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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
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 OSTI2025

Study of the Design and Characteristics of a Modified Pulsed Plasma Thruster with Graphite and Tungsten Trigger Electrodes

The paper presents experimental results for a modified pulsed plasma thruster (PPT) with solid propellant, using a coaxial anode–cathode design. Graphite from pencil leads served as propellant, and a tungsten trigger electrode was tested to reduce carbonization effects. Experiments were performed in a vacuum chamber at 0.001 Pa, employing diagnostics such as discharge current/voltage recording, power measurement, ballistic pendulum, time-of-flight (TOF) method, and a Faraday cup. Current and voltage waveforms matched an oscillatory RLC circuit with variable plasma channel resistance. Key discharge parameters were measured, including current pulse duration/amplitude and plasma channel formation/decay dynamics. Impulse bit values, obtained with a ballistic pendulum, reached up to 8.5 μN·s. Increasing trigger capacitor capacitance reduced thrust due to unstable “pre-plasma” formation and partial pre-discharge energy loss. Using TOF and Faraday cup diagnostics, plasma front velocity, ion current amplitude, current density, and ion concentration were determined. Tungsten electrodes produced lower charged particle concentrations than graphite but offered better adhesion resistance, minimal carbonization, and stable long-term performance. The findings support optimizing trigger electrode materials and PPT operating modes to extend lifetime and stabilize thrust output.

Faraday cup↗
DOE OSTI2022

Towards a Benchmark Experiment with the Compact Nuclear Power Source

The Compact Nuclear Power Source (CNPS) was a high-assay low enriched uranium (HALEU) tristructural isotropic (TRISO)-fueled, graphite-moderated microreactor constructed in 1987 at Los Alamos National Laboratory. The reactor, conceived as a power source for short-range radar stations, was designed to be "walk-away safe," and was cooled by heat pipes and ambient air. Though the project was formally cancelled after the fuel and moderator material had been received, its potential to meaningfully advance the body of critical and integral data was evident, and critical experiments with a mock-up of the reactor proceeded at TA-18 until its disassembly in 1991. The reflector and some components of the core would later go on to see service as part of the New Production Reactor Modular High-Temperature Gas-Cooled Reactor (NP-MHTGR) critical experiments. In light of the progress made by groups like Westinghouse and X-Energy towards contemporary graphite-moderated microreactors, the system remains an attractive candidate for the basis of a benchmark experiment even today. Uncertainties remain, however, in some features of the system–inconsistencies in the dimensions and composition of the reactor’s components as described in literature and implemented in current computational models. The identification, quantification, and to the extent possible, the minimization of these uncertainties is a crucial task on the path towards creating a benchmark based on the CNPS system. The present work seeks to initiate this process with a survey of available documentation and improvements to the accuracy of CNPS neutronics models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGC Experiment Status

AGC Experiment Status to include: History and status of the AGC-4, Review of the AGC experiment, AGC graphite grades and samples, The AGC-4 capsule and specimens, and HDG-1 and HDG-2. Description of the Experiment, AGC graphite grades and samples, AGC-1 Test Train, AGC-1 Test Train Design Features, New AGC Irradiation Schedule (2018) and update in schedule, Irradiation and disassembly history, present and future status, and shipment and initial PIE. Initial PIE strategy, recovered specimens, PIE results, and status of experiments.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Design and Analysis of a 250 MW Plate-Fuel Reactor for Nuclear Thermal Propulsion

Nuclear thermal propulsion (NTP) is a viable option for deep space missions due to its high thrust and lightweight system. Promising fuel options considered for the high operating temperature of NTP include graphite composite fuel and tungsten cermet fuel. As there remain uncertainties in the fabrication and performance of these two fuel elements, alternative designs using fuel plates and more typical fuels based on proven nuclear technologies are also being considered. This paper provides a summary of thermal hydraulic and neutronic analyses of the straight-plate fuel assembly design used for a 250 MW NTP core. The summary of technical findings from a series of scoping studies might be useful for future NTP engine designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

ZiaCore Critical Experiment Demonstrates Key Technologies for Nuclear Energy Systems

ZiaCore is a LANL Laboratory Directed Research and Development (LDRD) project focused on developing and demonstrating key technologies for future nuclear energy systems. The project itself was split into three tasks: 1) Design of the ZiaCore Reactor, a UO2 fueled, graphite and zirconium-hydride (ZrH) moderated, heat pipe cooled micro-reactor 2) Development of the ZrH and heat pipes components 3) Performance of a critical experiment with a representative portion of the ZiaCore reactor incorporating the ZrH and heat pipes developed and made at LANL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
DOE OSTI2020

Design and Analysis of a 250 MW Plate-fuel Reactor for Nuclear Thermal Propulsion

Nuclear thermal propulsion (NTP) system is a viable option for deep space missions considering its high thrust and lightweight. Promising fuel options being considered for the high operating temperature of NTP include the graphite composite fuel and the tungsten cermet fuel. As there remains uncertainties in the fabrication and performance of these two fuel elements, alternative designs using fuel plates and more standard fuels that are based on proven and tested nuclear technologies are being considered. This paper provides a summary of thermal hydraulic and neutronic analyses of the straight fuel plate concept used for a 250 MW NTP core. The summary of technical findings from a series of scoping studies might be useful for the future NTP engine designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Chicago Pile-1 paved the way for nuclear science and a lab in Los Alamos First self-sustaining nuclear chain reaction was nearly 80 years ago

On a bitter-cold winter day, 43 scientists gathered at an abandoned squash court at the University of Chicago where they would ultimately enable a secret lab in Los Alamos to change the world just years later. It was December 2, 1942. The group, led by Italian physicist and Nobel laureate Enrico Fermi, stacked graphite bricks, piling 57 layers that totaled more than 770,000 pounds. Later named Chicago Pile-1, their goal was to create the world’s first self-sustaining, controlled nuclear chain reaction. Inside the approximately 20-feet-tall pile were smaller blocks of uranium and control rods that, when removed, would cause the reaction to go critical – meaning create a nuclear chain reaction. It was roughly $1 million worth of materials, equivalent to nearly $16 million today, and a concept that a nuclear chain reaction would allow the weaponization of the atom. “Its success would be the crucial proof needed to know it would be possible to create an atomic bomb,” said LANL Historian Roger Meade (C-NR). “This was the precursor to the Lab we have today, nearly 80 years later.”

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Crosslinked Polyethyleneimine Gel Polymer Interface to Improve Cycling Stability of RFBs

Redox flow batteries are considered a promising technology for grid energy storage. However, capacity decay caused by crossover of active materials is a universal challenge for many flow battery systems, which are based on various chemistries. In this paper, using the vanadium redox flow battery as an example, we demonstrate a new gel polymer interface (GPI) consisting of crosslinked polyethyleneimine with a large amount of amino and carboxylic acid groups introduced between the positive electrode and the membrane. The GPI functions as a key component to prevent vanadium ions from crossing the membrane, thus supporting stable long-term cycling. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) measurements were conducted to investigate the effect of GPI on the electrochemical properties of graphitic carbon electrodes (GCFs) and redox reaction of catholyte. X-ray photoelectron spectroscopy (XPS) and 1 H nuclear magnetic resonance (NMR) spectra demonstrated that the crosslinked GPI is chemically stable for 100 cycles without dissolution of polymers and swelling in the strong acidic electrolytes. Results from inductively coupled plasma mass spectrometry (ICP-MS), Fourier-transform infrared (FTIR) spectroscopy, and energy-dispersive X-ray (EDX) spectroscopy proved that the GPI is effective in maintaining the concentration of vanadium species in their respective half-cells, resulting in improved cycling stability because of it prevents active species from crossing the membrane and stabilizes the oxidation states of active species.

Lim, Hyung-Seok↗
DOE OSTI2020

AGR-5/6/7 Test Train Disassembly & Tools Phases I & II

This document provides the process for qualification of the AGR-5/6/7 Test Train Disassembly equipment for remote operation in the HFEF Main Cell. The test train is composed of five welded stainless steel capsules that are expected to be transported in two sections. The first section will be Capsules 1 and 2. The second section will include Capsules 3, 4 and 5. The capsules range in length from about 9 inches to about 12 inches. Inside the capsules is a graphite holder that contains multiple fueled AGR compacts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Tritium Transport Phenomena in Molten-Salt Reactors

In this work, we review phenomena relevant to tritium transport in molten-salt reactors, which produce tritium from lithium and beryllium salts at significantly higher levels than other reactor types. Modeling of such phenomena began following MSRE operations, and these early models attempted to predict measured tritium distributions in the MSRE, accounting for turbulent mass-transport processes (using established heat transfer correlations), permeation through a variety of metal structures such as heat exchanger tubes, and transport to and from bubbles introduced into the salt by gas sparging. The models reasonably reproduced the MSRE data, but did so best when the permeability of structures was reduced by about a factor of 1,000. This issue does not appear to have been conclusively resolved, and all of the more recent attempts to model tritium transport in molten salts appear to make use of the same methodology. MSRE remains, however, essentially our only source of integral tritium transport data relevant to MSRs. Here, we generalize the MSRE approach to permeation in order to include potential rate-limiting effects at interfaces, as well the effects of added hydrogen. Appropriately non-dimensionalized, this system of equations identifies two dimensionless numbers whose relative values clearly delineate the conditions under which mass transport, surface effects, permeation, and hydrogen swamping are expected to become rate-limiting. We also describe the preliminary conceptual design of a forced convection FLiBe loop, into which tritium would be introduced for the purpose of providing validation data for such a model. The primary purpose of this is to investigate the coupled transport phenomena described above and identify those that are rate-limiting in MSRs. Additional test section configurations are described that would address other transport phenomena relevant to MSRs, including bubbly flows and graphite interactions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

DOE ART-GCR Program Overview

To provide Advanced Reactor Technologies Gas-Cooled Reactor Program: Overview of ART-GCR and Graphite program objectives, status, and activities. Also Identify research areas and outcomes that will benefit stakeholders and clients (HTGR designers, suppliers, regulators, DOE-NE, etc.), and to Identify remaining R&D gaps and future needs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗