A case study was conducted on a mechanistic model development that predicted tensile strength deterioration with thermal aging of 9Cr-1Mo-V steel in supporting the 60-year design life expected for advanced nuclear reactors. For property prediction beyond practical testing times, mechanistic modeling is highly desired, as it taps into the physics of structure-property relationships and therefore can generate reliable results for extrapolation. Meanwhile, as mechanistic models are often complicated, reflecting the intricacy of microstructure and strengthening mechanisms, pitfalls that are difficult to detect often exist. This paper discusses latent pitfalls that are common in mechanistic modeling or specific in this 9Cr-1Mo-V case development through using the American Society of Mechanical Engineers verification and validation in computational solid mechanics (ASME V&V 10) standard for evaluating credibility of modeling in materials engineering. Suggestions are also made for enhancing reliability of microstructure-based modeling.

Development of the ASME Materials Properties Database was initiated in the early 2010s to support the ASME Codes and Standards. As information technologies advance at an accelerated pace with the artificial intelligence era on the horizon, the ASME Materials Properties Database must be further modernized from a database to a knowledgebase to ride the wave of digital information revolution and effectively support the ASME Codes and Standards in the new era. This paper is intended to provide an overview of the ASME Materials Properties Database and discuss a roadmap for its future development to facilitate understanding of and participation from different sectors of the Codes and Standards community. It first reviews the basic concepts of data, information, knowledge, database, and database system; as well as the pros and cons in different types of data management, and then discusses the path forward for a desired evolution of the database into a self-explanatory and machine-readable knowledgebase that is consistent with human cognitive processes for the Codes and Standards development and furthermore provides resources for data processing and analysis to reach an eventual goal of streamlining the Codes and Standards development from the initial inquiry, throughout data submission, analysis, ..., to Codes and Standards rule establishment for final publication.

The long-term exposure of 9Cr-1Mo-V (Grade 91) steel to elevated temperature can have a significant effect on reducing its yield and tensile strength. The yield and tensile strength changes, in turn, have potential implications to the assurance of the integrity of components constructed with this material over their design or intended lifetime. The ASME Boiler & Pressure Vessel Code (BPVC), Section III, Division 5 (III-5, high temperature reactors) provides tabulated reduction factors for Grade 91 yield and tensile strength as a function of exposure temperature and time up to 300,000 h. ASME BPVC III-5 s intent to extend these factors to an exposure duration of 500,000 h, the lack of available historic information to support the existing factors, and the recent development of a physics-based prediction model for ASME BPVC application are prime motivation for this study. This paper describes results of the conventional time-temperature Hollomon-Jaffe parameter, strength reduction ratio prediction method using an updated, extensive Grade 91 unaged and related aged material strength database. The method, previously used by Oak Ridge National Laboratory in its evaluation of Grade 91 and likely used in development of the existing BPVC III-5 reduction factors, provides strength reduction ratio predictions useful for general component fitness-for-service assessments and for computing BPVC III-5 reduction factors as defined. Specific reduction factors to 500,000h at 650 degrees C applicable to ASME BPVC III-5 are computed from the strength reduction ratios.

As the Boiler and Pressure Vessel Code continues to gain popularity worldwide, it is not uncommon that many users are mystified by the 1.1 factor used to establish the ultimate tensile strength above room temperature in developing the related design stresses of Section II Part D Stress Tables. Questions often arise about the origin and purpose of the factor, the reason for it not being applied to the yield strength, its due considerations regarding the tested tensile strength value of an alloy batch when accepting the batch for construction use, and more. Not having historic background or reliable sources for explicit explanations, some users tend to misinterpret the factor, abuse its application, incorrectly define alloy acceptance or qualification criteria, and mistakenly manipulate the safety margin in component design. To help the Code users who struggle with these frequent and confusing issues, particularly those in the nuclear industry where rigorous criteria are required for component design and alloy acceptance or qualification, this paper is intended to demystify the 1.1 factor and facilitate knowledgeable interpretation and use of the Section II Part D Stress Tables as well as relevant Mandatory Appendices. A brief review is first given on the background of the 1.1 factor, followed by a summary of its application in the Stress Tables. The provenance and purpose of the factor are then discussed in detail with graphic examples. Finally, current applicability and necessity of the factor are considered through a demonstration using example alloys.

In consideration of structural alloy property deterioration during long-term exposure to elevated temperatures, the yield and ultimate tensile strength reduction factors are provided in the Boiler and Pressure Vessel Code Section III for nuclear reactor component design and operation analysis. Because the Gen IV reactor requirement of 40 similar to 60 years of service life makes it difficult to acquire such long exposure test data for developing the reduction factors, they must be derived from test data with relatively short exposure by predictive methods considered to be reasonably reliable. A novel approach with a physically-based model has recently been proposed for application to development of reduction factors for 9Cr-1Mo/16-V In the model, contributors to the tensile strength are first identified and related to definite microstructural features of the alloy, then some physically-based methods are employed to simulate the microstructural evolution, and finally the model is assembled with test-data-calibrated parameters to generate the yield and ultimate tensile strength reduction factors covering elevated temperature exposure for up to 57 years. The approach is undoubtedly a trailblazing development that will, if proven reliable, lead to a paradigm shift in predicting thermal aging behavior of many other alloys. Its debut application to Section III, however, concerns nuclear safety and naturally warrants objective, impartial, and thorough technical scrutiny. In the present paper, the novel and conventional approaches are discussed. Necessary improvements to the novel approach are recommended for its application to nuclear structural component design and analysis, and for its potential expanded use to other alloys.

To assist the Nuclear Regulatory Commission in its decision making on endorsement of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code Section III, Division 5 (2017 Edition) for development of advanced non-light water reactors, the following Division 5 portions were reviewed: Article HBB-2000 Material; Article HCB-2000 Material; Article HGB-2000 Material; Mandatory Appendix HBB-I-14 Tables and Figures; and, Nonmandatory Appendix HBB-U Guidelines for Restricted Material Specifications to Improve Performance in Certain Service Applications. In addition to the 2017 Edition, the same parts of the 2019 Edition have also been reviewed as indicated in various sections of the report. This review was conducted by a collaboration of national laboratory and private sector participants with significant industrial experience, including some heavy lifting and deep diving from Clarus Consulting, LLC., all intended to achieve an objective, independent, and practical perspective. The report provides recommendations, descriptions of the evaluation methods, and the source references for the data used. To build confidence required for endorsement of the Code, this review was conducted as a verification and validation of the above Code contents. The objective of verification is to ensure that the Code is free of error – direct or implied; contains the information needed for its use, including proper coverage of the Code-specified materials for the intended application, and completeness and adequacy of references to other portions of the Code. The objective of validation is to authenticate that the Code tabulations and graphs represent design inputs consistent with what are determined using rules and methods specified by the Code. The authentication process used data that were assembled and/or generated independent of Code development, while the methods of analysis followed Code-specified methods where appropriate. The designated portions for this review cover the five alloys codified for high temperature reactor applications in Division 5, i.e. 316 SS, 304 SS, 800H, 2¼Cr-1Mo, and 9Cr-1Mo-V, regarding their general requirements, permitted specifications and design stress intensity values for pressure-retaining applications, deterioration in service, fatigue acceptance test, permissible weld materials, tensile and yield strength, expected minimum stress-to-rupture values (including for Alloy 718), weld stress rupture factors, permissible materials for bolting use, and restricted specifications in certain service applications. Additionally, stress intensity values for bolting materials including 316 SS, 304 SS and alloy 718 were reviewed. Analysis and discussion are also provided on contents outside of these designated Code portions where it was deemed relevant and necessary to develop a technically sound understanding of issues relating to the designated portions. Due to unavailability of sufficient test data on welds during the review period, the weld stress rupture factors in Tables HBB-I-10.14A to E, which cover a total of ten tables for the five alloys welded with twenty-eight different weld metals (some with similar properties), have been deferred to a future review effort. The review identified mainly two types of issues. The first type includes instances where the Code is found factually incomplete or incorrect, such as obsolete materials specifications listings, missing tabulation of stresses for bolting. Changes to the Code are recommended in these cases. The second type of issue includes instances where the Code tabulations and graphs are found to be less conservative than the review analysis results. In these cases, recommendations are made for further review and consideration where the difference in conservatism exceeds 10%, which is our threshold for questioning technical adequacy, meriting a risk assessment by the Nuclear Regulatory Commission and/or reactor designers. It is noted that this effort has been executed using all available data and established methods of analysis, including methods and criteria specified and used by the Code. As such, the findings that are presented in quantitative detail, in a format for convenient comparison with the Code, and with identification of where further review is recommended, should provide a sound technical basis for decisions about quantifying the implications of the reduced design margins and technical adequacy/inadequacy to form a basis for conditioning specific Code tabulation values on endorsement. Recommendations for specific changes to the Code, however, entail design conservatism considerations beyond the scope of this review effort, and are not made in this report.

Molten halide salts are being considered as working fluids for nuclear and concentrated solar power applications. High temperature molten fluoride and chloride salts are known to preferentially attack and deplete Cr in alloys, which leads to the use of high-Ni low-Cr alloys in test facilities for advanced molten salt technology. Alloy C-276 is a commercially available Ni alloy that has adequate Cr contents and is qualified to the maximum temperature of 677 degrees C (1,251 degrees F) in the Boiler and Pressure Vessel Code. The alloy has good corrosion resistance to acids, is resistant to stress-corrosion cracking, and has long track records of use in the chemical industry. Therefore, it has been considered as a structural material for test facilities that require operations at 700 degrees C (1,292 degrees F) or greater to develop high-temperature molten salt technology. To meet the requirements, predictions of the Maximum Allowable Stress above the usage temperatures permitted by the Boiler and Pressure Vessel Code were developed with experimental data as an extension to the current code design values. Analysis showed that above current Codified maximum temperature, strength of the alloy is mainly controlled by creep rupture life under the average stress, although the Sc creep rate criterion is close to the F-avg.S-avg rupture criterion. This paper presents the intended test facilities and the design requirements, alloy selection considerations, literature review, data analysis, and proposed allowable stress extension based on some creep test data for C-276 at temperatures greater than 677 degrees C (1,251 degrees F). Further research activities are also briefly mentioned.

The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. This paper gives a summary of important results from the PIRT panel discussions and report.

Abstract To construct advanced non-light water reactors (ANLWRs) operating in the temperature range above that for the traditional light water reactors (LWRs), Alloy 316H is one of the candidate materials because of its inexpensiveness, significant service experience, and qualification for nuclear applications by the American Society of Mechanical Engineers (ASME). However, during the life span at temperatures expected for the ANLWRs, the alloy is likely to experience thermal embrittlement that has not been a concern for the traditional LWRs. To prepare for the development, the possibility of adverse thermal embrittlement effects on Alloy 316H performance in the ANLWRs must be evaluated and a technical basis regarding thermal embrittlement, if necessary, must be established for structural integrity analysis to provide reasonable assurance of adequate nuclear safety protection. In this paper, current technical basis for nuclear applications of Alloy 316H deterioration from thermal aging is briefly introduced. The likelihood of adverse thermal embrittlement effects on Alloy 316H performance is evaluated through historical data on microstructural and mechanical property evolution. Characterization of thermal embrittlement is then discussed, followed by a review of predictive models and trend curves for alloy embrittlement. Based on the review and evaluation, technical gaps for addressing thermal embrittlement issues are identified and gap-filling actions are recommended for establishing a technical basis to enable adequate consideration of thermal embrittlement in Alloy 316H applications to the ANLWRs.

For commercial development and deployment of the molten salt reactor, a structural alloy that provides both strength at high temperature and resistance to very corrosive molten salt environment is required. To meet this requirement, a survey is conducted on domestic and international candidate alloys. Alloy N turns out to be the sole frontrunner in readiness for qualification to enable the desired deployment within an estimated 10 years. A review of the qualification for commercial nuclear applications indicates that Alloy N has met a large portion of the requirements. Gaps in the qualification are also identified. A search for historical data is underway to retrieve information needed for filling the gaps and upgrading the qualification. Scope of the discovered historical data is briefly discussed and strategic planning for research and development pathway is suggested to ensure successful evolution in commercial deployment of the molten salt reactor system.

To construct advanced non-light water reactors (ANLWRs) operating in the temperature range above that for the traditional light water reactors (LWRs), Alloy 3 I 6H is one of the candidate materials because of its inexpensiveness, significant service experience, and qualification for nuclear applications by the American Society of Mechanical Engineers (ASME). However, during the life span at temperatures expected for the ANLWRs, the alloy is likely to experience thermal embrittlement that has not been a concern for the traditional LWRs. To prepare for the development, the possibility of adverse thermal embrittlement effects on Alloy 316H performance in the ANLWRs must be evaluated and a technical basis regarding thermal embrittlement, if necessary, must be established for structural integrity analysis to provide reasonable assurance ofadequate nuclear safety protection. In this paper, current technical basis for nuclear applications ofAlloy 316H deterioration from thermal aging is briefly introduced. The likelihood of adverse thermal embrittlement effects on Alloy 316H performance is evaluated through historical data on microstructural and mechanical property evolution. Characterization of thermal embrittlement is then discussed, followed by a review ofpredictive models and trend curves for alloy embrittlement. Based on the review and evaluation, technical gaps for addressing thermal embrittlement issues are identified and gap -filling actions are recommended for establishing a technical basis to enable adequate consideration of thermal embrittlement in Alloy 316H applications to the ANLWRs.

The Integrated Computational Materials Engineering (ICME) approach provides a new paradigm for improving the performance of existing materials or discovering and developing new materials. It focuses on developing effective connections between isolated engineering fields, to bring quantitative processing-structure-property relationships and abundant validated data that populate the knowledge base for accelerating the research of new materials while reducing the cost of development. The data exchanging interfaces play a key role in building such ICME connections among different materials models, simulation tools and individual organizations. With implementations of information exchanging interface between different materials database, and interface between database and applications, this article concludes that in building effective ICME applications, 1) standards-compliant interface can improve the exchange efficiency; 2) popular web service enables automated online materials data transfer; 3) customized data interoperation scripts can provide flexibility and productivity.

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.

Materials database interoperability has been of great interest in recent years for information exchange in support of research and development (R&D). In response to data and knowledge sharing needs of the Gen IV International Forum (GIF) for global collaboration in nuclear energy R&D, the European Commission JRC Institute for Energy and Transport (JRC-IET) and the Oak Ridge National Laboratory (ORNL) established a materials database interoperability project that developed techniques for automated materials data exchange between systems hosted at the two institutes-MatDB Online at JRC-IET and the Gen IV Materials Handbook at ORNL, respectively. The work to enable automated exchange of data between the two systems leveraged the XML data import and export functionalities of both systems in combination with recently developed standards for engineering materials data. The preliminary results of data communication between the two systems demonstrated the feasibility and efficiency of materials database interoperability, which constructs an interoperation framework that can be seamlessly integrated into the high-throughput First Principles material databases and thus advance the discovery of novel materials in fuel cell applications.

The database for Alloy 800H was reviewed with respect to the initiation of tertiary creep and subsequent rupture. Data covered temperatures in the range of 600 to 1000°C and times to 100,000 hours. The establishment of the time to initiate tertiary creep presented a problem because of the non-classical behavior of creep often exhibited in the creep curves. Stress values based on the current tertiary creep and rupture criteria were compared for several heats of material considering both classical and non-classical creep data. Fits to tertiary creep showed more scatter and it was found that 80% of the minimum stress for tertiary creep was generally lower than 67% of the minimum stress to rupture. Using only data from classical curves did produce stress values higher than those from the rupture criterion.

To support the ASME Boiler and Pressure Vessel Codes and Standard (BPVC) in modern information era, development of a web-based materials property database is initiated under the supervision of ASME Committee on Materials. To achieve efficiency, the project heavily draws upon experience from development of the Gen IV Materials Handbook and the Nuclear System Materials Handbook. The effort is divided into two phases. Phase I is planned to deliver a materials data file warehouse that offers a depository for various files containing raw data and background information, and Phase II will provide a relational digital database that provides advanced features facilitating digital data processing and management. Population of the database will start with materials property data for nuclear applications and expand to data covering the entire ASME Code and Standards including the piping codes as the database structure is continuously optimized. The ultimate goal of the effort is to establish a sound cyber infrastructure that support ASME Codes and Standards development and maintenance.