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ASME Conference Presenter Attendance Policy and Archival Proceedings

2017;():V01BT00A001. doi:10.1115/PVP2017-NS1B.
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This online compilation of papers from the ASME 2017 Pressure Vessels and Piping Conference (PVP2017) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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: Probabilistic and Risk-Informed Methods for Structural Integrity Assessment (Joint With MF-17)

2017;():V01BT01A001. doi:10.1115/PVP2017-65225.

The PFM approach has been widely used to evaluate the integrity of reactor pressure vessel (RPV) in nuclear power plant. Since the 1980s, a number of probabilistic fracture mechanics (PFM) analysis codes have been developed to perform the probabilistic analysis for RPV, and these codes are continuously updated by reflecting recent irradiation shift model, database of fracture toughness and compendia of stress intensity factors. The author developed a PFM analysis program for RPV, PROFAS-RV (PRObabilistic Failure Analysis System for Reactor Vessel), recently, which can evaluate failure probability of RPV using recent RTNDT shift model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx A16 code as well as required basic functions of PFM program. In this paper, the failure probabilities of boiling water reactor (BWR) for cool-down and low temperature over pressurization (LTOP) transient are calculated by using the own PFM analysis code, PROFAS-RV. This work was conducted as part of an international collaborative study. The effects of key parameters such as transient, fluence level, Cu and Ni content, initial RTNDT and RTNDT shift model on the failure probability are systematically compared and reviewed. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial RTNDT. However, the effect of Cu and Ni content is negligible for the very low fluence of 0.02×1019 n/cm2 because there is no additional irradiation embrittlement. The effect of initial RTNDT on the failure probability is more significant for the lower fluence region in both transients. The failure probability of LTOP transient is lower than that of cool-down transient, and the probability of failure with irradiation shift model of 10CFR50.61a is larger than that of R.G.-1.99 rev. 2 at the fluence ranges 0.2×1019 n/cm2 to 0.5×1019 n/cm2.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A002. doi:10.1115/PVP2017-65262.

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life.

Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a).

This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A003. doi:10.1115/PVP2017-65384.

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals.

There are several codes available for addressing the requirements of GDC-4. This paper addresses three of these codes: (1) xLPR 2.0; (2) PROLOCA; and (3) PROMETHEUS. Each of these codes is described and applied to a representative plant where active degradation mechanisms have been found. Conclusions about the design, results, and interpretation of the results is then provided. In all cases the probability of failure of the pipe is found to be extremely low when the crack inspections and leak detection systems are modeled.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A004. doi:10.1115/PVP2017-65921.

In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the standards developed by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the statistical distributions of various influence factors related to ageing due to the long-term operation. At Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, JAEA has developed a guideline for the PFM based structural integrity assessment of RPVs to promote the applicability of PFM in Japan and achieve the objective that an engineer/analyst who familiar with the fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body (general requirements), explanation (guidance), and several supplements. The technical basis for PFM analysis is also provided, and the new information and better fracture mechanics models are included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A005. doi:10.1115/PVP2017-65950.

A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. One of the verification methodologies of PFM codes is that the result evaluated by using each function of PFM code is compared with a theoretical value. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A006. doi:10.1115/PVP2017-66003.

Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessments of reactor pressure vessels (RPVs) by considering the inherent probabilistic distributions of various influence factors. For practical applications, several evaluation models are improved, and have been implemented into the current PASCAL code. In this paper, the improvements of PASCAL are introduced firstly, such as the evaluation method for underclad cracks, treatments of the complicated welding residual stress distribution, and evaluation models for the warm pre-stressing effect. In addition, the effects of these improvements on failure probability or failure frequency of RPVs are investigated by performing PFM analyses for domestic RPVs using PASCAL. From the analysis results, the effects of the improved evaluation models are discussed.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A007. doi:10.1115/PVP2017-66004.

A probabilistic fracture mechanics (PFM) analysis code called PASCAL has been developed by the Japan Atomic Energy Agency to evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on Japanese data and Japanese methods published for or provided in Japanese codes and standards. To verify this code, benchmark analyses were carried out using the FAVOR code, which was developed in the United States and has been utilized in nuclear regulation. The results of these analyses confirmed with great confidence the applicability of PASCAL to failure probability and frequency evaluation of Japanese RPVs. An outline of PASCAL, the benchmark analysis conditions and analysis results are reported in this paper.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A008. doi:10.1115/PVP2017-66101.

Technical requirements for analytical evaluation of in-service Zr-2.5Nb pressure tubes in CANDU(1) reactors are provided in the Canadian Standards Associate (CSA) N285.8. The evaluation must address all in-service degradation mechanisms including the presence of in-service flaws. Flaws found during in-service inspection of CANDU Zr-2.5Nb pressure tubes, including fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws, dummy bundle bearing pad fretting flaws, erosion-shot flaws and crevice corrosion flaws, are volumetric and blunt in nature. These in-service flaws can become crack initiation sites during pressure tube operation and potentially lead to pressure tube failure. Any detected flaws that do not satisfy the criteria of acceptance as per CSA N285.4 must be analytically evaluated to justify continued operation of the pressure tube. Moreover, the risk of pressure tube failure due to presence of in-service flaws in the entire reactor core must be assessed.

A review of assessment of the risk of pressure tube failure due to presence of in-service flaws in CANDU reactor core is provided in this paper. The review covers the technical requirements in the CSA N285.8 for evaluating degradation mechanisms related to flaws in the reactor core. Current Canadian industry practice of probabilistic assessment of reactor core for pressure tube failure due to presence of in-service flaws is described, including evaluation of flaws for crack initiation, subsequent crack growth to through-wall penetration, and pressure tube rupture due to unstable crack growth prior to safe shutdown of the reactor. Operating experience with the application of probabilistic assessment of reactor core for the risk of pressure tube failure due to presence of in-service flaws is also provided.

Topics: Pressure , Failure , Risk
Commentary by Dr. Valentin Fuster
2017;():V01BT01A009. doi:10.1115/PVP2017-66102.

On-going assessments of the impact of active degradation mechanisms in US nuclear power plants previously approved for leak before break (LBB) relief are being performed with, among other methods, the extremely low probability of rupture (xLPR) code being developed under a memorandum of understanding between the US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) [1]. This code finished with internal acceptance testing in July of 2016 and is undergoing sensitivity and understanding analyses and studies in preparation for its general release. One of the key components of the analysis tool is the integration of inspection methods for damage and the impact of leak detection on the risk of a pipe rupture before the pipe is detected to be leaking. While it is not impossible to detect a crack or defect that is less than 10% of the pipe wall thickness current ASME code does not give credit for inspections identifying a crack of this size. This study investigates the impact of combining the probabilistic analysis results from xLPR with a pre-existing flaw to determine the change, if any, to the risk. Fabrication flaws were found to have a statistically significant impact on the risk of rupture before leak detection.

Topics: Manufacturing , Risk
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in ASME Codes and Standards

2017;():V01BT01A010. doi:10.1115/PVP2017-65102.

Material selection decisions for advanced reactor concepts are frequently based on simple consideration of required wall thickness for a particular component and the resultant cost based on averaged cost per unit, usually by weight. However, this approach does not take into consideration the overall impact of other material properties on design feasibility. An example would be the interrelated roles of thermal conductivity, thermal expansion and creep-strength on the design of components to withstand cyclic and sustained loading. The problem is that this would nominally require a detailed design and loading definition. However, as presented herein, a meaningful comparison can be achieved by selective evaluation of the ratios of the material properties required to achieve a particular performance goal for a particular design objective; for example, the relative ability to accommodate axial thermal gradients in a pressurized cylindrical vessel. This paper covers the development of such critical parametric ratios for a number of component elements and loadings and illustrates their application.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A011. doi:10.1115/PVP2017-65274.

Software is being developed to aid assessment procedures of components under specified loading conditions in accordance with the elevated temperature design requirements for Class A components in ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB, Subpart B (HBB). There are many features and alternative paths of varying complexity in HBB. The initial focus of this program is a basic path through the various options for a single reference material, 316H stainless steel. However, the program will be structured for eventual incorporation all of the features and permitted materials of HBB. This paper focuses on a description of the overall program, particular challenges in developing numerical procedures for the assessment, and an overall description of the approach to computer program development.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A012. doi:10.1115/PVP2017-65309.

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to Class 2 and 3 systems, so that safety issues regarding short-term system operation are minimized. The first version of the Code Case was published in 2000. Since then, there have been four revisions to the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the four revisions have been previously published.

There is currently work underway to employ the methods given in N-513 for a new and separate Code Case for higher pressure piping applications. This paper provides the technical basis for the proposed Code Case that includes a structural integrity evaluation and consideration of potential jet thrust forces. In addition, discussion is provided on additional Code Case requirements considering the application to higher pressure systems in order to bolster defense-in-depth. Note that the proposed Code Case still maintains the temperature limit given in N-513.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A013. doi:10.1115/PVP2017-65399.

In 1974, the Level D Service Limits for Section III, Division 1, Class 1 components were published in Non-Mandatory Appendix F titled “Rules for Evaluation of Service Loading with Level D Service Limits”. Over the past 40 years, the scope of Appendix F has been expanded to be applicable to certain Class 1, Class 2 and Class 3 components and supports in Division 1 as well as in Division 3 and Division 5. With each addition, the organization and implementation of the rules in Appendix F became more cumbersome for the user and consistency between the Appendix and the Code Books1 was not maintained. At the same time, the use of these rules has evolved to the point where the non-Mandatory Appendix is essential the default for Level D Service Limits. Starting in the 2017 Code edition, the component design rules will reference Mandatory Appendix XXVII when Design by Analysis is used to determine Level D Service Limits. This paper describes the methodology utilized to convert Non-Mandatory Appendix F to Mandatory Appendix XXVII which includes the history of the Level D Rules in the ASME Code, the philosophy taken to convert Non-Mandatory Appendix F to Mandatory Appendix XXVII, and an overview of the new Appendix XXVII. The approaches to ensure identical safety margins are maintained and the basis for adding or clarifying three allowable stress limits are also included.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A014. doi:10.1115/PVP2017-65400.

In 1974, the Level D Service Limits for Section III, Division 1, Class 1 components were published in Non-Mandatory Appendix F titled “Rules for Evaluation of Service Loading with Level D Service Limits”. Over the past 40 years, the scope of the Appendix F has been expanded to be applicable to certain Class 1, Class 2 and Class 3 components and supports in Division 1 as well as in Division 3 and Division 5. With each addition, the organization and implementation of the rules in Appendix F became more cumbersome for the user and consistency between the Appendix and the Code Books1 was not maintained. At the same time, the use of these rules has evolved to the point where the non-Mandatory Appendix is essential the default for Level D Service Limits. Starting in the 2017 Code edition, the component design rules will reference Mandatory Appendix XXVII when Design by Analysis is used to determine Level D Service Limits. In particular, the component design rules, or rules specific to design of components and not Design by Analysis, were removed from Appendix XXVII and placed in the appropriate Code Book. This approach resulted in noteworthy updates to the support rules in Subsection NF, the core support rules in Subsection NG, the valve rules in NB-3500, and the piping rules in NB/NC/ND-3600. The detailed approach used to incorporate the component design rules into each Code Book are presented in this paper.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A015. doi:10.1115/PVP2017-65418.

An ASME Section III Division 5 code case, N-861, for the evaluation of strain limits based on the elastic-perfectly plastic (EPP) methodology has recently been published. A key feature of the EPP methodology is the application of the EPP finite element analysis method with a pseudo yield stress to bound component response under elevated temperature cyclic service. The simplified inelastic approach in Division 5, Appendix HBB-T that is based on the elastic analysis results is not applicable at the elevated temperature range where creep and plasticity cannot be distinguished and unified viscoplastic model is required to describe the deformation behavior. The EPP strain limits code case overcomes such limitations. It also has the distinct advantage that stress classification, which is required by the simplified inelastic approach, is not needed. Thus, it is ideally suited for modern-day finite element technology. The conservatism of the EPP strain limits code case was verified for some simple geometries. In this paper, a viscoplastic constitutive model calibrated to the experimental data for 316H stainless steel is used to conduct a full inelastic analysis. The calculated strain accumulation is compared with that obtained from using the EPP code case approach.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A016. doi:10.1115/PVP2017-65455.

Combined load and displacement controlled test results are being used to develop and verify a simplified component design methodology for elevated temperature service. A key feature of the proposed design methodology is the use of rules for the evaluation of strain limits that are based on the application of elastic-perfectly plastic (EPP) analysis method with a pseudo yield stress to bound component response to applied loadings. The resulting strain ranges would then be used as input to the evaluation of creep-fatigue damage. This paper describes testing with two separate servo-hydraulic machines that are electronically coupled in parallel to represent combined loading effects on a two-bar test configuration.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A017. doi:10.1115/PVP2017-65457.

The Simplified Model Test (SMT) is an alternative approach to determine cyclic life at elevated temperature and avoids parsing the damage into creep and fatigue components. The Elastic-Perfectly Plastic (EPP) combined integrated creep-fatigue damage evaluation approach incorporates the SMT data based approach for creep-fatigue damage evaluation into the EPP methodology to avoid the separate evaluation of creep and fatigue damage and to eliminate the requirement for stress classification as in current methods; thus greatly simplifying evaluation of elevated temperature cyclic service. The conceptual basis of the SMT approach is that if the effects of plasticity, creep and strain redistribution in the SMT specimen result in a stress-strain hysteresis loop that envelopes the hysteresis loop at the peak strain location in the component, then the SMT results can be used to assess the cyclic damage in the component.

The original SMT concept (Jetter, 1998) considered that the effects of sustained primary stress loading could be safely neglected because the allowable local stress and strain levels were much higher than the allowable sustained primary stress levels. This key assumption requires experimental verification. The influence of the internal pressure on SMT creep-fatigue life is demonstrated and the effect of primary load on the SMT design approach is discussed.

Topics: Creep , Fatigue , Alloys , Testing
Commentary by Dr. Valentin Fuster
2017;():V01BT01A018. doi:10.1115/PVP2017-65754.

The ASME Materials Properties Database has been under development in the past few years to support the ASME Codes and Standards under the supervision of the Boiler and Pressure Vessel Code Committee on Materials. With the guidance of its Working Group on Materials Database, the project has completed the Phase I development for the Data File Warehouse that offers a depository for various files containing ASME Code Week records, materials test data from codification inquiries, and information associated with code rules development. While the database is in operation, the development has continued into Phase II to create a relational Digital Database that offers customized and relational schemas with advanced software functionalities and tools facilitating digital data processing and management. This paper discusses the current status of the project including its development management, database components and features, business operation, and future growth. Some issues and prospective resolutions for meeting the needs and requirements from Codes and Standards are also discussed.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A019. doi:10.1115/PVP2017-65943.

Generally, post-weld heat treatment is applied to decrease welding residual stress and improve the mechanical properties and microstructure of weldment, and its performance has been recommended for many years [1, 2]. However, current steel-making technology has improved significantly and, steel toughness levels have generally improved substantially [1]. Additionally for several quenched and tempered steels, it is reported that in some cases, mechanical properties such as tensile strength and impact toughness are degraded after post-weld heat treatment [3]. In addition, for large steel assemblies, post-weld heat treatment can be expensive, so that there is an economic incentive to avoid post-weld heat treatment [2]. The research presented here suggests a way to exempt post-weld heat treatment for SA-508 Grade 1A material, which is used for pressure vessels in nuclear power plants, by considering both mechanical properties and residual stress to simplify the welding procedure. Weldments made of 120 mm thick SA-508 Grade 1A should be post-weld heat treated, according to current ASME BPV Code. In order to increase the PWHT exemption thickness to 120 mm, we performed mechanical tests using welding coupons without PWHT; the test results satisfied current mechanical property criteria. We present a residual stress acceptance criterion based on brittle fracture criteria in this research.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A020. doi:10.1115/PVP2017-66069.

Alloy 617 (UNS N06617) 52Ni-22Cr-13Co-9Mo is the leading candidate material for the intermediate heat exchanger for the very high temperature reactor (VHTR). An ASME Task Group on Alloy 617 Qualification has drafted a Code Case for Alloy 617 to allow construction of components conforming to the requirements of Section III, Division 5, Subsection HB, Subpart B “Elevated Temperature Service” for service when Service Loading temperatures exceed the temperature limits established in Subsection HB, Subpart A.

There are two categories of allowable stress levels in Division 5, Subsection HB, Subpart B. The first category, identified as So, applies to Design Loadings and the second category, identified as Smt, applies to Service Loadings. Both categories are based on the lesser of time dependent and time independent properties. The time dependent properties control at higher temperatures where creep effects are significant.

The Design condition time-independent allowable stresses are determined per the criteria given in Appendix 5 of Section II, Part D and are based on extrapolated 100,000-hour creep properties. The time dependent allowable stress values for Service Loadings, St, as a function of time and temperature, up to the maximum design lifetime and use temperature, are required for the applications of the Service condition limits. The criteria for establishing St for base metal for each specified time, t, are based on the lesser of

1. 100% of the average stress required to obtain 1% total strain

2. 80% of the minimum stress to cause the initiation of tertiary creep, 67% of the minimum stress to cause rupture

3. Minimum stress-to-rupture curves are also required in the applications of the Service condition limits.

In this paper, the technical background for time dependent allowable stress values for Alloy 617 is presented. These values are being recommended to the ASME Code committees for approval, but are subject to change upon feedback from Code committees.

Topics: Alloys , Stress
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Chinese Codes and Standards

2017;():V01BT01A021. doi:10.1115/PVP2017-65018.

It is well known that foundation settlement of tank is particularly severe, and can produce distortion and stress of the tank, especially differential settlement around the circumference of the foundation below the shell of large-volume tank. The settlement standards involving European EEMUA 159-2003, American API 653-2009, and Chinese codes SH/T 3123-2001, SY/T 5921-2011 for in-service assessment of large-scale storage tank were reviewed and discussed. Finite element model for strength assessment of large-scale oil storage tank was developed based on actual field data of tank foundation settlement. The whole stress distributions and deformation of seven large-scale oil storage tanks in a depot in China were analyzed under the conditions of the practical pressure test through finite-element method. It also provides a comparison between an analytical model based on settlement criteria and a finite element model that replicates field operating loading and settlement conditions of storage tanks. A basis for comparison between models was established from the maximum allowable settlement and stress values. It was found that results from settlement standards of tank in China and other countries were more conservative than those from FEA, and SY/T 5921 in China made most stringent requirements for the tank settlement. The evaluation indicators of differential foundation settlement around the tank circumference are unreasonable in standards and rules mentioned above, the structural response of tank such as stress and deformation under foundation settlement should be considered sufficiently.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A022. doi:10.1115/PVP2017-65052.

The paper briefly gives a summary of standard developments on fiber-reinforced plastic pressure vessels home and abroad. The management and basic technical requirements of FRP vessels are presented in the latest edition Chinese code Supervision Regulation on Safety Technology for Stationary Pressure Vessel (TSG 21-2016). Primary contents of China National Standard General Requirements of Fiber Reinforced Plastics Pressure Vessel (draft standard for approval) are introduced. Comparisons and investigations on FRP are conducted based on difference between China National Standard and ASME BPV CODE X -2015 Fiber-Reinforced Plastic Pressure Vessels, focusing on application scope, design qualification, procedure qualification, inspection and so on. The research will lay a solid foundation for Chinese development in the FRP fields in the future.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A023. doi:10.1115/PVP2017-65092.

A series of numerical simulation about gas-solid erosion for feed type tee have been taken out. The gas-solid two phase flow was formed in the tee with the solid particles coming from the top of the tee pipes and air blowing from the left side. Tee pipes erosion situation was simulated by DPM model in Fluent software. The serious erosion location in the tee pipes was analyzed with different speeds of solid and air. The reasonable distribution method of the particle velocity and gas velocity was put forward and the particles were remained in the intermediate position of the pipes. So the collision with the wall was reduced, and the pipeline erosion rate was slowed down, in addition, the service life of pipes was prolonged.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A024. doi:10.1115/PVP2017-65100.

Cold-stretching is an effective way to make the pressure vessels lighter. A test vessel with austenitic stainless steel was designed and manufactured based on the current criterion, and the cold-stretching process was carried out. The deformation of each part of the vessel under the pressure of each stage were measured by the resistance strain gauge in its measurement range. The finite element method is used to calculate the deformation trend of the whole process of cold-stretching. In addition, the strain of complex deformation at nozzle and its surrounding was analyzed by FEM.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A025. doi:10.1115/PVP2017-65279.

The fitness-for-service assessment to fire damage for pressure vessels was considered to be based on the material deterioration and performance degradation associated with heat exposure. The identification of thermal damage zones after exposure to fire was proposed and provided in the API 579-1/ASME FFS-1 Standard. However, the more explicit quantitative relationships between the heat exposure conditions and the performance degradation degree of the pressure vessels suffering fire were not reported in detail with the thermal damage zone metallurgical analysis, which was not available in the Standard. Therefore, the present research was conducted on the influences of fire suffering test and heat exposure, under different thermal conditions, on the micro-structure evolution and mechanical performance of austenitic stainless steels and carbon steels for pressure vessel equipment. And the metallurgical analysis results described some typical appearances in micro-structure observed in the materials experienced to fire and heat exposure. Moreover, the quantitative degradation of mechanical properties was investigated via multiple testing means such as mechanical tensile test at room temperature and low temperature, the Charpy impact testing, the torsion testing, and the hardness measurement. The present research provided data accumulation of material deterioration and performance degradation was believed to be benefited to the fitness-for-service assessment of pressure vessel after exposure to fire. The material thermal degradation mechanism and the fitness-for-service assessment process of fire damage behavior was further discussed.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A026. doi:10.1115/PVP2017-65282.

The technical evaluation core of pressure piping components after exposure to fire was performed by considering various factors affecting the pipelines subjected to fire damage, including the determination of material deterioration degree, and the fitness-for-service assessment of the piping equipment. Based on the analysis results and testing data, the safety status and performance loss of the pressure piping after exposure to fire could be obtained. On this view, the influences of fire test and heat exposure as the fire-condition simulation on the grade X70 pipeline steel (API Spec.5L), which was widely used for piping equipment, was carefully investigated. According to the division of fire zone with different factors of temperature and fire suffering preservation, the material micro-structure deterioration and performance degradation were analyzed in detail, after that the grade X70 pipeline steel was heat-treated in muffle furnace to simulate the fire exposure under different conditions of the temperature, heat holding period and cooling mode. In the present foundational research, the mechanical tensile tests, Charpy impact tests, hardness measurements and the metallographic examinations were then conducted in detail. Thus, the data accumulation of performance degradation was believed to benefit to the fitness-for-service assessment of piping components of the grade X70 pipeline steel. Moreover, the material thermal degradation mechanism and the fitness-for-service assessment process of the grade X70 pipeline steel after exposure to fire was further discussed.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A027. doi:10.1115/PVP2017-65430.

Partial safety factor (PSF) is a reliability approach for considering the variance of parameters in flaw assessment procedure in major fitness-for-service (FFS) codes, such as recent API579 and BS7910 codes, but is still not adopted in Chinese FFS code GB/T 19624-2005. This study investigated the derivation method for PSFs based on GB/T 19624 procedure. The limit state equations for PSFs calculation were proposed based on GB/T 19624 level 2 failure assessment diagram (FAD). The distribution of random variables was determined according to China’s domestic features. The first order reliability method (FORM) and second order reliability method (SORM) were employed as reliability analysis methods, and the calculated results were both compared with that simulated using Monte Carlo method. The PSFs at different target reliability levels were established and compared with that in API 579 and BS 7910. The method proposed in this study provides a basis for introducing PSF approach into Chinese FFS code.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A028. doi:10.1115/PVP2017-65492.

In the past, shakedown evaluation was usually based on the elastic method that the sum of the primary and secondary stress should be limited to 3Sm or the simplified elastic-plastic analysis method. The elastic method is just an approximate analysis, and the rigorous evaluation of shakedown normally requires an elastic-plastic analysis. In this paper, using an elastic perfectly plastic material model, the shakedown analysis was performed by a series of elastic-plastic analyses. Taking a shell with a nozzle subjected to parameterized temperature loads as an example, the impact of temperature change on the shakedown load was discussed and the shakedown loads of this structure at different temperature change rates were also obtained. This study can provide helpful references for engineering design.

Topics: Nozzles , Shells
Commentary by Dr. Valentin Fuster
2017;():V01BT01A029. doi:10.1115/PVP2017-65494.

The thermal load is one of important design condition that should be considered carefully in engineering practice. In most instances, the heat source is located inside the vessel, which causes a temperature gradient along the thickness, especially when the thickness is large. In this case, secondary stress should be considered and thermal ratcheting should be checked. In this paper, a thin-walled ellipsoidal head with heating spiral was studied. In this structure, temperature is uniformly distributed along the thickness but changes alternately between hot and cold along the meridional direction. This has a significant effect not only on the head itself but also on the nozzle. For the nozzle, its elastic support condition has been changed and then its stress distribution will also be changed. In this paper, several cases have been calculated and some laws are established. Finally, some useful conclusions and suggestions are proposed for engineering design.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A030. doi:10.1115/PVP2017-65577.

The understanding of hydrogen embrittlement (HE) is of significant importance and fundamental interest owing to its negative effects on industrial metallic materials. The effect of solute H on the void coalescence and growth needs to be clarified. Using molecular dynamics simulation, the evolution of preexisting nano voids is studied in the presence of H atoms. As the per unit area concentration of trapped H atom on void surface reaches 0.45 /Å2, the movement of void is observed. It proceeds along with the interdiffusion of H and Fe atoms around the voids. Strain-mediated diffusion of H atoms from void surface to the zone between nearest voids occurs at first. Then the Fe atoms are affected by migrated H and diffuse in the opposite direction following the principle of energy minimization. Such mechanism can help us understand the formation of high pressure bubble at nano scale. Based on this useful information, some methods could be obtained to prevent the growth of voids further, such as strengthening the stability of metal lattice around voids by dopant etc.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A031. doi:10.1115/PVP2017-65582.

Hydrogenation reactor, a typical equipment in petrochemical industry, usually works in tough condition, such as high temperature, high pressure, with hydrogen gas as medium. 2.25Cr-1Mo is widely used as reactor material. However, with the increase of operating condition, a better material is needed. At present, 2.25Cr-1Mo-0.25V is proved having a better mechanical property in high temperature than that of 2.25Cr-1Mo. Hence, it is very important to study the hydrogen impact on 2.25Cr1Mo0.25V. This paper aims to study the relationship between H atom and metal crystal from microscopic view. Based on the first-principles calculation, the convergence analysis of parameters, the adsorption of H atom on Fe, V and their surfaces have been discussed. The results show that the parameter values of simple crystal surface (110) are less than surface (100), such as energy cutoff, k-point sampling, especially the number of slab layers. Tetrahedral-site is the stable site when H atom exists in bbc Fe, V lattice. And quasi three-ford site is the stable status when atomic H absorption on Fe(110) and V(110).

Topics: Hydrogen , Iron
Commentary by Dr. Valentin Fuster
2017;():V01BT01A032. doi:10.1115/PVP2017-65613.

Tubesheet is the main part of high pressure heater, which is very thick based on chinese code GB151 for the design of heat exchangers. Increased tubesheet with large thermal stress are not conducive to manufacture, heat transmission and detection. The stress and structure of tubesheet are so complex that the time costs too large during the analysis design, and stress classification exists uncertainty. Limit load method contributes to tubesheet lightweight. 3-D finite element model used for analysis design should be simplified reasonably. In this paper, the effect of mechanical model on limit load analysis of high pressure heater tubesheet conforming to the design-by-analysis code is researched. It is found that the tubesheet could pass the plastic collapse assessment, and the thickness of tubesheet could be decreased. The difference between the equivalent sold tubesheet model and the whole tubesheet model exists during plastic collapse assessment. Though the local stress distribution is different, the limit load results occurred plastic collapse by the equivalent sold tubesheet model is close to that by the whole tubesheet model. The limit load occurred plastic collapse is influenced by max circular diameter of tube layout little. The reason is attributed to original tubesheet owning enough rigidity related to thickness, and high stress appeares on the inner wall of jointing of tubesheet with head. The equivalent sold tubesheet model could be used for primary evaluation of limit load, and the whole tubesheet model is suited for partial analysis. The results provide some reference for the design-by-analysis of high pressure heater tubesheet.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A033. doi:10.1115/PVP2017-65617.

The 2.25Cr-1Mo-0.25V steels are widely used in the petroleum chemical industry for the manufacturing of pressure vessels. The multi-pass welding is a critical type of fabrication in hydrogenation reactor. However, very complicated residual stresses could be generated during the multi-pass welding process. The presence of residual stresses could have significant influence on the performance of welded product. In the present work, the transient temperature distribution and residual stress distribution in welding of 2.25Cr-1Mo-0.25V steel are analyzed by using numerical method. An uncoupled thermal-mechanical two-dimensional (2-D) FEM is proposed under the ABAQUS environment. The transient temperature distribution and the residual stress distribution during the welding processes are determined through the finite element method. A group of experiments by using the blind-hole method have been conducted to validate the numerical results. The results of 2-D model agree well with the experiment. The result shows that the maximum welding stress generated at heat affected zone (HAZ) both at the top and bottom surface whether to transverse stress or longitudinal stress.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A034. doi:10.1115/PVP2017-65629.

In this paper, the four-roll plate bending process of 2.25Cr-1Mo-0.25V steel at elevated temperature is investigated by numerical simulation. This 3-D simulation is finished by using the elastic-plastic dynamic explicit finite element method (FEM) under the ANSYS/LS-DYNA environment. The strain softening behavior of 2.25Cr-1Mo-0.25V steel at elevated temperature is presented and discussed. The stress-strain relationship of the steel plate is modeled using a piecewise linear material model, with the stress-strain curve obtained through tensile tests. The plate bending process with a plate thickness of 150 mm is investigated. The amount and position of maximum plastic deformation are analyzed. The present study provides an important basis for the optimization of bending parameters and further investigation of the effect of high-temperature deformation on the resistance to hydrogen attack of 2.25Cr-1Mo-0.25V steel.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A035. doi:10.1115/PVP2017-65632.

Cladding titanium-steel plate is often used in design of pressure vessel. Using large finite element software, models of cladding titanium-steel tube sheets of tubular heat exchangers are established, to investigate the effect of different tube patterns, including triangle tube pattern and square tube pattern on cladding titanium-steel tube sheets. What is more, stress distribution characteristics of cladding titanium-steel tube sheets in different tube patterns are obtained under different tube diameters. Results show that the effect of tube patterns on cladding titanium-steel tube sheets mainly concentrated in tube layout area. Compared with triangular tube pattern, square tube pattern could effectively improve the phenomenon of stress concentration of cladding titanium-steel tube sheet at the connection of base layer and cladding layer, and then improve performance of cladding titanium-steel tube sheet. However, increase of tube diameters makes an adverse effect on performance of cladding titanium-steel tube sheets in both tube patterns. Through reasonable decrease of tube diameter could enhance the performance of cladding titanium-steel tube sheet. Results provide references to study and design heat exchanger with cladding steel tube sheet.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A036. doi:10.1115/PVP2017-65651.

Natural gas transmission pipeline is prone to internal corrosion due to the combination of corrosive impurities in the pipe (such as CO2, H2S and chlorides) and applied pressure of the pipeline, which seriously affects the safe operation of the pipeline. In this work, the corrosion behavior of a typical X70 pipeline steel was investigated by using potentiodynamic polarization and electrochemical impendence spectroscopy (EIS). The polarization and EIS data under different CO2 partial pressures (0–1 atm), H2S concentrations (0–150 ppm), chloride concentrations (0–3.5 wt%) and tensile stress (0–400 MPa) were obtained. The results show that corrosion rate increases with the increase of CO2 partial pressure and chloride concentration, respectively, while first increases and then decreases with the increase H2S concentrations. The corrosion rate is less affected by elastic tensile stress. In addition, a quantitative prediction model for corrosion rate of natural gas pipeline based on adaptive neuro-fuzzy inference system (ANFIS) was established by fitting the experimental data which maps the relationship between the key influencing factors (i.e. CO2 partial pressure, H2S concentration, chloride concentration and tensile stress) and the corrosion rate. The prediction results show that the relative percentage errors of the predicted and experimental values are relatively small. The prediction accuracy of the model satisfies the engineering application requirement.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A037. doi:10.1115/PVP2017-65684.

The 2.25Cr-1Mo-0.25V steel, which has excellent thermal mechanical properties under high pressure and temperature, is widely applied in the pressure equipment. Previous researches show that crack generated in the Heat Affected Zone (HAZ) is one of the main failure mechanisms for the high pressure equipment. Consequently, this work aims at investigating the fracture behavior of HAZ by fracture toughness test. A set of specimens was manufactured with the welding current of 580 A, then, specimens were tempered at 705 °C for 32 h. Along with the fracture toughness test, the fracture properties (e.g. crack initiation, propagation and fracture) of specimens were monitored by acoustic emission (AE). Comparing the AE amplitude of each specimen, the crack initiation point was verified by the first peak of AE result, and the analysis of deformation work at the initiation crack point has also been carried out to investigate fracture properties and the fracture toughness of HAZ of the 2.25Cr-1Mo-0.25V steel. Then the relationship between deformation work and crack length has been studied, and the fracture toughness of HAZ could be obtained. Analyzing the fracture characteristics of experiment process and fracture toughness of HAZ has a significant guidance for the further applications of 2.25Cr-1Mo-0.25V steel.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A038. doi:10.1115/PVP2017-65687.

Characteristic safety parameter refers to the parameter that reflects the inherent safety margin of pressure equipments subjected to certain failure mechanism. It has three main characteristics. Firstly, it is sensitive to the change in failure mechanism. Secondly, the safety of pressure equipments can be guaranteed by controlling this parameter. Thirdly, it is easy to measure. By real-time monitoring of this characteristic safety parameter, the quantitative assessment of the structural integrity and furhter the diagnosis and warning on the safety of in-service pressure equipments can be realized. In this paper, the definition of characteristic safety parameter is given first for the pressure equipments subjected to several typical failure modes. After that, the selection principle, measurement technique and determination of its critical value, etc., are then introduced by analyzing typical examples. In combination with the technical concepts of the Internet of Things and Big-Data, some research suggestions are proposed with respect to the remote monitoring and diagnosis techniques based on the characteristic safety parameter, including the sensing measurement, monitoring and analysis of big data, real-time diagnosis and early warning of safety condition, etc.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A039. doi:10.1115/PVP2017-65695.

The local post weld heat treatment by electric heating method is widely used to eliminate welding residual stress in processes of manufacture and maintenance of pressure equipment. The key point of local post weld heat treatment is to choose a reasonable heated band width and insulated band width. But the criterions to determine the minimum heated band width and insulated band width are different according to Chinese, European and American standards, which are GB/T 30583-2014, EN 13445-4: 2009 and AWS D10.10/D10.10M :1999, respectively. Taking the local post weld heat treatment for the circumferential butt weld between two thick cylinders both with a 115 mm thickness as an example, numerical simulation is used to compare the wall temperature distribution of the cylinders during the heat preservation stage when the heated band width and insulated band width are chose according to the above three standards, and the numerical simulation was verified by the tested temperature from one field experiment. The results show that the numerical calculation method can accurately predict the wall temperature of the cylinders during the local heat treatment, and the wall temperature of the surfaces on which the heaters are arranged according to the three standards all well meets the requirement of the heat treatment, but the wall temperature of the surfaces without the heaters cannot meet the temperature requirement. So double-side heating and double-side insulating are suggested to be adopted during local post weld heat treatment.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A040. doi:10.1115/PVP2017-65712.

Wall temperatures are a necessary part of the design conditions for pressure vessels. Thermal analysis by numerical simulation is widely used to compute the wall temperatures and thermal stresses of the pressure vessels, and insulation layers outside the pressure vessels are normally included in numerical models to predict more accurate results. However, modeling and meshing the insulation layers introduces more work to designers and more cost to enterprises. In this paper, a simplified calculation method for the wall temperatures of the pressure vessels with insulating layers is presented. An equivalent convective heat transfer coefficient between the outside walls of the pressure vessels and circumstances is derived by theoretical analysis, which takes into account the thermal resistances of the insulation layers and the convective heat transfer coefficient between them and the circumstances, then it can be used in numerical models of the pressure vessels without the insulation layers to compute the wall temperatures. Results of comparative analysis show that the wall temperatures of the pressure vessels by the simplified calculation method agree very well with those using the numerical simulation with the insulation layer models.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A041. doi:10.1115/PVP2017-65906.

Finite element analysis model of magnetic flux leakage testing outside pipeline was established with analysis software in this paper. The leakage magnetic field characteristic curves of defects were obtained using this model. Also, the simulation on different diameter and depth of the cylindrical defects was carried out. And influence on flux leakage field, such as depth of defects corrosion defects was gotten. Then, detection device which can do inspection on pipeline with different diameters was manufactured based on the theory and finite element calculation results. The experiments with this device were carried out on the wall thickness of 8 mm and 6 mm pipeline with artificial defects above or under. And the results showed good agreement with simulation which meets engineering application of pipeline inspection.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A042. doi:10.1115/PVP2017-65957.

Leakage risk management is very important for propellant storage because of its poisonousness. Numerical method was used to study the leakage, volatilization and dispersion of liquid propellant. The results of CFD (Computational Fluid Dynamics) simulation show that leaked liquid UDMH (unsymmetrical dimethylhydrazine) will jet on the ground under the pressure of storage tank and then it disperses forward slowly with the liquid film. Therefore, it is necessary to set a safety cofferdam around the storage tank. Based on the Gaussian dispersion model, the dispersion process of liquid propellant in pools under different atmospheric conditions was simulated. The results show that atmospheric stability and wind speed influence the risk distance after UDMH leaking. The less stability of atmosphere and lower wind speed , the longer the risk distance is. Hence, the field atmosphere condition should be considered when the risk distance is determined.

Topics: Propellants , Leakage
Commentary by Dr. Valentin Fuster
2017;():V01BT01A043. doi:10.1115/PVP2017-66042.

In this investigation, fracture toughness behavior of high strength low alloy (HSLA) steel welded joint was studied using acoustic emission (AE) monitoring. For the design of new structures and for the safety and reliability analyses of operating components, fracture toughness (KIC) values of materials play an essential role. Acoustic emission technique (AET) has been used for determination of fracture toughness based on some observable changes of AE evolutions. However, the occurrence of appreciable plasticity in materials, the friction between the crack surfaces and mechanical noise could generate high emission and may result in some difficulties in precise determination of fracture toughness. Thus, the objective of this study is to propose a new approach to evaluate fracture toughness values and to characterize the fracture process based on AE entropy. Specimens were selected from 2.25Cr-1Mo-0.25V steel welded joint which were thermally aged at 978 K for 8 h. The AE signals generated during fracture processes were recorded and the corresponding AE entropy was calculated based on the probability amplitude distribution from each original AE waveform. The point of crack initiation was identified by the occurrence of sudden rise of AE entropy and the corresponding critical load was used to estimate fracture toughness value. The estimated values obtained from the proposed new approach were compared with those determined by the methodology proposed by compact tension specimen testing according to ASTM standard E399. The results showed that the estimated values were in close agreement with those gained from ASTM standard. It was concluded that AE entropy was an effective parameter to estimate fracture characteristics and fracture toughness values.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A044. doi:10.1115/PVP2017-66123.

Fillet welds in small-diameter pipe socket of pressure vessels always have complicated structures and groove types, which make it easy to produce porosity, lack of fusion, incomplete penetration and other flaws during welding. Therefore, nondestructive testing is a significant and meaningful approach to ensure the quality of welding for pressure vessels’ safety. Ultrasonic testing is the main method for nondestructive testing of pipe fillet welds. However, it is difficult to distinguish between the interference wave and the flaw echo, or to recognize the defect signal, while utilizing conventional ultrasonic testing technology. Additionally, the coupling effect is bad for traditional rigid probe on the concave surface when the probe is inserted into the small-diameter pipe to do the inner scanning. To obtain a good coupling effect, flexible phased array technology was put forward, with a bendable probe made from flexible materials. The probe could be bent and inserted into the inner pipe for longitudinal wave scanning, giving a good matching with the inner wall and replacing the traditional rigid probe. Besides, it is more convenient to conduct the ultrasonic testing, and the focal law could be changed easily according to the curve shape of the inner pipe, without replacing the probe. Thus, scanning and dynamic focusing in multiple angles and directions can be carried out, and the position, distribution and size of the flaws could be displayed intuitively combined with real-time imaging technology. This technology is able to obtain better coupling and detecting effects and solve the technical problem for concave ultrasonic inspection of fillet welds.

Topics: Welded joints , Pipes
Commentary by Dr. Valentin Fuster
2017;():V01BT01A045. doi:10.1115/PVP2017-66162.

The underground storage tank usually has small volume and typically cylinder shape. A magnetic flux leakage (MFL) inspection device was developed for underground storage tank. In order to be suitable for curvature plates, the magnetization structure was designed, which contains three substructure and can cause the non-uniform magnetic field appeared on the cross section. The impact by the non-uniform magnetic field and uneven placement sensors were analyzed. The MFL testing requires effective signal progressing methods. A signal circuit without high-pass filter was used in hardware signal progressing. A zero-phase filter was adopted in the digital signal processing. The defect was characterized by peak-peak value of defect signal. The channel independent calibration method was carried out. An experimental underground tank with a diameter of 2.4m was established. The experimental results for cracks with different depths, lengths and widths were presented, and the results proved that the presented signal progressing was effective.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A046. doi:10.1115/PVP2017-66172.

ASME VIII-2 code [1] and EN 13445 standard [2] give the design criterion against plastic collapse and gross plastic deformation respectively. Duan et al. [3] propose a new criterion combining the advantages of both. This paper introduces two characteristics of the new criterion, and points out that it is a dual criterion against gross plastic deformation and collapse. The Load and Resistance Factor Design form of the new criterion and an application example are given.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A047. doi:10.1115/PVP2017-66220.

Understanding of hydrogen penetration into α-Fe plays an important role in revealing the mechanism of hydrogen embrittlement in Fe-based alloys. This work aims to investigate the penetration process of hydrogen into α-Fe by molecular dynamics simulation method, including how hydrogen changes from molecular to atomic form and how hydrogen atoms enter into the sub-surface. Potential energy difference and atom density are calculated to describe the characteristics of H-Fe interactions and to analysis the invasion process. The simulation results provide an atomic-scale insight into the hydrogen invasion process.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A048. doi:10.1115/PVP2017-66249.

GIS, Gas Isolation Switchgears, is the closed combination of electrical appliances, which is filled with SF6 gas as a dielectric and insulation medium. The gas pressure is up to 0.5 MPa, so the GIS busbars shells are pressure vessels, which should have high reliability. Research about horizontal busbars shell showed that crack may appeared on the connection between the saddle support and the shell, then the gas leakage taken place. In this paper, the GIS busbars system in a certain substation is studied, the reaction force of the supports under the action of interior pressure, gravity, and temperature are calculated used as Autopipe. Then the stress distributions of the busbars shells are analyzed according to the supporting reaction force. At last, the safety of high voltage GIS busbars shell was analyzed.

Topics: Safety , Shells
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in European Codes and Standards

2017;():V01BT01A049. doi:10.1115/PVP2017-65067.

The paper will cover the general approach followed by nuclear code RCC-M [1] of AFCEN, the French Society for Design, Construction and In-Service Inspection Rules for Nuclear Island Components, in codes and standards setting, from the technical and organizational points of views. The latest issue of RCC-M code edited in beginning of 2017 will be explained and commented. Main new topics of activity of RCC-M subcommittee will be also addressed: improvement in nonlinear analysis methods, technical qualification of active equipment, material and manufacturing techniques changes, principles of demonstration of code conformity with regulation, comparison with other codes and standards. The presentation highlights how the industrial experience is currently integrated into the RCC-M codes, and how code members had demonstrated the conformity with European essential safety requirements for pressure equipment. Processes for updates, interpretations and inquiries will be also addressed.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A050. doi:10.1115/PVP2017-66073.

The fatigue assessment of welded joints in different engineering disciplines is usually based on nominal, structural or notch stresses on one hand (elastic concept using component fatigue curves of load controlled test data) and local strains on the other hand (elasto-plastic concept using material fatigue curves of strain-controlled push-pull test data of un-notched and polished standard specimens). The concepts of the first mentioned group are implemented in widespread standards and recommendations such as [1] to [3]. The fatigue assessment procedure of the European standard for unfired pressure vessels (EN 13445-3, Clause 17 & 18 and related annexes) [4] is currently under revision with one focus on the elaboration of user friendly fatigue assessment options for welded components [5]. The current state of the art focuses on the application of an adapted structural hot spot stress approach to the fatigue assessment of welded pressure equipment [5]. Although this is a significant step forward, the implementation of a notch stress approach can furtherly increase the fatigue assessment options by detailed weld seam analysis. The paper focuses on respective methodological proposals and application examples of typical welded joints. The finite element analysis as part of the procedure has to be harmonized with the requirements of the assessment procedure. Of course, the compatibility of the hot spot stress approach and a notch stress approach has to be guaranteed for individual examples. The direct comparison of the different approaches allows for a qualitative evaluation of methods. The application of an appropriate master fatigue curve FAT100 and the limitations with regard of stress/strain ranges in the low cycle fatigue (LCF) regime as well as the fatigue assessment of welded joints with mild weld toe notches is the subject of special considerations. The latest recommendations of German Welding Society (DVS) [6] constitute a reference for the last two subjects raised.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A051. doi:10.1115/PVP2017-66199.

The purpose of this work is to establish a method for safety assessment with respect to plastic collapse, which is consistent for design assessment in accordance with Section III of the ASME Boiler and Pressure Vessel Code and assessment of in-service defects.

In the Swedish nuclear power industry, a procedure is used for assessment of in-service defects that is based on failure assessment diagrams according to the R6-method combined with a safety evaluation system corresponding to Section III and XI of the ASME Code. The Swedish procedure was originally based on R6 Revision 3 combined with the 1995 edition of the ASME Code. In recent years the procedure has been updated to correspond to current versions of R6 and ASME. These updates include a transition to the updated definition of flow stress as well as structural factors against plastic collapse that was included in the 2002 Addenda of ASME 2001 Section XI.

Following these updates, discrepancies were found between plastic collapse analyses made in accordance with Section III and assessments of in-service defects. Due to this, a project was initiated to make the procedure for assessment of in-service defects consistent with Section III analyses. An in-depth study of the flow stress definitions and structural factors in Section XI was initiated. This study found that the structural factors are based on approximations to get material independence. These approximations uses mean values based on a number of materials which was found to be quite inaccurate for many materials used in the Swedish nuclear power industry and can differ up to about 15 % compared to Section III analyses.

A method has been developed to calculate material specific structural factors that gives a near perfect match between plastic collapse according to Section III and the Swedish procedure for assessment of cracks. The method is based on the methodology for structural factors in Section XI, but excludes the assumptions made to get material independent structural factors. The method is presented and comparisons show the agreement to Section III and highlight cases of large discrepancies in Section XI.

Topics: Safety , Design , Collapse
Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Japanese Codes and Standards

2017;():V01BT01A052. doi:10.1115/PVP2017-65314.

This paper describes examinations of the screening criteria of fracture assessment methods for pipes having a circumferential surface flaw. The failure mode screening criteria for ferritic steel pipes are defined in Appendix E-11 of the Rules on Fitness-for-service for Nuclear Power Plants [1] (hereinafter, the FFS Codes) of the Japan Society of Mechanical Engineers (JSME) and ARTICLE C-4000 of ASME Section XI [2] (hereinafter, ASME Sec. XI). However, these screening criteria were assessed under limited conditions; they are not intended for detailed evaluation. Essentially, the fracture mode depends on a stress-strain curve, a J-R curve, flaw shape (length and depth), pipe shape (R/t), etc. First, among these parameters, these that have a large effect on fracture strength evaluation were selected. The important parameters for fracture strength are a stress-strain curve, a J-R curve, flaw depth. Next, failure modes according to these parameters were classified. Finally, a screening method to determine whether the failure mode is Limit Load (LL) or Elastic-Plastic Fracture Mechanics (EPFM) for all pipe materials was proposed. The proposed method is a general method not limiting the applicable pipe materials.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A053. doi:10.1115/PVP2017-66092.

Materials made of Alloy 82/182/600 used in light-water reactors are known to be susceptible to stress corrosion cracking. It is known that the depth a of some cracks due to primary water stress corrosion cracking is larger than the half-length c. The stress intensity factor solution for cracks plays an important role to predict crack propagation and failure. However, Section XI of the ASME Boiler and Pressure Vessel Code does not provide the solutions for cracks with large aspect ratios a/c.

In this study, closed-form stress intensity factor influence coefficients for deep surface cracks in plates are discussed. The crack tip stress distribution is represented by a fourth degree polynomial equation. Influence coefficient tables obtained by using finite element analysis in previous studies are used for curve fitting. The closed-form solutions for the coefficient were developed at the surface points and the deepest points of the cracks with aspect ratio a/c ranged from 1.0 to 8.0. The solutions for the points where the stress intensity factor reaches maximum were also investigated.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A054. doi:10.1115/PVP2017-66183.

In the 2012 version of the JSME Fitness-for-Service Rules for Nuclear Power Plants, the procedure to calculate screening parameter, SC, which is used for selecting the analysis method (limit load controlled by plastic collapse, elastic-plastic fracture mechanics, or linear elastic fracture mechanics), has been revised to reflect a semi-elliptical surface crack. Both limit load solution and stress intensity factor solution are needed to calculate SC, and the solutions for a semi-elliptical surface crack are different from those for a fan-shape surface crack. In this study, the effect of the difference in crack shape on SC is investigated. Through the results on the sensitivity analysis, the adequacy of the evaluation procedure of SC is ascertained.

Commentary by Dr. Valentin Fuster

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

2017;():V01BT01A055. doi:10.1115/PVP2017-65659.

A primary repair option of Light Water Reactors (LWR) components is welding. However, it is known that welding on steels that have been exposed to neutron irradiation [i.e. Reactor Pressure Vessels (RPV) in PWR and Reactor Internals in BWR] can result in Helium Induced Cracking (HeIC). Helium forms from neutron transmutation reactions of Boron (B) and Nickel (Ni) during operation of the plant.

In order to address this issue and establish verified methods for weld repair of irradiated RPV and Reactor Internals materials in Japanese power plants, an investigation denominated WIM (Welding of Irradiated Materials) Project was conducted; the WIM project was carried out between the years of 1997 and 2004 in an analytical conservative manner, correlating the results of weld repaired irradiated materials with the concentration of helium and the heat input used while welding. It was concluded that, under determined conditions, the irradiated materials were able to be successfully welded in accordance with the requirements established in both the JSME and ASME Code Cases. In the light of such discovery, the necessity of establishing and a new code case and revising the standard JSME Rules on Fitness-for-Service concerning the weld repair of irradiated RPV and Reactor Internals steels is currently under investigation.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A056. doi:10.1115/PVP2017-65942.

This paper discusses the current status of ASME Section XI Code activities associated with two methods employing non-metallic materials for repair or replacement of metallic piping: high-density polyethylene (HDPE) and carbon fiber reinforced composites (CFRC).

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2017;():V01BT01A057. doi:10.1115/PVP2017-66017.

Primary water stress corrosion cracking (PWSCC) is a degradation process that has plagued nickel alloy components and welds in the nuclear industry for decades. Numerous mitigation techniques have been developed over the years that help reduce the potential for cracking in nickel alloy components exposed to the primary water environment. One such method is Laser Peening (LP), which improves the stress properties and helps to reduce the potential for crack initiation. The LP process has been applied in Japan to both boiling water reactors (BWR) and pressurized water reactors (PWR) for stress corrosion mitigation. The first application of LP in the US for the nuclear industry was applied in the fall of 2016 to the bottom mounted instrumentation (BMI) nozzles of a PWR. The bottom mounted nozzles are made from Alloy 600 tubing and attached with Alloy 82/182 welds, which are known to be susceptible to PWSCC. In order to prevent crack initiation, it is important for the peening mitigation process to induce sufficient compressive stress on the surface of the susceptible materials. However, it is not practical to take stress measurements directly on the reactor components in order to verify compression. Thus, the magnitude of compression induced by the LP process was verified prior to the application at the plant using mockups of the BMI nozzles. As a part of the qualification process, test coupons were peened and stress measurements were taken using X-ray diffraction (XRD). The results of the stress measurements demonstrate that sufficient surface compression was achieved by the LP process in order to provide PWSCC mitigation. This paper presents and discusses key stress measurement results taken during the qualification process for the first application of LP at a U.S. nuclear plant. Although not directly applicable in this case, the guidance in ASME Code Case N-729 Mandatory Appendix II and MRP-335 for PWR upper head nozzles was generally followed.

Topics: Laser hardening
Commentary by Dr. Valentin Fuster
2017;():V01BT01A058. doi:10.1115/PVP2017-66164.

Water Jet Peening (WJP) has been widely applied to nuclear power plants in Japan as one of mitigation techniques against Stress Corrosion Cracking (SCC) initiation [1]. WJP utilizes high pressure water flow including numerous cavitation bubbles and improves surface residual stress of susceptible materials used in reactor internals from tensile stress to compressive stress without significant plastic deformation, hardening, heating and furthermore retrieval of foreign materials.

An inspection relief for the Primary Water SCC (PWSCC) concerned components, by means of peening technique application, has been discussed among PWR owners in the US for about last 10 years. The topical report on PWSCC mitigation by surface stress improvement (Material Reliability Program (MRP)-335, revision 3-A) was published through the above activities by Electric Power Research Institute (EPRI) MRP [2]. The target components, where PWSCC is concerned, are listed as Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs), such as Control Rod Drive Mechanism Nozzle (CRDMN), and dissimilar metal welds (DMWs) of Reactor Coolant System (RCS) nozzles, and performance criteria for peening are defined in the topical report.

Moreover, the technical basis for PWSCC mitigation by surface stress improvement (MRP-267, revision 2) was published by EPRI MRP [3].The report details numerous data for each peening technique which show the effectiveness in mitigating the PWSCC initiation and its sustainability, i.e. state of stress. The report also includes the process control; covering nozzle diameter, water flow rate, application time, jet stand-off, impingement angle and stationary nozzle time for WJP [3].

RPVHPNs inner diameter (ID), such as CRDMN ID, is in narrower areas than the other target components of peening techniques. Hence the WJP nozzle should be set appropriate condition, e. g. sufficient stand-off distance or angle of the WJP nozzle, in line with the MRP-267 in order to ensure the stress improvement effect by WJP. Further, the reactor pressure vessel head, which has the RPVHPNs including the CRDMNs, is placed on the refueling floor and under atmosphere condition during outage, and therefore, the CRDMNs have to be filled with water by plugging etc. for WJP application on CRDMN ID. Thus the CRDMN ID becomes a closed narrow chamber.

In such a closed narrow chamber, water flow might become complex and disturb the cavitation collapse on the target surface, resulting in decreased stress improvement. Additionally, WJP has been rarely applied in a narrow closed water chamber, and only a few residual stress measurement data are available for such a WJP treated specimen. For the above reason, we has conducted a WJP test utilizing the water chamber and measured the residual stress of the test coupon simulating the CRDMN ID before and after WJP application as our own research.

As a result, an improvement in residual stress was ensured even in an application of WJP in a closed narrow water chamber, which assumes CRDMN ID configuration, and created a depth over the performance criteria (0.01” (0.25 mm) in depth) stated in MRP-335 [2].

As an another applicability study, we developed a WJP tool for Bottom Mounted Instrument (BMI) Nozzles and confirmed that the residual stress of BMI ID and Outer Diameter (OD) can be improved . The background of this study is that BMI nozzle is under discussion for inspection relief as one of the components which are concerned about PWSCC. Especially, BMI ID is narrow area for WJP application; on the other hand it does not need to become a closed chamber since the reactor pressure vessel, which has the BMI Nozzles on the bottom head, is filled with water during outage.

As a result, it is ensured that the residual stress for BMI ID and OD is improved by WJP to a depth of at least 0.2mm which is deeper than the performance criteria for the depth of compressive residual stress of Austenitic Stainless Steel in Japan (3.9 × 10−3” (0.1mm) in depth).

Commentary by Dr. Valentin Fuster

Codes and Standards: Structural Integrity of Pressure Components

2017;():V01BT01A059. doi:10.1115/PVP2017-65090.

The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ∼ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1–3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 × 1019 n/cm2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A060. doi:10.1115/PVP2017-65108.

ASTM A 335-Grade P91 components of steam generators may be critical because of possible steel microstructure changes and/or embrittlement due to the FATT increase during service at high temperature: both phenomena may worsen the material creep behavior globally. Operation temperatures below 600°C such as in the worked case considered herein should be less critical; nevertheless, the worked case plan has included additional controls on microstructure also to have a reference for the future. Present study considers for the worked case steam generator the creep analysis of high-temperature-section (superheater / reheater) two components, outflow tubing and manifold: they may be critical because of the long continued service (110,000 hours or twelve years) and loading conditions (maximum operation temperature and applied stress at the intersection). Aim of the work is to compare life results from the Italian creep code with those predicted by the API 579-1; it also checks compatibility of results from the polynomial models in Italian, ECCC and API 579-1 procedures. Life results based on the Italian-code polynomial function are consistent with those based on the polynomial function proposed in ECCC: With preliminary stresses from pressure formulas, life estimates are a bit more conservative than the ECCC model’s. Finally, life results obtained through the API 579-1 Level 3 assessment appear consistent with those predicted by the Italian creep code, ECCC recommendations application.

Topics: Creep , Steel , Boilers
Commentary by Dr. Valentin Fuster
2017;():V01BT01A061. doi:10.1115/PVP2017-65141.

Deterministic assessment codes can contain large safety factors that give very conservative results. By applying probabilistic analysis to these deterministic assessments, an implicitly accepted probability of failure can be determined. The probability of failure is implicit because it is calculated with the parameter values resulting in a state that is deterministically accepted by the code [2]. When these probabilities are compared for similar deterministic assessments, the excess conservatism can be shown and possibly reduced.

During the present study a probabilistic analysis of the critical crack length initiation was performed. Such analysis led to the formulation of a corrective action proposal to the Master Curve approach given in BS7910:2013 Annex J.

Firstly a deterministic calculation was performed with the Kr-Lr method to define the Critical Length of a through-wall circumferential crack present in a nuclear reactor’s piping. The value of Kmat used in the Kr-Lr method was calculated for a probability of 0.05 and with T0 directly measured (T0 a unique value).

The second step was to pass to probabilistic calculation. Here Kmat was calculated from both T0 directly measured and T0 estimated by Charpy-V tests (T0 as a distribution). The results from these calculations gave the probability of a crack being equal to the Critical Crack Length. Moreover, these results showed that the Tk safety margin introduced in BS7910:2013 Annex J introduce an excess conservatism.

Results from the probabilistic calculations were then compared to the implicitly accepted failure probability Pf (5%) that results from deterministic analysis (T0 considered as a single value) to account for the effects of T0 distribution. An optimized Tk was then found to account for the real uncertainty of the statistical distribution.

Finally, excluding a dependency on the yield stress, the Tk optimization method was generalized. A new correlation for the Tk safety margin is proposed.

Topics: Safety , Optimization
Commentary by Dr. Valentin Fuster
2017;():V01BT01A062. doi:10.1115/PVP2017-65263.

This paper describes the development of a new computer code called Leak Analysis of Piping - Oak Ridge (LEAPOR) which calculates estimates for the leakage rate of water escaping from postulated through-wall cracks in a piping segment of a nuclear power plant cooling water system. The ability of nuclear power plant control and safety systems to detect a piping leak prior to breakage is a fundamental requirement of the leak-before-break concept. The design and assessment of leak-detection systems, therefore, requires the determination of through-wall crack leakage rates covering a significant range of operating and flow conditions. For the primary use case of pressurized water reactors, the coolant is subcooled liquid-phase water at high pressures and temperatures, and the leakage flow regimes can range from adiabatic flow boiling (“flashing”) with non-equilibrium vapor generation inside the crack to orifice flow of a subcooled liquid with vapor generation occurring outside of the pipe. The thermohydraulic Henry-Fauske model (with extensions) for non-equilibrium flashing flow through “tight cracks” has been implemented into LEAPOR.

A primary driver in the development of LEAPOR has been that its Software Quality Assurance (SQA) requirements included evaluations for correctness, consistency, completeness, accuracy, source code readability, and testability. The new code should be prepared to successfully meet the criteria of formal SQA audits. The attributes of maintainability, portability, and extensibility also informed LEAPOR’s layered software architectural design.

The paper presents the results of verification and validation studies carried out with LEAPOR where verification by benchmark comparisons to the results of an independently developed leak rate code and validation against experimental data are described.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A063. doi:10.1115/PVP2017-65458.

Threaded closures for pressure vessels have been in use for decades. Much work has been done to develop safe threaded closures. Threaded closures are very advantageous when there is a need for opening the vessel at intervals for maintenance purposes.

Heat Exchangers are a typical application where there is a need for opening the vessel to get good access to the inside and outside of the tubes for mechanical cleaning, thus maintaining heat transfer efficiency. These are known as Screw Plug Heat Exchangers and are basically U-tube heat exchangers. The tube side normally operates at high pressure and temperature and is closed by a threaded end closure.

Two problems are often encountered in screw plug heat exchangers. These are:

1. Leakage through the gasket at the tubesheet causing intermixing of shell side and tube side fluids, which is unacceptable

2. Jamming of the threaded plug due to deformation of channel barrel

In an earlier paper (PVP2016-63137) these problems were studied for a vessel designed to ASME Section VIII Div. 1. It was found that leakage through the tubesheet gasket could be eliminated by changing the gasket to a grooved metal gasket with covering layers as defined in ASME B16.20.

Preventing leakage from the tubesheet gasket is extremely necessary to get the ultra-low sulphur requirements for clean fuel.

In the work reported in this paper, a procedure for obtaining leak-free performance on a vessel designed to ASME Section VIII Div. 2 was developed and verified using a prototype.

Code formulae for calculation of thickness of various parts normally consider only the need to limit the component stress to be within allowable limits defined in the Code. Allowable stresses for Section VIII Div. 2 construction may be about 18 % higher than the allowable stress for Section VIII Div. 1 construction at design temperature, thereby allowing thinner sections for the same design conditions.

As the thinner sections would deform more, the likelihood of jamming of the end cover could be more severe in ASME Section VIII Div. 2 constructions. Hence this study was additionally undertaken to verify the adequacy of the earlier proposed design methodology, i.e., use of an additional steel ring shrunk fit to the end of the channel to prevent flaring of the channel and jamming of screw threads, for Section VIII Div. 2 constructions.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A064. doi:10.1115/PVP2017-65521.

Clause UG - 39 of ASME Section VIII Division 1 [1] provide rules for compensation of openings in flat stayed/ flat unstayed heads having fitted nozzles.

The rules provided in Clause UG - 39 and its sub clauses apply to all openings other than small openings covered by UG - 36 (c)(3)(a) and provide rules for compensation of openings to those geometries which confirms to the geometric limitations specified therein.

The rules provided in Clause UG - 39 of ASME Section VIII Division 1 are based on area replacement method. This method is also elaborated in WRC Bulletin 335 Aug 1988[4]. The conclusion of this bulletin is applicable to ASME Section VIII Div 1, ASME Section I, ASME B 31.1 and ASME Section III Class 2 and 3. This method requires that the metal cut out by an opening be replaced by reinforcement within a prescribed zone around the opening. This methodology is relatively simple and vast majority of the piping and pressure vessels with openings conforming to this methodology have given satisfactory service.

In Code [1], as such there appears to be no restriction on the location of the nozzle opening, i.e., a header flat head pierced concentrically or eccentrically to locate the nozzle opening as long as the required area is obtained and the stresses are within allowable limits. While both these alternatives would be acceptable in Code [1] constructions, the actual stresses at the header flat heads/nozzle junction may vary considerably.

The work reported in this paper was undertaken to make a comparative study on the effect of unstayed flat head pierced concentrically or eccentrically by using ASME Section VIII Division 1 and to study the stress pattern in both the cases using Finite Element Analysis (FEA) as a referral methodology.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A065. doi:10.1115/PVP2017-65859.

The 2014 Edition of ASME B31.3, Process Piping [1], introduced significant changes to the post weld heat treatment (PWHT) requirements for P-No. 1 carbon steel materials. In particular, PWHT is no longer a mandatory requirement for any wall thickness provided that multi-pass welding is employed for wall thicknesses greater than 5 mm (3/16 of an inch) and a minimum preheat of 95°C (200°F) is implemented for wall thicknesses greater than 25 mm (1 inch). Detailed fracture mechanics analyses have shown that the lack of a mandatory PWHT requirement for thicker P-No. 1 components may result in a significant increase in risk for brittle fracture failures due to near-yield level weld residual stresses. Given the concern throughout the pressure vessel and piping community regarding potential brittle fracture failures, this updated PWHT guidance is examined.

Impact testing requirements and exemption curves were introduced in the 1987 Addenda [2] of ASME Section VIII Division 1 (VIII-1) [3] in Paragraph UCS-66 and extended into ASME Section VIII Division 2 (VIII-2) [4]. During the VIII-2 rewrite in 2007 [5], the available technical and historical basis for the UCS-66 exemption curves was examined and improved to reflect modern fracture mechanics standards. The result of that effort was a systematic approach that can be modified for particular geometries and assumed flaws, if desired. The method used the most modern, fracture mechanics approach for welds in API 579-1/ASME FFS-1, Fitness-For-Service, (API 579) [6] based on the failure assessment diagram (FAD). As a result of explicitly accounting for weld residual stress, two separate sets of exemption curves are provided in VIII-2 [4]; one set for as-welded components and another set for PWHT components. In this paper, a similar approach is summarized to generate exemption curves by establishing newer as-welded and PWHT curves using the Fracture Toughness Master Curve (Master Curve) as documented in upcoming Welding Research Council (WRC) Bulletin 562 [7]. The increased propensity for brittle fracture in as-welded components versus PWHT components is clearly highlighted using this approach. The Master Curve, in conjunction with the elastic-plastic fracture mechanics employed in API 579 [6] provides a means to develop exemptions curves anchored in state-of-the-art fracture toughness technology that can be directly tied to different reference flaw sizes. Additionally, commentary on the appropriateness of the current ASME B31.3 [1] PWHT requirements is offered and the effectiveness of using weld preheat in lieu of PWHT as permitted in the National Board Inspection Code (NBIC) [8] is examined using simplified computational weld analysis.

Commentary by Dr. Valentin Fuster
2017;():V01BT01A066. doi:10.1115/PVP2017-66272.

Thermal analyses of girth welded joints of clads have been carried out using 2D and 3D finite element analysis (FEA) by using the engineering software Abaqus v.2016 (Dassault Systèmes). Transient temperature curves have been generated for different cladding thicknesses (of stainless steel and mild steel). The welding of the two dissimilar materials has been carried out in-house with the aid of a Tungsten Arc weld with dynamic measurement of the temperature profile in the vicinity areas of the welding track using high temperature thermocouples. Comparison of the measured temperature versus the simulation results shows close agreement.

Commentary by Dr. Valentin Fuster

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