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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 OSTI2020

Design and strategy for next-generation silicon carbide composites for nuclear energy

Silicon carbide (SiC) ceramic-based composites continue to be attractive material options for fusion in-vessel components and fission reactor core structures because of their exceptional high-temperature capability and favorable neutronic properties. As performance data accumulates, the limitations of the current generation of nuclear-grade SiC composites are becoming more apparent. These limitations mainly involve strength degradation during high-dose neutron irradiation. In this paper, we discuss several options for improving the performance of the next generation of SiC composites to enhance the radiation resistance, along with new experimental results on neutron irradiation resistance. The main emphasis is on the selection of the fibers and the design and development of alternative interphase layers for advanced composites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Multichannel gas-uptake/evolution reactor for monitoring liquid-phase chemical reactions

The design of a headspace pressure-monitoring reactor for measuring the uptake/evolution of gas in gas–liquid chemical transformations is described. The reactor features a parallel setup with ten-reactor cells, each featuring a low working volume of 0.2–2 ml, a pressure capacity from 0 to 150 PSIa, and a high sensitivity pressure transducer. The reactor cells are composed of commercially available disposable thickwalled glassware and compact monolithic weld assemblies. The software interface controls the reactor temperature while monitoring pressure in each of the parallel reactor cells. Reactions are easy to set up and yield high-density gas uptake/evolution data. This instrument is especially well suited to acquire quantitative time-course data for reactions with small quantities of gas consumed or produced.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Identification of a network of nonlinear interactions as a mechanism triggering the onset of edge localized modes

Edge localized modes (ELMs) remain a critical issue for the lifetime of in-vessel components of fusion reactors such as ITER. We study ELMs triggered when the pedestal parameters are below the peeling-ballooning limit. These ELMs occur while the pedestal recovers after an ELM crash. We identify a mechanism by which a non-linear perturbation involving many active triads—a network of non-linear interactions – leads to the triggering of these ELMs. Each active triad leads to the non-linear transfer of energy between three waves, such that the network of non-linear interactions lead to a transfer of energy between many waves. The triggering of this non-linear perturbation systematically coincides with a pedestal perturbation induced by the neutral beam injection beams and localized near the magnetic surface of safety factor q = 5. Among the non-linear interactions detected during the ELM by a bi-coherence analysis, the most active triad involves a slow frequency that is qualitatively consistent with a geodesic acoustic mode located at the same location, i.e., near the q = 5 surface.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Status Summary of FY16 Atom Probe Tomography Studies on UCSB ATR-2 Irradiated RPV Steels

The University of California Santa Barbara-2 RPV Steel Irradiation experiment was awarded in 2010 by the Nuclear Science User Facility (formerly ATR NSUF) through a competitive peer review proposal process. The experiment involved irradiation of nearly 1300 samples distributed over 13 capsules. The major objective of this experiment was to better understand embrittlement behavior of reactor pressure steels at doses beyond which available data exists yet may be achieved if reactor operating licenses are extended beyond 60 years. The experiment was instrumented during irradiation and active temperature control was used to maintain the temperature at the design temperature. Six samples were selected from a large matrix of materials to perform atom probe tomography (APT) to look at formation of high dose phases. The nature and formation behavior of these phases is discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Oxide Dispersion Strengthened Ferritic Steel Wire Feedstock Development for Large Format Additive Manufacturing - CRADA 620 (Abstract)

This CRADA project funded through DOE’s INFUSE program seeks to demonstrate the viability of fabricating large, complex parts from oxide dispersion strengthened (ODS) steel with advanced manufacturing. Exhibiting excellent radiation tolerance and high mechanical performance at elevated temperatures, ODS steel is a promising structural material candidate for near-plasma components in fusion energy systems. Its use, however, has been limited by a lack of manufacturability. This project will seek to produce ODS steel wire through a solid-state shear assisted extrusion process and then demonstrate that the wire can undergo controlled local melting while being welded with the final part sufficiently retaining the beneficial properties of ODS steel. This would allow the use of wire-arc additive manufacturing (WAAM) to manufacture large-scale ODS parts, even though ODS is currently only available as a powder. WAAM is a promising technique for producing components like the replaceable ARC vacuum vessel in CFS’ fusion reactor design. Meanwhile, this project will also expand PNNL’s capability in producing custom wire feedstock with friction extrusion, enabling downstream large-scale manufacturing with WAAM and solid-state based additive manufacturing.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Oxide Dispersion Strengthened Ferritic Steel Wire Feedstock Development for Larger Format Additive Manufacturing (CRADA Final Report)

This CRADA project funded through DOE’s INFUSE program sought to demonstrate the viability of fabricating large, complex parts from oxide dispersion strengthened (ODS) steel with advanced manufacturing. Exhibiting excellent radiation tolerance and high mechanical performance at elevated temperatures, ODS steel is a promising structural material candidate for near-plasma components in fusion energy systems. Its use, however, has been limited by a lack of manufacturability. This project sought to produce ODS steel wire through a solid-state shear assisted extrusion process and then demonstrate that the wire can undergo controlled local melting while being welded with the final part sufficiently retaining the beneficial properties of ODS steel. This would allow the use of wire-arc additive manufacturing (WAAM) to manufacture large-scale ODS parts, even though ODS is currently only available as a powder. WAAM is a promising technique for producing components like the replaceable ARC vacuum vessel in CFS’ fusion reactor design. Meanwhile, this project will also expand PNNL’s capability in producing custom wire feedstock with friction extrusion, enabling downstream large-scale manufacturing with WAAM and solid-state based additive manufacturing. The project achieved its goals of developing tooling and fixturing to produce ODS wire at smaller diameters than previous projects. Several small lengths of wire of 1.5 mm and 2.5 mm diameter in the range of 2.5-30 mm long were produced at tool temperatures that are known to cause ODS particle coarsening (~1200 °C). Fixtures and tooling for longer (>1 m) wires were developed but further process development is needed reduce tool temperature during extrusions and to increase wire length needed for WAAM testing and development.

36 MATERIALS SCIENCE↗
NASA NTRS1983

Ultrasonic Monitor

MicroUltrasonics PLR-1000 is a refined microprocessor-controlled version (usable on bolts, plates, liquids and gases) of the P2L2 developed by Langley Research Center. New technique is for nondestructive measurement of residual stress in various types of structures, for example, nuclear pressure vessels, pipes in nuclear reactors, offshore platforms, bridges, railroad tracks and wheels, aircraft wings, and engines. The instrument produces sound tone pulses that travel through a test specimen. PLR-1000 precisely measures speeds as stress increases speed of sound, tone changes, making precise measurements possible.

Source record↗
NASA NTRS2014

Alteration of Basaltic Glass to Mg/Fe-Smectite under Acidic Conditions: A Potential Smectite Formation Mechanism on Mars

Phyllosilicates of the smectite group including Mg- and Fe-saponite and Fe(III)-rich nontronite have been identified on Mars. Smectites are believed to be formed under neutral to alkaline conditions that prevailed on early Mars. This hypothesis is supported by the observation of smectite and carbonate deposits in Noachian terrain on Mars. However, smectite may have formed under mildly acidic conditions. Abundant smectite formations have been detected as layered deposits hundreds of meters thick in intracrater depositional fans and plains sediments, while no large deposits of carbonates are found. Development of mildly acidic conditions at early Mars might allow formation of smectite but inhibit widespread carbonate precipitation. Little is known regarding the mechanisms of smectite formation from basaltic glass under acidic conditions. The objective of this study was to test a hypothesis that Mars-analogue basaltic glass alters to smectite minerals under acidic conditions (pH 4). The effects of Mg and Fe concentrations and temperature on smectite formation from basaltic glass were evaluated. Phyllosilicate synthesis was performed in batch reactors (Parr acid digestion vessel) under reducing hydrothermal conditions at 200 C and 100 C. Synthetic basaltic glass with a composition similar to that of the Gusev crater rock Adirondack (Ground surface APXS measurement) was used in these experiments. Basaltic glass was prepared by melting and quenching procedures. X-ray diffraction (XRD) analysis indicated that the synthesized glass was composed of olivine, magnetite and X-ray amorphous phase. Samples were prepared by mixing 250 mg Adirondack with 0.1 M acetic acid (final pH 4). In order to study influence of Mg concentration on smectite formation, experiments were performed with addition of 0, 1 and 10 mM MgCl2. After 1, 7 and 14 day incubations the solution composition was analyzed by Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and the altered glass and formed phyllosilicates were examined by XRD analysis. Mineralogical changes were significant in Adirondack incubated with 10 mM MgCl2 at pH 4 and heated at 200 C. X-ray diffraction analysis revealed formation of phyllosilicate during 14 day incubation (Figure 1). Smectite was confirmed as the phyllosilicate after treatments with glycerol and KCl and heating to 550 C. The position of 02l (4.60 A) and 060 (1.54 A) diffraction bands were indicative of trioctahedral smectite such as saponite. Analysis of solution composition demonstrated that aqueous concentration of Mg decreased from 10 mM to approx.4 mM after 7 day incubation likely due to saponite formation. Smectite also formed in Adirondack incubated with 0 mM MgCl2 at pH 4 and heated at 200 C. However, diffraction peak positions of 02l (4.52 A) and 060 (1.51 A) suggested formation of dioctahedral nontronite. The 100degC Mg and Fe(II) treated basaltic glass experiments are ongoing and results will be presented.

Peretyazhko, Tanya↗
DOE OSTI2018

U.S. Efforts in Support of Examinations at Fukushima Daiichi - 2018 Evaluations

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station [Daiichi]. Clearly, the accident signatures for each of these units differed. Some of this uncertainty about debris end-state can be attributed to a lack of information related to cooling system operation and cooling water injection during the events. There is also uncertainty in our understanding of phenomena affecting: (a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs) [compared with Pressurized Water Reactors (PWRs)], and (b) accident progression after vessel failure (ex-vessel progression) for BWRs and PWRs. These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Assembly and Delivery of Rabbit Capsules for Irradiation of Reinforced Radiation Resistant SiC-SiC Composites in the High Flux Isotope Reactor

The development of reinforced radiation resistant composites presents a critical challenge for in-vessel component application in advanced high temperature reactors. In its pursuit of this, Physical Optics Corporation (POC) is interested in developing a silicon carbide (SiC)-based ceramic matrix composite (CMC) with improved thermal conductivity in a high neutron radiation environment. The US Department of Energy (DOE) is supporting research efforts to develop Advanced Reactor Technology. Ultimately, the results of this project will determine the viability of using SiC matrix composites in hot structures in advanced high temperature reactors. The first objective of this project is to irradiate torsion and bend bar specimens in the High Flux Isotope Reactor (HFIR) and develop an initial data set to evaluate the evolution of composite mechanical strength, dimensional changes, and thermal conductivity. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of three rabbit capsules to the HFIR. Rabbits of both torsion and bend bar configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Considerations for Maritime Nuclear Technologies, Economic Viability and Public Acceptance

The Maritime Nuclear Application Group (MNAG) is a working group convened by the National Reactor Innovation Center at Idaho National Laboratory (INL), the American Bureau of Shipping, and Morgan, Lewis, and Bockius LLP. This report documents an MNAG examination of considerations relevant to implementing nuclear technology in commercial maritime applications. In general, two types of use case are examined: maritime nuclear power plants and nuclear reactors used on board shipping vessels for propulsion and other ship needs. The report finds that there may be economic benefits related to maritime nuclear technologies, including the flexible deployment of maritime nuclear reactors, which would allow them to complement land-based nuclear projects, and operational differences for nuclear cargo ships that may lead to an overall increase in revenue. High-level analyses in this report show that, based on general small modular reactor and microreactor cost estimates developed by INL, maritime nuclear reactors may be economically competitive for electricity production in remote regions and for use in the propulsion of large cargo ships. Besides economic viability, public acceptance will be key to implementing maritime nuclear technologies. The report discusses the public’s current perception of nuclear technologies. Engaging with the public will be important to improving this perception. The report discusses some key benefits and risks associated with maritime nuclear technologies. Benefits include the creation of jobs, the production of reliable energy, and the potential to improve air quality. Risks that concern the public are the potential for radioactive releases during operation and decommissioning, as well as those related to waste management. Communicating the benefits and the risks of maritime nuclear technologies will be essential to improving public perception.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Final Report on Development of Microwave Resonant Cavity Transducer for Fluid Flow Sensing: Development of Microwave Cavity Flow Meter for Advanced Reactor High Temperature Fluids

We are investigating a microwave resonant cavity transducer for flow sensing in the vessel of a high temperature fluid advanced reactor (AR), such as a molten salt cooled reactor (MSCR) or a sodium fast reactor (SFR). This transducer is a hollow metallic cylindrical cavity, with one of the flat walls of the cylinder flexible enough to undergo microscopic deflection due to dynamic fluid pressure. Membrane deflection leads to a shift in the microwave resonant frequency, which can be detected with a spectrum analyzer. We have developed a continuum electromechanics model of the microwave resonant cavity transducer performance. The deflection of the membrane is calculated using equations from the theory of plates and shells for deflection of radially constrained thin circular disk under uniform load. The microwave frequency shift is calculated using the equations from the microwave cavity perturbation theory. The transducer model was validated with proof-of-concept experimental data of water flow sensing with a Brass 360 right circular cylinder with 22.2mm diameter and 203µm thick wall. The cavity was excited through a subwavelength hole in TE 011 mode with resonant frequency f ≈ 17.8GHz. Subsequently, we have performed a preliminary proof-of-principle test of flow sensing in high temperature liquid sodium in environment. For this test, we have developed a cylindrical resonator with the same dimensions as for the Brass 360 cavity. The resonator for high temperature liquid sodium test was machined from stainless steel 316 and electroplated with silver on the interior surfaces. We have also developed and insertion probe consisting of a 50cm WR-42 brass waveguide enclosed in a protective SS316 tube. The cavity was excited through a subwavelength hole on the side of the wall of the cylinder in the TE 011 mode with resonant frequency f ≈ 17.8GHz. The liquid sodium setup consists of a cylindrical vessel with a center feed line, where a transducer inserted through the top cap of the vessel measures velocity of the impinging liquid jet. The sodium flow sensing study was performed in a liquid sodium vessel at 340°C temperature and ambient pressure. The flow rate was changed by varying sodium pump power, and a frequency shift of several hundred kHz was observed. The transducer was removed after 70 days of testing. No visual damage was observed, and the electromagnetic response remained the same. As a calibration experiment, we have assembled a water vessel with a center feed, and with dimensions similar to those of the liquid sodium setup. Water flow sensing was performed before and after liquid sodium test. The responses in both cases are similar and agree with COMSOL computer simulations. Because long-term (multi-year) experimental tests of transducer resilience to harsh environment are not practical, we have developed a probabilistic model of creep to estimate transducer resilience to the harsh environment. The probabilistic model considers diffusion creep under the condition of high temperature and low stress, where the stress and temperature are allowed to be random variables with Gaussian distributions. Using the probabilistic model, we estimate inelastic membrane deflections due to creep for several temperature ranges. We conclude that for temperatures less than 650°C, creep has negligible long-term effect on the transducer performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. In this work, the analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Status Report on Ex-Vessel Coolability and Water Management

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗