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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

AGR-1 Daily As-run Thermal Analyses

This report documents the results of thermal analyses to predict the daily as-run temperatures for the AGR-1 experiment. Control gas gaps and compact-graphite holder gas gaps were modeled to change linearly with time. The control gas gap widths were determined by comparison to measured R/B data. The compact graphite holder gas gaps were determined from PIE measurements. Daily heat rates for each compact and component in the models were input from daily as-run physics analyses. Daily gas compositions and component fast neutron fluences were also input. Six different finite element models were created for the six different AGR-1 capsules. Each capsule had a different gas gap that was implemented to control the temperature of the experiment. Capsules on the top and bottom had larger gas gaps, while capsules in the middle were smaller.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A Code to Code Benchmark for HTGR Fuel Element Depletion

The current lack of well-qualified experimental measurements for spent HGTR fuel elements limits the validation of advanced doubly heterogeneous depletion methods. Because of this shortage of data, a benchmark was developed as the first, simplest phase in a planned series of increasingly complex set of code-to-code benchmarks. The intent of this benchmark was to encourage submission of a wide range of computational results for depletion calculations in a set of basic fuel cell models. Comparison of results using independent methods and data should provide insight into potential limitations in various modeling approximations. The benchmark seeks to provide the simplest possible models, in order to minimize the effect of competing and potentially offsetting phenomena that might mask weaknesses in given methods. Twenty-one sets of results were submitted by twelve participants internationally. Various approaches were used for the treatment of microscopic grains dispersed in fuel elements in the second and third parts of the benchmark, including explicit treatment of grains distributed within a regular lattice, a randomized distribution of explicit grains, reactivity-equivalent physical transformation and a two-pass homogenization with disadvantage factors. The broad agreement appears to validate each of the double heterogeneity treatments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

The Degradation of Strength under Varying Oxidizing Conditions for Nuclear Graphite

Strength behavior of fine and medium grain size nuclear graphite grades are presented after being subjected to varying oxidation conditions. This study provides valuable information for design and eventual licensing considerations for the oxidation behavior of graphite components within a gas-cooled High Temperature Reactor design and the eventual structural integrity of graphite core components. These tests address the underlying oxidized strength of nuclear graphite components for oxidized mass losses up to and beyond the current limits recommended for graphite components in the ASME code (maximum mass loss = 10%) over a range of oxidation temperatures (550° C – 750° C). Preliminary results generally demonstrate that low temperature oxidation can result in higher strength reductions than oxidation at higher temperatures for similar mass loss levels. Since the graphite component strength is a critical parameter necessary to determine the core structural integrity, determining the actual residual strength after oxidation is essential for determining actual component oxidation limits within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

AGR-5/6/7 Fuel Fabrication Feedstock Chemical Purity Specifications

This specification has been prepared to provide fuel fabricators with minimum chemical purity specifications intended to ensure that fuel components have consistent properties and that impurity levels in the as-fabricated fuel is sufficiently low to minimize in-pile TRISO particle failures and to enhance fission product retention. Process chemical purity specifications are given for the following processes: • Kernel forming and sintering • TRISO particle coating • Matrix preparation • TRISO overcoating • Compact carbonization and heat treatment

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pronghorn Theory Manual

This manual is intended to provide a thorough description of the governing equations; the relations used for material, heat transfer, and fluid flow parameters; and the numerical method utilized in Pronghorn. Detailed derivations of the governing equations are given, beginning from first principles in order to be as explicit as possible about assumptions made along the way and guide any future methods extensions. Additional explanatory or tangential information that is not required for detailed understanding of the governing equations used in Pronghorn is shown in gray boxes in the text. The convention used in this document is to represent all class names (or file names) in typewriter font.

42 ENGINEERING↗
NASA NTRS1995

The Clementine Mission

The Clementine Mission is summarized from purposes of the mission through details of the instruments. The mission was a joint Department of Defense and NASA mission to use lighter instruments and other equipment developed by the military for space exploration. New lightweight items on Clementine include batteries, optical sensors, reaction wheels, star trackers, release systems, and inertial measurement units. Other items include a compression chip set, a solid state data recorder, autonomous navigation, and a 32-bit RISC computer.

space↗
DOE OSTI2016

Communicating with residential electrical devices via a vehicle telematics unit

A method of communicating with residential electrical devices using a vehicle telematics unit includes receiving information identifying a residential electrical device to control; displaying in a vehicle one or more controlled features of the identified residential electrical device; receiving from a vehicle occupant a selection of the displayed controlled features of the residential electrical device; sending an instruction from the vehicle telematics unit to the residential electrical device via a wireless carrier system in response to the received selection; and controlling the residential electrical device using the sent instruction.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Optimizing vehicle recharging to maximize use of energy generated from particular identified sources

Implementations of the present invention contemplate utilizing a communicative infrastructure represented by a telematics unit of an electric vehicle and a telematics service provider (TSP) communicatively coupled to the telematics unit to control an amount of power reserved for a charging event of the vehicle. Specifically, the TSP utilizes information provided by the telematics unit, electrical power utilities, and electrical power metering devices to determine the characteristics of electrical energy available for charging the vehicle. The TSP can utilize the characteristics of the available energy along with additional information, such as charging preferences established by a subscriber of the TSP and vehicle use schedules, to determine an amount of power to reserve for the charging event. The amount of power reserved for the charging event may be determined based on cost, greenhouse gas emissions, a combination thereof, and on other factors.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2018

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

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DOE OSTI2017

AGR-5/6/7 Fuel Specification

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) program is developing and qualifying fuel for a first of a kind high temperature gas-cooled reactor (HTGR) and associated systems to generate electrical power and produce hydrogen or process heat. One of the critical technology development areas of the program is nuclear fuel development and qualification. Fuel qualification is needed to support licensing, operation, and near term deployment of an HTGR reactor.5 Toward this end, the Advanced Gas Reactor (AGR) program includes manufacturing fuels for a series of irradiation experiments in the Advanced Test Reactor (ATR) at INL.5 The fifth, sixth, and seventh irradiation experiments, called “AGR 5/6/7,” will use fuel manufactured entirely in engineering scale equipment to demonstrate the performance of fuels fabricated in a production environment. This fuel will be composited from multiple low enriched uranium carbide/oxide (LEUCO) kernel forming runs, multiple tristructural isotropic (TRISO) coating runs, and multiple overcoating runs. The fuel will be composited into lots as kernels, TRISO particles, and compacts. Previous AGR fuel specifications1,2,8,9 provide the basis for most of the following fuel specifications, which are specific to the AGR 5/6/7 experiment, and may not be immediately applicable to fuel produced for an HTGR reactor. Fuel specifications will be influenced by the final reactor design.

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DOE OSTI2018

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation IV (GenIV) pre-irradiation leach-burn-leach round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

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DOE OSTI2017

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2018, the activities for the INL ART TDO fuel development are focused on final characterization to support the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea.

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DOE OSTI2017

AGR-5/6/7 Fuel Fabrication Feedstock Chemical Purity Specifications

This specification has been prepared to provide fuel fabricators with minimum chemical purity specifications intended to ensure that fuel components have consistent properties and that impurity levels in the as-fabricated fuel is sufficiently low to minimize in-pile TRISO particle failures and to enhance fission product retention. Process chemical purity specifications are given for the following processes: • Kernel forming and sintering • TRISO particle coating • Matrix preparation • TRISO overcoating • Compact carbonization and heat treatment

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DOE OSTI2017

Support for INL ART TDO Advanced Gas Reactor Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) since the inception of the program. In this effort, ORNL has fabricated tri-structural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2 and AGR-3/4 experiments and for university studies, and provided characterization, coating, and compaction support for Babcock and Wilcox (B&W). In FY 2017, the activities for the Very-High Temperature Reactor (VHTR) Technology Development Office (TDO) fuel development are focused on preparations for the AGR-5/6/7 experiments. With the successful transfer of fuel fabrication technology to B&W, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing support for fabricating surrogate fuel materials international collaborations. ORNL will also provide analytical support to B&W for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

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DOE OSTI2020

NDMAS System and Process Description

The U. S. Department of Energy (DOE) has made a significant investment in research to develop the next generation of reactor technologies as well as to improve the performance and lengthen the life cycle of existing nuclear reactors. Data collected to demonstrate new concepts may also be used in the future to support licensing of these technologies. Provenance of these data must be preserved. The Nuclear Data Management and Analysis System (NDMAS) was established to manage and preserve data collected by fuels and materials research conducted by the high-temperature, gas-cooled reactor program. The scope of NDMAS is expanding to include other nuclear research programs that have the shared need to preserve the provenance of research data. Nuclear research funded by DOE is conducted by Idaho National Laboratory (INL), universities, other national laboratories, foreign research partners, and private companies. This research will generate a large amount of data from a variety of sources over a period of many years. Managing the data generated by the research and development projects presents a significant challenge for retaining data integrity and availability.

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DOE OSTI2016

Statistical Sampling Plan for AGR-5/6/7 Fuel Materials

Kernels, coated particles, and compacts will be manufactured to predefined specifications for irradiation in the AGR 5/6/7 experiments. Many tests will be performed for quality assurance and most of these tests correspond to criteria that have to be met with 95% confidence, based on random samples. This report describes the statistical acceptance testing methods and gives minimum sample sizes for kernels to achieve 95% confidence. Coated particles and compacts will be addressed in revisions to this report. The methods used for sample size determination are described in detail in INL/EXT 05 00349, Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials (Einerson 2005). The statistical approach is to collect a representative sample, apply an acceptance test to the sample, and accept or reject the product based on this test for each criterion. The criteria or “product specifications” are contained in SPC-1352, “AGR 5/6/7 Fuel Specification,” Revision 1 and are duplicated in this plan.

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DOE OSTI2020

OECD/NEA COUPLED NEUTRONIC/THERMAL-FLUIDS BENCHMARK OF THE MHTGR-350 MW CORE DESIGN: RESULTS FOR PHASE-I EXERCISE 1

This report presents the results from code-to-code comparisons for a prismatic modular high-temperature gas-cooled reactor 350-MW design for steady-state neutron transport calculations. Integral parameters as well as power and flux distributions are provided. Differences between solutions from diffusion and transport operators are analysed. Effects of control rod homogenization between full block and one sixth block are discussed.

42 ENGINEERING↗