the initiator locations.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A012. doi:10.1115/PVP2016-63440.

In the recent years, small punch testing (SPT) techniques has made great progress in China. The SPT was studied to estimate the tensile properties, the fracture toughness and ductile-to brittle transition temperature, and the creep behavior. In 2012, a standard of small punch testing was issued in China and the application of SPT in power generation and petrochemical industry has become a prime candidate. The present paper concentrates on progress of technique and standardization and industrial acceptance in assessing the structure integrity in China. China has carried out close cooperation with Material & Metallurgical research Ltd in Czech to compare Standards between China and EU code. The size of specimen and jig was researched and compared each other. The influence of jig and test machine was researched and improved the specific requirement of jig and test machine. The evolution of stress state of deformation process of specimen in SPT was clarified. The results showed that at initial stage the elastic bending stress is predominant and then the stress state dominated by membrane stress with the decreasing of elastic bending stress and the increasing of punch displacement. The reason for introducing the specimen thickness h2 to the equation for correlating yield load of SPT with yield strength, and the reason for introducing specimen thickness h to equation for correlating the maximum load of SPT with tensile strength were provided respectively. The correlation equation of ductile to brittle transition temperature and SPT energy transition temperature TSP was established and it was successfully used to evaluate embrittlement of hydrogenation reactor. Small punch creep testing by reverse finite element simulation was carried out and used to evaluate the creep life in power generation industry. Fracture toughness and Master curve using SPT by reverse finite element simulation combined with local approach was studied.

Topics: Testing , China
Commentary by Dr. Valentin Fuster
2016;():V01BT01A013. doi:10.1115/PVP2016-63455.

Based on the good experiences gained by using small specimens made of ferritic RPV materials, the Master Curve fracture toughness approach was applied to determine the fracture mechanical properties of oxide dispersion strengthened (ODS-) materials. A ferritic ODS-alloy (Fe-14Cr-1W-Ti-Y2O3) has been produced through the powder metallurgical production path via hot extrusion and hot isostatic pressing (HIP). Optimized oxide dispersion strengthened (ODS)-alloys have a promising potential to meet the foreseen requirements of components in future Gen IV power plants due to their high creep strength and swelling resistance under irradiation at elevated operational temperatures. The fracture toughness was characterized with mini 0.2T C(T) specimens in different material orientations (R-L / L-R) in the ductile-brittle and upper shelf region in the un-irradiated state, accounting especially for the ODS-material’s anisotropy as one key effect of manufacturing. Despite all tests were performed in orientation required by ASTM standards E 1921 and E 1820 not all validity criteria (e.g. height of yield strength, evenness of the crack, admissible K during testing or admissible stable crack growth) were met by the ODS-material: consequently, a valid T0 value and a standard-compliant Master Curve could not be determined for the ODS-material in the transition region especially in the respective R-L orientation, also due to a comparably low fracture toughness over the whole evaluated temperature range. Promising fracture toughness properties were obtained in the crack growth direction perpendicular to the prior main deformation (extrusion) direction, where a KJQ value of 196 MPa√m at T = 22°C was measured. Within the ductile regime, only a JQ = J0.2BL technical initiation toughness value could be calculated and at T = 22°C, a comparably large JQ of 137kJ/m2 is obtained for specimens with crack growth direction perpendicular to the extrusion direction, while in extrusion direction the toughness is again low.

In addition two further ODS-materials (14YWT and PM2000) were tested and compared to the alloys above. Non-conformances of ODS relating to the material requirements in ASTM standards E1921 and E1820 were finally detected and explained.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A014. doi:10.1115/PVP2016-63607.

The tendency to reduce specimen size for fracture toughness characterization of structural materials is gaining a spectacular interest in particular for irradiated materials. Indeed, the miniaturized compact tension (mini-CT) with a size of 10×10×4.2 mm is becoming very popular. With such a small volume, a large number of mini-CT specimens can be extracted from broken Charpy impact specimens and therefore makes this geometry very attractive and consequently several round robin exercises are organized to qualify this geometry. SCK•CEN has gained a lot of experience since the first usage of this geometry more than a decade ago. This geometry was qualified and tested in both unirradiated as well as irradiated condition.

This paper overviews the SCK•CEN experience using the mini-CT geometry for fracture toughness characterization. In particular, it discusses the reliability of this geometry in comparison to large specimens such as the precracked Charpy and compares the advantages and limitations of such a geometry to derive some recommendations.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A015. doi:10.1115/PVP2016-63615.

In this study the analysis of fracture toughness test data has been performed in terms of estimation of the proper T0 value for several WWER-1000 RPV materials in unirradiated condition. The surveillance test data for the standard and reconstituted specimens were included in the analysis. It was found that a reference temperature T0 for reconstituted specimens is 31°C higher on average in comparison to the standard specimens. The possible reason is a high level of the stress intensity factor Kmax during the cycle at the stage of completion of crack tip sharpening for standard specimens.

Furthermore, the Charpy impact and fracture toughness test data for standard and reconstituted specimens have been compared considering the known relationship between the reference temperature T0 and the transition temperature T28J which corresponds to the Charpy energy level of 28 J.

Another objective of this study was to compare the RPV metal embrittlement rate for the two reactor pressure vessels using surveillance test data from standard and reconstituted fracture toughness specimens. The analysis has shown that test data for the reconstituted specimens is consistent with the test data for the standard specimens with regard to the embrittlement rate.

Topics: Metals , Embrittlement
Commentary by Dr. Valentin Fuster
2016;():V01BT01A016. doi:10.1115/PVP2016-63762.

The Master Curve approach is a powerful tool to evaluate material-specific fracture toughness of ferritic steels, such as RPV steels, using a limited number of specimens. However, preparing a sufficient number of standard fracture toughness test specimens is difficult for irradiated RPV steels of existing surveillance programs. Utilization of miniature specimens that can be machined from broken halves of standard Charpy specimens is a possible solution to address this issue. CRIEPI has been working on the test technique utilizing a miniature C(T) (Mini-C(T)) specimens, whose dimensions are 4 × 10 × 9.6 mm (0.16 inch thickness specimen). The basic applicability of the Mini-C(T) Master Curve approach has been confirmed [1] for the base metals of typical Japanese RPV steels. International round robin tests confirmed the reproducibility of fracture toughness data obtained by Mini-C(T) specimens [2–4]. Ensuring the applicability of the Mini-C(T) Master Curve approach to weld metals and heat affected zone materials is of great importance to meet the future demand from the RPV surveillance programs for over 40 or 60 years’ reactor operation. For a weld metal deposit, we verified that valid reference temperature, To, can be estimated using the Mini-C(T) specimens and the statistics of the fracture toughness data [5] show good conformity to the assumption of the Master Curve method [6].

In the present paper, fracture toughness of a weld joint, which consists of two different heats of RPV plate material was examined. Five sets of Mini-C(T) specimens taken from two base metals, their heat affected zones (HAZ) and weld metal deposit, were subjected to the fracture toughness test. 0.5T-C(T) specimens taken from similar locations were also subjected to the fracture toughness tests to investigate specimen size effect. All the Mini-C(T) data sets taken from base metal, HAZ and weld metal were eligible for the determination of valid To with each 12 to 16 Mini-C(T) specimens. The relevance of the specimen size correction in the Master Curve method was confirmed for two base metals and a weld metal. The fracture toughness data for HAZ materials gave a reasonable agreement with the specific Weibull distribution assumed in the Master Curve method. Nevertheless, To values of four data sets of HAZ materials, including two Mini-C(T) datasets and two 0.5T-C(T) datasets, showed larger variation than that of the base metals or the weld metal. The crack initiation sites of HAZ specimens were all within so-called fine grain HAZ. However the HAZ width near the crack initiation site was dependent on the individual specimens. Higher fracture toughness tended to be gained from the specimens with narrower HAZ width. The resulting To values for HAZ material were close to or lower than that for base metals. The results suggest that the HAZ material gives equivalent or higher fracture toughness than in base metals.

Topics: Heat , Metals
Commentary by Dr. Valentin Fuster
2016;():V01BT01A017. doi:10.1115/PVP2016-63795.

The effect of warm pre-stressing (WPS) on fracture toughness was evaluated for a reactor pressure vessel steel. Various types of thermomechanical loadings were applied to 1T-CT specimens. The results were compared with predictions from several analytical WPS engineering models. The specimen size effect was subsequently investigated under the load-unload-cool-fracture transient condition using 1T-, 0.4T-, and 0.16T-CT specimens. Analyses of the plastic zone distribution and residual stress were conducted to identify the difference in the WPS effect among the specimens.

Commentary by Dr. Valentin Fuster

Codes and Standards: Probabilistic and Risk-Informed Methods for Structural Integrity Assessment

2016;():V01BT01A018. doi:10.1115/PVP2016-63141.

Fatigue curves are receiving nowadays an increased level of attention in the wake of experimental campaigns showing that the original ASME III mean air curve, also known as the Langer curve [1], does not represent accurately part of the recently obtained laboratory data.

EDF, VTT and E.ON have been working towards a relevant fatigue assessment strategy. The three organizations recently exchanged HCF databases, providing a common benchmark to test and compare the various analysis methods. Following the 2014 PVP paper [2], several statistical approaches are being investigated. A special focus is given to methods able to properly account for run-out data points, which do not have the same statistical significance as failed data points.

Besides, it has to be noted that only a limited number of material grades are used in NPP primary loop components and in each case, the material batches are identified and specified in detail. Therefore, a more accurate and relevant fatigue assessment might be obtained by splitting large datasets that generally mix various testing conditions and material grades. A comparison is made between a “mixed” approach and a “separated” one, in which the fatigue assessment is performed successively for two or more subsets, e.g. associated with two testing temperature ranges and/or steel grades. Both “mixed” and “separated” strategies are applied to the EDF and the E.ON databases containing fatigue data in different temperatures for non-stabilized and stabilized austenitic stainless steels. The resulting data scatters are compared and the significance of these statistical approaches to fatigue assessment is discussed.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A019. doi:10.1115/PVP2016-63622.

In order to address the risks associated with the operation of ageing pressure boundary components, many assessments incorporate probabilistic analysis methodologies for alleviating excessive conservatism of deterministic methodologies. In general, deterministic techniques utilize conservative upper bound values for all critical parameters. Equally, defense-in-depth assessments for the nuclear industry employ probabilistic methods in order to estimate potential risks associated with unanticipated events to demonstrate adequate margins associated with the licensed activity.

Probabilistic approaches typically invoke the Monte-Carlo (MC) approach where a set of critical input variables, assumed independent, are randomly distributed and inserted in deterministic computer models. Estimates of results from probabilistic structural integrity assessments are then compared against assessment criteria, at times, based on the assumption that these results follow normal distributions. However, this assumption is not always valid, as normality depends both on the initially assumed distributions of the input variables and linearity, or lack thereof, of the deterministic model. In particular, the characteristic of a system function (either a linear or a non-linear system function) and the sampling region of input parameters affect the level of normality of the MC simulation results.

As a proof of principle, a specific case study is presented. A system function is chosen based on the steady-state thermal creep of Zr-2.5Nb Pressure Tube (PT), instead of a full deterministic computational model, to show whether it can give rise to MC results that deviate from normality. The consequence of the deviation from normality when compared against assessment criteria is briefly discussed. It is noted that this study does not deal with analysis of Probabilistic Safety Assessments, also known as PSAs.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A020. doi:10.1115/PVP2016-63655.

For the assurance of fitness-for-service of CANDU Pressure Tubes (PTs), guidelines and acceptance criteria are provided in Canadian Standard Association (CSA) N285.8-15, Technical requirements for in-service evaluation of zirconium alloy pressure Tubes in CANDU reactors. With respect to the assessment of risk of operation associated with degradation mechanisms and aging of the PTs in the entire core of a given reactor Unit, Clause 7 of CSA N285.8 allows Licensee’s to use either a deterministic or probabilistic method to assess the likelihood of PT failures. When a probabilistic method is used, the Licensee is obligated to demonstrate that the combined frequency of PT failure(s) over the evaluation period, due to the various potential degradation mechanisms, is less than the maximum acceptable frequency provided in Table C.1 of CSA N285.8-15.

The maximum acceptable frequency provided in Table C.1 of CSA N285.8-15 was developed in the early-1990’s based on reactor operating experience and knowledge at that time, Station Siting Guides and Consultative Regulatory Guide C-006 (Revision 1). A task group was established by the CSA N285.8 Technical Steering Committee to re-evaluate the allowable failure frequencies to confirm that they remain relevant given the current state of knowledge and the additional evaluation tools available. This paper provides Canadian Nuclear Safety Commission staff views regarding the technical basis for revisions to the allowable frequencies based upon current industry practices in conducting probabilistic core assessments.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in ASME Codes and Standards

2016;():V01BT01A021. doi:10.1115/PVP2016-63559.

Proposals for high temperature design methods have been developed for primary loads, creep-fatigue and strain limits. The methodologies rely on a common basis and assumption, that elastic, perfectly plastic analysis based on appropriate properties reflects the ability of loads and stress to redistribute for steady and cyclic loading for high temperature as well as for conventional design. The cyclic load design analyses rely on a further key property, that a cyclic elastic-plastic solution provides an upper bound to displacements, strains and local damage rates. The primary load analysis ensures that the design load is in equilibrium with the code allowable stress, taking into account: i) The stress state dependent (multi-axial) rupture criterion, ii) The limit to stress re-distribution defined by the material creep law. The creep-fatigue analysis is focused on the cyclic creep damage calculation, and uses conventional fatigue and creep-fatigue damage calculations. It uses a temperature-dependent pseudo “yield” stress defined by the material yield and rupture data to identify cycles which will not cause creep damage > 1 for the selected life. Similarly the strain limits analysis bounds cyclic strain accumulation. It also uses a temperature-dependent pseudo “yield” stress defined by the material yield and creep strain accumulation data to identify cycles which will not cause average (membrane) inelastic strain > 1% for the design life. The paper gives an overview of the background and justification of these statements, and examples.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A022. doi:10.1115/PVP2016-63730.

The goal of the Elastic-Perfectly Plastic (EPP) combined integrated creep-fatigue damage evaluation approach is to incorporate a Simplified Model Test (SMT) data based approach for creep-fatigue damage evaluation into the EPP methodology to avoid the separate evaluation of creep and fatigue damage and eliminate the requirement for stress classification in current methods; thus greatly simplifying evaluation of elevated temperature cyclic service.

The EPP methodology is based on the idea that creep damage and strain accumulation can be bounded by a properly chosen “pseudo” yield strength used in an elastic-perfectly plastic analysis, thus avoiding the need for stress classification. The original SMT approach is based on the use of elastic analysis. The experimental data, cycles to failure, is correlated using the elastically calculated strain range in the test specimen and the corresponding component strain is also calculated elastically. The advantage of this approach is that it is no longer necessary to use the damage interaction, or D-diagram, because the damage due to the combined effects of creep and fatigue are accounted in the test data by means of a specimen that is designed to replicate or bound the stress and strain redistribution that occurs in actual components when loaded in the creep regime.

The reference approach to combining the two methodologies and the corresponding uncertainties and validation plans are presented. Results from recent key feature tests are discussed to illustrate the applicability of the EPP methodology and the behavior of materials at elevated temperature when undergoing stress and strain redistribution due to plasticity and creep.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Chinese Codes and Standards

2016;():V01BT01A023. doi:10.1115/PVP2016-63058.

The fire process was simulated by the heat treatment to the Steel SPV490 of atmospheric storage tank, thereby obtaining the metal specimens in different fire temperature, holding time, and cooling modes. And as the temperature increases, the microscopic structure of Steel SPV490 changes under different working conditions, which could be shown in optical microstructure pictures after doing the interception, inlay, polishing, finishing to the specimens. The result shows that, the mechanical properties of the Steel SPV490 for storage tank changes as the temperature rising from the microscopic view. Nodulizing of the cementite in pearlite occurs, and the strength decreases when the high strength steel SPV490 of large atmospheric storage tanks under air cooling condition below 700 °C, however, it equivalents to the normalizing process, as the sorbite occurs in the steel, and the strength increases a bit when the temperature is above 900 °C. The water-cooling of steel SPV490 above 900 °C equivalents to the process of quenching. The occurrence of martensitic substantially increases the strength and the brittleness, and the elongation decreases rapidly.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A024. doi:10.1115/PVP2016-63086.

In the ultrasonic test of girth or longitudinal butt joints in boiler, pressure vessel and pipe, there are multiple defects perpendicular to the length of the weld in a straight line. It is inappropriate of considering only the two adjacent defects in a line in quality classification without considering the defects located in different depth while in same lateral position. Hereby, we make tensile strength test of the rectangle samples with different numbers of side drilled holes, then do collection and analysis about the test result, so that we can get the relationship between elongation and tensile strength of the specimen under external force.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A025. doi:10.1115/PVP2016-63194.

The Chinese Safety Technical Regulation of Special Equipment TSG Z6002-2010 Examination Rules for Welding Operators of Special Equipment is introduced in the paper. Administrative requirements, technical requirements, testing requirements of TSG Z6002-2010 and ASME BPVC Sec.IX-2015 are compared. Requirements and responsibilities of organization who conducts the welder or welding operator performance qualification test are discussed. The rules for expiration and renewal of welder or welding operator performance qualification are also discussed. The essential variables for the welder or welding operator performance qualification and the ranges qualified under TSG Z6002-2010 and ASME BPVC Sec.IX are compared. The requirements of type of test and test coupons are compared.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A026. doi:10.1115/PVP2016-63260.

The efficiency of conventional boiler/steam turbine fossil power plants has strong relationship to the steam temperature and pressure. At present, steam temperatures of the most efficient fossil power plants are now in the 600°C range. Higher-strength materials are needed for upper water wall tube of boilers with steam pressure above 24 MPa. A high-strength 2.5%Cr steel recently approved by ASME code as T23 is the preferred candidate material for this application. Due to its superior properties, T23 steel is typically not post-weld heat treated. However, after several years running there are a lot of incident reports for T23 tubes especially the breakage of weldment in the ultra-supercritical power plant. This is cause for concern for T23 tubes weldment used under high temperature environments. Previous studies showed that the residual stress may play an important role to the performance of spiral water wall tube.

In this paper, the distribution of residual stress in T23 tube weldment has been investigated in detail. Inner wall cracks were found at the butt-jointed seam region of spiral water wall tubes by radiographic testing after one year’s operation. Failure analysis of the spiral water wall tube cracking was conducted by chemical composition analysis, mechanical testing and finite element analysis in this paper. It was found that localized residual stress after the weld process caused concentrated stress, which is the primary reason for failure. Our studies illustrate the necessity of post weld heat treatment for the T23 tubes used under high temperature.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A027. doi:10.1115/PVP2016-63332.

Welding of 15NiCuMoNb5-6-4 and austenitic stainless steel thickness pipes is necessary in the primary loop pipes installation for the platform. In the process of formulating welding procedure specification of the pipeline pre-welding between 15NiCuMoNb5-6-4 and Z3CN20-09M, some macroscopical cracks appeared repeatedly on the surface of bend-test specimens, while other performances of welding joints such as mechanical properties at room temperature, high temperature tensile properties and impact properties could meet the relevant codes requirements. Focus on present problem, chemical composition analyses, metallographic microstructures analyses and scanning electron microscope observation were adopted to find the welding failure root cause. Finally, INCONEL®152M and 52M with 30% chromium were chosen to be the welding materials because of good ductility dip cracking resistance. It was found that the performance of the welding joints could satisfy the codes and engineering requirements.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A028. doi:10.1115/PVP2016-63458.

Volume defect is the common defect in high temperature pressure pipeline. Those defects have a great influence on stress redistribution of the pipes in high temperature, and affect the integrity and safety operation of high temperature components. In this paper, the defects were regularized, and a high temperature creep model was established. Based on this model, creep behavior of high temperature pressure pipeline with volume defect was studied, and the stress concentration of the main feature points in the defect was researched. Then, defects interference effect was discussed, and the critical interference distance was given. The results provide theoretical support to the safety assessment of high temperature structure with defects.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A029. doi:10.1115/PVP2016-63532.

As one of the carbon-free energies, hydrogen is considered as an important energy carrier in the 21st century. The minimum ignition energy of hydrogen is the lowest among flammable gases, hence, hydrogen leaking from pinholes, narrow gaps, or broken pipes can be ignited by ignition sources such as static electricity. Understanding of the characteristics of the hydrogen jet release is crucial for the better design and the applications. In this paper, the hydrogen under-expanded jet and flammable envelope were studied by simulation method. The effects of the nozzle shape and release pressure on the under-expanded jet, the downstream shock structure and flammable envelope were investigated. The simulation results showed that the nozzle geometry had great influence hydrogen under-expanded jet. And the maximum flammable length increased with the increasing of aspect ratio. For the split nozzle, with the increasing of pressure the hydrogen diffusion region of minor axes in the near-to-nozzle field increased but decreased of major axes.

Topics: Nozzles , Hydrogen
Commentary by Dr. Valentin Fuster
2016;():V01BT01A030. doi:10.1115/PVP2016-63580.

For petrochemical pressure vessels subjected to complex media environment, the competition, inhibition, promotion and interference of multiple failure mechanisms have been discussed by several failure case analyses and experimental investigation. The main factors that influence formation of dominant failure mechanism are analyzed and the judgment principles of the dominant failure mechanism are raised in the case of interaction of multiple failure mechanisms. In this paper, relevant mechanisms are also discussed, e.g. intergranular corrosion and intergranular stress corrosion failure mechanisms of austenitic stainless steel, failure mechanisms of austenitic stainless steel when Cl and alkaline environment exist concurrently and failure mechanism of austenitic stainless steel when medium such as Cl, CO2, H2S, H2O, etc. exist concurrently.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A031. doi:10.1115/PVP2016-63794.

A finite element calculation and a leakage rate test were carried out in order to investigate the tightness performance of a corrugated gasket bolted flanged connections system. Based on a detailed finite element model of the bolted flange connection, gasket stresses and deformations were determined for room temperature operating conditions. Leakage testes were then performed to acquire leakage data at different gasket compression stresses and different internal pressures. The data was fit to a leakage rate formula. Additional analysis was performed to investigate the influence of temperature.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A032. doi:10.1115/PVP2016-63959.

NGV’s (Natural Gas Vehicle) are known for their energy-saving and environment-friendly advantages. The high-pressure cylinder for automotive vehicles (hereinafter cylinder) is the main energy supply unit of an NGV. Therefore, the Life time or Life Cycle of the cylinder is closely related to vehicle safety performance. Pressure cycle test is, as a test that simulates the cylinder filling process, the most realistic and effective method to evaluate cylinder Life time or Life Cycle. To simulate the actual situation of cylinder use, there are two types of Pressure cycle tests: Pressure cycle test under filling conditions and Pressure cycle test under overload conditions (LBB Mode). To meet the market demand for reduced vehicle mass, most cylinder manufacturers in China tend to reduce cylinder weight by improving cylinder material. strength and reducing cylinder wall thickness. Few manufacturers, however, pay attention to the relation between cylinder Life time or Life Cycle and cylinder thickness reduced by strength improvement. In this paper, Pressure cycle tests are conducted on cylinders with the same specification but various wall thickness values to calculate and analyze the Life time or Life Cycle values. This paper is trying to discover the inherent law between cylinder material. strength, wall thickness and Life time or Life Cycle, to put forward the viewpoint that analysis design or test verification can be adopted in cylinder wall thickness design, to build the wall thickness design model for a widely-used cylinder model, and to lay the theoretical basis for lightweight cylinder design under safe conditions.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A033. doi:10.1115/PVP2016-63960.

Austenite stainless steel weld overlay cladding is widely used for the equipments working in pressurized hydrogen environment such as hydrogenation reactors. Surface cracking is a basic failure mode of the weld overlay cladding, and also a quite difficult problem for safety assessment and defect elimination. In this paper, two cases of surface cracking of austenite stainless steel weld overlay cladding are introduced. Firstly, the inspection results and cracking causes are analyzed in detail. Secondly, two kinds of treatment methods for the defects are introduced. Finally, some suggestions for inspection and assessment of surface cracking of weld overlay cladding are proposed.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in European Codes and Standards

2016;():V01BT01A034. doi:10.1115/PVP2016-63038.

The paper describes the general approach followed by AFCEN, the French Society for Design, Construction and In-Service Inspection Rules for Nuclear Island Components, in developing the RCC-M code from the technical and organizational points of views. The changes that the RCC-M Subcommittee has introduced into the 2016 Edition of RCC-M code are explained and commented upon. The main activities of the RCC-M Subcommittee have been in achieving and demonstrating conformity of the code with French regulation and making comparisons with other codes and standards. The paper highlights how industrial experience is being integrated into the RCC-M code and how conformity with the essential safety requirements of the European Pressure Equipment Directive has been demonstrated. The processes for updates, interpretations or inquiries are also addressed.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A035. doi:10.1115/PVP2016-63084.

Gross failure of certain components in nuclear power plant has the potential to lead to intolerable radiological consequences. For these components, UK regulatory expectations require that the probability of gross failure must be shown to be so low that it can be discounted, i.e. that it is incredible. For prospective vendors of nuclear power plant in the UK, with established designs, the demonstration of “incredibility of failure” can be an onerous requirement carrying a high burden of proof. Requesting parties may need to commit to supplementary manufacturing inspection, augmented material testing requirements, enhanced defect tolerance assessment, enhanced material specifications or even changes to design and manufacturing processes.

A key part of this demonstration is the presentation of the structural integrity safety case argument. UK practice is to develop a safety case that incorporates the notion of ‘conceptual defence-in-depth’ to demonstrate the highest structural reliability. In support of recent Generic Design Assessment (GDA) submissions, significant experience has been gained in the development of so called “incredibility of failure” arguments. This paper presents an overview of some of the lessons learned relating to the identification of the highest reliability components, the development of the structural integrity safety arguments in the context of current GDA projects, and considers how the UK Technical Advisory Group on Structural Integrity (TAGSI) recommendations continue to be applied almost 15 years after their work was first published. The paper also reports the approach adopted by Horizon Nuclear Power and their partners to develop the structural integrity safety case in support of the GDA process to build the UK’s first commercial Boiling Water Reactor design.

Topics: Safety , Reliability
Commentary by Dr. Valentin Fuster
2016;():V01BT01A036. doi:10.1115/PVP2016-63095.

With plans for new nuclear build being prominent in the UK in recent years, the licensing aspects have obviously been of paramount importance for the requesting parties. It has been particularly important for the requesting parties to properly understand the overall UK licensing framework and the detailed differences with respect to that of their country of origin. One of the main detailed differences is in the areas of fracture mechanics and Inspection Qualification and how the information relating to these two technical disciplines is linked in order to evaluate the defect tolerance of relevant plant components.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A037. doi:10.1115/PVP2016-63227.

RCC-MRx Code (ref 1) is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER.

This code has been issued in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) in 2012, and a new edition was published at the end of 2015.

A significant work has been performed for this edition to improve the code in order to facilitate its use and understandability, and also to have a better fit with the feedbacks of the users.

In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers...) through a workshop called CWA phase 2.

This paper gives an overview of the realized work and also will identify the work to be done for an opening of a standard such as RCC-MRx code.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Japanese Codes and Standards

2016;():V01BT01A038. doi:10.1115/PVP2016-63210.

Elastic-plastic finite element (FE) analysis is performed to determine the plastic behavior of the reactor pressure vessel (RPV) inner surface caused by rapid cooling during pressurized thermal shock (PTS) events. However, as the J-integral is not path-independent for elastic-plastic material in the unloading process, it is necessary to apply a suitable correction method using elastic material. In addition, it is also necessary to consider the effect of the welding residual stress appropriately.

Therefore, we investigated the stress intensity factor derived from FE analysis based on a model consisting of elastic-plastic cladding and linear elastic low-alloy steel with subsequent plastic zone correction, since the stress level of low-alloy steel remains within the elastic region except the crack front during a PTS event. Furthermore, we examined whether the stress mapping method is applicable for reflecting the effect of welding residual stress in FE analysis, even though the plastic strain generated during welding is ignored.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A039. doi:10.1115/PVP2016-63443.

For structural integrity assessment on a reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, thermal histories of coolant water and heat transfer coefficient between coolant water and RPV are important influence factors. The former is determined on the basis of thermal-hydraulics (TH) analyses simulating PTS events and the latter is derived from Jackson-Fewster correlation using TH analysis results. Using these factors, subsequently, loading conditions for structural integrity of RPVs are evaluated by structural analyses. Nowadays, three-dimensional TH and structural analyses are recognized as precise methods for assessing structural integrity of RPVs. In this study, we performed the TH and structural analyses using a three-dimensional model including cold-leg, downcomer and beltline region of RPV in order to evaluate loading conditions during a PTS event more accurately. Distributions of temperature of coolant water and heat transfer coefficient on the surface of RPV were calculated by TH analysis. Loading condition evaluation was then performed by structural analysis using these values and taking the weld residual stress due to weld-overlay cladding and post-weld heat treatment into consideration. From these analyses, we obtained histories and distributions of loading conditions at the reactor beltline region of RPV. Based on the analysis results of loading conditions, we discussed the conservativeness of current structural integrity assessment method of RPV prescribed in the current codes through the comparison of the loading conditions due to a PTS event.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A040. doi:10.1115/PVP2016-63822.

It has been accepted that neutron irradiation embrittlement of reactor pressure vessel is caused by irradiation-induced formation of solute clusters (SCs) and matrix damages (MDs). In the present study, to analyze the contribution of chemical composition contained in SCs to irradiation embrittlement at high fluence region, statistical analysis using the Bayesian nonparametric (BNP) method was performed for Japanese PWR surveillance data. The significance of P, Si and Mn contents, which are not necessarily included in embrittlement correlations unlike the Cu and Ni content, was evaluated. The BNP method can learn the complexity of the statistical model itself from the input data and infer the predicted data with individual probability distribution of predict condition. The result suggested that irradiation embrittlement was most affected by the Si content in three examined elements at high fluence region.

Commentary by Dr. Valentin Fuster

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

2016;():V01BT01A041. doi:10.1115/PVP2016-63094.

The improvement of residual stress in component surface by shot peening was applied mainly to the core shroud for the purpose of preventing stress corrosion cracking (SCC) in BWR Nuclear Power Plants (NPP’s) in the late 1990’s. Subsequently, advanced peening techniques using pressure shock wave impact, such as Water Jet Peening (WJP) and Laser Peening (LP), were developed. Subsequently, PWR NPP’s also adopted these techniques to prevent Primary Water Stress Corrosion Cracking (PWSCC) in Japan.

These peening techniques were implemented with performance demonstration tests conducted in advance by JAPEIC (Japan Power Engineering and Inspection Corporation). Operators also planned on-site applications, including risk management. This risk refers to the risk that components integrity or outage schedule may be affected by peening. In fact, Tsuruga-2 faced a difficult situation, despite the preparation for the risk that cracks might be detected during pre-peening inspection. It was beyond anything that could have been expected.

This paper presents the experiences of peening technique applications at Tokai-2 (2 loop, 1,100 MWe BWR) and Tsuruga-2 (4 loop, 1,160 MWe PWR), analyzing common risk factors present in general application cases based on specific experiences. As a result, high-risk possibilities are indicated depending on risk management, and the necessity is shown for controls that take into consideration materials, design, operation time and other factors. Additionally, the relationship between the risk management and Codes & Standards is also referenced.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A042. doi:10.1115/PVP2016-63769.

Stress corrosion cracking (SCC), though infrequent, is often detected in nuclear power reactor system piping and components. A number of approaches have been developed and successfully deployed for SCC repair and mitigation such as full structural weld overlay (FSWOL), optimized weld overlay (OWOL), and mechanical stress improvement process (MSIP). While these approaches are proven technologies and have served the industry well, a new strategy, excavate and weld repair (EWR), provides yet another option for repair or mitigation of SCC. The EWR approach excavates a portion of the outer part of the butt weld. The excavation is then filled with a weld metal with demonstrated SCC resistance. The EWR approach would require less welding compared to a weld overlay and may be the best option for large bore butt welds where restricted access may make FSWOL, OWOL, or MSIP impractical. For the situation where a flaw is detected and removal or reduction of the flaw to acceptable size is necessary for continued service, the approach would permit a local partial arc EWR where only a portion of the butt weld circumference is removed and repaired. While the partial arc EWR is not a full mitigation, it would provide the needed preparation time for a more permanent repair during a subsequent refueling or maintenance outage. ASME Code Case N-847 was developed to provide examination, design, installation, and preservice/inservice inspection requirements for the EWR repair and mitigation approach. This paper provides a background, description and the technical bases for the EWR case case.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2016;():V01BT01A043. doi:10.1115/PVP2016-63902.

The NRC Issued a Regulatory Issue Summary (RIS 15-10) which was an inquiry to the ASME Code Committee to review the in-service inspection requirements for Alloy 600 full penetration branch connections. This NRC request resulted in the initiation of two PWROG projects, PA-MSC-1283 and 1294.

PA-MSC-1283 is a Fracture Analysis to evaluate the crack growth rate of the various existing A600 configurations, heat treat conditions and operating loads. The results from this study would provide the technical basis for in-service inspection method and frequency.

PA-MSC-1294 is a project to consider contingency repair techniques and provide a bounding analysis for all nozzle configurations of the selected repair technique. The selected repair approach to be analyzed was a weld buildup referred as a Branch Connection Weld Metal Buildup (BCWMB) in the ASME Code Case (N-853) developed to provide the rules for design, implementation and inspection. Included in this project was a proof-of-principle phase to demonstrate that the repair methodology is implementable and assist in establishing the potential on component repair duration and anticipated dose.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2016;():V01BT01A044. doi:10.1115/PVP2016-64007.

As the existing light water reactor (LWR) fleet ages, the weldability of structural materials used to construct the reactor pressure vessels (RPVs) and reactor internals is diminished. The decrease in the weldability in austenitic and ferritic materials is attributed to the formation of helium in the material microstructure. Helium (He) generation occurs during the service life of irradiated reactor internals from neutron transmutation reactions of boron and nickel in these materials. Welding on irradiated materials, if performed without appropriate consideration of fluence exposure and helium generation, can result in a heat affected zone cracking phenomenon termed helium induced cracking (HeIC). The heat input associated with welding is a major factor affecting the coalescence of the generated helium along grain boundaries. As the material cools, the tensile stresses generated from welding can cause cracking to occur along grain boundaries weakened by helium bubble coalescence. In some cases, the preferred or only method of repair or replacement of a reactor internal component is welding. For components located in regions of low thermal fluence, the welding process implementation may be relatively straightforward and only heat input control may be required. However, in high thermal fluence regions, weld repair of irradiated reactor internal components is complicated by the presence of high concentrations of helium and significant care must be taken in welding process selection and heat input control. This paper highlights envisioned applications for weld repair on irradiated reactor internals. It also summarizes recently completed guidance published by the EPRI Materials Reliability Program (MRP) and EPRI BWR Vessel and Internals Project (BWRVIP) which provides an improved basis for plants to assess the weldability of components at various locations within the reactor.

Topics: Maintenance , Welding
Commentary by Dr. Valentin Fuster
2016;():V01BT01A045. doi:10.1115/PVP2016-64008.

This paper will discuss the ASME Code Committee activities involved in the incorporation of surface stress improvement (SSI) into ASME Code Cases N-770-4 and N-729-5. ASME Code Cases N-770 [1] and N-770-1 introduced several mitigation approaches for dissimilar metal weld (DMW) locations in PWR primary system piping and provided inspection relief for locations that were mitigated. The initial approaches contained in N-770 and N-770-1 included mechanical stress improvement and weld overlay methods that have a global stress relief effect to achieve a very low tensile surface stress state or a compressive stress state at the weld inside surface to halt crack initiation, as well as growth of acceptably sized cracks. The weld overlay mitigation methods are also effective because they introduce PWSCC-resistant material, i.e., Alloys 52, 152, or their variants. (The initial approaches also included Alloy 52/152 weld inlay and weld onlay, methods that do not require stress improvement but do require access to the weld inside surface.)

While the mechanical stress improvement and weld overlay methods address the majority of the DMW locations in the primary piping system, there are locations that cannot be treated by these approaches due to the weld geometry or access limitations for the needed equipment. Additionally the dissimilar metal J-groove welds in the reactor pressure vessel head penetration nozzles (RPVHPN) could not be addressed at all by the approaches developed for DMW locations. To address the industry need to mitigate the unfavorable DMW geometries and locations along with the RPVHPN locations, the use of surface stress improvement (SSI) was studied and documented in EPRI reports Materials Reliability Program (MRP)-267 [2], “Technical Basis for Primary Water Stress Corrosion Cracking by Surface Stress Improvement,” and MRP-335 [3], “Topical Report for Primary Water Stress Corrosion Cracking by Surface Stress Improvement.” These reports formed the technical basis for the SSI-related changes made in Code Cases N-770-4 and N-729-5. Along with the technical bases noted, support from the international community in terms of operational experience with SSI in their power plants was invaluable in providing the necessary understanding, context, and confidence to committee members.

The ASME “Task Group High Strength Nickel Alloy Issues” (TGHSNAI) was assigned the task of revising the existing Code Cases, N-770 [1], “Alternate Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities” and N-729 [4], “Alternate Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds.” To incorporate the SSI approach into these Code Cases, the first action was to determine whether the SSI process was considered to be a peening process as defined by ASME Section III NB-4422 criteria. This required the submittal of an Interpretation of NB-4422 to determine if SSI techniques were considered a peening process under ASME Section III. The interpretation (Interpretation III-1-13-03), documented in ASME File 12-1192 [5], specified that SSI was not considered peening by Section III. This interpretation provided the framework by which SSI could be directly applied to ASME Section XI inspection criteria without the need to first revise ASME Section III NB-4422. SSI (peening) was first incorporated into Code Case N-770 [1] to provide a mitigation alternative for locations unable to be addressed by the methods addressed thus far. The revision to Code Case N-770 [1] does not provide guidance for the application of SSI activities but rather, it provides the process performance criteria and the inspection guidance following the application of SSI and establishes the pre-application inspection acceptance criteria. Following the approval of SSI in Code Case N-770 [1] addressing the DMW in the primary coolant piping system, the SSI approach was applied to the partial penetration dissimilar metal J-groove welds in RPVHPNs in Code Case N-729 [4]. The application to RPVHPNs provides the industry with a valuable asset preservation tool while significantly lowering the safety risks associated with primary water stress corrosion cracking (PWSCC) and degradation from borated water leakage for the RPVHPNs.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A046. doi:10.1115/PVP2016-64032.

Plant operating experience with Alloy 600 reactor pressure vessel top head penetration nozzles in U.S. PWRs shows that the inspection intervals prescribed by ASME Code Case N-729-1 have been successful in managing the PWSCC concern. No through-wall cracking has been observed in the U.S. after the first in-service volumetric or surface examination was performed on all CRDM or CEDM nozzles in a given head. The current inspection intervals have facilitated identification of any PWSCC in its early stages, with small numbers of nozzles affected and substantial margins to leakage at the five affected heads operating at Tcold. MRP-395 demonstrated through both deterministic and probabilistic analyses that the inspection intervals of Code Case N-729-1 remain valid to conservatively address the PWSCC concern.

This paper supplements MRP-395 with additional deterministic crack growth analyses coupled with assessments of the PWSCC indications detected in heads operating at Tcold. The supplemental deterministic assessments presented in this paper demonstrate the acceptability of a 36-month volumetric or surface inspection interval for heads with previously detected PWSCC and that operate at Tcold. Until Code Case N-729-5 is approved by U.S. NRC, use of the 36-month interval in the U.S. for such heads would require review and approval by U.S. NRC of a relief request submitted by the licensee.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A047. doi:10.1115/PVP2016-64041.

The level of stress at which stress corrosion cracking can occur in nickel-based alloys is a key factor in assessing the effectiveness of methods that have been proposed to mitigate such cracking. As various mitigation methods have different approaches and capabilities, the stress limit specified for each case has varied. This paper will review the available data on this subject and develop a consistent basis for defining the stress levels that would be required to initiate stress corrosion cracking in these materials. This paper may be applied to improve the consistency of the requirements of the various mitigation methods involving stress improvement.

Topics: Stress , Indium alloys
Commentary by Dr. Valentin Fuster

Codes and Standards: Structural Integrity of Pressure Components

2016;():V01BT01A048. doi:10.1115/PVP2016-63137.

Threaded closures for pressure vessels have been in use for decades. Much work has been done to develop convenient, safe and economical threaded closures. Threaded closures are used when there is a need for opening the vessel either for maintenance or as part of its operation.

Heat Exchangers are a typical application where there is a need for opening the vessel and cleaning the tubes at regular intervals to maintain the heat transfer efficiency. These are known as Breech Lock or Screw Plug Exchangers. These are basically U-tube exchangers. The channel side operates at high temperature and pressure and it has a threaded end closure. In some designs, the shell side may also be at high pressure. The tube bundle is removable without having to dismantle the channel or disconnect the nozzles from the pipeline. Thus screw plug exchangers help to reduce fabrication cost and reduce time for in-service maintenance.

The major problem encountered with the use of such end closures are 1) Jamming of the threaded plug, due to deformation of the channel barrel. Thus the opening of the end closure by unscrewing becomes a difficult task. With the increase in operating temperatures and pressures, the problems become more severe, due to which, users are not inclined to use these type of end closures.

A study was undertaken to assess the reasons for bulging of the end of the channel which caused jamming of the screw threads and also for leakage through the gasket.

By shrink fitting a ring over the end of the channel, the deformation was reduced, enabling easy opening of the cover.

2) The leakage through the gasket between the shell and tubesheet, causing the intermixing of shell and tube-side fluids. This on analysing was found that the additional forces were acting on the gasket due to thermal expansion of the internals. This led to changing to a gasket that could withstand the forces and pressure.

Leakage through the gasket was prevented by analysing the additional forces acting on the gasket due to thermal expansion of the internals and changing to a gasket that could withstand the forces and pressure.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A049. doi:10.1115/PVP2016-63221.

Pressure vessels designed for the internal pressure are required to be subjected to hydrostatic test after completion of their fabrication. Two methods are provided in the ASME Sec. VIII Div.2 to calculate the minimum test pressure. One method is based on the Maximum Allowable Working Pressure of the vessel or the MAWP and the other is based on the highest permissible pressure of the vessel at the test condition. This latter pressure is also known as the ‘calculated’ test pressure or the pressure based on the Maximum Allowable Pressure in new and cold condition or the MAP. In either case, the induced primary membrane stress, Pm, shall satisfy the below condition Display Formula

Pm0.95*Sy
where, Sy is the yield strength of the material at the test temperature.

It is common to notice in many User’s Design Specifications (UDS), the requirement to test the vessels at the ‘calculated’ test pressure.

In the year 2007, major changes were made to ASME Sec.VIII Div.2. Some changes that had relevance to the hydrostatic testing were the change in the basis of the allowable stress at room temperature to the yield strength for most materials, increase in the factor on the test pressure basis to 1.43 and the corresponding increase in permissible limits on the induced primary membrane stress to 0.95*Sy.

It is noted in pressure vessels, built with cylindrical shells and(/or) with hemispherical dished heads with materials whose maximum allowable stresses at room temperature were dependent on the yield strength, that the membrane stress induced when subjected to ‘calculated’ test pressure exceeded the Code limit of 0.95*Sy. Increasing the thickness of the shell and(/or) head did not result in containing the stresses below 0.95*Sy as such an increase in thickness resulted in proportional increase in the MAP values and therefore, further increase of the test pressure.

The objective of this paper is to first derive the actual membrane stress induced in the vessel subjected to the ‘calculated’ test pressure and provide an engineering solution to the problem of induced stress exceeding the Code limit in order to achieve Code as well as the UDS compliance.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A050. doi:10.1115/PVP2016-63278.

Using Risk-Based Inspection (RBI) and Fitness for Service (FFS) approaches to manage the integrity of pressure equipment has been the industries best practice for almost a decade. However, there had never been a procedural link between these two approaches in a way that when one performed FFS analysis on a defect, one could update the risk accordingly. This paper proposes a quantitative method to refresh the risk calculated in the RBI process when FFS analysis is completed on a locally thinned area. The proposed approach applies a probabilistic technique by considering the Remaining Strength Factor (RSF) from API 579-1/ASME FFS-1 as the limit state equation and assuming the corrosion rate as a distribution variable to estimate the unconditional probability of failure. This value is then modified using a Bayesian updating method allowing for the conditional probability to represent a new failure likelihood which could be utilized in the RBI planning.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A051. doi:10.1115/PVP2016-63330.

The fatigue behavior in steam turbine valve is generally considered to be associated with thermal transients during start and stop phase. However, the recent analysis has shown that steam parameter fluctuations under in-service steady state operation could induce significant increase in the damage. In this paper, the effect of in-service steam parameter fluctuations on stress-strain behavior in the valve was analyzed based on FE analysis under in-service data. The rain-flow cycle counting method was applied to get effective stress-strain cycle numbers and cycle amplitudes to classify the types of fatigue cycle based on in-service data. The creep-fatigue damage during steady state operation of the valve was estimated by using R5 (Volume 2/3) high temperature assessment procedure together with the FE results. Frequent steam pressure fluctuations at steady state operation were identified as the most influential factor for the fatigue life of the steam turbine valve.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A052. doi:10.1115/PVP2016-63612.

Clause UG-27 of ASME Section VIII Division 1 [1] provides rules for calculating the thickness of shells under internal pressure. Mandatory Appendix-2 of Code [1] provides rules for design of bolted flanged connections.

In certain high pressure and high thickness pressure vessels having a cylindrical shell with bolted cover flange, Manufacturers avoid a separate end flange welded to the shell, as the construction becomes bulky. Instead of the same, Manufacturers provide tapped holes in shell wall parallel to axis of the cylindrical shell. The cover is directly bolted to these tapped holes provided in the shell.

This type of construction may be economical as compared to welding a conventional flange to the end of the shell. However this type of construction is not covered in the Code [1].

When such tapped holes are provided in the cylindrical shell, generally the total metal thickness provided at the tapped hole location meets UG-27 requirement of the Code [1]. However due to the tapped holes, the thickness from inside surface of vessel to inside surface of tapped hole is less than the required thickness of UG-27.

It is therefore required to analyze the stresses due to these tapped holes in the shell thickness to ensure that Code [1] allowable stresses are not exceeded.

The work reported in this paper was undertaken to determine the effect of internal pressure on the stresses in a cylindrical shell having tapped holes parallel to axis of the cylindrical shell.

Topics: Stress , Pipes , Cylinders
Commentary by Dr. Valentin Fuster

Codes and Standards: Technical Harmonization and Emerging Code and Standards

2016;():V01BT01A053. doi:10.1115/PVP2016-63040.

Different pressure vessel and piping design codes and standards have adopted different fatigue analysis methods. In order to make some contribution to current efforts to harmonize international design codes and standards, a review of fatigue analysis methods for a number of selected nuclear and non-nuclear design codes and standards has been carried out. The selected design codes and standards are ASME Boiler and Pressure Vessel Code Section III Subsection NB and Section VIII Division 2, EN 12952, EN 13445, EN 13480, PD 5500, RCC-M, RCC-MRx, JSME, PNAEG and R5. This paper presents the initial review results.

The results of the study could be used as part of the on-going work of the Codes and Standards Task Force of the World Nuclear Association (WNA) Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group.

Topics: Design , Fatigue life
Commentary by Dr. Valentin Fuster
2016;():V01BT01A054. doi:10.1115/PVP2016-63313.

Creep-fatigue assessments require inputs for both creep damage and fatigue damage. Then these damage terms are combined using some form of interaction diagram to enable estimation of the creep-fatigue life. Similarly, creep-fatigue crack growth assessments need separate calculations of creep crack growth, fatigue crack growth and again possibly some allowance for interaction in order to obtain the total crack growth. A consequence of these approaches is that uncertainties in creep calculations, uncertainties in fatigue calculations and uncertainties in the form of interaction all lead to significant uncertainties in the overall lifetime or crack growth assessment.

This paper first briefly describes how creep-fatigue and creep-fatigue crack growth assessments are performed using the UK R5 procedure and contrasts the methods with those in other codes. Then, the paper presents the guidance in R5 on addressing the many uncertainties in these assessments and discusses the use of probabilistic methods in order to avoid over-conservative lifetime and crack growth estimates.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A055. doi:10.1115/PVP2016-63350.

Recent experimental results on creep-fracture damage with minimum time to failure (minTTF) varying as the 9th power of stress, and a theoretical consequence that the coefficient of variation (CV) of minTTF is necessarily 9 times that of the CV of the stress, created a new engineering requirement that the finite element analysis of pressure vessel and piping systems in power generation and chemical plants be more accurate with an allowable error of no more than 2% when dealing with a leak-before-break scenario. This new requirement becomes more critical, for example, when one finds a small leakage in the vicinity of a hot steam piping weldment next to an elbow. To illustrate the critical nature of this creep and creep-fatigue interaction problem in engineering design and operation decision-making, we present the analysis of a typical steam piping maintenance problem, where 10 experimental data on the creep rupture time vs. stress (83 to 131 MPa) for an API Grade 91 steel at 571.1 C (1060 F) are fitted with a straight line using the linear least squares (LLSQ) method. The LLSQ fit yields not only a two-parameter model, but also an estimate of the 95% confidence upper and lower limits of the rupture time as basis for a statistical design of creep and creep-fatigue. In addition, we will show that when an error in stress estimate is 2% or more, the 95% confidence lower limit for the rupture time will be reduced from the minimum by as much as 40%.

Topics: Creep , Stress , Design , Errors , Rupture
Commentary by Dr. Valentin Fuster
2016;():V01BT01A056. doi:10.1115/PVP2016-63489.

This paper will compare two major In-Service Inspection Nuclear Codes: the Boiler and Pressure Vessel Code Section XI developed by ASME on the one hand and French RSE-M developed by AFCEN on the other hand. These two Codes cover totally or partially:

- General Introduction

- In-service and pre-service inspection program

- In-service inspection techniques and qualifications

- Surveillance in Operation

- Flaw evaluation

- Repair-Replacement

- Quality Assurance and Documentation

- Appendices and Code Cases

A general presentation of the two Codes will be developed and followed by a general comparison.

To conclude, the paper will consider two particular examples for detailed comparison:

- K stress intensity factor handbooks comparison for circumferential flaws in cylinder: ASME Code Appendix A3000 and RSE-M Code Appendix 5-4, last published editions

- analysis of longitudinal crack in steam generator tubes

Commentary by Dr. Valentin Fuster
2016;():V01BT01A057. doi:10.1115/PVP2016-63600.

Axial compression allowable stress for pipe supports and restraints based on linear elastic analysis is detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF. The axial compression design by analysis equations within NF-3300 is replicated from the American Institute of Steel Construction (AISC) using the Allowable Stress Design (ASD) Method which was first published in the ASME Code in 1973. Although the ASME Boiler and Pressure Vessel Code is an international code, these equations are not familiar to many users outside the American Industry. For those unfamiliar with the allowable stress equations, the equations do not simply address the elastic buckling of a support or restraint which may occur when the slenderness ratio of the pipe support or restraint is relatively large, however, the allowable stress equations address each aspect of stability which encompasses the phenomena of elastic buckling and yielding of a pipe support or restraint. As a result, discussion of the axial compression allowable stresses provides much insight of how the equations have evolved over the last forty years and how they could be refined.

Topics: Pipes , Compression
Commentary by Dr. Valentin Fuster
2016;():V01BT01A058. doi:10.1115/PVP2016-63715.

In most finite-element-analysis codes, accuracy is achieved through the use of the hexahedron hexa-20 elements (a node at each of the 8 corners and 12 edges of a brick element). Unfortunately, without an additional node in the center of each of the element’s 6 faces, nor in the center of the hexa, the hexa-20 elements are not fully quadratic such that its truncation error remains at h2(0), the same as the error of a hexa-8 element formulation.

To achieve an accuracy with a truncation error of h3(0), we need the fully-quadratic hexa-27 formulation. A competitor of the hexa-27 element in the early days was the so-called serendipity cubic hexa-32 solid elements (see Ahmad, Irons, and Zienkiewicz, Int. J. Numer. Methods in Eng., 2:419–451 (1970) [1]). The hexa-32 elements, unfortunately, also suffer from the same lack of accuracy syndrome as the hexa20’s.

In this paper, we investigate the accuracy of various elements described in the literature including the fully quadratic hexa-27 elements to a shell problem of interest to the pressure vessels and piping community, viz. the shell-element-based analysis of a barrel vault. Significance of the highly accurate hexa-27 formulation and a comparison of its results with similar solutions using ABAQUS hexa-8, and hexa-20 elements, are presented and discussed. Guidelines are proposed for selection of better elements.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A059. doi:10.1115/PVP2016-63890.

A finite element method (FEM)-based solution of an industry-grade problem with complex geometry, partially-validated material property databases, incomplete knowledge of prior loading histories, and an increasingly user-friendly human-computer interface, is extremely difficult to verify because of at least five major sources of errors or solution uncertainty (SU), namely, (SU-1) numerical algorithm of approximation for solving a system of partial differential equations with initial and boundary conditions; (SU-2) the choice of the element type in the design of a finite element mesh; (SU-3) the choice of a mesh density; (SU-4) the quality measures of a finite element mesh such as the mean aspect ratio.; and (SU-5) the uncertainty in the geometric parameters, the physical and material property parameters, the loading parameters, and the boundary constraints. To address this problem, a super-parametric approach to FEM is developed, where the uncertainties in all of the known factors are estimated using three classical tools, namely, (a) a nonlinear least squares logistic fit algorithm, (b) a relative error convergence plot, and (c) a sensitivity analysis based on a fractional factorial orthogonal design of experiments approach. To illustrate our approach, with emphasis on addressing the mesh quality issue, we present a numerical example on the elastic deformation of a cylindrical pipe with a surface crack and subjected to a uniform load along the axis of the pipe.

Commentary by Dr. Valentin Fuster

Codes and Standards: Use of Modern FEA Methods for Code Assessment

2016;():V01BT01A060. doi:10.1115/PVP2016-63118.

The Linear Matching Method Framework (LMMF) consists of a number of simplified direct methods for generating approximate inelastic solutions and answering specific design related issues in pressure vessel design codes using standard finite element codes. Currently, all the LMM procedures have been implemented in ABAQUS through user subroutines with powerful user-friendly plug-in tools. The LMM ABAQUS user subroutines and plug-in tools for structural integrity assessment are now in extensive use in industries for the design and routine assessment of power plant components. This paper presents a detailed review and case study of the current state-of-the art LMM direct methods applied to the structural integrity assessment. The focus is on the development and use of the LMMF on a wide range of crucial aspects for the power industry. The LMMF is reviewed to show a wide range of capabilities of the direct methods under this framework, and the basic theory background is also presented. Different structural integrity aspects are covered including the calculation of shakedown, ratchet and creep rupture limits. Furthermore, the crack initiation assessments of an un-cracked body by the LMM are shown for cases both with and without the presence of a creep dwell during the cyclic loading history. Finally an overview of the in house developed LMM plug-in is given. Its implementation in ABAQUS finite element solver through an intuitive Graphical User Interface is presented. The efficiency and robustness of these direct methods in calculating the aforementioned quantities are confirmed through a numerical case study, which is a semi-circular notched (Bridgman notch) bar. A 2D axisymmetric finite element model is adopted, and the notched bar is subjected to both cyclic and constant axial mechanical loads. For the crack initiation assessment, different cyclic loading conditions are evaluated to demonstrate the impact of the different load types on the structural response. The creep dwell impact is also investigated to show how this parameter is capable of causing in some cases a dangerous phenomenon known as creep ratcheting. All the results in the case study demonstrate the level of simplicity of the LMMs but at the same time accuracy, efficiency and robustness over the more complicated and inefficient incremental finite element analyses.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A061. doi:10.1115/PVP2016-63154.

Stress classification at shell and nozzle interface has always been an interesting and challenging problem for Engineers. Basic shell theory analyses shell stresses as membrane with local bending stresses developed at locations of discontinuity and load applications. Since in a shell structure, bending stresses develop to mainly maintain compatibility of deformation and membrane stresses to equilibrate the applied load, a simple stress classification will be to categorize the bending stresses as secondary stresses. This is because by definition, secondary stresses develop to maintain compatibility of deformation and primary stresses develop to maintain equilibrium with the applied load. This simplified analysis can result in errors as in real world 100% primary stress as well as 100% secondary stress is rare if not impossible. The widespread use of Finite Element Analysis has made this problem become even more challenging. In this paper the work done by Chen and Li [1], using the two step primary structure method has been used to analyze the problem of stress classification of a shell and nozzle. This paper is a continuation of the author’s previous work on this topic [21]. In the previous paper, the sensitivity of modelling and the effect of the same on the results were investigated. However, the various approaches adapted in the paper [21], were not exactly in the true spirit of the method i.e in all the models, stresses in the vessels and nozzles were checked separately and compared against the stresses in the vessel and nozzle in the original model where by “original “model we mean the model with the vessel and nozzle modelled together i.e. connected along the space curve of intersection in all six degrees of freedom. The spirit of the method requires that the comparison has to be with reference to maximum M+B stresses in the original and reduced structure ( a “reduced” structure means where the vessel and the nozzle are not connected along some degrees of freedom along the space curve of intersection) and not individually in the vessels and nozzles and the M+B stresses have to be evaluated anywhere on the structure and not just at and close to the space curve of intersection. It is because of these reasons that [21] in not exactly in spirit of the method. In other words, the development of this paper was motivated by the fact that the previous paper did not use the exact spirit of the method and hence to investigate how its exact implementation changes results. This is the approach followed in this paper. A point to note; not in spirit of the method does not necessarily mean that the approach taken in [21] was not correct. It’s just that it was not in line with the way this method was defined by Chen and Li [1] and the present authors used their subjective approach to the problem. Additionally, this paper investigates the effect of geometric parameters like D/T, d/t and t/T on the results which was not investigated in the previous paper.

Commentary by Dr. Valentin Fuster
2016;():V01BT01A062. doi:10.1115/PVP2016-63665.

Neuber’s rule is commonly applied in fatigue analysis to estimate the plasticity of purely elastic FEA results. In certain cases, this is more efficient than running elastic-plastic models. However, the applicability of Neuber’s rule is not well understood for complex models and may not always be appropriate. In this paper, the applicability of Neuber’s rule is investigated. The background of Neuber’s rule is discussed, theoretical limitations are derived, and algorithmic outlines of the procedures are presented. Neuber’s plasticity correction procedure is applied to both the Ramberg-Osgood elastic-plastic constitutive relation and the advanced Chaboche isotropic/kinematic nonlinear hardening relation. Throughout the manuscript, the aspects of each model are discussed from an educational perspective, highlighting each step of the implementation in sufficient detail for independent reproduction and verification. This level of detail is often absent from similar publications and, it is hoped, may lead to the wider dissemination of Neuber’s rule for plasticity correction. The final component of the paper presents a multiaxial correction of the Chaboche hardening model. To the best of the authors’ knowledge, this is the first published application of Neuber’s rule to the multiaxial plasticity correction of the Chaboche combined isotropic/kinematic hardening model. Examples are used to illustrate the behavior of the method and to present some of the commonly overlooked components when assessing the applicability of Neuber’s method.

Commentary by Dr. Valentin Fuster

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