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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Structural Design and Modeling of MARVEL Primary Coolant System Using the ASME Section III, Division 5, Code

This paper presents the structural design and supporting analysis for the Microreactor Applications Research Validation and Evaluation (MARVEL) primary coolant system (PCS) using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Division 5, rules. MARVEL is a liquid metal–cooled microreactor intended to provide experimental capabilities for the rapid testing and development of microreactor technologies. The PCS utilizes high-temperature sodium-potassium liquid metal as the primary coolant and operates at a design temperature of 570°C, necessitating the consideration of creep-related failure mechanisms. The base metal for the PCS is 316H stainless steel, and the weldments are made with a 16-8-2 filler. The design approach incorporates the current base code rules along with ASME code cases N-924, N-861, and N-862 to address primary load, ratcheting, and creep-fatigue evaluations, respectively. The reactor’s operation involves complex thermal and mechanical interactions due to natural convective flow and differential thermal expansion between components. In conclusion, this paper discusses the structural engineering challenges encountered, such as managing thermal stresses in the distribution plenum and guard vessel, and outlines the strategies implemented to meet the code requirements, including design modifications and operational constraints.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

Consistency of fatigue crack growth behavior of pipeline and low-alloy pressure vessel steels in gaseous hydrogen

Here, this study investigates the fatigue crack growth rate (FCGR) behavior of pipeline and low-alloy pressure vessel steels in high-pressure gaseous hydrogen. Despite a broad range of yield strengths and microstructures ranging from ferrite/pearlite, acicular ferrite, bainite, and martensite, the FCGR in gaseous hydrogen remained consistent (falling within a factor of 2–3). Steels with higher fractions of pearlite, typical of older vintage pipeline steels, exhibited modestly lower crack growth rates in gaseous hydrogen compared to steels with lower fractions of pearlite. Crack growth rates in these materials exhibit a systematic dependence on stress ratio and partial pressure of hydrogen, as captured in the recently published fatigue design curves in ASME B31 code case 220 for pipeline steels and ASME BPVC code case 2938 for pressure-vessel steels.

08 HYDROGEN↗
DOE OSTI2022

MECHANICAL PERFORMANCE EVALUATION OF THE PRINTED CIRCUIT HEAT EXCHANGER CORE EXPERIMENTS UNDER TENSION AND PRESSURE LOADING

The printed circuit heat exchanger (PCHE) has small channels with high surface area, making them an efficient solution for next-generation nuclear plants (NGNPs). These PCHEs are fabricated through a diffusion bonding process. This fabrication step changes the microstructure of wrought metal plates. The current ASME design code does not support the PCHE design for NGNPs due to a lack of test data. Hence, there has been initiative towards elevated temperature mechanical property characterization of the diffusion bonded material. One of the most common channel shapes is a semicircular channel with sharp corners. These corners act as a stress riser at the diffusion bonding interface. Evaluating elevated temperature mechanical performance of diffusion bonded material in the presence of stress risers is an essential step towards the ASME code development of PCHE design. This study selected two specimen geometries: the first is a PCHE bar specimen for tensile loading with three rows and three columns of channels, and the second is a lab-scaled PCHE with six rows and eight columns of channels. A set of elevated temperature monotonic and cyclic tests were conducted on the PCHE bar specimen to evaluate the mechanical performance under axial tensile loadings to study the failure mechanism. The lab-scaled PCHE specimens were tested under overpressure loads at room temperature, and pressure creep and pressure creep-fatigue loadings to mimic the realistic loading conditions observed in typical NGNPs. The X-ray scans of channeled specimens show interesting observations. The test results and observations are presented in the paper.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Draft ASME Boiler and Pressure Vessel Code Cases and Technical Bases for Use of Alloy 617 for Constructions of Nuclear Component Under Section III, Division 5

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Code Standard Evaluation for ASME AG-1-2019, Section FL

ASME AG-1 Section FL Deep Bed Sand Filters includes a requirement to use Special Test Sections to provide a representative assessment of the filter efficiency. The SRPPF project design basis is to test the entire filter bed (versus a representative sample) following the same basis methodology to test the existing filters at the Savannah River Site. Using a combination of: 1) the filter testing experience at SRS since the early 1970’s, 2) the methodology found in ASME AG-1 for testing high capacity HEPA filter housings and 3) requirements for stack or duct sampling systems published by the Environmental Protection Agency (EPA), the Special Test Sections methodology for Deep Bed Sand (DBS) filters in ASME AG-1 Section FL has been replaced with a methodology to test the entire DBS filter system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Application of the ASME Boiler and Pressure Vessel Code in the Design of SRF Cavities at Fermilab

Jacketed Superconducting Radio Frequency (SRF) cavities structurally comprise of an inner niobium vessel surrounded by a liquid helium containment vessels. The pressure of the helium bath and/or its volume might be such that a jacketed SRF cavity shall be considered a system of pressure vessels. Thus, methods described in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) should be used to analyze the structural soundness of jacketed SRF cavities. This paper will report the use of the set of rules developed at Fermilab for the design of SRF cavities, such as jacketed 1.3 GHz cavities for LCLS-II HE and jacketed Single Spoke Resonator type~2 (SSR2) for PIP-II, to ensure a similar level of safety as prescribed by the ASME BPVC.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Application Of The ASME Boiler And Pressure Vessel Code In The Design Of SSR Cryomodule Beamlines For PIP-II Project At Fermilab

This contribution reports the design of the main components used to interconnect SRF cavities and superconducting focusing lenses in the SSR Cryomodule beamlines, developed in the framework of the PIP-II project at Fermilab. The focus of the present contribution is on the design and testing of the edge-welded bellows according to ASME Boiler and Pressure Vessel Code. The activities performed to qualify the bellows to be assembled in cleanroom, for operation in high vacuum, cryogenic environments, and their characterization from magnetic standpoint, will also be presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Technical Qualification of New Materials for High Efficiency Coal-Fired Boilers and Other Advanced FE Concepts: Haynes® 282® ASME Boiler and Pressure Vessel Code Case

This DOE Fossil Energy project, FEAA117, addressed materials issues relevant to qualifying and deploying a Ni-base alloy for a new application in an advanced ultra-supercritical coal-fired boiler. The goal was the deployment of Haynes International alloy Haynes® 282® for applications in superheaters, reheaters, and steam delivery pipes, by completing base metal, cross-weld and all-weld metal mechanical testing needed for an ASME Boiler and Pressure Vessel Code Case and the associated microstructural analyses needed for assurance of boiler-relevant lifetimes. The alloy also is of interest for other applications including supercritical CO₂ power generation cycles. Tensile testing was completed from 20°-927°C on the three base metal alloy 282 heats, two sets of cross-weld specimens and all weld metal specimens. A total of 148 creep tests with over 565,000 h of cumulative testing was completed on base metal, cross-weld and all weld metal specimens. This project was performed with cost share from Haynes International.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2022

Simplified Formulas for External Pressure Design

Design of vessels for external pressure currently requires a chart-based solution or analytical approaches which are not necessarily intuitive. In this paper, we propose simple formulas for the external pressure evaluation of pipes and other cylindrical pressure vessels. We present a conceptual comparison between the elastic and elastic-plastic stability of structural columns and that of cylindrical vessels of long, intermediate, and short length. Their common features allow an accurate and straightforward approach for external pressure design. The approach is also extended to spherical caps, conical vessels, and formed heads.We compare the method presented to the current acceptance criteria from various design codes, including the ASME Boiler and Pressure Vessel Code Section VIII, Code Case 2286, and EN 13445-3, as well as codes for steel and aluminum structures. In further discussion, the simplified method is compared against the results of more than 500 experiments on the buckling of cylindrical and spherical vessels published over the past two centuries.This simple but accurate approximation is conceptually intuitive, analytically straightforward, and shows potential utility in pressure vessel design codes, as well as piping design codes such as B31 that currently reference ASME VIII for external pressure design.

Kirkland, Will↗
DOE OSTI2023

A summary of the mechanical properties data developed in FY 2023 by ANL, INL and ORNL to support the data package development for the A709 Code Case

This report provides the status of tensile, creep, fatigue, and creep-fatigue testing to date conducted at Argonne National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to generate the data package required to qualify Alloy 709 in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. The Division 5 Class A Alloy 709 Code Case requires data generated from a minimum of three commercial heats. Extensive mechanical properties data have been generated on two commercial heats, and some initial test data have been generated from the third commercial heat. The room temperature tensile test results for the three commercial heats met the specification minimum of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section II, Part A, SA-213/SA-213M for Grade TP310MoCbN (UNS S31025) seamless tubing. These three commercial heats of Alloy 709 are suitable for generating the code case data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A Structural Design Approach Tailored for the Rapid Preliminary Design of Microreactor Components

High-temperature microreactors can play a role in developing reliable, portable energy sources for off-grid remote locations, microgrid concepts, and industrial process heat. Portability and passive safety criteria tend to skew microreactor structural component designs toward complex geometries, high thermal stresses, and design bases with large numbers of startup/shutdown cycles. Current design rules, as typified by Section III of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code, are less than optimal for these conditions, particularly for preliminary component designs where developers need to rapidly consider a large number of potential component configurations. This paper presents a design method targeted toward rapid, efficient evaluation of preliminary component designs using modern finite element analysis. The new method retains key connections with the ASME Code rules and design data while streamlining the design approach. This paper presents the design method, several verification examples illustrating the similarities and differences between the new method and the current ASME rules, and the application of the new approach to the evaluation of a test article mimicking key features of a heat pipe–cooled microreactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Equivalent Safety Evaluation of the 2021 Edition of the ASME Boiler and Pressure Vessel Code to Editions Incorporated by Reference in 49 CFR 192, 193, and 195

Federal safety standards for natural gas and hazardous liquid pipelines and liquid natural gas facilities incorporate rules and requirements for boiler and pressure vessel design and fabrication into 49 CFR Parts 192, 193, and 195 through the IBR process. The equivalent safety evaluations documented in this report demonstrate that boilers and pressure vessels in pipeline facilities that are designed and fabricated in accordance with rules and requirements specified in the 2021 edition do not violate the fundamental safety assumptions stated or implied in the 2007 edition of the ASME BPVC and therefore provide equivalent safety.

03 NATURAL GAS↗
DOE OSTI2024

Initial Alloy 709 constitutive models for use with the ASME design by inelastic analysis and EPP+SMT design methods

This report details a preliminary inelastic constitutive model describing the behavior of Alloy 709. This model will serve two purposes: (1) integration into Nonmandatory Appendix HBB-Z of the ASME Boiler & Pressure Vessel Code Section III, Division 5 and (2) extrapolating cyclic test data to difficult to measure conditions for formulating improved creep-fatigue design methods. For both applications, the model must accurately capture the material behavior across a wide range of temperatures and a variety of test conditions, both monotonic and cyclic. For this purpose we adopt a universal model form under consideration to standardize the description of high temperature constitutive models in the ASME Code. This report briefly restates that model form and how we calibrate the model against the test data, summarizes the test database, and validates the final, trained model by comparison to the experimental tests.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Simplified inelastic constitutive models for ASME Section III, Division 5 design by inelastic analysis

This report describes the development of simplified, universal constitutive model that captures the high temperature monotonic and cyclic behavior of a range of commonly-used high temperature materials. The goal of the work is to provide a simple, universal constitutive model to replace the current bespoke models for Grade 91, 316H, and Alloy 617 included in Nonmandatory Appendix HBB-Z of the ASME Boiler & Pressure Vessel Code, and to extend this model to cover Alloy 800H. We initiated this work in response to feedback from reactor vendors and other Code users requesting simplified models, compared to the current models, that are easier to implement and use in commercial finite element analysis software. This report describes the completion of this effort by developing a model to correct the defects in standard model forms presently used for high temperature material modeling, described in past work, developing and implementing new numerical methods to train this model against test data, and then actually training the model for the four materials. The report provides a complete mathematical description of the model along with the tabulated material coefficients for the four materials. The final step will be to formulate an ASME Code change to introduce the new models into the Code.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Draft Rules for Alloy 617 Creep-Fatigue Design Using an EPP+SMT Approach

The report describes an improved new creep-fatigue design method for structural components in high temperature nuclear service. The new method uses an integrated elastic-perfectly plastic (EPP) analysis and Simplified Model Test (SMT) approach that greatly simplifies the design process by avoiding the separate evaluation of creep and fatigue damages and eliminating the requirement of stress classification. The report includes draft design rules, presented in a format compatible with an ASME nuclear Code Case, as well as a commentary on the rules and the validation of the rules using pressurized SMT (p-SMT) test data. Areas of potential improvement have been identified for further evaluation. The report also develops EPP+SMT design charts for Alloy 617 at temperatures between 800°C and 950°C by extrapolating high strain range, short hold test data to low strain range and long hold times, typical for structural components in high temperature nuclear service. The design charts are currently limited to 10,000 cycles to avoid excessive extrapolation outside the test database. More robust alternate extrapolation procedures will be considered in future work. The report includes several verification problems comparing the new EPP+SMT design method with current ASME creep-fatigue design methods. The comparison demonstrates that the EPP+SMT method significantly reduces the over-conservatism in current methods. Finally, the report includes a fully-documented sample problem with multiple load cases detailing the application of proposed EPP+SMT design rules.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Initial Development of Variable Design Lifetimes and Creep-Fatigue Evaluations for ASME Section III, Division 5, Class B Code Rules

This report summarizes the initial development of new ASME Section III, Division 5, Class B rules to address the gaps identified for new high temperature reactor designs. The overall objective is to introduce the design-by-analysis approach, the variable design lifetimes for primary load design, and the strain limits and creep-fatigue damage evaluations for Class B components. It is also desired that the new evaluation procedures do not require stress classification and stress linearization. Further, the data requirements to support the development of design parameters for the new Class B rules would not be overly burdensome, and less extensive than those supporting Class A rules. We have identified the Elastic, Perfectly Plastic (EPP) method as the design by-analysis approach for primary load design, and developed allowable stress criteria for variable design lifetimes, up to 500,000 hours, that only require data typical for the allowable stresses in Section II, Part D, Tables 1A and 1B. We have also adapted the elastic analysis approach from HBB-T-1400 and Section III, Mandatory Appendix XIII, XIII-2400 and XIII-3520(e) for the new Class B creep-fatigue evaluation procedure. A new elastic follow-up-based Isochronous Stress Strain Curve (ISSC) stress relaxation procedure has been developed to provide adequate conservatism commensurate with Class B constructions. No stress classification and stress linearization are required in the new procedure. Future work needed to complete the development is also summarized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗