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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 OSTI2012

Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Type A Verification Report for the Brookhaven Graphite Research Reactor Engineered Cap, Brookhaven National Laboratory Upton, New York

The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Graphite Research Reactor (RFT) Spectrometric Investigation During Dismantling - 17006

Dismantling of research reactors are carried out at the National Research Centre 'Kurchatov Institute'. Technical start-up of the 20 MW capacity experimental graphite water cooling reactor (RFT) was started up in 1952. During experimental work, some accidents, followed by a pollution of a graphite blocks by fission products of nuclear fuel, took place. After 10 years working period the reactor was shut down. The fuel was removed and the reactor was partially dismantled. Reactor graphite stack (cylinder 380 cm high and 260 cm in diameter) was isolated. RFT reactor dismantling was began in 2015. Video investigation of reactor channels in graphite stack was made before dismantling. Graphite in channels near active zone of reactor was destroyed. Measurements of exposure dose rate inside the channels near active zone showed value up to 2 Sv. The samples were taken from different layers of stack during graphite blocks removing from the reactor. Complex α-, β-, γ- spectrometric analysis of this samples was made. The graphite blocks were contaminated by fuel components and its fission products. It was found that {sup 137}Cs and {sup 90}Sr were the main contamination radionuclides in graphite. After graphite block removing and reactor vessel bottom cleaning, exposure dose rate reached 5-7 mSv/h at upper level of the reactor vessel and at the bottom ∼40 mSv/h at the bottom. Gamma-spectrometric analysis of reactor vessel was made. These measurements showed that activated in iron reactor vessel {sup 60}Co made the main contribution to exposure dose rate. After reactor vessel dismantling exposure dose rate reached 2 mSv/h at upper level of the reactor vessel and∼ 3 mSv/h at the bottom. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1965

Carbon and graphite research

Structural changes accompanying creep deformation of pyrolytic carbon and electron micrography and fractography of glassy carbons

ELECTRON MICROSCOPY↗
NASA NTRS1972

CMB-13 research on carbon and graphite

Research and development are reported on phenolic resin binders, coke flour, binder soaking, hot molding glass-like carbons, and particle packing.

Smith, M. C.↗
DOE OSTI2020

Vacancy-driven variations in the phonon density of states of fast neutron irradiated nuclear graphite

Research examining radiation damage and its effects in graphite began in the 1940’s during the development of moderated nuclear reactors. Interest in this topic is expanding because of emerging applications associated with fullerenes and carbon nanostructures as well as its long-standing use as a fission moderator. In this work, we report the measurements of the full phonon density of states of irradiated nuclear graphite using inelastic neutron scattering experiments at room temperature. The samples were previously exposed to different levels of neutron doses and irradiation temperatures. The phonon density of states of perfect and defected graphite supercells including different configurations of vacancies and interstitials were calculated using first-principles direct method. Despite high neutron irradiation doses, the scattering measurements show that the induced damage is localized, and the layered structure is preserved. In conclusion, a comparison of the measured phonon densities of states of irradiated samples with those calculated of defected supercells indicates that the main changes observed are mainly attributed to formation of vacancies.

36 MATERIALS SCIENCE↗
NASA NTRS1977

Research on graphite reinforced glass matrix composites

The results of research for the origination of graphite-fiber reinforced glass matrix composites are presented. The method selected to form the composites consisted of pulling the graphite fiber through a slurry containing powdered glass, winding up the graphite fiber and the glass it picks up on a drum, drying, cutting into segments, loading the tape segment into a graphite die, and hot pressing. During the course of the work, composites were made with a variety of graphite fibers in a glass matrix.

Bacon, J. F.↗
NASA NTRS1972

CMB-13 research on carbon and graphite

The research on graphite and carbon for this period is reported. Topics discussed include: effects of grinding on the Santa Marie graphites, properties and purities of coal-tar, resin-bonded graphite, carbonization of resin components, and glass-like carbon filler.

Smith, M. C.↗
NASA NTRS1977

Research on Graphite Reinforced Glass Matrix Composites

This report contains the results obtained in the first twelve months of research under NASA Langley Contract NAS1-14346 for the origination of graphite-fiber reinforced glass matrix composites. Included in the report is a summary of the research by other investigators in this area. The method selected to form the composites consisted of pulling the graphite fiber through a slurry containing powdered glass, winding up the graphite fiber and the glass it picks up on a drum, drying, cutting into segments, loading the tape segment into a graphite die, and hot pressing. During the course of the work, composites were made with a variety of graphite fibers in a C.G.W. 7740 (Pyrex) glass matrix. The graphite fibers used included Hercules HMS, Hercules HTS, Thornel 300S, and Celanese DG-102 and, of these, the Hercules HMS and Celanese DG-102 graphite fibers in C.G.W. 7740 gave the most interesting but widely different results. Hercules HMS fiber in C.G.W. 7740 glass (Pyrex) showed an average four-point flexural strength of 848 MPa or 127,300 psi. As the test temperature was raised from room temperature to 560 C in argon or vacuum, the strength was higher by 50 percent. However, in air, similar tests at 560 C gave a severe loss in strength. These composites also have good thermal cycle properties in argon or vacuum, greatly increased toughness compared to glass, and no loss in strength in a 100 cycle fatigue test. Celanese DG-102 fiber in C.G.W. 7740 glass (Pyrex) had a much lower flexural strength but did not suffer any loss in this strength when samples were heated to 560 C in air for 4 hrs.

Bacon, J. F.↗
DOE OSTI2016

Experimental research of graphite behavior under the air ingress accident condition of High Temperature Gas-cooled Reactor - 18752

As the major component of both cladding and grid spacer materials of High Temperature Gas-cooled Reactor (HTGR), graphite plays an important role on heat removal during the severe accident sequence. However, its heat transfer behavior in high temperature condition may be affected by the oxidation of graphite when the air or steam ingresses in an accident. In this research, an experiment was conducted to analyze the graphite heat transfer behavior in high temperature oxidable environment. An assembly including two graphite rings was used as the test section. It was heated up by a tungsten wire, which is installed inside the inner graphite ring and be joule heated. Temperature was measured by several K-type thermos-couples in certain positions. Argon gas was injected in advance to avoid the unwanted oxidation by the oxygen in the air. Air was injected into the gap between two graphite rings once the temperature of the inner ring reached target point to create the oxidable environment. Based on the results, a rapid increase of inner graphite ring's temperature can be observed after starting injected the air, and a rapid decrease after stopped the injection. It might due to the chemical reaction heat during the graphite surface oxidation. By comparing the steady state before and after air injection, the inner specimen s temperature is a little bit lower after the oxidation, which suggests that its ability of radiation heat transfer might be enhanced a little bit by oxidation. By visualized the test section, coarser surface at the inner graphite ring can be observed after the oxidation. In addition, about 8.9% mass loss of the oxidized graphite can also be found from the posttest examination. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

A Review of Finite Element Method Models for Nuclear Graphite Applications

Graphite components serve as structural support and as a neutron moderator in multiple nuclear power stations and is also a prospective material in Generation IV reactor concepts. The environment (e.g. high temperature and intense radiation) within a graphite core changes the geometry, microstructure and material properties of the graphite components, that could limit the lifetime (i.e. safe operation) of a reactor. Some of the assessments made for graphite reactor core components include Finite Element Method (FEM) simulations. These simulations are used to estimate the stress state and strain state and probability of failure of irradiated graphite components. FEM also forms part of the validation of measurements or as a complementary technique to experimental procedures in graphite research. A comprehensive overview of FEM methodologies is provided in this review, including: FEM software, basics steps to set up a FEM model, and analytical solutions used to validate simulations for this field of study. In conclusion, the objective of this paper is to inform researchers in the different available FEM techniques and tools used in the research of nuclear graphite.

42 ENGINEERING↗
NASA NTRS1978

Research on graphite reinforced glass matrix composites

A composite that can be used at temperatures up to 875 K with mechanical properties equal or superior to graphite fiber reinforced epoxy composites is presented. The composite system consist of graphite fiber, uniaxially or biaxially, reinforced borosilicate glass. The mechanical and thermal properties of such a graphite fiber reinforced glass composite are described, and the system is shown to offer promise as a high performance structural material. Specific properties that were measured were: a modified borosilicate glass uniaxially reinforced by Hercules HMS graphite fiber has a three-point flexural strength of 1030 MPa, a four-point flexural strength of 964 MPa, an elastic modulus of 199 GPa and a failure strain of 0.0052. The preparation and properties of similar composites with Hercules HTS, Celanese DG-102, Thornel 300 and Thornel Pitch graphite fibers are also described.

Bacon, J. F.↗
NASA NTRS1980

Research on graphite reinforced glass matrix composites

High levels of mechanical performance in tension, flexure, fatigue, and creep loading situations of graphite fiber reinforced glass matrix composites are discussed. At test temperatures of up to 813 K it was found that the major limiting factor was the oxidative instability of the reinforcing graphite fibers. Particular points to note include the following: (1) a wide variety of graphite fibers were found to be comparable with the glass matrix composite fabrication process; (2) choice of fiber, to a large extent, controlled resultant composite performance; (3) composite fatigue performance was found to be excellent at both 300 K and 703 K; (4) composite creep and stress rupture at temperatures of up to 813 K was limited by the oxidative stability of the fiber; (5) exceptionally low values of composite thermal expansion coefficient were attributable to the dimensional stability of both matrix and fiber; and (6) component fabricability was demonstrated through the hot pressing of hot sections and brazing using glass and metal joining phases.

Prewo, K. M.↗
NASA NTRS1981

Research on graphite reinforced glass matrix composites

A broad group of fibers and matrices were combined to create a wide range of composite properties. Primary material fabrication procedures were developed which readily permit the fabrication of flat plate and shaped composites. Composite mechanical properties were measured under a wide range of test conditions. Tensile, flexure mechanical fatigue, thermal fatigue, fracture toughness, and fatigue crack growth resistance were evaluated. Selected fiber-matrix combinations were shown to maintain their strength at up to 1300 K when tested in an inert atmosphere. Composite high temperature mechanical properties were shown to be limited primarily by the oxidation resistance of the graphite fibers. Composite thermal dimensional stability was measured and found to be excellent.

Prewo, K. M.↗
DOE OSTI2017

AGC 2 Irradiated Material Properties Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

36 MATERIALS SCIENCE↗
DOE OSTI2019

AGC-3 Irradiation Creep Strain Data Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete irradiated properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

42 ENGINEERING↗