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Update and Improve Subsection NH –– Alternative Simplified Creep-Fatigue Design Methods

Description: This report described the results of investigation on Task 10 of DOE/ASME Materials NGNP/Generation IV Project based on a contract between ASME Standards Technology, LLC (ASME ST-LLC) and Japan Atomic Energy Agency (JAEA). Task 10 is to Update and Improve Subsection NH -- Alternative Simplified Creep-Fatigue Design Methods. Five newly proposed promising creep-fatigue evaluation methods were investigated. Those are (1) modified ductility exhaustion method, (2) strain range separation method, (3) approach for pressure vessel application, (4) hybrid method of time fraction and ductility exhaustion, and (5) simplified model test approach. The outlines of those methods are presented first, and predictability of experimental results of these methods is demonstrated using the creep-fatigue data collected in previous Tasks 3 and 5. All the methods (except the simplified model test approach which is not ready for application) predicted experimental results fairly accurately. On the other hand, predicted creep-fatigue life in long-term regions showed considerable differences among the methodologies. These differences come from the concepts each method is based on. All the new methods investigated in this report have advantages over the currently employed time fraction rule and offer technical insights that should be thought much of in the improvement of creep-fatigue evaluation procedures. The main points of the modified ductility exhaustion method, the strain range separation method, the approach for pressure vessel application and the hybrid method can be reflected in the improvement of the current time fraction rule. The simplified mode test approach would offer a whole new advantage including robustness and simplicity which are definitely attractive but this approach is yet to be validated for implementation at this point. Therefore, this report recommends the following two steps as a course of improvement of NH based on newly proposed creep-fatigue evaluation methodologies. The first step is to modify the current approach by ...
Date: October 26, 2009
Creator: Asayama, Tai
Partner: UNT Libraries Government Documents Department

Update and Improve Subsection NH - Simplified Elastic and Inelastic Design Analysis Methods

Description: The objective of this subtask is to develop a template for the 'Ideal' high temperature design Code, in which individual topics can be identified and worked on separately in order to provide the detail necessary to comprise a comprehensive Code. Like all ideals, this one may not be attainable as a practical matter. The purpose is to set a goal for what is believed the 'Ideal' design Code should address, recognizing that some elements are not mutually exclusive and that the same objectives can be achieved in different way. Most, if not all existing Codes may therefore be found to be lacking in some respects, but this does not mean necessarily that they are not comprehensive. While this subtask does attempt to list the elements which individually or in combination are considered essential in such a Code, the authors do not presume to recommend how these elements should be implemented or even, that they should all be implemented at all. The scope of this subtask is limited to compiling the list of elements thought to be necessary or at minimum, useful in such an 'Ideal' Code; suggestions are provided as to their relationship to one another. Except for brief descriptions, where these are needed for clarification, neither this repot, nor Task 9 as a whole, attempts to address details of the contents of all these elements. Some, namely primary load limits (elastic, limit load, reference stress), and ratcheting (elastic, e-p, reference stress) are dealt with specifically in other subtasks of Task 9. All others are merely listed; the expectation is that they will either be the focus of attention of other active DOE-ASME GenIV Materials Tasks, e.g. creep-fatigue, or to be considered in future DOE-ASME GenIV Materials Tasks. Since the focus of this Task is specifically approximate methods, the authors ...
Date: June 27, 2009
Creator: Abou-Hanna, Jeries J.; Marriott, Douglas L. & McGreevy, Timothy E.
Partner: UNT Libraries Government Documents Department

New Materials for NGNP/Gen IV

Description: The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.
Date: December 18, 2009
Creator: Swindeman, Robert W. & Marriott, Douglas L.
Partner: UNT Libraries Government Documents Department

Review of Current Experience on Intermediate Heat Exchanger (IHX) and A Recommended Code Approach

Description: The purpose of the ASME/DOE Gen IV Task 7 Part I is to review the current experience on various high temperature reactor intermediate heat exchanger (IHX) concepts. There are several different IHX concepts that could be envisioned for HTR/VHTR applications in a range of temperature from 850C to 950C. The concepts that will be primarily discussed herein are: (1) Tubular Helical Coil Heat Exchanger (THCHE); (2) Plate-Stamped Heat Exchanger (PSHE); (3) Plate-Fin Heat Exchanger (PFHE); and (4) Plate-Machined Heat Exchanger (PMHE). The primary coolant of the NGNP is potentially subject to radioactive contamination by the core as well as contamination from the secondary loop fluid. To isolate the radioactivity to minimize radiation doses to personnel, and protect the primary circuit from contamination, intermediate heat exchangers (IHXs) have been proposed as a means for separating the primary circuit of the NGNP (Next Generation Nuclear Plant) or other process heat application from the remainder of the plant. This task will first review the different concepts of IHX that could be envisioned for HTR/VHTR applications in a range of temperature from 850 to 950 C. This will cover shell-and-tube and compact designs (including the platefin concept). The review will then discuss the maturity of the concepts in terms of design, fabricability and component testing (or feedback from experience when applicable). Particular attention will be paid to the feasibility of developing the IHX concepts for the NGNP with operation expected in 2018-2021. This report will also discuss material candidates for IHX applications and will discuss specific issues that will have to be addressed in the context of the HTR design (thermal aging, corrosion, creep, creep-fatigue, etc). Particular attention will be paid to specific issues associated with operation at the upper end of the creep regime.
Date: February 2, 2010
Creator: Spencer, Duane & McCoy, Kevin
Partner: UNT Libraries Government Documents Department

A Review & Assessment of Current Operating Conditions Allowable Stresses in ASME Section III Subsection NH

Description: The current operating condition allowable stresses provided in ASME Section III, Subsection NH were reviewed for consistency with the criteria used to establish the stress allowables and with the allowable stresses provided in ASME Section II, Part D. It was found that the S{sub o} values in ASME III-NH were consistent with the S values in ASME IID for the five materials of interest. However, it was found that 0.80 S{sub r} was less than S{sub o} for some temperatures for four of the materials. Only values for alloy 800H appeared to be consistent with the criteria on which S{sub o} values are established. With the intent of undertaking a more detailed evaluation of issues related to the allowable stresses in ASME III-NH, the availabilities of databases for the five materials were reviewed and augmented databases were assembled.
Date: December 14, 2009
Creator: Swindeman, R. W.
Partner: UNT Libraries Government Documents Department

Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

Description: The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.
Date: May 7, 2007
Creator: O’Donnell, William J. & Griffin, Donald S.
Partner: UNT Libraries Government Documents Department

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

Description: The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules for possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life ...
Date: January 27, 2010
Creator: Brust, Dr. F. W.; Wilkowski, Dr. G. M.; Krishnaswamy, Dr. P. & Wichman, Mr. Keith
Partner: UNT Libraries Government Documents Department