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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Comprehensive Margin Assessment of the ASME Section III, Division 5, Class A Primary Load Design Rules

This report provides a comprehensive margin assessment of the ASME Section III, Division 5, Subsection HB, Subpart B rules for the design of high temperature reactor components against load controlled stress limits. These rules, often called the Code primary load design rules, provide protection against creep rupture and plastic collapse under steady conditions. The method adopted here is to compare the deterministic ASME design life, according to the primary load design rules, against the expected, statistical service life of the component, considering creep rupture as the relevant failure mode. The report provides the design margin of the ASME rules in terms of the probability of premature failure -- the probability that the component fails in service before reaching the ASME design life. To complete this assessment, the report describes the development of a statistical creep life assessment procedure accounting for variations in the component loading, material creep rate, and the material rupture time. This includes a novel method for correlating creep rupture data using a Gaussian process accounting for heat-to-heat variation in the Larson-Miller time-temperature parameter. The report applies the complete margin assessment process to quantify the design margin inherent in the ASME primary load design rules as a function of temperature for relevant component geometries and materials. The overall conclusion is that the ASME rules are conservative, providing designs with very small probability of premature failure, but some optimization of the Code design rules is possible to provide a more uniform design margin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Complete the Planned FY24 Design Parameters Development at INL to Support the Drafting of the A709 Near-Term Code Case

This report presents the thermophysical properties measurements conducted in the Fiscal Year 2024 to extend the database up to 1000?. The new values are compared against properties of high Molybdenum steels in the ASME Section II Part D tables. This report also shows the comparison between new measurements against the previously measured values in literature and provides recommended thermophysical properties for A709 code case. These values will support the thermophysical property tables in near term code case of A709.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Performance Test Code for Determining and Reporting Annual Heat Rate and CO2 for Coal Fueled EGUs

This material is based upon work supported by the Department of Energy under Award Number DEFE0031933. The US Department of Energy (DOE) has retained the services of ASME Standards Technology LLC (ASME ST-LLC) to develop a standardized test method and calculation protocol for determining and reporting annual/long-term heat rate for coal-fueled plants electricity generating units. ASME ST-LLC has retained the services of McHale & Associates, Inc. and Licata Energy & Environmental Consultants, Inc. to provide the technical input to this study. This code applies to hydrocarbon fueled Rankine cycle power generating facilities of any size. It is intended to be used for the determination of continuous power output and heat rate of a generating facility in its operating load cycle and in its operational lineup during the reporting period.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2020

Evaluation of ASCE 4-16 and AISC 43-18 (Draft) for use in the Risk-Informed Performance-Based Seismic Design of Nuclear Power Plant Structures, Systems, and Components

This report describes the assessment of ASCE Standards 4-16 and 43-18 (Draft) for use in the Risk-Informed Performance-Based (RIPB) seismic design of structures, systems, and components (SSCs) at nuclear power plants. This work was performed for the U.S. Nuclear Regulatory Commission (NRC) Office of Regulatory Research (RES), to support potential endorsement of these industry standards for the design of nuclear power plants based on the RIPB approach. Currently, the NRC endorses a deterministic design approach for demonstrating the design adequacy of SSCs based on the Standard Review Plan (NUREG-0800) and NRC Regulatory Guides. In the RIPB approach, the design criteria are developed to achieve a target performance goal, which is defined by the annual frequency of occurrence of the design basis earthquake (i.e., Seismic Design Category) and the acceptable level of structural performance (i.e., Limit State) for the SSCs. ASCE 4-16 provides methods for performing a seismic analysis of structures to obtain the seismic response of these structures (e.g., building displacements, accelerations, in-structure response spectra) which are used in the design of the SSCs. It also provides methods for performing seismic analysis of SSCs to determine the seismic demands (e.g. member forces and displacements) needed to design individual SSCs. ASCE 43-18 (Draft) provides the criteria for the seismic design of SSCs using the seismic demands developed in ASCE 4-16. The use of ASCE Standard 43-18 (Draft), along with ASCE 4-16, provides the criteria for the seismic design of the SSCs. ASCE 43-18 (DRAFT) relies on other consensus codes and standards such as ACI 349 for reinforced concrete, AISC/N690 for steel structures, ASME Section III for pressure-retaining mechanical components and Containments, and IEEE-344 for Class 1E equipment. The goal of this technical review was to assess the adequacy of the provisions in these standards for use by the NRC in developing regulatory guidance for design of SSCs in nuclear power plants, based on the RIPB approach. The research reported herein describes the basis for acceptance of the new standards and identifies areas where additional staff guidance is needed for the seismic design of SSCs at nuclear power plants. This technical review has determined that ASCE 4-16 and ASCE 43-18 (Draft) provide an appropriate framework for the seismic design of SSCs at nuclear power plants using a Risk-Informed Performance-Based approach. However, some of the criteria therein warrant exceptions, qualifications, and/or clarifications.

42 ENGINEERING↗
DOE OSTI2020

Update on ASME AG-1 and Revision of the DOE Nuclear Air Cleaning Handbook - 20536

The American Society of Mechanical Engineers AG-1 Code on Nuclear Air and Gas Treatment is updated every two years. This paper provides an overview of issues being discussed within AG-1 committee meetings and revisions that will be included in the next version of the code. Additional information is provided on issues being discussed ranging from media velocity limits, control of airflow using actual versus and standard cubic feet per minute. Navigation and identification of applicable high-efficiency particulate air (HEPA) filter code sections can be confusing for those unfamiliar with AG-1. A general reorganization of filtration code sections clarifies geometry, flow, and performance envelopes for nuclear grade HEPA filters. The initial step consists of establishment of Section FN covering qualification of conventional, ceramic, metal, and high strength fibrous glass media. Future filter sections are currently expected to be characterized by geometry and flow direction (axial vs. radial). The Department of Energy's (DOE) Nuclear Air Cleaning Handbook (NACH) is also undergoing revision. An update on the status of that effort, along with issues and changes being considered. The original charter of the AG-1 Code was specific to commercial nuclear power plants, and to safety-related air cleaning systems. This has evolved over the years to include normal non-safety ventilation systems and process gas systems. Corollary guidance for DOE facilities has been available in several editions of the Nuclear Air Cleaning Handbook, starting with ORNL-NSIC 65. The latest edition of the Handbook, 2003, is undergoing revision and modifications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Assessing the ASME Section III, Division 5, Class A Primary Load Design Rules Against Creep Notch Effects

This report assesses the ASME Section III, Division 5, Subsection HB, Subpart B rules covering the design and construction of high temperature Class A nuclear reactor components for their robustness against creep notch effects. The creep notch effect combines the effect of multiaxial stresses on material creep deformation and damage. This report considers both effects, including the potential for creep mechanism shifts affecting the extrapolation of design data from high stress, short-time experimental data to low stress, long time operating component conditions. The report summarizes the history of and literature on multiaxial creep and surveys the current ASME design rules dealing with multiaxial effects. The report then describes the results of two dedicated numerical studies, one assessing the robustness of the ASME rules against uncertainty in extrapolating from uniaxial creep test data to multiaxial component conditions and a second study examining the potential effects of a mechanism shift from high stress, dislocation-mediated creep to diffusion-dominated creep at lower stresses. The final conclusion of the report is that the ASME rules adequately guard against multiaxial creep failure, though there are several aspects of the Code that could be optimized to provide less over-conservative design predictions or to provide a more consistent design margin as a function of temperature and stress. In a few areas, particularly the potential for mechanism shifts outside the currently-available experimental data, the Code rules should be revaluated as additional experimental data and new modeling and simulation results become available

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

MRP: Design Rules for Refractory Metals

Presentation on Design rules for refractory metals to be given at the Joint ART Materials/AMMT Program Review at DOE Headquarters, Germantown, MD, June 5-8, 2023. Includes microreactor program work for structural materials, failure points of interest, technology, technology maturation and de-risk designs, refractory metals, plan for developing code case as prototype by advanced reactor developers which meets design requirements, and also provides ASME Section III rules.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2014

Calibration of the NASA Glenn Research Center 16 in. Mass-Flow Plug

The results of an experimental calibration of the NASA Glenn Research Center 16 in. Mass-Flow Plug (MFP) are presented and compared to a previously obtained calibration of a 15 in. Mass-Flow Plug. An ASME low-beta, long-radius nozzle was used as the calibration reference. The discharge coefficient for the ASME nozzle was obtained by numerically simulating the flow through the nozzle from the WIND-US code. The results showed agreement between the 15 and 16 in. MFPs for area ratios (MFP to pipe area ratio) greater than 0.6 but deviate at area ratios below this value for reasons that are not fully understood. A general uncertainty analysis was also performed and indicates that large uncertainties in the calibration are present for low MFP area ratios.

Davis, David O.↗
NASA NTRS2012

Calibration of the NASA GRC 16 In. Mass-Flow Plug

The results of an experimental calibration of the NASA Glenn Research Center 16 in. Mass-Flow Plug (MFP) are presented and compared to a previously obtained calibration of a 15 in. Mass-Flow Plug. An ASME low-beta, long-radius nozzle was used as the calibration reference. The discharge coefficient for the ASME nozzle was obtained by numerically simulating the flow through the nozzle from the WIND-US code. The results showed agreement between the 15 in. and 16 in. MFPs for area ratios (MFP to pipe area ratio) greater than 0.6 but deviate at area ratios below this value for reasons that are not fully understood. A general uncertainty analysis was also performed and indicates that large uncertainties in the calibration are present for low MFP area ratios.

Davis, David O.↗
DOE OSTI2023

Theory of an Automatic Seepage Meter and Ramifications for Applications

A new approach for measuring fluxes across surface water—groundwater interfaces was recently proposed. The Automatic Seepage Meter (ASM) is equipped with a precise water level sensor and digital memory that analyzes water level time series in a vertical tube inserted into a streambed. The ability to infer flux values with high temporal resolution relies on an accurate interpretation of water level dynamics inside the tube. Here, we reduce the three-dimensional hydrodynamic problem that describes the ASM water level in a variety of field conditions to a single ordinary differential equation. This novel general analytical solution for estimating ASM responses is more comprehensive and flexible than previous approaches and is applicable to the entire range of field conditions, including steady or transient stream stages, evaporation, rainfall, and noise. For example, our analysis determines the timing of the nonmonotonic ASM response to a monotonic linear stream stage variation and explains previously used empirical parabolic approximation for estimating fluxes. We present algorithms for simultaneous inference of vertical interface flux and hydraulic conductivity values together with an example code. We quantify how the accuracy of parameter estimation depends on test duration and noise amplitude and propose how our analysis can be used to optimize field test protocols. On this basis, changing the ASM geometry by increasing the radius and decreasing tube insertion depth may enable ASM field test protocols that estimate interface flux and hydraulic conductivity faster while maintaining desired accuracy. Potential applications of joint parameter estimation are suggested.

47 OTHER INSTRUMENTATION↗
NASA NTRS2000

The NPARC Alliance Verification and Validation Archive

The NPARC Alliance (National Project for Applications oriented Research in CFD) maintains a publicly-available, web-based verification and validation archive as part of the development and support of the WIND CFD code. The verification and validation methods used for the cases attempt to follow the policies and guidelines of the ASME and AIAA. The emphasis is on air-breathing propulsion flow fields with Mach numbers ranging from low-subsonic to hypersonic.

Slater, John W.↗
DOE OSTI2023

SPC-2772 Rev 2 General Build to Print Fabrication Specification for Nuclear Use Items

This specification is for use with Safety Significant structures, systems, and components. The scope is for build to print fabrication including, but not limited to, machining, forming, welding, cutting, assembling, etc. Specific scope and associated requirements will be identified in drawings and/or additional procurement documents. ASME B31 fabrication is not included in the scope of this specification. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
NASA NTRS1988

Design of low Reynolds number airfoils. I

The low Reynolds number airfoils designated ASM-LRN-003 and -007 have been designed for high section L/D ratios using Drela's (1985) design-and-analysis code; close to 70-percent laminar flow is maintained on the upper surfaces, and 100-percent on the lower, at coefficients of lift of 1.0-1.3, assuming optimum laminar separation and transition control on the upper surface by means of suitable turbulators. If peak performance is critical, airfoils of this type with an undercut front lower surface and a correspondingly sharper leading edge may be resorted to.

Pfenninger, W.↗
DOE OSTI2023

FY23 status report on the development of new ASME Section III, Division 5 Class B rules

This report summarizes the work done in Fiscal Year 2023 on the development of the new American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 5, Class B rules to address the gaps identified for high temperature reactor designs. The summary of the design-by-analysis strain limit evaluation and creep-fatigue damage assessment is presented. The proposed design-by-analysis creep-fatigue damage calculation approach uses a new elastic follow-up-based Isochronous Stress Strain Curve stress relaxation procedure. This approach captures the elastic follow-up generated due to interactions of components with adjacent components, supports, and other connections in the power plant. A set of sample problems are selected to validate the proposed design-by-analysis rules for creep-fatigue damage assessment. The proposed Class B rules are evaluated against the Class A elastic design rules, and the experimental data obtained from a family of Simplified Model Test based key-feature test results. The proposed Class B creep-fatigue damage assessment methodology yields conservative design cycles estimates compared to the experimental results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Oak Ridge National Laboratory Technical Input for the Nuclear Regulatory Commission Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Glass Additive and Subtractive Manufacturing of a Fiber Optic Rotary Encoder for Downhole Pressure Sensing

This article presents a fiber optic rotary encoder-based pressure sensor for downhole applications. The rotary encoder, which is composed of an encoding pad and an all-glass optical fiber sensing head, converts the rotation angles of a pressure transducer to digital codes. The optical fiber only serves as the telemetry channel to directly transmit the data in digital format such that the environmental temperature influence and the long-time drift are minimized. The glass additive and subtractive manufacturing (ASM) techniques are used to embed the multichannel optical fibers into a bulk-fused silica glass substrate. Because the fused silica glass shares a similar material property to the optical fiber (e.g., thermal expansion coefficient), the optical fiber sensing head is well sealed and fixed even at an elevated temperature. The picosecond laser blackening technique is used to fabricate the high reflectivity contrast encoding pad with high accuracy. Here, the proposed pressure sensor was manufactured and experimentally verified to have a signal-to-noise ratio (SNR) of 23 dB and a linear pressure response with a root-mean-square error (RMSE) of 0.37. During the 400-h 250 °C long-term stability test, there was no digital code variation. In addition, a mathematical model to study the relationships between the sensor’s performances and design parameters was established. According to the model, a 12-bit encoder (0.02% pressure sensitivity) can be achieved with a radius of 20 mm.

02 PETROLEUM↗
NASA NTRS1994

Computational control of flexible aerospace systems

The main objective of this project is to establish a distributed parameter modeling technique for structural analysis, parameter estimation, vibration suppression and control synthesis of large flexible aerospace structures. This report concentrates on the research outputs produced in the last two years. The main accomplishments can be summarized as follows. A new version of the PDEMOD Code had been completed based on several incomplete versions. The verification of the code had been conducted by comparing the results with those examples for which the exact theoretical solutions can be obtained. The theoretical background of the package and the verification examples has been reported in a technical paper submitted to the Joint Applied Mechanics & Material Conference, ASME. A brief USER'S MANUAL had been compiled, which includes three parts: (1) Input data preparation; (2) Explanation of the Subroutines; and (3) Specification of control variables. Meanwhile, a theoretical investigation of the NASA MSFC two-dimensional ground-based manipulator facility by using distributed parameter modeling technique has been conducted. A new mathematical treatment for dynamic analysis and control of large flexible manipulator systems has been conceived, which may provide an embryonic form of a more sophisticated mathematical model for future modified versions of the PDEMOD Codes.

Sharpe, Lonnie, Jr.↗
DOE OSTI2023

A709 Procurement and ASTM Specification Status

This is a presentation regarding A709 Procurement and ASTM Specification Status for the Joint ART Materials/AMMT Program Review meeting. The A709 team consists of Argonne National Laboratory, Idaho National Laboratory, Oak Ridge National Laboratory, and Subject matter expert John Grubb and Richard Wright. Discussion topics will discuss incorporation of 709 into ASTM Standard A 240 for plate material in the solution annealed condition with a precipitation treatment as a supplemental requirement is underway, Incorporation of Alloy 709 into ASME, Section II, Specification SA 240 will be made upon the completion of the ASTM task, The required commercial scale heats of Alloy 709 for property characterization to provide the technical basis for Code qualification have been procured, and Laboratory testing of properties from these three heats is underway as will be discussed separately at this review meeting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗