3 - Materials

The required thickness of welding tees is neither specified in ASME (2012, “Factory-Made Wrought Buttwelding Fittings,” American Society of Mechanical Engineers, New York, Standard No. B16.9-2012) nor is a clear calculation method provided in codes such as ASME (2016, “Process Piping,” American Society of Mechanical Engineers, New York, Standard No. B31.3-2016). This can lead to uncertainty regarding the pressure capacity of a tee fitting, particularly one that has suffered from erosion or corrosion. Code methods including area replacement (ASME, 2016, “Process Piping,” American Society of Mechanical Engineers, New York, Standard No. B31.3-2016) or pressure-area (ASME, 2015, “Boiler and Pressure Vessel Code Section VIII Division 2,” American Society of Mechanical Engineers, New York, Standard No. BPVC-VIII-2-2015; BSI, 2014, “Unfired Pressure Vessels Part 3: Design,” BSI, London, UK, Standard No. BS EN 13445-3) do not directly account for the effect which the curvature of the crotch region may have on the stress state in the tee. The approach adopted in this work is to liken the geometry of the tee crotch to the intrados of a torus or pipe bend. The shell theory applicable to the torus is adapted for the tee in order to derive a relationship for circumferential membrane stress. An equivalent tube radius is assigned by determining the local radius of shell curvature in the plane passing through the crotch center of the curvature. The actual stresses in the tee crotch are significantly reduced by the adjoining straight portions. This effect is difficult to quantify theoretically and has thus been investigated by means of finite element analysis (FEA)-based assessments. An empirical relationship was then established providing a conservative correlation between the theoretical stresses and the program calculated local stress intensities.

To develop rationalized maintenance plans for nuclear power plants, the characteristics of each plant must be considered. For sodium-cooled fast reactor (SFR) plants, constraints on inspections exist due to the specialty that equipment retaining sodium must be handled, which is one of the important points that must be considered in maintenance rationalization. In this study, we propose a maintenance optimization scheme, which is a design support tool, using risk information to develop a maintenance strategy based on the system based code (SBC) concept. The SBC concept intends to provide a theoretical procedure to optimize the reliability of structure, system and components (SSCs) by administrating every related engineering requirements throughout the life of the SSCs from design to decommissioning. “The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Code Case, N-875, Alternative In-service Inspection Requirements for Liquid-Metal Reactor Passive Components” was developed based on the SBC concept. The basic procedure of the Code Case has also been implemented in “The ASME Boiler and Pressure Vessel Code, Section XI, Division 2,” which provides requirements for Reliability and Integrity Management programs for nuclear power plants, including advanced reactors. The purpose of this study is to establish detailed procedures for the maintenance optimization scheme based on the procedure in Code Case N-875. Furthermore, a quantitative trial evaluation of the core support structure of the next SFR under development in Japan is also performed using the maintenance optimization scheme.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code currently only allows five materials for use in construction of nuclear components for high temperature service. These are: 2.25Cr-1Mo and V-modified 9Cr-1Mo steels, Types 304 and 316 stainless steels and the high nickel Alloy 800H. Since 2005, the US high temperature gas-cooled reactor program has been characterizing elevated temperature mechanical properties of Alloy 617 as the leading candidate construction material for the intermediate heat exchanger. After analysis of these experimental results, along with historical data and additional results available through the Generation IV International Forum, Very High Temperature Reactor, Materials Program Management Board Materials Handbook, a draft ASME Code Case to allow nuclear construction with Alloy 617 for temperatures up to 1750°F (954°C) has been developed. This report contains the Code Case for Low Temperature Service Construction of Section III, Division 5, Subsection HB, Subpart A, Class A and Subsection HC, Subpart A, Class B components, which has been approved in Section II, Materials, and Section III, Rules for Construction of Nuclear Facility Components. Supporting technical justification for the low temperature Code Case is also included. This Code Case allows use of Alloy 617 up to 800°F (425°C). This report also contains an updated draft of a Section III, Division 5, Subsection HB, Subpart B, Class A Code Case for Alloy 617 to qualify it for use in construction of nuclear components up to 1750°F (954°C) for service life up to 100,000 hours. The draft contained in Appendix 4, subject to editorial revision and approval by the ASME Special Task Group on Alloy 617 Code Qualification, will be submitted for approval by letter ballot by the appropriate ASME Committees. The technical justification supporting the Code Case is presented in Appendix 5 of this report. This background document is part of the information package that will be submitted with the Code Case for ballot.

Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

Reviewed by: Chronicles of Mechanical Engineering in the United States ed. by Thomas H. Fehring and Terry S. Reynolds Peter Liebhold (bio) Chronicles of Mechanical Engineering in the United States Edited by Thomas H. Fehring and Terry S. Reynolds. New York: American Society of Mechanical Engineers, 2021. Pp. xvi + 405. A fiftieth anniversary is a milestone of longevity that many organizations never achieve. It takes a lot to maintain leadership continuity, overcome economic challenges, and keep membership involved for half of a century. To mark the golden anniversary of the History and Heritage Committee of the American Society of Mechanical Engineers (ASME), the society commissioned Chronicles of Mechanical Engineering in the United States. This book is a collection of over seventy vignettes, mostly biographical and thematic with a few landmark stories. In a surprising organizational decision, 90 percent of the entries—many written by Fritz Hirschfeld between 1976 and 1980—are reprints from the last fifty years of the ASME journal Mechanical Engineering. The book editors, Thomas Fehring and Terry Reynolds, contribute short chapter heads. Organized into eight topical chapters, Fehring and Reynolds do not argue a grand theory of engineering through their selection of articles. The simple goal of this book is to celebrate the achievements of engineers and the ASME. Other historians have followed a different approach: Bruce Sinclair's Centennial History of ASME explored the role of engineers and professional societies through the tensions of professionalizing the group. Eugene Ferguson's Engineering and the Mind's Eye railed against engineering's dependency on academic mathematical modeling and suggested the need for practical experience. Henry Petroski argued in Success through Failure that engineering success leads to overconfidence and catastrophe and that humility and insight stem from disaster. The intended audience for Chronicles of Mechanical Engineering is a bit mysterious. Fehring and Reynolds explain the articles lack academic rigor (footnotes, bibliography, etc.) and are not for historians. They suggest the collection of stories presents an important learning opportunity for the general public. But what if the sum is less than the parts? The real audience here is ASME members reifying the importance of their organization. Most historians agree that technology played a critical role (both positive and negative) in the development of the United States. As the nation matured and society transitioned from agrarian to industrial, engineers acted as innovators and agents for change. In this crucible of transformation, engineers were at the center, developing machines, engines, and the infrastructure of modern society. A book that broadly tells the history of engineering is a great opportunity for presenting epiphany moments, helping readers learn from the past and understand the complexity of life moving forward. [End Page 649] A challenge for the book is whether old versions of history are still useful. Some of the entries may be old but are interesting and resonate today. Frank Wicks's "First Flights" uses the birth of powered flight to provide insight into the failure of a government-supported academic engineer (Samuel Langley) and the success of untrained independent tinkerers (Frank and Wilbur Wright). Charles Beardsley's retrospective on the Hydrolevel case is one of the few essays that are not celebratory. Instead, he details how the ASME, a voluntary standards-setting organization, got into serious antitrust troubles through conflicts of interest. Reprinting classics can be challenging, and not all the entries weather well. Written for audiences of the day, most do not address questions of race, class, or gender. In fact, one of the most surprising problems in this book is diversity. ASME should be painfully aware of STEM stereotypes and the need to bring women and people of color into their organization. This book could have reinvented the ASME through a revisionist history that highlights both old heroes and the contributions of people overlooked in the past. Today, ASME membership is 53 percent white, 16 percent Latino, 14 percent Asian, and about 82 percent male. But the book is mostly a "great man" history highlighting only two women (Lilian Gilbreth and Emily Roebling) and no people of color. The editors acknowledge that "the articles are generally not critical and that they focus primarily on engineering success rather than failure." The concept...

Fatigue is a common failure mode in mechanical components that are subject to repeated load cycles. In nuclear plants the cyclic loading is usually associated with variations of pressure and temperature. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provide fatigue evaluation procedures in ASME III Section NB-3220. It is required to calculate a Cumulative Usage Factor (CUF). The Code criterion for acceptance is that the calculated CUF must be less than one. The ASME method is deterministic and it is assumed that uncertainties are accounted for by largely un-quantified inherent safety margins in the assessment. The conservatism in the ASME III Code fatigue evaluation methods may come from the assessment procedure used and/or the fatigue design curve. In this paper the inputs to an ASME III fatigue assessment that is based upon a finite element determination of stresses are outlined and assessed. Based upon these inputs possible areas of conservatism within the analysis are then discussed and conclusion drawn on the inherent margins. In addition a probabilistic model that uses Monte Carlo (MC) methods is developed. By assuming the scatter in the fatigue data, used to generate the design fatigue curve, to be the most significant random variable, the MC model was used to calculate the conditional probability of initiation associated with CUFs from a number of plant representative examples.

The American Society of Mechanical Engineers (ASME) Performance Test Codes (PTCs) have provided the power industry with the premier source of guidance for conducting and reporting performance tests of their evolving base technologies of power producing plants and supporting components. With an overwhelming push for renewable energy in recent years, ASME PTCs are in the development of similar standards for the testing of concentrating solar thermal technologies based power plants by the formation of a committee to develop “PTC 52, Performance Test Code on Concentrated Solar Plants”, on July 2009. The U.S. Department of Energy’s (DOE) SunShot Initiative goal is to reduce costs and eliminate market barriers to make large-scale solar energy systems cost-competitive with other forms of energy by the end of the decade. The ASME PTC-52 similarly removes critical barriers hindering deployment and speeds the implementation of concentrating solar power technologies by reducing commercial risk by facilitating performance testing procedures with quantified uncertainty. As with any commercialization of power producing technologies, clearly defining risk and providing methods to mitigate those risks are essential in providing the confidence necessary to secure investment funding. The traditional power market accomplishes this by citation of codes and standards in contracts; specifically ASME PTCs which supply commercially accepted guidelines and technical standards for performance testing to validate the guarantees of the project (Power Output, Heat Rate, Efficiency, etc.). Thus providing the parties to a power project with the tools they need to ensure that the planned project performance was met and the proper transfer of funds are accomplished. To enable solar energy systems to be fully embraced by the power industry, they must have similar codes and standards to mitigate commercial risks associated with contractual acceptance testing. The ASME PTC 52 will provide these standard testing methods to validate Concentrating Solar Power (CSP) systems performance guarantees with confidence. This paper will present the affect that solar resource variability and measurement accuracies have on concentrating solar field performance uncertainty based on calculation methods like those used for conventional fossil power plants. Measurement practices and methods will be discussed to mitigate that uncertainty. These uncertainty values will be correlated to the levelized cost of electricity (LCOE), and LCOE sensitivities will be derived. The results quantify the impact of resource variability during testing, test duration and sampling rate to annual performance calculation. These uncertainties will be further associated with costs and risks based on typical technology performance guarantees. The paper will also discuss how the development of standard measurements and calculation methods help to produce lower uncertainty associated with the overall plant result, which is already being accomplished by ASME PTCs in conventional power genreation.

In Memoriam: Dr. Clive L. Dym (1942–2016)

Fatigue design of Class-1 piping of NPP is carried out as par Section-III of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code. The fatigue design criteria of ASME are based on the concept of safety factor, which does not provide means for management of uncertainties for consistently reliable designs. In this regard, a work is taken up to estimate the implicit reliability level associated with fatigue design criteria of Class-1 piping specified by ASME Section III, NB-3650. The methodology employed for reliability evaluation is FORM, HORSM and MCS, where limit state function is derived using mean fatigue curve developed by Argonne National Laboratory. A number of important aspects related to reliability of various piping product and joints are discussed, and implicit reliability level is evaluated.

Section III, Division 1 and Section VIII, Division 2 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code provide procedures for demonstrating shakedown using elastic-plastic analysis. While these procedures may be used in place of elastic analysis procedures, they are typically employed after the elastic analysis and simplified elastic-plastic analysis limits have been exceeded. In using the Section III, Division 1 and Section VIII, Division 2 procedures for elastic-plastic shakedown analyses, three concerns are raised. First, the Section III, Division 1 procedure is vague, which can result in inconsistent results between analysts. Second, the acceptance criteria contained in both procedures are vague, which can also result in inconsistent results between analysts. Lastly, differences in the procedures and acceptance criteria can result in demonstration of component elastic-plastic shakedown under Section III, Division 1 but not under Section VIII, Division 2. The authors presume that the ASME Code intends to provide similar design and analysis conclusions, which may not be a correct assumption. To demonstrate these concerns, a nozzle benchmark design subject to a representative thermal and pressure transient was evaluated using the two Code elastic-plastic shakedown procedures. Shakedown was successfully demonstrated using the Section III, Division 1 procedure. However, shakedown could not be demonstrated using the Section VIII, Division 2 procedure. The conflicting results seem to indicate that, for the nozzle design evaluated, the Section VIII, Division 2 procedure is considerably more conservative than the Section III, Division 1 procedure. To further assess the conservative nature of the Section VIII, Division 2 procedure, the nozzle benchmark design was evaluated using the same thermal transient, but without a pressure load. While shakedown was technically not observed using the Section VIII, Division 2 acceptance criteria, engineering judgment concluded that shakedown was demonstrated. Based on the results of all the evaluations, recommendations for modifications to both procedures were presented for consideration.

A blowout preventer (BOP) is a large valve that encases an oil well on the surface. The valve may be closed while drilling if overpressure from a reservoir causes formation fluids such as oil and natural gas to back up within the wellbore and endanger the rig. A pipe ram is one of the critical components in the BOP system, which is designed to seal around a drill pipe, restricting the flow in the annulus between the outside of the drill pipe and the wellbore. The pipe ram design family, which contains hundreds of pipe ram designs, intends to cover different configurations to suit a variety of dill pipe sizes. To assure their structural integrity under service loads, these pipe rams need to be designed to meet the stress requirements per the American Petroleum Institute (API) Specification 16A and ASME (American Society of Mechanical Engineers) Boiler and Pressure Vessel Code Section VIII, Division 2. With a conventional manual design workflow, it would take a significant amount of time and effort to complete all the designs for the product family with code compliance. Therefore, a scalable solution is highly desirable for delivery of customer-needed ram blocks with much shorter lead time. MDO (Multi-Disciplinary Optimization) involves multi-code integration, CAE (Computer-Aided Engineering) workflow automation, design space exploration, and optimization. MDO-enabled, automated design optimization is becoming increasingly popular in both scientific and engineering communities. In this paper, a methodology for integration of CAD (Computer Aided Design), FEA (Finite Element Analysis), cost, and optimization packages to enable FEA automation and design optimization is presented. A BOP pipe ram is adopted as the use case. The ram block geometry was parameterized before being imported into an FEA package. The FEA workflow was automated such that once a set of geometric parameters are given, the preprocessing, solving, and postprocessing steps can be automatically completed. As part of the FEA postprocessing, stress linearization analysis per the API and ASME BPVC codes has also been automated, which had never been done in the past. A manufacturing cost analysis package can also be used to consume the parameterized geometry for automatic manufacturability assessment and cost predictions. The stress analysis and cost analysis workflows are conducted separately but also orchestrated by an MDO package. Reports that contain the analysis results are sequentially generated for various design permutations. The MDO-enabled automated design and cost analysis approach could substantially enhance efficiency and consistency in performing FEA and cost studies and producing analysis reports. It is also the backbone for automated design optimization, which could significantly improve the product performance and reliability and, meanwhile, minimize the development and product cost.

Due to its exceptional high-temperature properties, Alloy 709 (UNS 31025), an advanced austenitic stainless steel, has been selected by the U.S. Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies Program for qualification as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 for high temperature reactor applications. This initiative encompasses a comprehensive staged qualification plan for code cases with progressively longer design lives of 100,000, 300,000, and 500,000 hr, as long-term test data become available. A significant portion of this effort focuses on creating a data package that is essential for establishing material-specific design parameters for Alloy 709, intended for use in sodium fast reactors, molten salt reactors, and gas-cooled reactors. Data generation for the initial 100,000-hr code case was completed at the end of 2024. A key component of this data package involves conducting fatigue and creep-fatigue testing at elevated temperatures to develop fatigue design curves and the damage envelope for the creep-fatigue interaction diagram (D-diagram), both of which are common to all three code cases. Strain-controlled fatigue and creep-fatigue tests have been conducted on three commercial heats of Alloy 709 plates at the Oak Ridge National Laboratory and the Idaho National Laboratory. The comprehensive code case testing matrix includes strain ranges from 0.25 to 3%, hold times of up to 10 hr, and testing temperatures of 21, 427, 649, 760, 816, and 954 °C (70, 800, 1200, 1400, 1500, and 1750 °F). This paper summarizes the results of cyclic testing and provides a comprehensive analysis of the test results. It outlines the methodology for developing temperature-dependent fatigue curves and presents draft fatigue design curves for inclusion in the code cases. Furthermore, creep-fatigue test data with various tension hold times are used to support the development of the creep-fatigue interaction diagram. Both sets of design parameters are crucial for the code case data package.

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case N-830. This revision incorporates a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all the fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J-R) from the lower shelf through the upper shelf. A paper presented at the 2014 ASME Pressure Vessel and Piping Conference described most of these models; a companion paper at this conference describes the J-R model. This paper also supports the WGFE effort by performing an assessment of the appropriateness of Wallin’s Master Curve model to represent toughness data on the lower shelf, and by comparing the Master Curve with the current Code KIc curve on the lower shelf. The work presented in this paper supports the following conclusions: 1. The Master Curve provides a reasonable representation of cleavage fracture toughness (KJc) data at lower shelf temperatures. A statistical evaluation of a large database demonstrates that the Master Curve works well to temperatures approximately 140 °C below To or, equivalently, approximately 160 °C below RTTo. 2. The percentile of cleavage fracture toughness data falling below a KIc curve indexed to RTTo varies considerably with temperature. At lower shelf temperatures as much as half of the data lie below the KIc curve, while at temperatures close to RTTo this percentage falls to approximately ≈ 1.5%. The current guidance of Nonmandatory Appendix A to Section XI to use structural factors of √10 or √2 is one means of addressing this inconsistency. 3. The inconsistent degree to which the KIc curve, with or without structural factors, bounds fracture toughness data cannot be fixed within the current Code framework for two reasons: the KIc curve does not reflect the actual temperature dependence shown by the fracture toughness of ferritic RPV steels, and the ratio of a mean or median toughness curve to a fixed percentile bound is not a constant value. It is for these reasons that in the next revision of Code Case N-830 the ASME WGFE is moving away from use of the KIc curve coupled with structural factors and, instead, is adopting models of fracture toughness that represent both the temperature trends and the scatter in the data with high accuracy.

Methods for the quantification of numerical uncertainty have been a subject of interest to the American Society of Mechanical Engineers (ASME) and the mechanical engineering community as a whole for over a decade. During this time period, ASME has promulgated three statements of standards for the reporting of numerical uncertainty in archival publications (Journal of Fluids Engineering). This paper summarizes the work that has gone into the specification of these standards and the continuing effort in formulation of methods and procedures for quantifying numerical uncertainty. Specifically, this paper discusses the efforts of the ASME V&V 20 Committee (Verification and Validation in Computational Fluid Dynamics and Heat Transfer) to lay a foundation and structure to verification and validation for fluid flow and heat transfer simulations. Issues and methods related to code verification and in particular solution verification are presented and discussed in the context of the recently released V&V20 Standard.

Fossil power plant high-energy piping systems operated at high temperatures are subject to creep damage, which is a time-dependent phenomenon. Traditional guidelines, such as the American Society of Mechanical Engineers (ASME) B31.1 Power Piping Code, were developed for plants having design lives in the 25–30 yr regime. Since many of these systems are being operated beyond 200,000 hr, it is important to reconsider the methodology of creep damage analysis to assure reliable long-term operation. Seven high-energy piping systems were evaluated in this study. The analysis of a minimum piping system life due to creep considered two approaches. The first approach used the traditional ASME B31.1 flexibility analysis guidelines. The second approach considered more detailed multiaxial stress state types of evaluations. The various equivalent stress methods used all six load components from the flexibility analysis. In nearly every case, the equivalent stress methods predicted significantly higher stresses. Consequently, the equivalent stress methodology results in 14 to 97 percent lower time to rupture values as compared to the values predicted using ASME B31.1 stresses.