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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

DOE ART Graphite R&D Program

Graphite overview: Material science, supply chain facts, nuclear material science, licensing thoughts

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DOE ART Graphite R&D Introduction

Introduction of ART Graphite R&D to include oxidation activities, oxidation resistant graphite, model development, ASME component failure, ASME code development (design rules), ceramic composites, AGC update, molten salt intrusion, split-disk studies, wear testing, and concluding remarks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ART Graphite R&D Program Status - 2020

Structure of the program Graphite material issues of interest How do things fit together Status of the program Baseline AGC Experiment Update AGC-4 and HDG-1 Irradiation Oxidation Modeling ASME Code development Discussion of how the data can be used Modeling behavior of core components Licensing/code development Interaction with vendors Collaborations (domestic and international)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

ART Graphite R&D Introduction

Five Research Areas: Licensing and Code Graphite Research and Development Behavior Models Virgin Properties Mechanisms and Analysis Irradiation

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Status on Development of Graphite Analytical Tool (GAT)

The DOE-ART Graphite R&D program has been generating significant amounts of irradiated and unirradiated graphite data since 2006 when the program was part of the DOE NGNP (Next Generation Nuclear Plant) Project. This data includes critical irradiation creep and irradiated material property changes from the Advanced Graphite Creep (AGR) experiment as well as significant amounts of data on unirradiated material property values on several current nuclear graphite grades (Baseline program). Previously, Idaho National Laboratory has developed an internal analysis tool to assist with analysis of the unirradiated and irradiated data. The Graphite Analytical Tool (GAT) is intended to provide easy access to the graphite data in the form of comparing unirradiated and irradiated material property changes, comparison of material property differences between various nuclear graphite grades, and illustrate trends within the irradiated and unirradiated data generated within the DOE-ART Graphite R&D program. This report summarizes the progress to-date on the development of this analytical tool.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

DOE ART Graphite Program Intro Status 2023

General description of the ART Graphite Program with Status for 2023. This includes list of staff members, and brief description of behavior models, licensing & code, graphite R&D, As-Fab'd Properties, Machanism and Analysis, Irradiation. Also detail FY23 Activites, with AGC Update, Oxidation Activiteis, Baseline STatus, ASME Code Development, VIC Project Update, Molten salt studies. Slide included to highlight each employee contributions for work and papers and NEUPs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Graphite Licensing (ASME) and Testing (ASTM) Technical Assistance

DOE ART Graphite R&D program is currently involved in developing the structural code for licensing the new High Temperature Reactor (HTR) designs through the ASME Boiler Pressure Vessel Code (BPVC). The basic structural design code for graphite core components has been written, approved, and within the BPVC since 2012. However, several areas within the graphite code remain to be improved, specifically the addition of material property changes resulting from environmental degradation (neutron flux and oxidation). The incorporation of this data, which can significantly affect the graphite components and alter the structural integrity of the core during operation, is proving to be an important activity. How this data will be incorporated within the code and how it will be used for license applications will be critical to a number of HTR applicants in the future. These modifications to the code must take into account the unique irradiation and oxidation response of the wide spectrum of available nuclear graphite grades currently under consideration for HTR applications. Additionally, the ASME BPVC requires material property testing within the expected operational conditions of the HTR core including neutron radiation flux, elevated temperatures, and after chronic/acute oxidation of the graphite components. Currently, no standardized graphite testing methods are approved for these environmental conditions. New American Society for Testing and Materials (ASTM) standardized tests for elevated temperature testing, small sample testing, and testing after oxidation must be developed to support the requirements within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Effect of Sample Thickness on the Tensile Strength of Small Graphite Discs

This report formally documents the completion of the Advanced Reactor Technologies (ART) Level 3 Milestone (M3TG-24OR0501054), “Continue activities related to Split Disc-DIC - complete analysis of effect of sample thickness on one fine grain graphite,” due May 31, 2024. Details within this report outline the status of activities aimed at elucidating the effect of sample thickness on the tensile strength measurement using the small-disc, split-disc testing standard in support of the US Department of Energy’s ART Graphite R&D Program. The ASTM D8289, Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite, was developed to provide an alternative means for testing tensile strength on smaller specimens, which are compatible with available irradiation capsule volumes, and other compatibility measurements. ASTM D8289 specifies that acceptable specimen diameter can range from 6 to 12.7 mm and that the maximum allowed sample thickness should be half of the diameter. However, information is limited with respect to the effect of thickness on the measured splitting tensile strength. This report documents efforts to understand the effect of sample thickness on splitting tensile strength. The work involved testing Ø12.7 mm samples of fine-grain graphites 2114 and IG-110 of different thicknesses (6.35, 5, 4, and 3 mm). The digital image correlation (DIC) method was applied to the samples, along with the ASTM D8289 Standard, to help interpret the measured results.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Summary of US DOE R&D Activities on Graphite Oxidation (2006–2021)

The objective of the international collaboration between United States Department of Energy (U.S.-DOE) and Generation IV International Forum (GIF) is the development of the next generation of nuclear energy systems. The current GIF Project Arrangement (PA) on Materials (2018-2022) was revised in 2019 and extended for another 10 years (2020-2030). The Work Package 1 (“Graphite”) of the extended Project Plan (PP) on Materials specifies technical tasks and High Level Deliverables for research and development (R&D) activities related to using graphite in fuel elements, reflectors, and support structures of Very High Temperature Reactors (VHTR). The graphite tasks include specification and acquisition, qualification and development of new grades, characterization of properties, and development of behavior models. Specifically, Task 1.4 (“Graphite Oxidation Behavior”) outlines planned activities related to acute oxidation by air and chronic oxidation by impurities in the helium coolant. A final report on experimental data regarding graphite oxidation behavior is scheduled for 2022 (deliverable 3.1.1.4.a). In preparation of this deliverable, this document summarizes the R&D activities funded by U.S.-DOE from 2006 (the inception of the VHTR system arrangement) through present (2021). This report is being submitted to the GIF Graphite Working Group (GWG) to serve as input for the GWG high-level deliverable to the Project Management Board (PMB) of PA on Materials. Besides U.S.-DOE, other organizations participating to Task 1.4 of the current PA on Materials are: European Commission’s Joint Research Center (JRC), Korea Atomic Energy Research Institute (KAERI), and Japan Atomic Energy Agency (JAEA). U.S.-DOE is the main contributor on graphite oxidation R&D, with 85 % commitment of total funding during 2018-2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

ASME Code Development - Nonmetallics

ASME Code Development - Nonmetallics slides to include behavior models, licensing & code, graphite R&D Program, As-Fab'd Properties, Mechanisms and Analysis, and Irradiation. Discussion on where these fit into Irradiation behavior, Ceramic Composites, Design, and Component Failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1978

A perspective on composites

The development of composites in the United States in the 1960s and 1970s is briefly reviewed with emphasis on aerospace applications. Consideration is given to such highlights as the manufacture of boron filaments, the use of boron-epoxy composites for aerospace structures, graphite-epoxy composites, and current graphite R & D work.

Lovelace, A. M.↗