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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 163 records · Page 9

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NASA NTRS2003

Relative Abundances and Energy Spectra of C, N, and 0 as Measured by the Advanced Thin Ionization Calorimeter Balloon Experiment

We present results on the spectra and the relative abundances of C, N, and 0 nuclei in the cosmic radiation as measured from the Advanced Thin Ionization Calorimeter Balloon Experiment (ATIC) . The ATIC detector has completed two successful balloon flights from McMurdo, Antarctica lasting a total of more than 35 days. ATIC is designed as a multiple, long duration balloon flight, investigation of the cosmic ray spectra from below 50 GeV to near 100 TeV total energy, using a fully active Bismuth Germanate calorimeter. It is equipped with a large area mosaic of silicon detector pixels capable of charge identification from H to Fe. As a redundancy check for the charge identification and a particle tracking system, three projective layers of x-y scintillator hodoscopes were employed, above, in the middle and below a 0.75 nuclear interaction length graphite target.

Fazely, A. R.↗

NASA NTRS2003

ATIC Experiment: Elemental Spectra from the Flight in 2000

The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a silicon matrix charge detector, a 0.75 nuclear interaction length graphite target, 3 scintillator strip hodoscopes, and an 18 radiation length thick BGO calorimeter to measure the cosmic ray composition and energy spectra from approximately 30 GeV to near 100 TeV. In this paper, we present preliminary results from the first flight, which was a test flight that lasted for 16 days, starting on 12/28/00.

Ahn, H. S.↗

NASA NTRS2004

The Energy Spectra of Proton and Helium Measured from the ATIC Experiment

The Advanced Thin Ionization Calorimeter (ATIC) balloon experiment is designed to investigate the composition and energy spectra of cosmic rays at the highest energies currently accessible from direct measurements, the region up to 100 TeV. The instrument consists of a silicon matrix for charge measurement, a graphite target (0.75 nuclear interaction length) to induce hadronic fragmentation, 3 scintillator strip hodoscopes for triggering and helping reconstruct trajectory, and a BGO calorimeter (18 radiation lengths) to measure the energy of incident particles. ATIC had two successful Long Duration Balloon (LDB) flights from McMurdo, Antarctica: from 12/28/00 to 01/13/01 and from 12/29/02 to 01/18/03. We present the energy spectra of proton and helium extracted from the ATIC flights, over the energy range from 100 GeV to 100 TeV, and compare them with the results from other experiments at both the lower and higher energy ends.

Ahn, H. S.↗

NASA NTRS2006

A Detailed FLUKA-2005 Monte Carlo Simulation for the ATIC Detector

We have performed a detailed Monte Carlo (MC) calculation for the Advanced thin Ionization Calorimeter (ATIC) detector using the MC code FLUKA-2005 which is capable of simulating particles up to 10 PeV. The ATIC detector has completed two successful balloon flights from McMurdo, Antarctica lasting a total of more than 35 days. ATIC is designed as a multiple, long duration balloon Bight, investigation of the cosmic ray spectra from below 50 GeV to near 100 TeV total energy; using a fully active Bismuth Germanate @GO) calorimeter. It is equipped with a large mosaic of silicon detector pixels capable of charge identification and as a particle tracking system, three projective layers of x-y scintillator hodoscopes were employed, above, in the middle and below a 0.75 nuclear interaction length graphite target. Our calculations are part of an analysis package of both A- and energy-dependences of different nuclei interacting with the ATIC detector. The MC simulates the responses of different components of the detector such as the Simatrix, the scintillator hodoscopes and the BGO calorimeter to various nuclei. We also show comparisons of the FLUKA-2005 MC calculations with a GEANT calculation and data for protons, He and CNO.

Gunasingha, R. M.↗

DOE OSTI2025

Evaluating design safety margins in the American Society of Mechanical Engineers graphite core components design-by-analysis assessments

Graphite is an important material being used for core components in next-generation high-temperature gas-cooled nuclear reactors. The selection of graphite grade for a specific Designer is a complex task, dependent on reactor conditions, component functionality, and required reliability. The American Society of Mechanical Engineers (ASME) provides two semi-probabilistic design-by-analysis assessments to evaluate graphite core components against design reliability targets. The simplified assessment uses a 2-parameter Weibull distribution to describe the graphite grade’s tensile-strength distribution to establish component stress limits. The full assessment uses the 3-parameter Weibull distribution and a modified Weakest-Link Theory approach to calculate a component design probability of failure. The paper defines recommended assessment rules, which are the as-written simplified assessment and the full assessment with parameter lower bounds, the modulus update with threshold reduction, and the 2027 grouping rules. Code rules are applied to three grades: 2114, IG-110, and NBG-18. The baseline margin calculation is developed using the experimental tensile dogbone specimen. Percent margin is defined as the percent reduction in the median experimental load to obtain the allowable load per ASME assessments. Under the recommended rules, the SRC–1 margin in the simplified assessment ranged from 40.2 % to 52.7 % among the grades in this study and from 36.1 % to 49.8 % in the full assessment. The full assessment only decreases the margins by 2.5–4.5 % for the SRC-1 components and 0–1.5 % for the SRC-2 components for this baseline case. Margin is inversely related to material median strength (i.e., the strongest grade, 2114, has the lowest margin).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Graphite Degradation Modeling and Analysis

A graphite component in a nuclear reactor core is subjected to variety of stresses and can experience degradation during normal and off-normal operation. Understanding how a graphite component will behave in service is essential to ensuring core structural stability and safe reactor operation. This report summarizes a graphite modeling tool currently under development at Idaho National Laboratory. The model incorporates several of the anticipated stresses during service and includes the effects of oxidation and irradiation prior to turnaround. This tool is intended to be used to help assess the design of graphite components by utilizing design code rules found in Section III, Division 5 of the American Society of Mechanical Engineering Boiler and Pressure Vessel Code. Specifically, the tool uses the methodologies found within the Full and Simplified assessments from Article HHA-3000 to verify that a graphite component has an acceptably low probability of failure.

36 MATERIALS SCIENCE↗

NASA NTRS1991

Graphite Fiber/Copper Matrix Composites for Space Power Heat Pipe Fin Applications

High specific thermal conductivity (thermal conductivity divided by density) is a major design criterion for minimizing system mass for space power systems. For nuclear source power systems, graphite fiber reinforced copper matrix (Gr/Cu) composites offer good potential as a radiator fin material operating at service temperatures above 500 K. Specific thermal conductivity in the longitudinal direction is better than beryllium and almost twice that of copper. The high specific thermal conductivity of Gr/Cu offers the potential of reducing radiator mass by as much as 30 percent. Gr/Cu composites also offer the designer a range of available properties for various missions and applications. The properties of Gr/Cu are highly anisotropic. Longitudinal elastic modulus is comparable to beryllium and about three times that of copper. Thermal expansion in the longitudinal direction is near zero, while it exceeds that of copper in the transverse direction.

McDanels, David L.↗

NASA NTRS2018

Advanced Protective Coatings for Gr-Based Nuclear Propulsion Fuel Elements

A protective coating for a graphite (Gr) containing fuel element used in a nuclear thermal propulsion system includes a first layer that is configured to resist hot hydrogen attacks. The first layer has a coefficient of thermal expansion that is higher than a coefficient of thermal expansion of the Gr containing substrate. The coating also includes a plurality of second layers located between the first layer and the substrate. The second layers are configured to mitigate the differences in coefficients of thermal expansion between the first layer and the substrate to minimize debonding and exposure of the substrate to hydrogen attack. Preferably, the protective coating can comprise an outermost first layer including zirconium carbide (ZrC), a second layer including niobium (Nb), a third layer including molybdenum (Mo), and a fourth layer including molybdenum carbide (Mo.sub.2C) located adjacent to the substrate.

Raj, Sai V.↗

DOE OSTI2023

Trepanning Graphite Samples at a Shutdown Magnox Reactor: A US-UK Joint Exercise for Nuclear Verification

In April 2022, a team sponsored by the US DOE Office of Nuclear Verification (ONV) shipped trepanning equipment to the nuclear power station at Trawsfynydd, Wales, UK to collect samples from the graphite moderator of Reactor 1. The sampling campaign to collect 48 graphite samples was completed in five weeks beginning July 11. Half the samples will be analyzed in the US to determine the total energy production of the facility. The other half will be analyzed by Magnox Ltd. to support decisions for options for the ultimate decommissioning path for the facility. This paper provides a description of the equipment and the operations required to trepan, collect, and package the samples while controlling the small amounts of radioactive graphite dust entrained on and within the tool. The most important sample parameter needed is the precise location of the sample, to within ±6 mm in a fuel channel, while the tool is deployed to between 12 and 20 meters below the operations floor; our means of achieving this precision is given.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

NASA NTRS1987

Indirect measurement of N-14 quadrupolar coupling for NH3 intercalated in potassium graphite

A method for indirect measurement of the nuclear quadrupolar coupling was developed and applied to NH3 molecules in the graphite intercalation compound K(NH3)4.3C24, which has a layered structure with alternating carbon and intercalant layers. Three triplets were observed in the H-1 NMR spectra of the compound. The value of the N-14 quadrupolar coupling constant of NH3 (3.7 MHz), determined indirectly from the H-1 NMR spectra, was intermediate between the gas value of 4.1 MHz and the solid-state value of 3.2 MHz. The method was also used to deduce the (H-1)-(H-1) and (N-14)-(H-1) dipolar interactions, the H-1 chemical shifts, and the molecular orientations and motions of NH3.

Tsang, T.↗

DOE OSTI2025

PSA 2025 Presentation: "Modeling and Sensitivity Analysis of a Generation IV Pebble Bed Reactor Using MELCOR 2.2"

Accompanying the advancement of reactor technologies is the need for computational modeling and simulation to predict their behavior under normal operating conditions and accident scenarios. New Generation IV reactor designs which employ non-conventional fuel have a particular need for modeling the behavior and release of radionuclides and other material from the fuel. In this work, MELCOR version 2.2, a system-level safety and accident scenario code developed by Sandia National Laboratories, was used to model a 200-MWth pebble bed modular reactor and calculate the inventories of circulating and deposited graphite, metal dust, and elemental components released from the fuel elements. A base case modeling the reactor under standard operating conditions was calculated using MELCOR and the inventories were extrapolated to 30 years of operation time using a logarithmic regression fit. A sensitivity analysis was also performed in which several key parameters for the base case model were modified to explore the effect of these changes on the inventories calculated by MELCOR. A set of transient scenario simulations for a depressurized loss of forced cooling (DLOFC) accident were also performed. The results of the sensitivity analysis and transient simulations are reported and discussed in relation to the modeling techniques used for this study.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Deimos: HALEU TRISO Heated Critical Experiment Data

Deimos was the first critical experiment using high-assay low-enriched uranium (HALEU) TRistructural ISOtropic (TRISO) fuel in over 40 years. HALEU TRISO is the desired fuel form for many of the advanced reactor designs in development; however, very little experimental data are available for this fuel type. Deimos was designed to utilize existing HALEU TRISO fuel in a large graphite moderator to obtain nuclear and reactor physics data to fill the gaps surrounding this fuel type and enrichment. In addition to cold critical data, three separate heated experiments were conducted to measure the temperature reactivity coefficient for this type of system. These measured coefficients were then compared to simulated coefficients to a first level order of fidelity. This comparison showed very good agreement for the experiment where only the inner core was heated and good agreement for the other two configurations, which included heating portions of the outer core. Less agreement when the outer core was heated is attributed to potential heating in the beryllium reflector, which has a positive temperature reactivity coefficient and was unaccounted for in the first-order models. Future heated experiments with Deimos will include temperature monitoring of the beryllium reflector to account for beryllium heating in the simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2003

ATIC Experiment: Preliminary Results from the Flight in 2002

Abstract The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a Silicon matrix for charge measurement, a flared graphite target to induce nuclear interactions, scintillator strip hodoscopes for triggering and helping reconstruct trajectory, and a BGO calorimeter to measure the energy of incident particles. In this paper, we discuss the second flight, which lasted 20 days, starting on 12/29/02. Preliminary results from the on-going analysis of the data including the proton and helium spectra are reported.

Ahn, H. S.↗

NASA NTRSDate not supplied

Knudsen Effusion Mass Spectrometry: Principles and Applications

Knudsen Effusion Mass Spectrometry (KEMS) is a well-established technique in physical chemistry. First, we discuss the basic principles of the technique. It is based on a small enclosure which sets up a condensed/vapor near equilibria. A small, well-defined orifice allows the vapor to be sampled a mass spectrometer. Although the number of groups utilizing this method has decrease significantly from its height in 1960s, we will show that it is still useful for many problems in physical chemistry, materials science, geology, and nuclear science. Specific examples illustrate its versatility: Measurement of the enthalpy of formation of HfO(g); measurement of activities in the Ti-Al-O and Ni-Al systems; measurement of activities in several rare earth silicate systems; measurement of vapor pressures above olivine; dating of minerals with K/Ar method; and determination of adsorption/desorption behavior on nuclear fission products on graphite. These are only a small sampling of the many applications of this versatile technique.

Mass spectrometry, thermodynamics, high temperatur↗

DOE OSTI2025

Interesting Graphite Facts

Interesting facts about graphite. Graphite within HTRs and how long they will last, potential issues prediction, model building to predict expected behavior, and frequent and thorough inspection and monitoring strategy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Enhanced graphite neutron reflector with beryllium oxide inclusions

A nuclear reactor is disclosed including a reactor core and a reflector assembly surrounding the reactor core. The reflector assembly includes a stationary reflector component including a graphite support structure comprising a plurality of channels defined therein and a plurality of beryllium-oxide pins positioned in the channels.

Aleshin, Yuriy↗

DOE OSTI2024

Overview of Graphite Model Development

Overview of graphite model development. Discuss graphite in motel salt, potential degradation mechanisms, stress due to volumetric heating, and the role of different parameters. Also wear modeling, modeling salt infiltration into graphite, oxidation modeling, and the existing graphite models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Materials property changes in ETU-10 graphite due to neutron irradiation at elevated temperatures

Graphite grade ETU-10, from IBIDEN Co., Ltd. Has been irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The irradiation program was developed to provide a preliminary study the irradiation-induced property changes to the dimensions/volume, elastic properties, strength, electrical resistivity, coefficient of thermal expansion, and thermal diffusivity/conductivity over a range of temperatures and neutron exposures that may be relevant for future nuclear reactors. The irradiation envelope covers a range of irradiation temperatures (300°C–900 °C) and fluences (up to 40 × 10 25 n/m 2 [E > 0.1 MeV] or ~30 dpa) that would be relevant for advanced nuclear reactors. Further, the dimensional change was observed to be anisotropic for an isotropic graphite, the specimen dimensions, volume, Young's modulus, shear modulus, and strength all displayed a parabolic fluence dependence, the electrical resistivity had a rapid rise followed by a decrease and a later increase, at high fluence the mean coefficient of thermal expansions was similar for all irradiation temperatures, and thermal conductivity rapidly decreased followed by a continued loss.

36 MATERIALS SCIENCE↗