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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 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

Grain size control of rhenium strip

Ensuring the desired grain size in the pure Re strip employed by the SP-100 space nuclear reactor design entails the establishment of an initial grain size in the as-received strip and the avoidance of excessive grain growth during subsequent fabrication. Pure Re tapered tensile specimens have been fabricated and tested in order to quantify the effects of grain-boundary migration. Grain size could be rendered fine and uniform by means of a rolling procedure that uses rather large reductions between short intermediate anneals. The critical strain regime varies inversely with annealing temperature.

Schuster, Gary B.↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Design and testing of an enriched uranium fueled molten salt irradiation vehicle

Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has started being reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched-uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of and enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13cm 3 of UCl 3 -NaCl salt (93% enrichment) generating around 20 W/cm 3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525-900°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Irradiation of an enriched uranium (NaCl-UCl 3 ) fuel salt capsule, summary of nondestructive post-irradiation examinations, and solidification modeling

Molten salt reactors (MSRs) are gaining attention due to their potential for safe, carbon-free nuclear energy with reduced waste. However, licensing these reactors is hindered by limited experimental data on fueled salts, both pre- and post-irradiation. Here, the novel Molten-salt Research Temperature-controlled Irradiation (MRTI) vehicle was designed to address knowledge gaps in irradiating enriched‑uranium-bearing salts. The MRTI experiment irradiated 13 cm 3 of UCl₃-NaCl (93 % U-235) salt in the Neutron Radiography (NRAD) Reactor, achieving a burnup of 0.196 GWd/MTU over 390 h. Despite a heater failure, the thermocouple data suggested fission heat kept the salt molten. The MRTI assembly was remotely disassembled for nondestructive post-irradiation examination (PIE), which included precision gamma-ray scanning (PGS) and neutron radiography. Radiograph images showed the location of the salt and the solidification pattern. PGS results provided an early indication that activated materials of construction did not increase in relative intensity in the region of the capsule where the salt was in contact with the material of construction. Additionally, PGS data showed the presence of several gamma emitting fission products, such as Nb-95, Zr-95, Ru-103, Ce-141, and La-140, where Ru-103 had the highest counts at the bottom of the capsule. Computational fluid dynamics modeling supported observations of salt solidification patterns and proved to be a valuable tool to inform PIE activities. The MRTI experiment has thus far provided critical data and lessons learned for fuel salt PIE activities, essential for advancing the technical readiness of MSRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

ROADRUNNER MiniFuel Experiment: Readiness for Insertion

The Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments (ROADRUNNER) campaign is a collaborative irradiation experiment designed to evaluate the performance of uranium nitride (UN) fuel fabricated with controlled variations in density and impurity content. Oak Ridge National Laboratory’s MiniFuel irradiation platform and the High Flux Isotope Reactor (HFIR) are being used to irradiate six targets containing 36 UN miniature disks. These targets have been assembled to systematically investigate burnup and temperature effects under accelerated irradiation conditions. This report documents the finalized test plan, describes the experiment assembly and subcapsule integration process, and confirms the readiness of all ROADRUNNER targets for HFIR insertion. All targets have passed quality control procedures, and five have begun irradiation. The campaign will provide critical performance data to support fuel model validation and future UN qualification efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Infrastructure and Testing Needs for Molten Salt Reactor Safeguards

Molten salt reactors (MSRs) are an attractive advanced reactor technology because of their increased fuel utilization and efficiency, the ease of fueling, and the potential to use alternative fuel cycles, including thorium and breeder/burner scenarios. Since an MSR has never been domestically operated for commercial electricity production, minimal infrastructure exists to support performance and materials testing, reactor licensing (including research, test, and operating reactors), and technology development in general. These factors could present challenges for the commercialization of the proposed design technology. Several research and development efforts and investments have begun to identify and supplement the infrastructure to accommodate the molten salt concept, including the United States Department of Energy Office of Nuclear Energy Molten Salt Reactor Program. This report reviews safeguards technology and domestic infrastructure to support the evaluation of nuclear safeguards (primarily material control and accounting) measures and approaches for salt-fueled MSRs. Recommendations for potential priority infrastructure investments, including safeguards measurement test beds and future research, are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Model MC&A for Pebble Bed Reactors (Technical Direction No. 5 Task 2.6 Letter Report)

In preparation for non-light water reactor (non-LWR) activities, US Nuclear Regulatory Commission (NRC) staff are advancing risk-informed and performance-based licensing approaches and addressing key policy issues. One non-LWR reactor concept is a pebble bed reactor (PBR). This reactor design uses spherical fuel elements (pebbles) that are continually added to and removed from the reactor core. The free movement of the fuel in this design presents new challenges for material control and accounting (MC&A) programs. Therefore, an assessment of MC&A program features and measures for a PBR was performed to help NRC staff develop associated MC&A regulations or regulatory guides. The current regulatory framework for non-LWR fuel cycles excludes support for licensing reviews for MC&A programs for PBRs. Licensing reviews of an MC&A program for PBRs can be facilitated by (1) a model MC&A program for a PBR based on identification and assessment of MC&A program features and recommended measures for a reference PBR and (2) a methodology for assessing MC&A performance that can help assess different MC&A program features and measures. This report supports the NRC’s non-LWR Vision and Strategy Near-Term Implementation Action Plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Nuclear Data Management and Analysis System Plan

The United States Department of Energy Advanced Reactor Technologies Program was formed in Fiscal Year 2015 and encompasses the Next Generation Nuclear Plant Project and Very High Temperature Reactor (VHTR) Program as they were known previously. The VHTR Program was created to support design and licensing of the first VHTR nuclear plant. Data created for and used by the program must be qualified for use, stored in a readily accessible electronic form, categorized to assure the correct data are used, and controlled to prevent data corruption or inadvertent changes. The Nuclear Data Management and Analysis System was designed to support the data needs of the VHTR Program, at the time and now the Advanced Reactor Technologies Program. Since its inception, use of the Nuclear Data Management and Analysis System has expanded to support additional projects and programs with similar requirements for control, analysis, and availability of large data sets.

99 GENERAL AND MISCELLANEOUS↗
NASA NTRS2000

The Sugar Model: Catalytic Flow Reactor Dynamics of Pyruvaldehyde Synthesis from Triose Catalyzed by Poly-L-Lysine Contained in a Dialyzer

The formation of pyruvaldehyde from triose sugars was catalyzed by poly-L-lysine contained in a small dialyzer (100 MWCO) suspended in a much larger triose substrate reservoir. The polylysine confined in the dialyzer functioned as a catalytic flow reactor that constantly brought in triose from the substrate reservoir by diffusion to offset the drop in triose concentration within the reactor caused by its conversion to pyruvaldehyde. A 400 mM solution of poly-L-lysine contained in a 0.35 ml dialyzer placed in a 120 ml solution of triose substrate (pH 5.5, 40 C) generated pyruvaldehyde 11 -times faster than an a control reaction without the catalytic dialyzer. However, since the catalytic dialyzer's volume was 343-times smaller than the control reaction, the synthetic intensity (rate/volume) of pyruvaldehyde synthesis within the catalytic dialyzer was 3400-times greater than that of the control reaction and substrate solution. A similar result was obtained using a dialyzer with a 500 MWCO value. Acting as a catalytic flow reactor the polylysine catalytic dialyzer synthesized about 3.5 molecules of pyruvaldehyde per lysine residue in 7 days -- an amount of triose equal to twice the weight of the catalyst. At 7 days the catalytic activity of polylysine was 16% of its initial value, a result indicating catalyst-poisoning caused by reaction of pyruvaldehyde with the e-amino groups of polylysine. The dialyzer method of catalyst containment was selected it provides a simple, flexible, and easily manipulated experimental system for studying the dynamics and evolutionary development of confined autocatalytic processes related to the origin of life under anaerobic conditions.

Weber, Arthur L.↗
DOE OSTI2024

TRISO Fuel’s Safety Functions, Contributions to Reactor Safety, and Necessary Safety Limits

Safety functions are the actions, passive or active, that structures, systems, and components of nuclear facility that contribute to the safety of the workers, the public, or the environment. Well-defined safety functions are the foundation of a solid safety case for a reactor. For reactors that use TRISO-coated particles, the TRISO fuel plays an important part of the safety case because of its ability to contain radionuclides in the fuel itself. This ability enables the use of a functional containment strategy for the reactor where radionuclide retention is the primary safety function supported by the safety functions of controlling reactivity control and controlling heat rejection. This paper establishes at a deeper level the role that TRISO fuel plays in each of these safety functions and associated quality assurance and testing requirements for TRISO particle manufacturing to ensure these safety functions. Safety limits necessary to protect these safety functions include: operational limits, time at temperature limits, and fission gas release activity limits. In conclusion, this approach demonstrates the role that specific aspects of TRISO fuel play in protecting the safety of workers, the public, and the environment.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Irradiation Vehicles for Evaluating SiC/SiC Cladding Bowing Under Neutron Flux Gradients

Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are among the most promising candidates for long term accident-tolerant nuclear fuel cladding. A key challenge related to their deployment is lateral bowing caused by differential radiation-induced swelling under dose or temperature gradients, which could obstruct coolant flow or interfere with control rod/blade movements. Although previous modeling efforts have predicted bowing behavior in light-water reactor (LWR) environments, experimental validation remains limited, especially at prototypic LWR temperatures. This study addresses that gap by irradiating six reduced-length SiC/SiC cladding tubes (~600 mm) in the High Flux Isotope Reactor (HFIR) at ~300°C, which is representative of LWR conditions. The tubes will be housed in a sealed vessel with an inert gas gap to maintain target temperatures and prevent direct coolant contact. Arranged in three pairs, each set will receive a different radiation dose (low, medium, high), with the central pair receiving ~0.1 displacements per atom (dpa)—the expected dose for peak bowing. The experiment will determine the dose-dependent bowing behavior and validate predictive finite element models. In this work, the tubes are freely suspended from pins to allow for unconstrained bowing; however, we present a concept for introducing localized constraints to represent grid spacer effects. Post-irradiation examination will include swelling measurements and profilometry to assess bowing and compare the results with model predictions. This work aims to confirm the conditions under which maximum bowing occurs so as to improve the reliability of SiC/SiC performance models in nuclear applications.

Russell, Nick [ORNL] (ORCID:0000000296099820)↗
NASA NTRS1975

Bosch CO2 Reduction System Development

Refinements in the design of a Bosch CO2 reduction unit for spacecraft O2 production are described. Sealing of the vacuum insulation jacket was simplified so that high vacuum and high insulation performance are easily maintained. The device includes a relatively simple concentric shell recuperative heat exchanger which operates at approximately 95% temperature effectiveness and helps lower power consumption. The influence of reactor temperature, pressure, and recycle gas composition on power consumption was investigated. In general, precise control is not required since power consumption is not very sensitive to moderate variations of these parameters near their optimum values. There are two process rate control modes which match flow rate to process demand. Catalyst conditioning, support, and packing pattern developments assure consistent starts, reduced energy consumption, and extended cartridge life. Operation levels for four or five men were maintained with overall power input values of 50 to 60 watts per man.

Holmes, R. F.↗
NASA NTRS1977

Design of single-winding energy-storage reactors for dc-to-dc converters using air-gapped magnetic-core structures

A procedure is presented for designing air-gapped energy-storage reactors for nine different dc-to-dc converters resulting from combinations of three single-winding power stages for voltage stepup, current stepup and voltage stepup/current stepup and three controllers with control laws that impose constant-frequency, constant transistor on-time and constant transistor off-time operation. The analysis, based on the energy-transfer requirement of the reactor, leads to a simple relationship for the required minimum volume of the air gap. Determination of this minimum air gap volume then permits the selection of either an air gap or a cross-sectional core area. Having picked one parameter, the minimum value of the other immediately leads to selection of the physical magnetic structure. Other analytically derived equations are used to obtain values for the required turns, the inductance, and the maximum rms winding current. The design procedure is applicable to a wide range of magnetic material characteristics and physical configurations for the air-gapped magnetic structure.

Ohri, A. K.↗
NASA NTRS1981

Low current linearization of magnetic amplifier for dc transducer

A magnetic amplifier having two saturable reactor cores with a separate excitation winding on each connected in series opposition, a common control winding, and a common output winding, is adapted for use as a low level signal transducer. The separate excitation windings are excited in push-pull mode through a center tapped transformer, and at least one diode is included in series with a load resistor connected to the output winding. A resistor in series with the output winding and load resistor is connected between the center tap of the excitation transformer and the connection between the two excitation windings of the saturable cores. This series resistor provides a return current path for the output winding and allows the excitation windings of the saturable cores to operate as primary windings of transformers.

Nagano, S.↗
NASA NTRS1992

Removing Undesired Fine Powder From Silicon Reactor

Fluidized-bed reactor produces highly pure polycrystalline silicon particles with diameters approximately greater than 400 micrometers. Operates by pyrolysis of silane in reaction zone, which is bed of silicon seed particles fluidized by flow of silane and carrier gas. Above reaction zone, gas mixture flows rapidly enough to entrain silicon powders, but not larger seed and product particles. Entrained particles swept out of reactor. Applicable to other processes such as production of fine metal and ceramic powders where control of sizes of product needed.

Flagella, Robert N.↗
NASA NTRS1999

Feature Profile Evolution of SiO2 Trenches In Fluorocarbon Plasmas

Etching of silicon microstructures for semiconductor manufacturing in chlorine plasmas has been well characterized. The etching proceeds in a two-part process, where the chlorine neutrals passivate the Si surface and then the ions etch away SiClx. However, etching in more complicated gas mixtures and materials, such as etching of SiO2 in Ar/C4F8, requires knowledge of the ion and neutral distribution functions as a function of angle and velocity, in addition to modeling the gas surface reactions. In order to address these needs, we have developed and integrated a suite of models to simulate the etching process from the plasma reactor level to the feature profile evolution level. This arrangement allows for a better understanding, control, and prediction of the influence of equipment level process parameters on feature profile evolution. We are currently using the HPEM (Hybrid Plasma Equipment Model) and PCMCM (Plasma Chemistry Monte Carlo Model) to generate plasma properties and ion and neutral distribution functions for argon/fluorocarbon discharges in a GEC Reference Cell. These quantities are then input to the feature scale model, Simulation of Profile Evolution by Level Sets (SPELS). A surface chemistry model is used to determine the interaction of the incoming species with the substrate material and simulate the evolution of the trench profile. The impact of change of gas pressure and inductive power on the relative flux of CFx and F to the wafer, the etch and polymerization rates, and feature profiles will be examined. Comparisons to experimental profiles will also be presented.

Hwang, Helen↗
NASA NTRS2016

Hydrogen Purification and Recycling for an Integrated Oxygen Recovery System Architecture

The United States Atmosphere Revitalization life support system on the International Space Station (ISS) performs several services for the crew including oxygen generation, trace contaminant control, carbon dioxide (CO2) removal, and oxygen recovery. Oxygen recovery is performed using a Sabatier reactor developed by Hamilton Sundstrand, wherein CO2 is reduced with hydrogen in a catalytic reactor to produce methane and water. The water product is purified in the Water Purification Assembly and recycled to the Oxygen Generation Assembly (OGA) to provide O2 to the crew. This architecture results in a theoretical maximum oxygen recovery from CO2 of approx.54% due to the loss of reactant hydrogen in Sabatier-produced methane that is currently vented outside of ISS. Plasma Pyrolysis Assembly (PPA) technology, developed by Umpqua Research Company, provides the capability to further close the Atmosphere Revitalization oxygen loop by recovering hydrogen from Sabatier-produced methane. A key aspect of this technology approach is the need to purify the hydrogen from the PPA product stream which includes acetylene, unreacted methane and byproduct water and carbon monoxide. In 2015, four sub-scale hydrogen separation systems were delivered to NASA for evaluation. These included two electrolysis single-cell hydrogen purification cell stacks developed by Sustainable Innovations, LLC, a sorbent-based hydrogen purification unit using microwave power for sorbent regeneration developed by Umpqua Research Company, and a LaNi4.6Sn0.4 metal hydride produced by Hydrogen Consultants, Inc. Here we report the results of these evaluations to-date, discuss potential architecture options, and propose future work.

Abney, Morgan B.↗