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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
NASA NTRS2000

ASM-Triggered Too Observations of Kilohertz Oscillations in Three Atoll Sources

Three Rossi Timing Explorer (RXTE) observations were carried out for this proposal based on target of opportunity triggers derived from the All-Sky Monitor (ASM) on RXTE. We obtained short observations of 4U1636-536 (15ks) and 4U1735-44 (23ks) and a longer observation of 4U0614+091 (117ks). Our analysis of our observations of the atoll neutron star x-ray binary 4U1735-44 lead to the discovery of a second high frequency quasiperiodic oscillation (QPO) in this source. These results were published in the Astrophysical Journal Letters. The data obtained on the source 4U0614+091 were used in a comprehensive study of this source, which will be published in the Astrophysical Journal. The data from this proposal were particularly critical for that study as they lead to the detection of the highest QPO frequency every found in the x-ray emission from an x-ray binary which will be important in placing limits on the equation of state of nuclear matter.

Kaaret, P.↗
NASA NTRS2015

ASME Section VIII Recertification of a 33,000 Gallon Vacuum-jacketed LH2 Storage Vessel for Densified Hydrogen Testing at NASA Kennedy Space Center

The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) has been developed at NASA Kennedy Space Center in Florida. GODU-LH2 has three main objectives: zero-loss storage and transfer, liquefaction, and densification of liquid hydrogen. A cryogenic refrigerator has been integrated into an existing, previously certified, 33,000 gallon vacuum-jacketed storage vessel built by Minnesota Valley Engineering in 1991 for the Titan program. The dewar has an inner diameter of 9.5 and a length of 71.5; original design temperature and pressure ranges are -423 F to 100 F and 0 to 95 psig respectively. During densification operations the liquid temperature will be decreased below the normal boiling point by the refrigerator, and consequently the pressure inside the inner vessel will be sub-atmospheric. These new operational conditions rendered the original certification invalid, so an effort was undertaken to recertify the tank to the new pressure and temperature requirements (-12.7 to 95 psig and -433 F to 100 F respectively) per ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. This paper will discuss the unique design, analysis and implementation issues encountered during the vessel recertification process.

Liquefaction↗
DOE OSTI2021

Recommendation for Limiting Conditions for ASME BPVC Section III Division 5 Allowable Stress Criteria

Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) specifies rules for elevated temperature nuclear reactors. Time dependent allowable stresses are an important aspect of the high temperature design models and are related to the time to 1% strain, time to rupture, and time to onset of tertiary creep. The time to onset of tertiary creep criterion has an overly strong negative influence on the allowable stresses for some structural materials, likely due to very limited datasets and misidentification of microstructural changes rather than onset of significant creep damage, as tertiary creep is meant to represent. It is recommended that the onset of tertiary creep be determined through an alloy specific ratio between onset of tertiary creep and creep rupture time, which would give access to the significantly larger creep rupture time datasets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2021

ASME Code Rules and ASTM Standards Integration for Ceramic Composite Core Materials and Components 1

Fiber-reinforced ceramic matrix composites have many desirable properties for high-temperature nuclear applications, including excellent thermal and mechanical properties and reasonable to outstanding radiation resistance. Over the last 20 years, the use of ceramic composite materials has already expanded in many commercial nonnuclear industries as fabrication and application technologies mature. The new ASME design and construction rules under Section III, Subsection HH, Subpart B lay out the requirements and criteria for materials, design, machining and installation, inspection, examination, testing, and the marking procedure for ceramic composite core components, which is similar to the established graphite code under Section III, Subsection HH, Subpart A. Moreover, the general requirements listed in Section III, Subsection HA, Subpart B are also expanded to include ceramic composite materials. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in the American Society for Testing and Materials Committee C28 on Advanced Ceramics with a current focus on ceramic composite tubes. Details of the composites code, design methodology, and similarities to the graphite code, as well as guidance for the development of specifications for ceramic composites for nuclear application and recent standard developments, are discussed. The next step is to "close the gap" to support licensing aspects by validating the code with benchmarking data.

Geringer, Josina↗
DOE OSTI2022

Interim Creep, Fatigue and Creep-Fatigue Data from FY 2022 INL Testing of A709 with Precipitation Treatment for ASME Code Case Data Package

This report provides the status of creep, fatigue, and creep-fatigue testing that transpired in fiscal year 2022 at Idaho National Laboratory (INL). This testing is being conducted to develop the data package to qualify A709 in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). This would permit the use of A709 for elevated temperature nuclear construction. Preliminary results continue to demonstrate the improved creep and fatigue resistance of A709 compared to 316H stainless steel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
DOE OSTI2023

Qualifying nuclear graphite components using ASME guidelines

Qualifying nuclear graphite components using ASME guidelines, including Semi-probabilistic, probabilistic, and deterministic approaches to design, theory of methods, pre-assessment analysis inputs, post assessment analysis outputs, simplified assessment, full assessment, and application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ECAR-6564 Rev 1 MARVEL Project Primary Coolant System ASME BPVC Section III Division 5 Design by Analysis

This report demonstrates that the MARVEL Reactor Primary Coolant Boundary (herein referred to as the Primary Coolant System, or PCS) is designed to meet ASME BPVC Section III Division 5 elevated temperature service design-by-analysis criteria. The analysis approach that delineates division of responsibilities to meet project objectives is discussed herein. In short, Design and Service Level A, B, and C code calculations are detailed in ECAR-6580 [9] for the majority of the PCS with complex geometry, while the Lower Downcomers, Bottom Head, and Reactor Core Barrel are analyzed in ANL-24/36. Service Level D code calculations are detailed in this document.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-7547 Rev 0 MARVEL Intermediate Heat Exchanger Liner ASME Analysis

This ECAR analyzes the Secondary Coolant System liner, otherwise known as the Intermediate Heat Exchanger (IHX) Liner, as an ASME Section VIII, Division 1 pressure vessel. This analysis is only necessary in the event of a failure of the primary coolant boundary, in which case the IHX liner acts as a secondary confinement boundary to prevent primary coolant leakage. The liner main body or vessel is analyzed, as well as the liner flange ring.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

SPC-70731 Rev 0 MARVEL Reactor Project ASME BPVC Section III Division 5 Design Specification

This specification establishes the requirements for the design of the MARVEL Primary Coolant System (PCS) Boundary, Guard Vessel System (GVS) Boundary, Core Support Structure (CSS), and Internal Heat Exchanger (IHX) Liner to provide a complete basis for ASME BPVC Section III Division 5 construction. All information in this section is generic and applicable to all metallic components unless otherwise specified.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2015

Draft ASME Boiler and Pressure Vessel Code Section III, Division 5, Section HB, Subsection B, Code Case for Alloy 617 and Background Documentation

Alloy 617 is the leading candidate material for an intermediate heat exchanger for the very high temperature reactor. To evaluate the behavior of this material in the expected service conditions, strain controlled cyclic tests that include long hold times up to 240 minutes at maximum tensile strain were conducted at 850°C. In terms of the total number of cycles to failure, the fatigue resistance decreased when a hold time was added at peak tensile strain. Increases in the tensile hold duration degraded the creep fatigue resistance, at least to the investigated strain controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. The creep fatigue deformation mode is considered relative to the lack of saturation, or continually decreasing number of cycles to failure with increasing hold times. Additionally, preliminary values from the 850°C creep fatigue data are calculated for the creep fatigue damage diagram and have higher values of creep damage than those from tests at 950°C.

36 MATERIALS SCIENCE↗